ANL-81-13 MAStER - UNT Digital Library/67531/metadc283342/... · Distribution Category: Nuclear...

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ANL-81-13 /

MAStERFUEL CY

QUARTERLY

CLE PROGRAMS

'ROGRESS REPORT

October-December 1980

by

M. J. Steindler, Seymour Vogler, G. F. Vandegrift, Jacqueline Williams,

T. J. Gerding, L. J. Jv'rdine, J. K. Bates, J. E. Kincinas, W. J. Mecham,

R. H. Pelto, M. G. Seitz, R. A. Couture, N. M. Levitz, T. F. Cannon,

P. G. Deeken, R. E. Nelson, J. E. Parks, L. E. Trevorrow, C. G. Wach,

I. 0. Winsch, J. E. Fagan, Henry Lautermilch, and R. E. Brock

G N

AR

ARGONNE NATIONAL LABORATORY, ARGONNE, ILLINOIS

Prepared for the U. S. DEPARTMENT OF ENERGYunder Contract W-31-109-Eng-38 _

ANL-81-13

The facilities of Argonne National Laboratory are owned by the United States Government. Under theterms of a contract (W-31-109-Eng-38) among the U. S. Department of Energy, Argonne UniversitiesAssociation and The University of Chicago, the University employs the staff and operates the Laboratory inaccordance with policies and programs formulated, approved and reviewed by the Association.

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This report was prepared as an account of work sponsored by anagency of the United States Government. Neither the UnitedStates Government nor any agency thereof, nor any of theiremployees, makes any warranty, express or implied, or assumesany legal liability or responsibility for the accuracy, com-pleteness, or usefulness of any information, apparatus, product,or process disclosed, or represents that its use would not infringeprivately owned rights. Reference herein to any specific com-mercial product, process, or service by trade name, trademark,manufacturer, or otherwise, does not necessarily constitute orimply its endorsement, recommendation, or favoring by theUnited States Government or any agency thereof. The views andopinions of authors expressed herein do not necessarily state orreflect those of the United States Government or any agencythereof.

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Distribution Category:Nuclear Waste Management

(UC-70)

ANL-81-13

ARGONNE NATIONAL LABORATORY9700 South Cass Avenue

Argonne, Illinois 60439

FUEL CYCLE PROGRAMSQUARTERLY PROGRESS REPORT

October -December 1980

by

M. J. Steindler, Seymour Vogler, G. F. Vandegrift, Jacqueline Williams,T. J. Gerding, L. J. Jardine, J. K. Bates, J. E. Kincinas, W. J. Mecham,

R. H. Pelto, M. G. Seitz, R. A. Couture, N. M. Levitz, T. F. Cannon,P. G. Deeken, R. E. Nelson, J. E. Parks, L. E. Trevorrow, C. G. Wach,


Chemical Engineering Division

July 1981

DISCLAIMER

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Previous reports in this series

ANL-80-61ANL-80-76ANL-80-92ANL-80-114

October-Decer.oer 1979January-rear ch 1980

April-June 1980July-September 1980

TABLE OF CONTENTS

Page

ABSTRACT . .

SUMMARY................ .... . ........ . ..

I. DEVELOPMENT OF INTERIM HIGH LEkVL WASTE FORMS. . . . . . . . . .

A. Introduction .0. . . . . . . . . . . . . . . . . . . 0.0 .

2

6

6

67

7

1. Process Criteria2. Product Criteria

B. Possible Interim Waste Forms . .0.0 . ......

1. Fused Salt Entrapment . . . . . . . .2. Calcining Nitrate/Nitrite Suspensions3. Silicate-Agglomerated Calcine .4. Borate Glass . . . . . . . .5. Phosphate Immobilization . .6. Plans for Experimentation . . . . . .

C. Interim Waste Form Development: Materials Studies

1. Fused Salt/Sludge System.. . . . . .

2. Strength and Impact Fracture Behavior ofSintered Silicate and Pyrex . . . . . . . . . . .

II. NEUTRON ACTIVATION AND TRACER STUDIES.".. . . . . . . .

A. Introduction . . . . . . . . . . . . . . .

B. Radioactive Tracer Method Qualification.. . . . . .

1. Introduction2. Experimental Approach and Glass Preparation .3. Leaching Characterizations . .. 0.. ..0 .. .

C. Qualification of NAA Method . . . . . . . . . . . . . . . . .

1. Neutron Irradiation Effects . . . . . . . . . . . . . . .2. Leacn Experiments-Conditions, Results, and

Discussion

D. Weathering Experiments . . . . . . . . . . . . . . . . .

E. Service-Role Studies . . . . . . . . . . . . . . . . . . . .

III. BRITTLE FRACTURE STUDIES . . . . . . . . . . . . . . . .

A. Introduction . . . . . . . . . . . . . . . .. . . . . . . .

B. Status of Results Previously Reported on Measurementsof Fracture Particulates from Impacted BrittleMaterials. . . . . . . . . . . . . . . . . . . . . . . . . .

C. Correlation of Lognormal Particle Size Parameterswith Input Energy Density in Drop-Weight Experiments . . .

iii

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TABLE OF CONTENTS (contd)

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D. Effect of Grain Size on Size Distribution ofFracture Particulates . . . . . . . . . . . . . . . . . . . . 53

E. Microscopic Measurements of Particle Size toDetermine Shape Factors . . . . . . . . . . . . . . . . . . . 57

F. Calibration of BET Measurements and Sieves . . . . . . . . . 58G. Preliminary Consideration of Free-Fall Tests . . . . . . . . 58

IV. FLUIDS IN ROCK . . . . . . . . . . . . . . . . . . . . . . . . . 60

A. Development of Dynamic Stream Simulation Methodology . . . . 60

1. Penetration of Rock Salt by Water . . . . . . . . . . . . 60

2. Conclusions . . . . . . . . . . . . . . . . . . . . . . . 66

B. Determination of Residual Oil Saturation inDepleted Oil Reservoirs . . . . . . . . . . . . . . . . . . . 66

1. Geochemical Characterization Relevant toNuclear Logging . . . . . . . . . . . . . . . . . . . . . 67

2. A Proposed Method of Determining Residual Oil . . . . . . 67

V. TRACE ELEMENT TRANSPORT IN LITHIC MATERIAL BY FLUID FLOWAT HIGH TEMPERATURE . . . . . . . . . . . . . . . . . . . . . . . 70

A. Introduction . . . . . . . . . . . . . . . . . . . . . . . . 70

B. Ion Exchange on Kaolinite . . . . . . . . . . . . . . . . . . 70

1. Procedures . . . . . . . . . . . . . . . . . . . . . . . 712. Results . . . . . . . . . . . . . . . . . . . . . . . . . 723. Discussion . . . . . . . . . . . . . . . . . . . . . . . 74

C. Permeability of Big Horse Limestone . . . . . . . . . . . . . 75

1. Method . . . . . . . . . . . . . . . . . . . . . . . . . 752. Results . . . . . . . . . . . . . . . . . . . . . . . . . 76

D. Discussion . . . . . . . . . . . . . . . . . . . . . . . . . 77

E. Hydrologic Properties and Ground Water Composition

of Northern Illinois Granite . . . . . . . . . . . . . . . . 78

1. The Core Holder . . . . . . . . . . . . . . . . . . . . . 792. Design Problems . . . . . . . . . . . . . . . . . . . . . 7 93. Results . . . . . . . . . . . . . . . . . . . . . . . . . 79

VI. LIGHT WATER BREEDER REACTOR PROOF-OF-BREEDING ANALYTICALSUPPORT PROJECT . . . . . . . . . . . . . . . . . . . . . . . . . 80

A. Full-Scale Shear . . . . . . . . . . . . . . . . . . . . . . 80

B. Single-Unit Dissolver . . . . . . . . . . . . . . . . . . . . 81

C. Multiple Dissolver System . . . . . . . ... . . . . . . . . 82

iv

TABLE OF CONTENTS (contd)

Page

D. Scrap and Waste . . . . . . . . . . . . . . . . . . . . . . 83

1. Spray Calcination of Dissolver Solutions . . . . . . . . 832. Waste Shipment ......... . ......... 84

E. Computer System .......... . .......... 85

1. Hardware.4......... ... ....... . .. .852. Software . . . . . . . . . . . . . . . . . . . . . . . . 86

F. Blend Tanks . .......... ... ......... 86

1. Design ......... .... . ......... . 872. Blend Tank Decontamination .873. Sampling Error . . . . 87

G. Analytical . . . . . . . . . . . . . . . . . . . . . . . . . 87

REFERENCES.......................................... ........... 88

V

LIST OF FIGURES

No. Title Page

1. Strength of Fused-Salt Specimens as a Functionof their Wt % Fe203 . . . . . .. . . . . . . . . . . . . .- .- .- .- . 15

2. Strength of Sintered SS-C and SS-65 Compacts Preparedfrom As-Received Powders as a Function of Fe203 Content . . . . . 16

3. Impact-Induced Particle Size Distribution in Tests 68and 69 for Specimens Prepared from -200 +325 SizeFractions of SS-65 and Fe203 .-. . . . . . . . . . . . . . . . . .. . 20

4. Load-Time Trace for (a) Test 48 with Pyrex Specimen and(b) Test 71 with SS-65 Specimen Containing 33.3% Fe203.-...... 22

5. Flow Diagram of the General Approach Being Used forComparing SRL Frit 211 Waste Glass Leach Rate Results forNeutron Activation and for Radioactive Tracers Addition . . . . . 27

6. Schematic of 50 g Ingot of Radioactive Glass, Illustratingthe Code of the Wafers Cut from the Original Ingot . . . . . . . . 31

7. Normalized Element Losses from PNL 76-68 Glass Leachedat 90*C in Deionized Water . . . . . . . . . . . . . . . . . . . . 40

8. Normalized Silicon Loss from Activated and NonactivatedSamples of PNL 76-68 Glass at 90*C . . . . . . . . . . . . . . . . 41

9. Normalized Cesium Loss from Activated and NonactivatedSamples of PNL 76-68 Glass at 90*C . . . . . . . . . . . . . . . . 42

10. Pyrex Fraction Particulates as a Function of EnergyDensity in Drop-Weight Experiments . . . . . . . . . . . . . . . . 50

11. Particle Parameters as a Function of Energy Density . . . . . . . 51

12. Respirable Fraction V as a Function of Energy DensityV0

Density i for Pyrex Side Impacts. . ....... . . . . . . . . 52Vo

13. Particle Size Distribution of Sandstone Side-Impacted at1.2 J/cm3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53

14. Particle Size Distribution of Nepheline Syenite Rock,

Drop-Weight-Impacted at 1.2 J/cm3 . . . . . . . . . . . . . . . . 54

15. Particle Size Distribution of U02, Drop-Weight-Impactedat 1.2 J/cm3 . . . . . . . . . . . . . . . . . . . . . . . . . . . 55

vi

LIST OF FIGURES (contd)

No. Title

16. Particle Size Distribution of Fused Quartz,DroD-Weight-Impacted at 1.2 J/cm3 . . . . . . . . . .

17. Particle Size Distribution of Crystalline Quartz,Drop-Weight-Impacted at 1.20 J/cm3 . . . . . . . . . .

18. Schematic of the Experimental Setup to Study WaterPenetration into McNutt Salt Cores . . . . . . . . . .

19. Depiction of the Sleeve Confining Pressure duringthe Water-Penetration Experiment . . . . . . . . . . .

20. Specific Activity of the Segments of the Rock CoreRelative to their Depths in the Core . . . . . . . . .

21. Anticipated Ratios of Injected Solvent to ExtractedPetroleum as the Well is Pumped . . . . . . . . . . .

22. Effects of Prewashing and Equilibration Time onDistribution of Cs+ between Kaol:inite and O.1M NaHCO 3

Page

56

. . ...... . 56

. . .. ..... . 60

61

62

69

72

23. Isotherm for Sorption of Cs+ by Kaulinite in 0.1M NaHCO3 ,Expressed as Fractional Loading of Exchange Capacity . . . 73

7824. Flow of Water through a Limestone Core as a Function

of Pressure Difference, AP, and Temperature

vii

. .0.0.0

. .0 .a .0 .0

LIST OF TABLES

No. Title

1. Waste Solution Typical of West Valley Tank 8D2 . . . . .

2. Concentrations of Solid Constituents of SimulatedWest Valley Calcined Sludge . . . . . . . . . . . . . .

3. Experimental Conditions and Results of InstrumentedImpact Tests on Pyrex Specimens . . . . . . . . . . . .

4. Instrumented Impact Test Conditions and EstimatedFraction of Fines Generated during the Tests onSS-C and SS-65 Sodium Silicate Compacts ContainingVarious Amounts of Fe 2 03 . .-.- . .- . .. .- . . . . . . . . .

5. Pellet Size Distribution from Pelletization of Fe203in the Presence of Sodium Silicate Solution . . . . . .

6. Nominal Compositions of Simulated SRL Waste Glasses

Used in this Work . . . . . . . . . . . . . . . . . . .

7. Radioactive Isotopes Used and their Radiation Properties

8. "Mass" Balance of Radioactivity Added to SRL Frit 211**Glass and Amount of Radioactivity Measured afterMelting . . . . . . . . . . . . . . . . . . . . . . . .

9. Counting Statistics for Sections of Neutron-ActivatedPNL 76-68 Glass . . . . . . . . . . . . . . . . . . . .

10. Compositions of Leach-Tested Glasses . . . . . . . . . .

11. Leaching Test Conditions: MCC-1 Procedure . . . . . . .

12. MCC-1 Round Robin Leach Test Conditions and Results . .

13. Leachate Analysis for the MCC-1 Round Robin . . . . . .

14. Weights, Activities, and Lengths of Segments fromthe Four Rock Cores . . . . . . . . . . . . . . . . . .

15. Distribution Coefficient of Cs+ between Kaolinite andFresh O.1M NaHCO 3 . . . . . . . . . .*.*.*.*.*.*.*.*.*.

16. Permeability Measurements of Limestone Sample 7/12/80/12from the Notch Peak, Utah, Area . . . . . . . . . .

viii

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CHEMICAL ENGINEERING DIVISION

FUEL CYCLE PROGRAMSQUARTERLY PROGRESS REPORT

October-December 1980

by

M. J. Steindler, Seymour Vogler, G. F. Vandegrift, Jacqueline Williams,T. J. Gerding, L. J. Jardine, J. K. Bates, J. E. Kincinas, W. J. Mecham,

R. H. Pelto, M. G. Seitz, R. A. Couture, N. M. Levitz, T. F. Cannon,P. G. Deeken, R. E. Nelson, J. E. Parks, L. E. Trevorrow, C. G. Wach,


ABSTRACT

A program continues for the development of an interim wasteform that can be transported from facilities where waste is gener-ated to terminal waste processing. Waste forms being studied in-clude fused salt, simulated calcined sludge mixed with anhydroussilicates, and pellets made from Fe2O3 and aqueous silicate solu-tions.

Leach rate measurements were continued. A method of addingradioactive tracers to glass samples on a laboratory scale is de-scribed. Study of the leach resistance of several spiked, acti-vated, and unactivated waste forms is continuing. Studies havevalidated the neutron activation analytical method for measuringleach rates. Samples of Westinghouse alkoxide glass and PNL 76-68glass have also been leach-tested, and some results are presented.

In work characterizing the impact resistance of brittle wasteforms in terms of total fracture surface area and total amount ofparticles of respirable size, additional tests were made with Pyrexspecimens, and a possible alternative method of normalizing theweight fractions of the fracture particulate was used. This methodshows a correlation of surface area with energy input and that thegeometric mean of the lognormal particle size and the standard de-viation of lognormal distribution are both independent of energyinput. Preliminary results for Pyrex samples indicate that timeunder stress is critical for fracture in free-fall tests. Resultsare described that relate grain sizes and crystallinity of crystal-line and conglomerate materials to particle size distribution afterimpact.

As part of the characterization of fracture results, methodsare being developed of determining particulate shape factors con-sistent with sieve size analysis and with the parameters of thelognormal particle-size distribution.

In work to simulate a hydraulic breach of a nuclear waste re-pository located in bedded salt, experiments were performed to

1

2

study the flow of brine through cores of rock salt. In other work,logging techniques are being developed to measure the relativeamount of residual oil in a depleted oil reservoir.

Distribution coefficients of Cs+ between the clay mineral,kaolinite, and O.1M NaHCO3 solutions are being measured because ofthe interest in clay minerals as backfill in radioactive waste re-positories.

Testing and development of equipment for the destructive anal-ysis of full-length irradiated fuel rods from the LWBR is in pro-gress. This analysis will be in support of the LWBR Proof-of-Breeding Project.

SUMMARY

Development of Interim High Level Waste Forms

The goal of this program is to provide interim forms of high-level wastesfir facilities (usually with small waste volumes) having neither the inclina-cion nor facilities to fabricate borosilicate glass, the reference terminalwaste form. An interim waste form is a solid high-level waste which may beshipped to terminal processing facilities. The waste management program atthe Nuclear Fuel Services plant (NFS) may be able to utilize interim wasteforms such as those being evaluated for the Environmental Impact Statementfor removal of the NFS wastes.

Candidate materials for interim waste forms include entrapment of all ofthe material in fused salt, silicate-agglomerated calcine (i.e., calcine de-rived from the sludge), borate glasses, and immobilization in phosphate. Thefirst two are particularly applicable to the Nuclear Fuel Service wastes, andevaluation of the properties of these materials has been started. Experimentswill begin shortly to assess the suitability of phosphate as an interim wasteform.

A fused-salt mixture typical of the material found in NFS waste tankshas been prepared. Physical properties and the rate of dissolution were mea-sured.

Pellets of silicate-agglomerated calcine have been prepared, and both thefracture strength of the pellets and the particle size distribution upon impactof a specimen have been measured as a function of available impact energy.

Preliminary tests continue on pelletization of sludge, using aqueoussodium silicate solutions.

Neutron Activation and Tracer Studies

Experiments to qualify the spike/leach method of determining leach ratesare in progress. The spike/leach method is being qualified for SRL frit 211glass plus waste additives (SRL 211*) and PNL 76-68 glass. Nonactivated, ac-tivated, and spiked samples of SRL 211** glass have been obtained and sub-jected to leach tests of up to 28 days. Leaching of barium and cesium hasbeen observed from each form of SRL 211** glass.

3

Samples of PNL 76-68 glass (nonactivated and activated) have been leachedto determine the activity levels to be used in spiking experiments. Cesiumdominates the leachate from the activated samples, and methods of circumventingcesium interference are being investigated.

The neutron activation analysis (NAA) method of measuring leach rateshas been qualified. No effect of neutron irradiation of glasses on theleaching process has been observed, and the NAA method is very sensitive forseveral elements that are difficult to determine by other analytical techniques.

Service-role leaching of SYNROC samples is continuing, and qualificationof NAA for the analysis of crystalline waste forms (SYNROC) is in progress.Samples of Westinghouse alkoxide glass and PNL 76-68 glass (MCC-1 round robin)have also been leach-tested, and some results are presented.

Brittle Fracture Studies

Results of experimental tests in the current report period are summarizedand related to work previously reported. A new and arbitrary normalization ofparticle sizes of the fracture particulates of Pyrex test specimens was appliedto data measured in a series of drop-weight impact tests for a range of inputenergy densities from 0.21 to 2.4 J/cm3 . The fracture particulate was char-acterized by (1) the mass median Dg, (2) the standard deviation 0g, and (3)the ratio of the volume of material passing an 8-mm sieve V(8 mm) to the totalspecimen volume Vo. The values of Dg and a appeared to be nearly independentof energy density, and the ratio of V(8 mm)/Vo was found to be directly pro-portional to energy density. A preliminary correlation was also found in whichthe ratio of respirable fines V(10 um) to the original volume Vo was directlyproportional to energy density.

This arbitrary normalization method was also a plied to Pyrex fracturesize data obtained for conical and corner specimens and for side impacts ofcylinders in drop-weight and free-fall tests, all at an energy density 0.42J/cm3. For the side impacts, the values of Dg and ag were nearly identical,but the ratios V(8 mm)/Vo and V(10 um)/Vo were approximately a factor of twosmaller in the case of the free-fall test than in the drop-weight tests, in-dicating less fracture in free fall. A new type of Pyrex cylinder designedfor free-fall tests was made that has a hemispheric end for impact. Drop-weight tests of this specimen showed fracture particulate very similar to thatobtained from side impacts of cylinders. However, in a free-fall test with a18.2-m drop, very little fracture of the specimen occurred, and the specimenbounced about 4.6 m. Dynamic stress analysis indicated that the time understress was less for the free-fall test than for the drop-weight test. It isbelieved that time under stress i; critical for fracture and that more massivespecimens are required for free-fall tests khan for drop-weight tests. Newspecimens are being prepared for free-fall tests and for drop-weight tests.

New data on grain sizes of different materials impacted in the drop-weight device were used to evaluate the effects of crystallinity and grainsize on particle size distribution after impact. In the case of sandstoneand nephallne syenite, the grain sizes result in bimodal lognormal distribu-tions of fracture particulates. However, no effect of grain size was apparent

4

for microcrystalline U02 ceramic, vitreous SiO2, and nearly single-crystalSiO2. Computer-assisted linear regression is being developed to aid in theanalysis of particle-size distributions and in the correlations of parameters.

Because of the irregular shapes of fracture particulates, mean shapefactors--principally, the surface/volume shape factor--are required as partof the characterization of the fracture results. Statistical analysis ofmicroscopic measurements is being developed and calibration of particulatesurface-area measurements pursued to determine shape factors consistent withsieve size analysis and with the parameters of the lognormal particle-sizedistribution.

Approximate graphical lognormalities of fracture particulates have beenconsistently measured with the Coulter counter apparatus down to the lowerlimit of size detection (which is about 5 um). The general features of thebrittle fracture methodology appear promising.

A preliminary scale-model drop-weight impact test of a small canisteredPyrex specimen gave a particle size distribution (down to 5 ym) which closelymatched that of a bare specimen.

Fluids in Rock

Two studies of fluid behavior in rock were conducted this quarter. Thefirst study is a continuation of experimental work to determine the penetra-tion of water into cores of rock salt. This work is part of the developmentof groundwater-stream experiments that simulate a hydraulic breach of a nu-clear waste repository located in bedded salt. The second study is the de-velopment of logging techniques to determine the residual oil in a depletedoil i servoir,

Trace Element Transport in Lithic Material by Fluid Flow at HighTemperature

This report section describes the final work on Cs+-Na+ exchange andinitial work on pore water composition and hydrologic properties of rocks.

There is interest in the use of clay minerals as backfill in radioactivewaste repositories. Work presented this quarter (and previously) shows tiatthe distribution coefficent of Cs+ between the clay mineral kaolinite andO.1M NaHCO3 solutions depends on the cesium concentration, time, temperature,and the solid/solution ratio, but not on pH. Distribution coefficients ob-tained by the column elution technique differ from those obtained by batchexperiments. The dependence of distribution ratio on cesium concentration ismuch greater than would be expected if there is simple saturation of exchangesites. Apparently, there is more than one type of exchange site. The depen-dence on solid/solution ratio and the difference in results of column andbatch experiments are interpreted to mear. that the distribution coefficientdepends on the state of aggregation of the clay.

Work was continued on the hydrologic and mass transport properties ofrocks. First, the permeability of a sample of limestone from the Notch Peak,

5

Utah, area was measured. Great care was taken to ensure that the measurementswere accurate and that no leakage occurred between the outside of the coreholder and the rock.

second, a core holder was designed and tested which has no dead volumeto act as mixing chambers in the fluid stream. With this apparatus, elutionexperiments can be performed on rocks having small effective pore volume,without excessive dispersion due to the equipment. With this core holder,diffusion of salutes into the rock matrix can be studied. In favorable cases,pore water can be eluted from rocks and its composition determined. In thefirst experiments, halide-rich brine is being eluted from Precambrian graniteobtained from Northern Illinois drill hole UPH-3.

Light Water Breeder Reactor Proof-to-Breeding Analytical Support

Project

This project includes responsibility for the destructive analysis offull-length irradiated fuel rods from the LWBR. The results will be used bythe Bettis Atomic Power Laboratory (BAPL) in support of their nondestructiveassay of the end-of-life (EOL) core to determine the extent of breeding.

Activity is reported on four main subactivicies of this project: (1) thefull-scale shear (FSS), (2) the single-unit or prototype dissolver (SUD),(3) multiple dissolvers needed for the destructive analyses of full-length('3-m) irradiated fuel rods for the EOL campaign, and (4) scrap/waste dis-

posal. In addition, the Analytical Section is engaging in planning and pro-curement activities in support of this project. An integrated cor luter/dataacquisition System for the entire project is also being developed.

6

I. DEVELOPMENT OF INTERIM HIGH LEVEL WASTE FORMS(Seymour Vogler, G. F. Vandegrift, Jacqueline Williams,G. Bandyopadhyay,* S. M. Gehl,* T. M. Galvin,*J. E.Slattery,* W. J. Grajek* andR. B. Poeppel )

A. Introduction

High-level liquid radioactive wastes are often produced at facilities inwhich terminal processing of the waste is not practical because of the highcost >r hazards involved in such an operation. Under these circumstances,the waste can be incorporated into an interim solid waste form for shipmentto a terminal processing site. The interim form should require only simpleprocessing methods, so that extensive additional facilities will not be re-quired at the originating site. Additionally, the interim waste form (1)must satisfy shipping criteria, particularly in relation to transportationaccidents; (2) should be mechanically and chemically stable in the ambientenvironment; and (3) should be readily compatible with the terminal-waste-form processes. From this research, there should evolve a technology forpreparing interim waste forms compatible with transportation requirements andwith subsequent terminal waste form processing.

In work started orior to the beginning of the fiscal year (with fundingprovided by DOE), assistance was given in the preparation of an EnvironmentalImpact Statement for the removal of high-level liquid waste (HLLW) from '.heNuclear Fuel Services (NFS) reprocessing plant at West Valley, New York. Thewaste from the West Valley Nuclear Fuel Services plant is described below inSection B.l.a.

Currently, assistance is being given in the management of the wastes atthe West Valley plant; also, the literature is being reviewed to identifypossible interim waste forms. These possible interim waste forms were judgedon product and process criteria previously reported [VOGLER].

1. Process Criteria

a. Simplicity--the process should fit into the West Valley facil-ity, and should be simple to install and operate.

b. Safety--the process should be safe to install and operate.

c. Minimal R&D Needs--the process should be reasonably mature sothat little R&D is required.

d. Economic--the process should be economical in comparison withthe cost of final form manufacture.

e. The process should not extensively increase the voJume of wasteto be removed from the site.

*Member of the Materials Science Division, Argonne National Laboratory.

7

2. Product Criteria

a. The properties of the interim form should be compatible withthe process for producing the final waste form product--borosilicate glass.

b. The dispersibility of the waste form should be low, particu-larly in the event of a transportation accident.

c. The interim waste must show reasonable chemical and physicalstability because the waste may be stored for long periods before it is pro-cessed for final disposal. For example, the interim waste form should bestable in an ambient air environment, in the event some of the canistersshould leak. Stability in an ambient air atmosphere would also minimize theconsequences of canister upture during handling and shipping of the wastecanister. The interim waste form should be stable in the radiation fieldencountered during storage of canisters--at least to the extent that pressurebuildup is not sufficient to affect canister integrity. Any mechanical shockduring hands ng and shipping should not reduce the interim waste form to anexcessive number of particles in the respirable size range (<10 um).

d. The product should have thermal properties that allow heatgenerated by the decay of fission products to be adequately dissipated throughthe surface of the canister,

e. The waste fork should be capable of being packaged to meettransportation requirements for the waste-canister-cask system.

Because of the strong ties between this program and that for theWest Valley project, we are preparing a simulated NFS waste suspension as thefirst part of our laboratory studies of possible interim waste forms.

B. Possible Interim Waste Forms(Seymour Vogler, G. F. Vandegrift, and Jacqueline Williams)

Before individual possible interim waste forms are discussed, the mostlikely terminal waste form for high-level radioactive wastes, borosilicateglass is considered. There are definite limits to the quantities of certainconstitutents of the final product glass. If these limits [ERDA] are exceeded,there are immiscibility problems, excessive leach rates, and poor durability.The limits are as follows:

S102 25-40 wt %

B203 10-15 wt %

Alkali metal oxides 5-10 wt %

ZnO 0-20 wt %

Waste oxides 20-35 wt %

If NFS waste* (which contains about 80 wt % Na20) and in most other neu-tralized HLLW, alkali metal ion content is the main consideration in solubi-lizing the waste in borosilicate glass. For NFS waste, the dilution factor

*See the following section (1.a) for a description of this waste.

8

would have to be greater than 8/1 for borosilicate glass/waste to keep alkalimetal oxide concentration below 10%. If the NFS supernate (see Section l.abelow) were disposed of separately, the dilution factor would depend on thewaste oxide composition and would drop to the range, 5/1 to 3/1. For an in-terim waste form to be acc rtable, it should not substantially increase theseratios beyond 8/1 or 5/1.

In addition to the candidate interim waste forms discussed previously(fused salt entrapment, calcine, and silicate-agglomerated calcine) [STEINDLER],we have examined the use of borate glasses and phosphate glasses and bindersfor interim immobilization of HLLW. The following sections are summaries ofthe pertinent information acquired for each candidate.

1. Fused Salt Entrapment

a. Introduction

The waste from the West Valley Reprocessing Plant consists ofmaterial from two tanks (8D2 and 8D4). The contents of tank 8D2 (2.2 x 106 L)is the neutralized waste from the first cycle Purex extraction and consistsof (1) a supernatant liquid that is basically NaNO3/NO2 and (2) sludge(28,000 L) that is primarily Fe(OH)3 and FePO4. Tank 8D4 (45,000 L) containsTh(N03)4 and Al(N03)3 in nitric acid.

Entrapment in salt involves combining the sludge, supernate,and the contents of tank 8D4 and forming a uniform slurry. This slurry isthere fed to evaporators where the water is removed until a molten salt solu-tion and a suspension remain. This molten salt, containing all of the fissionproduct activity, is then used to fill the storage canisters. On cooling, amonolithic salt block forms.

b. Product Homogeneity

A serious problem may develop in using fused salt entrapmentif the canister becomes hotter than the melting point of the fused salt block.The phase diagram for the NaNO3/NaNO 2 system [BERGMAN] shows that at the pro-bable NaNO3/NaNO 2 ratio (between 2/1 and 1/1) of NFS waste, the fused saltwould be completely molten at 230-260 0 C.* If the fused salt should becomemolten, those fission product hydroxides/oxides not solubilized in the meltand having a density greater than that of the salt would drop to the bottomof the canister and form a hot spot. Since the density of the moltenNaN'03/NaNO2 in the canister would be between 1.9 and 1.7 g/ci 3 and the densityof strontium salts is in the range of 3.0-3.7, this situation is of real con-cern. This same problem could arise during the original casting of the salt,so that positive steps would have to be taken to ensure that there is a homo-geneous slurry at the time of solidification.

*A salt mixture consisting of 91% sodium nitrate/nitrite (nitrate:

nitrite = 7:3), 7.4% sodium sulfate, and 1.2% sodium hydroxide was found tomelt at 220*C.

9

c. Radiolysis of NaNO3/NaNO2

Sodium nitrate is sensitive to ionizing radiation, decomposingto sodium nitrite and oxygen; there is extensive literature on this point[JOHNSON-1970]. The yield curve for the decomposition of sodium nitratecontinually decreases with radiation dose. For example, after a dose of5 x 1021 eV/g, the calculated G value for oxygen formation (based on nitriteanalysis) is 0.067. With a dose of 50 x 1021 eV/g, the G value drops to0.041 [JOHNSON-1961]. Based on the estimated activity in the storage canis-ters, it is calculated that approximately 2 L/d of oxygen would be formed ineach canister. This could lead to pressurization problems during long periodsof storage. However, it has been speculated that the decrease in G value asa function of dose may be due to back-reaction of nitrite with oxygen. Withthe large amounts of nitrite present, the G value may be even lower than thosereported above. To resolve such speculation, it will be necessary to experi-mentally determine oxygen evolution rates for typical salts of interest.

2. Calcining Nitrate/Nitrite Suspensions

Simple calcining, in either fluidized beds or spray calciners, ofnitrate/nitrite salts that contain large amounts of sodium ion is difficultif not impossible. Sodium nitrate can exist in a molten-undecomposed stateover a wide temperature range of 300 to 850 C. In spray calcining, the wallsof the reactor become coated with a gummy mess of calcine and undecomposednitrate [PNL]; fluidized-bed reactors are plagued by agglomeration problems[NEWBY].

An interesting method of calcining nitrate-containing waste showspossibilities of solving the problems mentioned above [AARON]. The process,which was incorporated in the study of ceramic and cermet waste forms, isbased on calcining molten urea solutions and suspensions of nitrate-containingwaste. The HLLW, presumably water-free, is mixed with molten urea at 180 Cin ratios of urea/nitrate as low as 1/1. The urea solution/nitrate mixtureis then fed into a spray calciner by an ultrasonic spray nozzle. The exo-thermic reaction between urea and nitrate maintains the centerline temperatureequal to or slightly lower than the 800*C wall temperature, preventing buildupof material on the calciner walls.

The "Study of Ceramic and Cermet Waste Forms" program was terminatedat the end of FY 1980. However, the experimental data for the calcining partof this program look so encouraging that this work should be kept in mind ifcalcining becomes a necessary step for an interim waste form. We thereforeplan to learn more about its applicability to the concerns of our program.

3. Silicate-Agglomerated Calcine

Soluble silicates agglomerated with calcined waste is thought to bea viable interim waste form [VOGLER]. Agglomeration of the calcine wouldlower its potential dispersibility and leachability and would increase itsthermal conductivity. A literature search was initiated to learn of potentialoptions and problems for this process and product. The following is a summaryof pertinent information unearthed during this search.

10

a. The presence of large amounts of sodium salts in calcine pre-vents successful agglomeration of the calcined waste. If the concentrationof sodium ion is too high, sodium silicate (with no binding characteristicswhatsoever) precipitates instead of soluble silicates forming a cohesive gelto bind the calcine. This means, in the case of NFS waste, that supernatantsolutions must be separated from sludge before processing. rihe sludge mustthen be calcined and agglomerated with soluble silicates alone, and the su-pernate treated separately. It also means that the quantity of sodium ionpresent in the calcined sludge Is a very important parameter and must bestudied with simulated sludge.

b. The weight ratio of Si0 2 /Na 2 0 must be in the range of 2.5 to3.8 if soluble silicates are to be used as binders. In this range, there isa trade-off between two important characteristics for an interim waste form--binding ability and water solubility. At low ratios, water solubility is atits peak, as is binding ability; as the r:atio is raised, both parameters de-crease, the first for better, the second for worse. Clearly, the effect ofvarying this ratio must be studied and a compromise reached.

c. Sodium silicates are by far the most studied and utilized formof soluble silicates and are therefore the natural choice for use as an in-terim waste form. Because of special characteristics of quaternary ammoniumsilicates, they might be considered an alterr.tive to sodium silicates. Thereasons for this are:

(1) Quaternary ammonium silicate solutions maintain a lowviscosity over a much wider s'.lica range than do sodium silicates. The vis-cosity remains relatively low up to 50% silica.

(2) The quaternary ammonium ion is destroyed and volatilizedat 250-300*C. This may enhance bonding between the metal oxides in the cal-cine and silica, lowering the leachabili':y and the dissolution rates whileincreasing the physical durability of the agglomerate. Adding quaternaryammonium silicates does not add more alkali metal to the waste and, therefore,does not increase the dilution factor for borosilicate glass formation.

4. Borate Glass

Borate glasses may be undesirable an an interim waste form becausethey are notorious for forming two immiscible glass phases when metal ions arepresent in the melts. Following is a listing [RAWSON] of concentration rangesof metal ions that are miscible in IxO - B 2 03 glasses:

Range of Miscibility,Metal mol % Mx

Na 20 0 - 3866.5 - 71.5

BaO 17.0 - 39.8SrO 24.2 - 43.0La203 19.0 - 28.2CdO 39.1 - 55PbO 20.0 - 76.5B1 2 03 22.0 - 65.3

11

When there is more than one metal oxide in a borate glass, the concentrationrange of miscibility will be a complex function of the metal ions; the rangesof miscibility ar3 generally smaller. For an oxide mix such as calcined NFSwaste, miscibility of all metals at any B 2 03 to calcine ratio would be nearlyimpossible. Two-phase systems of borate glasses show much higher leachabil-ities and much lower chemical and physical durabilities than do one-phasesystems.

Even if these problems were solved or considered unimportant, thepresence of large amounts of B2 03 may require excessive dilution of thisinterim form to yield a borosilicate glass with a B2 03 concentration in therange, 10-15%. At the very least, careful monitoring of the B2 03 content ofthe waste would be needed to ensure a high-grade final form product.

5. Phosphate Immobilization

Phosphate, as a binder or as a glass, appears to be a strong an-tender as an interim waste form. The phosphate ion forms strong bonds withmost metal ions, and fir the most part, its salt have low aqueous solubilities.The strong bonds former between phosphate and metal ions lead to high loadingcapacities of metal o- des in phosphate glasses [RAWSONI:

Range of Miscibility,MXO mol % MxO

Na 2 0 0-60

BaO 0-58

SrO 0-56ZnO 0-68

CdO 0-62

Ag20 0-66T120 0-50

PbO 0-64

BeO 0-66

Phosphate and acid phosphate cements, used as dental cements, have low leach-ability and high compressive strength.

Compatiblity of phosphate glasses with final-form borosilicate glassseems to be no problem. The following two examples verify this statement.

a. The addition of P2 05 to lower the leachability of borosilicateglasses is often mentioned. When 6% P2 05 was substituted for S102 , the leachrate of high-alumina nuclear waste decreased by 1500% [BROTZMAN]. Substitu-tion of P2 05 for B2 03 also substantially decreased the leach rate of borosil-icate glasses [BROTZMANJ.

b. A borosilicate glass was made from BiPO4 process sludge; al-though the glass was opaque, it had a 137Cs leach rate of only5 x 10-7 g/cm2 -day. The glass contained 40% sludge, 53% Hanford frit,

12

2% Li20, and 5% Zr02; 25% of the sludge was P0 43- ion [CORNMAN]. Other reports

show that borosilicnte glasses containing 50% BiPO4 process sludge have a highleach resistance [KUPLER].

hese results indicate that a borosilicate glass which contains10-15% P205 should have as good or better characteristics than one with nophosphate present. An interim waste form containing 50% P205 would have to bediluted only three to five times to yield the final waste form.

Another attraction of phosphate glasses or cements is the simplicityof processing HLLW into these forms. The simplest method for small-scale userswould be by adding H3P04 or P205 to aqueous suspensions or solutions of HLLW.Formaldehyde can be added to destroy and remove nitrates. Water can be re-moved by evaporation. Phosphoric acid or P205 does not distill below a tem-perature of 8690C at 753 mm Hg, at which point a eutectic is formed having aconcentration of 92.1% P205, a concentration which is far above that neededto form phosphate cements. To make a phosphate-bound product, the suspensionat about 3000C could be dropped onto a cold plate to make small pellets--a pro-cedure that would have to be experimentally verified.

Because of the fairly high solubility of sodium phosphate, any largeamounts of sodium ion present may cause problems in leachability and durabilityin the presence of water. If experiments verify this, the supernate of NFSwaste would need to be treated separately, as it would be with silicate ag-glomeration (see Section B.3 above). An alternative to separate treatment orsludge and supernate would be to operate the process at a much higher temper-ature and thus produce phosphate glass. To make phosphate glass, a tempera-ture of 850-1000 *C is needed. Because the melt and its vapor are corrosiveat these temperatures, glass melting would have to be done in refractory-linedmelters. The glass could then be pelletized as was the cement or could bemade into monolithic blocks. Phosphate glass technology is well developed,and a suitable product for short-term storage is nearly assured.

For large-scale operation, it may be necessary to change the pro-cess for preparing a phosphate cement or glass interim waste form by calciningbefore phosphate addition. This process would be identical to the one sug-gested for silicate-agglomerated calcine except that phosphoric acid wouldbe used instead of a sodium silicate solution.

6. Plans for Experimentation

The composition we have chosen to simulate NFS waste is presentedin Table 1. We are prepared to add other constituents (e.g., Al and Eu salts)at a later date if their chemistry becomes suspect in a process being devel-oped. These suspensions will be prepared fresh before every experiment.

To prepare the final product, consisting of both supernate and aprecipitated oxide, an acidic solution containing the requisite metal cationswill be the starting material. To this acid solution will be added an amountof phosphoric acid to precipitate FePO4; an amount of concentrated NaOH solu-tion necessary to neutralize the nitric acid, to precipitate metal oxides/hydroxides, and to provide the requisite sodium hydroxide concentration; andfinally an amount of sodium nitrate and sodium nitrite salts necessary toprovide the concentrations of each compound shown in Table 1.

13

Table 1. Waste Solution Typical of West Valley Tank 8D2

Concentration in Concentration inCompound Suspension, g/L Suspension, M

NaNO 3

NaNO2

Na2 SO4

NaOH

KC1

Fe2 03 a

FePO4

Cr2 03 a

Ni~a

Nd20 3a

Mn02

Na 2 U20 7

SrOa

Ru02a

363

156

43.2

7.3

0.30

21.6

16.6

1.68

0.80

0.81

1.28

3.8

0.22

0.22

4.27

2.26

0.30

0.18

0.0039

0.135

0.111

0.011

0.011

0.0024

0.015

0.0059

0.0021

0.0016

Zr02a 0.65 0.0053

aAlthough the oxides are used for calculation purposes,the precipitates may be hydroxides or mixtures of oxidesand hydroxides.

Twenty liters of the acid solution will be prepared at once andkept as a stock. Simulated waste will be prepared from this solution asneeded, in volumes of 50-200 mL per experiment .

The immediate purpose of preparing this simulated NFS waste suspen-sion is to study the feasibility of using phosphoric acid as a binder orglass-former in preparing an interim waste form. Initial experiments will berun in glass vessels at temperatures below 400*C. In these experiments, theeffects of phosphoric acid and temperature on the distillate removed from thevessel (its volume and pH, HNO3 , HN02, and NOx concentrations) will be mea-sured, along with the fate of the phosphate material left in the vessel (itschemical composition and physical characteristics). The use of reagents suchas formic acid to destroy nitrate will also be tested.

14

Depending on the physical characteristics of this phosphate mate-rial, it will be either pressed or cut into pellets. These pellets will thenbe tested for chemical and physical durability.

C. Interim Waste Form Development: Materials Studies

(G. Bandyopadhyay,* S. M. Gehl,* T. M. Galvin,* J. E. Slattery,*W. J. Grajek,* and R. B. Poeppel*)

The interim forms that we are currently investigating are of general in-terest and may be applicable to any high-level waste. However, the emphasisat present is on determining their suitability for application to the NuclearFuel Services (NFS) high-level radioactive liquid wastes stored at the NuclearFuel Services Center at West Valley, NY. As stated above, most of the liquidstored at this site is a neutralized waste comprising (1) an insoluble sludgeconsisting of metal oxides and hydroxides and (2) a supernatant liquid con-taining primarily NaNO3 and NaNO2 .

The processes under study are (1) preparation of a fused salt/sludgemixture and (2) treatment of calcined sludge with anhydrous or aqueous sodiumsilicate binder. In most of the initial experiments [STEINDLER], Fe203 wasused as the simulated West Valley sludge. Additional experiments are cur-rently being performed using a more prototypic sludge composition (Table 2).

Table 2. Concentrations of SolidConstituents of SimulatedWest Valley CalcinedSludge

Constituent Weight %

Fe203 45.73

FeP04 35.23

Cr203 3.56

NiO 1.69

Nd203 1.73

Mn02 2.72

U0 2a 7.05

SrO 0.46

Ru02 0.46

Zr02 1.37

amn most scoping experiments, CeO2will be used in place of U02.

*MaterialsScience Division, Argonne National Laboratory.

15

The work to be discussed includes: (1) fabrication of fused salt samples con-taining Fe203 and simulated sludge and determination of the rate of solubilityand fracture strength of these samples, (2) fracture strength measurements ofsintered SS-C and SS-65 compacts (SiO2:Na2O ratios of 2.00 and 3.22, respec-tively) containing 0 to 90 wt % Fe203, (3) completion and preliminary analysisof a series of instrumented impact tests of Pyrex specimens and on sinteredSS-C and SS-65 compacts containing 0 to 33.3 wt % Fe203 , and (4) completion ofscoping experiments on pelletization of iron oxide power by aqueous sodiumsilicate solutions.

1. Fused Salt/Sludge System

In a prior quarterly report [STEINDLER], it was reported that a fusedsalt mixture simulating West Valley supernatant and containing 1.3% KOH,7.4% Na2 SOj,. and the remainder NaNO3/NO 2 (with a N03/N02 re tio of 7:3 and aNa/K ratio of 99:1) melted at 220*C and started to decompose at about 600*C.The salt cakes were deliquescent, dissolved rapidly in water, and had a frac-ture strength of 4400 1000 kPa. Additional fused-salt specimens were pre-

pared containing various weight fractions of Fe203. The specimens were frac-tured by the diametral compression technique to determine their strength.Figure 1 data indicate that Fe203 in the salt mixture reduced the fracturestrength of the samples to some extent.

Several additional fused salt specimens containing 5, 10, and 20 wt %sludge (Table 2) were prepared by the procedure described earlier [STEINDLER].Fused-salt specimens containing >20 wt % sludge could not be fabricated becauseof the high viscosity of such fused-salt/sludge mixtures.

' 0.6

0.4 -

S0.2z

w

010 20 30

Fe 2 03 , wt %

Fig. 1. Strength of Fused-Salt Specimens asa Function of their Wt % Fe203

16

The fused salt/sludge samples were tested for fracture strength andsolubility in static water at room temperature. Experiments are also contin-uing to determine their hygroscopic character by exposing the samples in acontrolled-humidity environment. The results of these tests will be reportedlater.

2. Strength and Impact Fracture Behavior of SinteredSilicate and Pyrex

Commercially available anhydrous, amorphous sodium silicate powders(SS-65 and SS-C, with Si02:Na2 0 ratios of 3.22 and 2.00, respectively), PQCorp., Valley Forge, PA., can easily be sintered to 90% theoretical densityat 600*C [STEINDLER]. Literature data [PQ CORP] show that 50% of -65 meshSS-65 and SS-C powders dissolved ir. static water at 25*C in 60 and 10 h,respectively, indicating that the solution rate of sodium silicates is (a)substantially lower than that of the fused salt and (b) strongly dependenton S102 : Na2 0 ratio.

Several experiments have been completed to determine the fracturestrength of SS-65 and SS-C compacts containing up to 90 wt % Fe 2 03 . Figure2 shows fracture strength as a faction of Fe2 03 concentration. The strengthof the compact was strongly dependent on the sodium silicate composition, es-pecially at the lower iron oxide contents. Furthermore, the strength of the

6 x 10 4 -

SiO2 : Na20

" 2.00: 1.00 (SS-C)

O0 3.22: 1.00 SS-65)4 x 104--

zw

H- 2x104 -

00 20 40 60 80

Fe2 03 , wt%

Fig. 2. Strength of Sintered SS-C and SS-65 Compacts Prepared fromAs-Received Powders as a Function of Fe203 Content. Thestrength was determined by the diametral compressiontechnique.

17

specimens containing only 10 wt % sodium silicate was similar to or higherthan that of the fused-salt specimens [STEINDLER].

Several instrumented impact experiments were performed on SS-65 andSS-C compacts containing 0 to 33.3 wt % Fe20 3 . To allow the impact fracture

behaviors of these compacts to be compared with that of idealized brittlematerials, additional tests were performed on cylindrical Pyrex specimens.In the following, we present the results of impact tests on Pyrex and thendiscuss test results for silicate specimens.

a. Impact Tests on Pyrex Specimens

The impact tests on the Pyrex specimens were performed to studythe effects on impact fracture behavior of several experimental parameterssuch as velocity of the impacting tup, the travel distance of the tup afterthe first impact, and the sample size. The tests and the relevant experi-mental parameters are listed in Table 3. The maximum available energy, Emax,was calculated from the relation

1 2E = -mymax 2

(1)

Table 3. Experimental Conditions and Resultsa of Instrumented ImpactTests on Pyrex Specimens

Tup Travel after Calculated

Maximum First Impact EnergySample Impact Available % of Absorbed

Test Dia, Height, Velocity, Energy, Sample by Sample,bNo. cm cm cm/s J cm Height J

59 2.47 1.27

60 2.47 1.27

61 2.47 1.27

62 2.47 1.27

63 2.47 1.27

64 2.47 1.27

65 2.47 1.27

66 2.54 2.57

67 2.54 2.57

120

120

120

150

150

200

200

120

120

58.4

58.4

58.4

91.2

91.2

162.2

162.2

58.4

58.4

0.95 74.8

0.63 49.6

0.32 25.2

0.32 25.2

0.63 49.6

0.63 49.6

0.64 50.4

1.93 75.1

1.93 75.1

aThe load-time trace and high-speed movie forfracture occurred as a single event.

bUncertainty, *10%.

every specimen indicae that

11.9

11.9

6.1

9.6

9.6

6.3

8.2

14.8

14.8

18

where m is the mass of the crosshead block including the impacting tup(81.1 kg) and v is the impact velocity. The equipment is designed so that vcan be set by the operator at any desired value. Prior to each experiment,the distance of tup travel after the first impact was also set. The energy,AE, absorbed by each specimen was calculated from the relation [IRELAND]:

EaAE = Ea (i (4Ea(2)

max

where Ea = vfP-dt, i.e., the area under the load-time curve, and

P = load force

Note that in tests 59 through 65 (Table 3), the specimen sizewas the same; however, impact velocity (and thus the maximum available energy)was varied from 120 to 200 cm/s and tup travel after initial impact variedfrom 0.32 to 0.95 cm. Despite these variations, the energies absorbed by thesamples during the impact tests were similar (~6-12 J). The load-time tracedata and the high-speed movies taken during these tests clearly indicate thatfor every specimen, impact fracture occurred in one step, i.e., the specimenexploded into many pieces as soon as the first impact occurred. The times toreach the maximum load in these impact tests were ~160-280 ps.

Tests 66 and 67 (Table 3) were performed with specimens largerthan those used in tests 59 through 65. The energies absorbed by the largersamples and the times to reach maximum load were somewhat greater than forthe earlier tests.

The relative insensitivity of the fracture behavior of Pyrexspecimens to experimental parameters can be related to the idealized brittlenature of the material. This idealization is supported by the observationthat fracture occurs instantaneously after the initial impact. The surfaceareas and the particle size distributions of the impacted Pyrex specimens arebeing measured to correlate them with fracture parameters.

b. Impact Tests on Silicate Specimens

Impact tests have been performed on SS-65 and SS-C compactscontaining various amounts of Fe203; the pertinent data are summarized inTable 4. Velocity of the impacting tup in all of these experiments was 120cm/s. The corresponding Emax was 58.4 J. Tup travel distance after firstimpact was set and the energy absorbed by the sample was calculated similarlyto the method described above for tests with Pyrex specimens.

The fraction of fines generated, shown in Table 4, was estimatedby the following procedure: the impact-induced particle-size distributionwas determined by sieve analysis. Since particle sizes smaller than 44 pm(325 mesh) could not be determined by this process, the data were extrapolatedto 10-pm size on a log-normal graph paper in which the weight percent smallerthan diameter, D, was plotted versus particle diameter, D. Such an extrapola-tion was possible because very often a straight-line relationship between

19

Table 4. Instrumented Impact Test Conditions and Estimated Fractionof Fines Generated during the Tests on SS-C and SS-65Sodium Silicate Compacts Containing Various Amounts ofFe2 03 . Velocity of impacting tup, 120 cm/s.

Tup Travel after CalculatedFirst Impact Energya Estimated

Sample % of Absorbed FractionTest Dia., Height, Sample by Sample, of FinesNo. Composition c' cm cm Height J x,10 um)

SS-Cb

SS-Cb

SS-65C

SS-65C

2.88

2.87

2.87

2.85

As-received SS-C:Fe203

88.9:11.1

85.7:14.3

66.7:33.3

(-200 +325) SS-C:Fe 2 03

88.9:11.1

85.7:14.3

85.7:14.3

66.7:33.3

66.7:33.3

As-received SS-65:Fe203

88.x.'L..1

85.7:14.3

66.7:33.3

(-200 +325) SS-65:Fe 203

88.9:11.1

85.7:14.3

66.7:33.3

2.93

2.95

3.23

2.87

2.95

2.97

3.21

3.21

2.95

2.94

3.22

2.86

3.03

3.22

1.55

1.57

1.94

1.94

1.51

1.46

1.47

1.58

1.49

1.46

1.43

1.44

1.47

1.65

1.43

1.51

1.712

1.48

1.23

1.25

1.62

1.31

0.88

0.83

0.84

0.95

0.86

0.83

0.80

0.81

0.83

1.02

0.80

0.88

1.08

0.85

79.4

79.6

83.5

67.5

58.3

56.9

54.0

60.1

57.7

56.9

55.9

56.3

56.5

61.8

55.9

58.3

63.2

57.4

53.3

57.9

25.3

9.7

12.3

11.2

15.6

4

2

1

51.4

56.0

57.9

57.4

52.3

17.0

9.1

4.0

36.7

7.6

12.6

10-8

10-8

10-5X10-10

x 10-3

x 10-4

x 10-4

5 x 10~7

d

6 x 10-3

4 x 10-2

4 x 10-2

2 x 10-2d

d

10-6

d

d

aUncertainty, 10%.

bS102 :Na20 ratio - 2.00.

cS102 :Na20 ratio - 3.22.

dEstimation of fines for these tests was not possible because of a large deviation from alog-normal relationship of the particle size distribution parameters.

40

41

42

43

47

72

73

44

52

53

54

55

46

70

71

45

68

69

20

these two parameters was obtained. However, as may be seen in Table 4, alarge deviation from the straight-line relationship was observed for severalof the compositions containing Fe2 03 , ( .-, see Fig. 3). Measurement ofparticle sizes as small as 10 pm by alternative direct methods is being at-tempted.

0

50

10'

0.I

0.01

SO TEST No. 68

. 0 TEST No. 69

- iI 11

10 100 1000

PARTICLE DIAMETER, D, p.m

Fig. 3. Impact-Induced Particle Size Distribution in Tests 68 and 69 forSpecimens Prepared from -200 +325 Size Fractions of SS-65 andFe203 .

The impact fracture behavior of SS-C and SS-65 compacts intests 40 through 43 (Table 4) was discussed in an earlier report [STEINDLER].The load-time trace and the high-speed movie taken during each of these tests(40 through 43) and the calculated energy absorbed by each sample as shown inTable 4 indicate several features:

(1) In all of these tesLs, the first impact was followed bycompaction of the specimens.

(2) The energy absorbed during the impact of SS-C specimens(test 40 and 41) was substantially higher than was ab-sorbed in SS-65 compacts (tests 42 and 43).

(3) The travel distance of the tup after the first impact cansubstantially influence the compaction and the fines gen-eration during the tests (e_.., see the data on tests 42and 43 in Table 4).

The remainder of the tests listed in Table 4 were performedwith widely varying compositions of the compacts. The tup travel distance

W-N

i; W

ow

LOW

L::zU

x z~W

Q .:

21

during these tests was kept relatively uniform. Some important trends ob-served in these data are briefly described in the following:

(1) The calculated energies absorbed during impact of thecompacts prepared from -200 +325 size fractions of SS-Cand Fe203 were substantially higher than absorbed by com-pacts prepared from as-received SS--C and Fe2 03 powder,from as-received SS-65 and Fe203 powder, or from -200 +325size SS-65 and Fe203. The reason for such differencesis not understood at this time. However, it is evidentthat the particle size of the starting powder and/or thecomposition of sodium silicate can significantly influencefracture behavior.

(2) We noted earlier [STEINDLER] that in several tests, theparticle size distribution data did not follow the usualstraight-line relationship in a log-normal plot, indi-cating that multiple-phase systems may respond differentlyfrom simple systems such as Pyrex, SS-C, or SS-65 com-pacts. The type of bonding of the phases could play animportant role in fines generation.

(3) In some experiments (tests 68, 70, 71, and 72) with SS-Cand SS-65 compacts containing Fe20 3 , the load-time traceindicated a one-step fracture. However, this behaviorwas substantially different from that observed for Pyrex,which also exhibited a single impact event. In the caseof Pyrex, the fracture was characterized by a sharp peakfollowed by instantaneous relaxation of load; the silicatespecimens, on the other hand, exhibited a broad peak witha gradual relaxation of load from its peak value (Fig. 4).

Additional impact experiments are being performed on SS-C andSS-65 specimens containing >50 wt % Fe203. The impact and the fracture pro-perties of these compacts are being correlated with the microstructural char-acteristics.

c. Sludge Pelletization with Aqueous Sodium Silicates

Several scoping experiments initiated earlier [STEINDLER] havebeen completed on the pelletizing of Fe203 powder with aqueous sodium silicatesolutions. In these experiments, weighed amounts of as-received Fe203 powderwere placed in a rotating glass jar. Sodium silicate solutions with Si02:Na20ratios of 3.22 (N solution) and 1.60 (B-W solution) and of various dilutionswere fed into the jar through a long copper tube containing small drilledholes. Fe203 agglomerates were formed as sodium silicate was mixed with thepowder. The ratio of the volume of added liquid to the unit weight of powderwas 3.33 cm3/g, and the rate of addition of the liquid was 2.25 cm3/min forall experiments.

For each test, the composition of the sodium silicate solution,the weight percent of pellets of each of several size fractions, and the amountof powder that stuck to the wall of the glass jar and did not form pellets aregiven in Table 5.

V.

Z

0

0-J

TIME (500 s/div)

N%

TIME (500 s/div)

(b)(a)

Fig. 4. Load-Time Trace for (a) Test 48 with Pyrex Specimen and (b) Test 71 with SS-65 Specimen

Containing 33.3% Fe203.

20

0Z

0

J

23

Table 5. Pellet Size Distribution from Pelletization of Fe2 03 in thePresence of Sodium Silicate Solution

Composition of Plesi ieFatowSilicate Solu- Pelles In Size Fraction, Wt Wt % Sticking

Test tiona Added to to Wall ofNo. 150 g Fe203 +1.3 cm -1.3 +1.0 cm -1.0 +0.6 cm -0.6 cm +6b -6 +18b -18 b Glass Jar

1-1 B-W sodium 15.39 23.36 11.34 9.92 0.95 0.61 36.43silicate:water(1:1 by volume)

1-2 B-W sodium 4.53 30.05 15.52 10.33 3.64 2.59 33.34silicate :water(1:1 by volume)



3-1 N sodium 13.53 22.67 11.87 13.33 2.00 1.20 35.40silicate:water(1:1 by volume)

3-2 N sodium - 20.07 25.47 10.20 3.00 1.53 39.73silicate:water(1:1 by volume)

4-1 N sodium 18.87 18.93 9.87 10.67 3.33 8.33 30.00silicate:water(1:2 by volume)

4-2 N sodium 6.67 19.33 13.33 3.33 0 24.00 33.34silicate:water(1:2 by volume)

aS102:Na20 ratios in B-W sodium silicate and N sodium silicate are 1.60 and 3.22, respectively.

bU.S. standard sieve size.

Note that each set of conditions was repeated (ej., tests 1-1and 1-2). Evidently, the reproducibility of the tests with respect to thesize of pellets was poor. However, in all tests, 33-40% of the originalpowder did not form pellets. The choice of B-W or N solution did not influ-ence pellet size distribution in any systematic manner. However, greaterdilution of the B-W solution (e.g., see tests 2-1 and 2-2) resulted in the

+-in. fraction being larger. The poor reproducibility of the tests indicates

that better control of the experiments will be required. A laboratory-sizepelletizer, which we have just received, is expected to provide significantlyimproved control during the experiments.

The pellets prepared in the above experiments were oven-driedat 100 C and in some cases were fired at 600*C for 4 h. Both the dried andthe fired pellets were tested for their mechanical strength in compression.Some interesting features were observed:

a. The strength of the oven-dried pellets made from B-Wsolution (SiO2:Na20 - 1.60) varied from 1 x 103 to

24

4 x 103 kPa, whereas those made from N solution(S10 2 :Na2 0 = 3.22) had a strength of'2 x 102 to8 x 102 kPa.

b. Firing of oven-dried pellets made from B-W solution didnot alter their strength.

These results clearly indicate that the composition of thesodium silicate solution used for pelletization must be selected carefully.In addition, processing of the pellets may not require a firing step; oven-drying alone may provide sufficient pellet strength for andling.

25

II. NEUTRON ACTIVATION AND TRACER STUDIES(L. J. Jardine, J. K. Bates, andJ. E. Kincinas)

A. Introduction

Dispersion of wastes into the biosphere is the principal potential hazardof nuclear waste disposal. Characterization of waste forms requires the useof sensitive analytical methods to obtain dispersion data of low-rate phenom-ena. Neutron activation analysis (NAA) has been shown to be one applicablemethod for measuring leach rates. Radioactive tracers can provide sensitivemethods for applications in which NAA is not possible or practical. Bothmethods are likely to be useful for characterization measurements other thanleach rates.

The general objective of this program is to develop techniques and qualifymethods that utilize neutron activation analysis and radioactive tracers forcharacterizing simulated waste forms. Comparison of characterization testresults obtained using these two methods and with other existing character-ization data will be used to Specify conditions under which such tests andthe resulting data may be extrapolated to fully radioactive specimens. Thecurrent focus is on leach-rate characterizations of simulated waste glassesand, to a lesser extent, advanced waste forms.

Major goals are completion of the development of the NAA method forleach-rate determinations, definition of the accuracy, sensitivity, and limi-tations of NAA, and establishment of the applicability of the test results tolarger-scale waste forms. Similar goals exist for the development of leachrate determinations based on the use of radioactive tracers.

B. Radioactive Tracer Method Qualification

1. Introduction

A series of experiments have been defined to qualify the use of NAAand radioactive tracers for measurement of the leach rates from SRL frit 211simulated waste glass (and later, possibly, from PNL 76-68). Four classes ofchemical elements present as minor (~5 wt %) elements in SRL defense wastesand of concern to radioactive waste management are the focus of the measure-ments. The four classes are alkali metals (Cs), alkaline earths (Sr and Ba),rare earths (Ce, Eu), and noble metals (Ru). Simulated waste glass containingradioactive tracers has been prepared; this glass is being leached directlyand also is submitted for neutron activation prior to leaching. Leach rateswill be measured, wherever possible, using four different techniques: (1)conventional chemical solution analysis (ICP,* AA,t flame emission) ofleachates from non-neutron-activated glass, (2) gamma-ray spectroscopy analy-ses of leachate solutions from neutron-activated glass specimen, (3) gamma-ray spectroscopic analyses of leachate solutions from glass specimen containingradioactive tracers, and (4) gamma-ray spectroscopic analyses of leachatesolutions from neutron-activated glass specimen containing radioactivetracers.

*Ion-coupled plasma emission spectroscopy.

tAtomic absorption.

26

The objective of these measurements is to obtain comparative leachrate data in order to define the reproducibility, sensitivity, limitations,and precision of leach rate measurements by the various methods.

2. Experimental Approach and Glass Preparation

A flow diagram of the experimental approach and the required ana-lytical measurements is shown in Fig. 5. A batch (300 g) of SRL frit 211simulated nonradioactive waste glass was melted for about 4 h at 1050*C in ahigh-purity alumina crucible, together with additives of cesium, barium,strontium, cerium, neodymium, europium, and ruthenium. The melt was thenpoured into water to make a frit . This 300-g batch of SRL 211 ** glass will beused as a source of glass for the future comparison tests, as indicated inFig. 5. The compositions of the starting SRL 211 and the SRL 211** glassesare given in Table 6, along with the composition of SRL 211*, used in earlierwork [STEINDLER].

Table 6. Nominal Compositions of Simulated SRLUsed in this Work

Note: Frit 211* was used in [STEINDLER].

Waste Glasses

Conc. of Element, wt %

211 Frit [SRL] 211* Frit 211** Frit

Element

Si 20.34 20 19.65Na 11.5 11.15 11.04Fe 10.28 10.08 9.75B 2.48 2.43 2.40Ca 3.78 2.78 3.66Mn 2.46 2.41 2.38Al 2.01 1.96 1.94Li 1.48 1.47 1.44Ni 1.34 1.31 1.29

Additive

Cs - 0.19 0.12Sr - 0.07 0.30Ba - 0.10 0.05

Ce - 0.52 0.73Nd - 0.63 0.79Eu - 0.003 0.01Ru - - 0.51Mg - 0.11 -

27

frit (SRL 211)

Non-Tracer Studie'

CsBaSrCe miNdEuRu

Melt Glass at1050 C, 4 h

frit (SRL

s Radioacti

nor elementadditives

211"")

ive Tracer Studies


NeutronActivation

Leach Leach

" Si, Cs, (Ba/Sr, * Si, Cs, (Ba/Sr,RE,' Ru) RE, Ru)

" 134Cs

"*(85Sr)

" 141Ce

* ('47Nd)" 152Eu, 154Eu

" (103Ru)


NeutronActivation

Leach Leach

" Si, Cs, (Ba/Sr, " Si, Cs, (Ba/Sr,RE, Ru) RE, Ru)

* 13Cs

* (855r), 133Ba" 139Ce

" 88Y

" 152Eu

Z 06Ru

" 134Cs/137Cs

* (85Sr)/ 133Ba

* 141Ce/139Ce

" (147Nd)

" 154Eu/152Eu/88Y

* (103Ruw/ 10 6Ru

" Solution analyses to be performed.* RE - rare earths.

Fig. 5. Flow Diagram of the General Approach Being Used forComparing SRL Frit 211 Waste Glass Leach Rate Resultsfor Neutron Activation and for Radioactive TracersAddition. Later, experiments with PNL 76-68 glassmay be done.

137Cs

(85Sr)

'33Ba'39Ce152Eu88Y

106Ru

traceradditives

_ _ _ _ _ _ _

28

A portion ('50 g) of the initial 300-g glass batch was melted in aplatinum crucible for about 2 h at 1050*C and then poured into a Pt-5 wt %Au mold. The glass was then annealed in the mold for two hours at 500*C,producing a 10-g button (~40 mm diameter x-7 mm thick disk) of frit 211**simulated SRL waste glass. The nonradioactive molded glass button was thensectioned with a low-speed Isomet saw (Model 11-1180) into pieces(~1/2 g of 15 x 15 mm x 1 mm) having a~400 mm2 surface area. Water was usedas the cutting lubricant. One-half of the cut specimens are being preparedfor leach tests, and the remainder have been neutron-activated at theUniversity of Illinois reactor. Leach tests of these two types of glassspecimen provide a data base that allows di;.ect comparisons of the leachrates of nonactivated and neutron-activated glass and of conventional solu-tion analyses with NAA techniques. These initial results are reported belowin Sections II.B.3.a. and II.B.3.b.

The steps in Fig. 5 involving radioactive tracers will produce a211** glass frit for two types of leach tests, i.e., leaching of radioactivetracers from glass (reported in Section II.B.3.c. below) and leaching ofneutron-irradiated glass containing radioactive tracers (reported in SectionII.B.3.d. below).

A 50-g batch of SRL frit 211** has been prepared from the initial300-g batch that contains about 300 pCi of radioactive isotopes. The radio-active isotopes, 13 7Cs, 13 3Ba, 139Ce, 152Eu, 88Y and 10 6Ru, were selected astracers (~1 pCi per gram of glass) to be added to the SRL frit 211**. Theradioactive properties of these isotopes are summarized in Table 7.Strontium-85 was not used in these initial experiments because of interferingy-rays from 1 0 6Rh. The tracers were purchased as HCl solutions, with 139Ceand 88Y being the only carrier-free isotopes. All solutions, after dilutionfrom vendor-supplied materials, are stored in Pyrex volumetrics, with theexception of the 137Cs which is stored in a polypropylene volumetric.

Aliquots (sizes given in Table 8) of these six stock solutions werethen added to a rectangular (17 g) platinum crucible approximately3.2 cm x 5.1 cm x 3.2 cm. The platinum crucible had been re-formed into arectangular shape by gentle hammering of a 50 cm3 oval platinum dish crucible.This shape facilitated the cutting of glass specimens for leach testing withtolerable losses. Table 8 shows the proportions and the calculated activitylevels loaded into the platinum crucible, based on vendor-supplied values.From the 9300X in the platinum crucible, two small aliquots (lOX and 20X) ofthe radioactive mixture were taken,.evaporated to dryness, and analyzed byy-ray counting. The average pCi in the crucible based on measured resultsfrom these two aliquots are shown in Table 8 in the "Before Melt" column.

Except for 1 3 7Cs and 1 06Ru, the measured pCi loaded into the cruci-ble agrees with the pCi calculated from the vendor-supplied values. The 33%lower value for 137Cs is assumed to be due to an error in the vendor-suppliedvalue of pCi present in the purchased stock tracer solution. The 20% highervalue for 10 6Ru is believed to be primarily due to uncertainty in the absoluteintensity of the two photons (i.e., 512 and 622 keV) associated with the 8decay; the disagreement with the vendor-supplied value is not regarded assignificant. In addition, because of the long counting times required for

29

Table 7. Radioactive Isotopes Used and their Radiation Properties

PrimaryRadioactive Decay y-radiation,

Tracer Half-Life Mode keV (photons/100 decays)

Alkali

137Cs 30.2y ~-661.7(85.0)

Alkaline Earth

8 5 Sr 64.7d ECa S4.0(99.3)

133Ba 10.7y EC 356.0(62), 81.0(31.7), 302.8(18.1),383.8(8.0), 276.4(7.0)

Rare Earth

139Ce 137.5d EC 165.9(79.9)

152Eu 14y 73% EC (B+), 1408.0(21.3), 121.8(29.1), 344.3(27.2),27% 8~ %4(14.8), 1112.1(13.8), 778.9(13.2)

88Y 106.6d EC, B+ 898.0(91.3), 1836.1(99.3)

Noble Metal

106Ru

( 10 6 Rh 30s) 368d 511.8(19), 622.2(9.8), 1050.5(1.6),616.3(0.82), 873.7(0.45), 1128.2(0.42)

aElectron capture.

these small aliquots of 10 6Ru activity, there is more uncertainty in the512-keV transition due to room background variations than in the 622-keVtransition. Thus, two values, derived from the two different transitions,are given in Table 8 to allow better estimation of the uncertainty.

About 30 g of the 50-g batch of SRL frit 211** (previously siftedto -14 mesh or <1.2 mm) was mixed with the about 9.3 mL of the above describedtracer solution directly in the platinum crucible, using a 0.3-cm glassstirring rod. The loaded crucible was then dried at about 140*C under a heatlamp and on a hot plate for several hours, after which it stood overnight ina hood.

The next day, the remaining about 20 g of SRL frit 211** was addedto the top layer of the platinum crucible and the entire crucible was placedin a 500-mL A1 2 03 secondary crucible having an A1 2 03 cover. The mixture andthe secondary crucible were then placed in a furnace, and its temperature washeld at 1050 C for 3 1/2 h to melt the glass. The power was then turned off,and the furnace was allowed to cool overnight naturally.

Weighing of the cooled ingot and platinum crucible indicated apossible but negligible weight gain (i.e., 20 mg in comparison to the 67-gtotal weight). The platinum was then cut with scissors and was removed by

30

Table 8. "Mass" Balance of Radioactivity Added to SRL Frit 211** Glassand Amount of Radioactivity Measured after Melting

Aliquot ,a Carrier, Calculated,b Measured, Before McltC Measured, After Meltd Loss,Isotope A pg pCi pCi (average) PCi (average) %

137Cs 300 4.5 60 40.1 0.5e 37.3 .38.9 0.5g (40 2) 37.8 0.411 (38 2) -

39.9 0.4

133Ba 2000 9.6 40 46.1 0.5 h43.1 0.544.2 0 .5g (45 2) 43.7 0.5 (44 3) -

46.5 0.5

139Ce 2000 1 40 41.9 0 .6e 39.5 0.641.4 0.6g (42 2) 41.9 0.6h (42 3) -

43.5 0.6

15 2Eu 2000 40 48 50.3 1.0e 49.9 0.548.2 1. 0g (49 2) 50.0 0.5 (50 3) -

52.5 0.5

88Y 1000 <0.04 51 49.0 2.0e 47.4 1.048.1 2 .0g (49 2) 50.5 1 . 0h (50 3) -

51.5 1.01

106Ru 2000 40 50.4 67.9 4 . 0e 48.5 2.0

63.0 4 . 0 g (66 7)1 52.6 2.0 (52 5)k <2553.6 2.0

58.9 3.8e 43.3 3056.0 3 . 8 g (58 6)i 46.6 3.0h (45 5 )i <25

47.7 3.0

Total 9300~95 289 <25

al - 0.001 mL of solution.

bBased on vendor-supplied values and initial dilutions used to prepare tracer stock solutions.

CThe measured average values based on Y-ray analyses of two separate aliquots of 10)' and 20)A each takenfrom platinum crucible before any glass frit additions or melting.

dThe measured average values based on Y-ray analyses of three representative about 0.45-g glass specimens

sectioned from the cooled melted glass and based on analyses normalized to entire mass (50 g) of glassmelt.eValue derived from 10A aliquot.

(Value from wafer (1,1).&V'alue derived from 20X' aliquot.

hValue from wafer (2,6).

iValue from wafer (3,11).iBased on absolute intensity value of 0.095 for 622-keV Y-ray.kBased on absolute intensity value of 0.19 for 514-keV Y-ray.

stripping pieces away from the 50-g glass ingot. All outer edges of the glassingot were removed by sectioning the ingot with an isomet saw into six majorpieces. These six pieces were further sectioned into 27 minor (~1-cm x 1-cm x~0.1-cm) pieces or wafers of about 0.5 g each. Figure 6 illustrates the sec-tioning and coding used to label the wafers cut from the ingot.

Of the 27 cut wafers, six were not usable, as indicated by the notesin Fig. 6. Ten wafers, (1,2), (1,3), (2,4), (2,7), (3,9), (4,14), (4,17),(5,21), (6,24), and (6,27), from locations throughout the ingot were cleaned,weighed, measured, wrapped in aluminum foil, and submitted for neutron acti-vation analysis. Three wafers, (1,1), (2,6), and (3,11), from three differentlocations were submitted (1) for y-ray analyses to establish the initial

31

1 2 3

4 5 6 1 4

4 5 6

Piece locations from original ingot

(Top) (Side) (End)

(Top) (Tcp)

1, 3 2,8 3,11 4,18 5,23 (,27

,12(?) 2,7 3,13(?) 4,17 5,22 6,26

2,6 4,16 5,21 --

1, 2 2,5 3,10 4, 15 5, 20 6,25

1, I 2,4 3,9 4, 14 5, 19 6, 24

(Bottom) (Bot tom)

Wafer locations cut from pieces

(Side) (Side)

Fig. 6. Schematic of 50 g Ingot (-5-cm long x ~3.8-cmwide x ~1.5 cm thick) of Radioactive Glass,Illustrating the Coding of the Wafers Cut fromthe Original Ingot

Notes: 1. The positions of major pieces 1and 3 could be interchanged sincethe precise location was not mon-itored; similarly, major pieces 4and 6 could be interchanged.

2. Two cut wafers, (1,12) and (3,13),associated with pieces 1 or 2 or3 exist for which precise loca-tions could not be assigned. Thesetwo wafers are assigned arbitrarylocations shown as (1,12) and(3,13).

3. Wafer (2,5) had a small (~0.3-cm x0.2-cm) triangular chip which waslost during cutting.

4. Wafers (4,18), (5,19) and (5,20)were fractured during the cuttingand cleaning procedures.

homogeneity of the tracers in the glass and (2) for 7- and 14-day leach testsat 900 C in deionized water to establish whether sufficient tracer had beenadded for measuring leach rates. The remaining eight wafers, (2,8), (3,10),(4,15), (4,16), (5,22), (5,23), (6,25), and (6,26) were prepared for y-ray

32

analyses to establish both the ins 1 radioactivity prior to leaching andthe degree of homogeneity of the radioactive tracers added to the glass. Thelatter eight wafers will be stored until the ten neutron-activated wafers arereturned; then, the entire set of 18 wafers will be leached at 90*C in deion-ized water using the modified MCC-1 leaching procedure* for radioactivity.

The results (Table 8) of the gamma-ray analyses of the three solidwafers (1,1), (2,6), and (3,11) agreed to 5% for all six isotopes. Thismeans that the tracers were apparently homogeneously distributed throughoutthe glass. These results were then used to convert measured activity in thethree about 0.5 g pieces of glass to that present in the entire 50-g batch(the Measured After Melt column of Table 8). Direct comparison of these re-sults with the Measured Before Melt column allows determination of the amountof radioactive tracers lost from the melt during glass fabrication. Thiscomparison shows that within the sensitivity of the measurements, only some106Ru was lost and that this was no greater than 25%. Counting of the Al203secondary indicated very small amounts of radioactivity there, namely, 13/Csand 10 6Ru, but in the same relative amounts. After removal of most of theglass from the platinum crucible with boiling acid solutions, it containeda larger (~"100X) amount of 10 6Ru than of 13 7Cs. Thus, the "loss" of 10 6Ruwas via some type of reaction with the platinum crucible during the melting.

3. Leaching Characterizations

The introduction of radioactive tracers into a waste form is onemethod proposed for the measurement of elemental release during leaching.This should be a particularly sensitive method; it is described in SectionII.B.2. of this report. Validation of the method requires that the useful-ness and limitations of the method be defined; this will be done in a seriesof leach tes-... Tests now in progress are described below.

a. SRL frit 211**

Leach tests on SRL 211** glass that has not been spiked or ac-tivated will serve to establish the basic leaching behavior of this glass.Analyses will be done by dc lasma spectroscopy for all constituents of theglass except for Cs, Na, and Li (which will be done by atomic absorption).These tests will follow the MWC-1 procedure [MCC], and their duration will beup to one year. These tests will serve as the basis on which the leaching ofspiked glass can be compared. The tests, started Dec. 22, 1980, will be re-ferred to in future reports as tests L-154 to L-169. The first 14 days oftests have been completed and leachates submitted for analysis.

b. SRL frit 211**, Activated

It is expected that several waste additives, i.e., Nd, Ce, Eu,Sr, and Ru, will have leach rates that can be measured only by NAA techniques.For this reason, a series of leach tests using the MCC-1 procedure are beingdone, using NAA. These tests will serve also as a basis of comparison with

*Developed by the Materials Characterization Center at Battelle PacificNorthwest Laboratories.

the spiked-glass results. The tests (L-134 to L-148) were started Dec. 18,1980. The first 14 days of tests have been completed. Preliminary resultssuggest that only cesium and antimony can be detected in leachate. If thisis so, the leach rates of the rare earths will have to be obtained afterincreasing the quantity of activated species in the glass.

c. SRL frit 211**, Spiked

Spiked glass (section II.B.2.) has been made and prepared forleach testing. These tests, scheduled to be run simultaneously with thespiked samples and activated samples (below), have not yet started. However,preliminary leach tests on three pirated spiked samples have been done to aidin program direction. These tests were done by MCC-1 procedures and were for7, 14, and 42 days. Radioactivity in the leachates from the 7-- and 14-daytests has been counted, and only barium and cesium were observed. These dataare consistent with the preliminary data for activated glass (L-134 to L-148)in that the rare earths, Y and Ru, are not released from the glass in verylarge amounts. Although this result is not totally unexpected, it does pre-sent a problem in measuring the leach rates of these elements. However, thisseries of tests will provide results for barium and cesium, allowing theamount of cesium added in any future tests to be substantially reduced.

d. SRL frit 211**, Spiked and Activated

By use of a spiked SRL 211** glass that has also been acti-vated, it will be possible to monitor the leach rates of two isotopes of thesame element, 1 34Cs and 13 7Cs or 13 9Ce and 14 1Ce. These results will provideadditional clarification of the behavior of radioactive tracers in glass.These tests will be run concurrently with the spiked glass tests.

The low element release noted for several elements in SRLfrit 211** probably results from the formation of a layer on the glass sur-face. This layer is quite distinctive and, by counting a flake removed froma leached-glass sample, is known to be depleted in cesium. Such a layer wasalso noted in the leaching of SRL 211* glass [STEINDLER] and for which theleach rates of all elements for times between 7 and 42 days were essentiallyzero. This suggests that it may be difficult to adjust the "spikes" in SRLfrit 211** glass to a level where the spikes can be detected in the leachate.For this reason, tests are also being done on PNL 76-68 glass. The leachrates of all elements of interest have been determined for this glass, pro-viding a basis for comparison.

e. PNL 76-68 Glass, Nonactivated and Nonspiked

These tests were done as part of the MCC round robin; resultsfrom many laboratories will be available for comparison.

f. PNL 76-68 Glass, Activated

These tests are being done to obtain the leach rates of therare earths and ruthenium, which generally cannot be detected by ICP analysis.Analysis by NAA requires a prior separation of cesium from the leachate toallow the other radioisotopes to be detected. Such a separation has been

3.j

34

demonstrated, and leach rates for Ce, Eu, Ru and possibly Nd may be available.These tests were started December 15, 1980, and are being done by the MCC-1procedure. Tests of 28 days have been completed and samples submitted foranalysis. The tests are identified as L-120 through L-126.

g. PNL 76-68 Glass, Spiked

Spiked 76-68 glass has not yet been made since the amount ofspike necessary for detection in the leachate is not currently known. Thesequantities will be determined from the separations being done on activatedPNL 76-68 glass (above).

h. SRL frit 131*, Nonactivated, Spiked, and Activated

The same scenario as used for SRL frit 211** will be appliedto SRL frit 131 with Savannah River waste additives (i.e., SRL 131*). A sup-ply of frit 131 has been requested, and these experiments will begin duringthe next report period.

From the above leaching experiments it should be possible tocompare leach rates as determined by spiking with leach rates determined byother methods of detection.

C. Qualification of NAA Method

Neutron activation analysis (NAA) is one method of simultaneous,multiple-element leach rate measurement being utilized by Argonne NationalLaboratory. The method and procedure have been described previously [FLYNN].Briefly, they involve irradiating a solid waste form with neutrons to produceradioisotopes of many waste form constituent elements directly in the wasteform matrix. The waste form is then leached in solution under standardizedconditions and the leachate is analyzed for released radioisotopes, usingnuclear instrumentation and counting procedures.

Since the NAA procedure described above had not previously been appliedto the leaching of nuclear waste forms, the purpose of this program has beento demonstrate the usefulness and limitations of the method, as well as toinvestigate any problem areas. One potential problem (since the solid isirradiated prior to leaching) is the measured leach rate being affected byirradiation. These effects include matrix damage due to recoiling nucleiproduced from gamma ray emissions after thermal neutron capture, fast neutronscattering, fast neutron reactions, and fission reactions. In addition,effects due to attenuation by the waste form matrix of both incident neutronsand emitted gamma radiations may be important.

The above issues have been addressed and reported [BATES-1980]. Themethod is also described in detail in a report [BATES-1981]. Results fromthese two sources that have not been reported in the previous quarterlyreports [STEINDLER] are discussed below.

1. Neutron Irradiation Effects

When a solid waste form is immersed in a neutron flux in the pro-cess of producing a specimen for leaching studies, nuclear interactions

35

convert stable nuclides in the waste form matrix to radionuclides, whichlater decay and are identified by their characteristic radiation emissions.The types of nuclear interactions that occur in a solid depend on the energyof the incident neutrons and on the nuclear charge and reaction cross sectionsof the target nuclei.

In the present studies, irradiations are conducted with a flux com-posed mainly of thermal neutrons so that neutron capture (n,r) is the dominantreaction. However for low-Z elements which have a low coulomb barrier,charged-particle emission may also occur. The smaller fraction of the fluxcomposed of neutrons above thermal energies can produce charged-particle re-actions, (n, p) or (n, a), inelastic scatter reactions, (n, n'), and multipleneutron emission reactions, (n, 2n). Fission reactions, producing high-energyfission fragments, can also occur in irradiated samples containing thoriumand actinides. Finally, the intense gamma radiation fields present in reactorcores, though interacting primarily with electrons in the solid, also havethe potential of causing (r, p), (r, n) or (r, 2n) reactions and induce damageeffects from recoiling nuclei.

All of the above reactions have the potential to affect both thematrix-forming and matrix-modifying bonds in the solid. Leach rate measure-ments based on NAA involve the detection of radionuclides that have at sometime been involved in a nuclear reaction process. A further assessment of thepotentially more damaging interactions is given below.

a. Lattice Damage Effects

The primary method of activation is by neutron capture. Whenthis occurs, a gamma ray is emitted and simultaneously the radionuclide under-goes a recoil. The recoil energy is inversely proportional to the energy ofthe emitted gamma ray. These recoil energies typically vary from 100 to 500eV for lighter nuclei to 20 to 100 eV for heavier nuclei. Since matrix bondenergies are in the range, 5-10 eV, nuclei involved in these reactions may bedislocated.

A possibility of matrix damage may also result from interactioninvolving the fast components of the neutron flux. For example, (n, n') re-actions may cause the dislocation of target nuclei due to scattering effects.For nuclei of mass number less than 20, neutron energies above 300 eV are pro-bably sufficient to cause dislocations, while for nuclei of mass number lessthan 150 the coresponding neutron energy is about 2000 eV.

It is important to note, however, that for both of the previ-ously described lattice damage scenarios, there is no independent evidenceindicating that such damage actually occurs in simulated glass waste forms,let alone alters the measurement of leach rates. It is possible that any dam-age is annealed out of the lattice due to the elevated temperatures achievedduring irradiation. The question is simply addressed by comparing the leachresults observed for irradiated and nonirradiated samples. Such tests aredescribed in Section C.2 below.

b. Neutron Attenuation Effects

Another potential irradiation effect that must be consideredfor borosilicate glasses is attenuation of the incoming neutron flux due to

36

reaction with boron. This effect is important since it can lead to nonuniformdistribution of activated nuclides. It is a result of boron (20% 1 0 B) beingpresent in glasses at concentration levels up to 10% and because 10B has alarge, 3800-barn thermal neutron cross section for the 10B (n, a) 7Li reac-tion. This reaction could result in nonuniform thermal neutron flux in a solidsample and if a nonuniform radioisotope distribution should occur in asample, the calculated fraction of element release would be incorrect.

The magnitude of this effect, which is dependent on sample sizeand composition, has been estimated by assuming 10% B203 in the solid sample.For irradiation depths of 1/2 mm (a sample thickness of 1 mm), about 15% ofthe thermal neutron flux is removed by (n, a) reaction between the surface andthe center of the specimen. This value increases to about 50% for 2-mm irra-diation depths. The effect is a nonuniform distribution of radionuclei as aresult of thermal neutron capture reactions. Nuclei produced from fast neutronreactions would not be affected since 10B has a very low cross section forfast neutrons and would not perturb this constituent of the flux.

To determine whether this effect is actually observed, severalexperiments were completed. Initially, five pieces of PNL 76-68 glass, ofdimensions 10 x 20 x 1 mm, were stacked together as in a loaf of sliced breadand the pieces were identified sequentially. The entire stack was then irra-diated as one unit for four hours at a flux of 3 x 1013 nth/s-cm2 . Subse-quently, the individual pieces of glass were analyzed for selected activitiesrepresentative of five different radioisotopes (6 Zn, 13 4Cs, 14 1Ce, 152Eu,154Eu), using gamma counting (Table 9). The activities were normalized forthe mass of each sample. The piece at the center had 10% less activity due to134Cs than did the end pieces. The corresponding decreases for 6 5Zn, 141Ce,15 2Eu, and 15 4Eu were 20, 25, 30, and 20%. The second and fourth pieces, whichhad similar activities, had activity levels lower than those of the end piecesbut greater than that of the center piece. This test was repeated using anothersource of PNL 76-68 glass, and essentially identical results were obtained.

These results are consistent with the neutron capture crosssections for activation by thermal and epithermal (>0.5 eV) flux components(Table 9). Cesium-134 is less affected by thermal neutron attenuation becauseit has a larger epithermal flux cross section; 15 2Eu is most affected becauseit has a greater dependence on activation by thermal neutrons.

The piece from the center of the stack from each experimentwas then autoradiographed. In this experiment, the center pieces were laidflat on a section of beta-sensitive film. The exposed film was then scannedwith a densitometer. No variations in light transmission greater than 8%were observed. This lack of exposure variance results since 1 34Cs, which isthe dominant beta-emitter darkening the film, is produced throughout thecentermost piece, as explained above.

These tests, combined with theoretical calculations, indicatethat the decrease in neutron flux due to interactions with the boron presentin the glass should not be a significant factor if the dimensions of theirradiated solids are controlled. For this reason, the NAA test procedure[BATES] requires that the irradiated solids samples be no thicker than 1 mmfor waste forms for which neutron attenuation is estimated to be an importanteffect.

37

Table 9. Counting Statistics for SectionsPNL 76-68 Glass

of Neutron-Activated

Normalized Count Rate, dpm x 10 7 /g of sampleSample

No. Mass, g 6 5 Zn 1 3 4 Cs 1 4 1 Ce 1 5 2 Eu 154Eu

Test 1

1 0.5499 2.3 5.3 2.9 8.7

2 0.7122 1.9 4.9 2.2 6.83 0.6868 1.8 4.7 2.0 6.04 0.5540 1.9 4.6 2.0 6.15 0.5932 2.1 5.0 2.3 7.4

ratioa 0.82 0.91 0.76 0.69

Test 2

1 0.3215 1.7 4.4 2.4 6.1 7.6

2 0.4137 1.5 4.0 1.9 4.9 6.5

3 0.2366 1.4 3.9 1.7 4.3 6.0

4 1.1831 1.6 4.1 1.9 5.2 6.6

ratiob 0.80 0.90 0.70 0.70 0.79

Neutron Capture Cross Sections

Othermal,barns 0.8 27 0.57 9200 390

i ntermediate, barns 1.8 360 0.43 3300 1635

aRatio is the normalized count rate for sample 3 dividedthe normalized count rates for samples 1 and 5.

aRatio is the normalized count rate for sample 3 dividedcount rate for sample 1.

by the average of

by the normalized

2. Leach Experiments-Conditions, Results, and Discussion

To assess the overall effect of neutron irradiation on the measure-ment of leach rates, the ultimate question to be answered is whether leachingfrom an activated sample differs from leaching from a nonactivated sample.To address this point and to assess the utility of the method, a series ofleach tests were done on the following glasses: Pacific Northwest Laboratory(PNL) 76-68, Savannah River Laboratory (SRL) 211, SRL 211 with additives (i.e.,SRL 211*), and No. 7740 Pyrex. The additives to SRL 211 frit are representa-tive of minor elements present in SRL wastes, and the compositions of all theglasses are given in Table 10. Preliminary results from these experimentswere discussed previously [STEINDLER].

A matrix of leach tests utilizing the static leach test procedureof the Materials Characterization Center (MCC-1) [MCC], as modified for NAA

38

Table 10. Compositions of Leach-Tested Glasses

Oxide in Glass, Glass Typewt % PNL 76-68a PNL 76-68 (U)b SRL 211*c SRL 211C Pyrex No. 7 7 4 0 b

Si

Na

Fe

B

Ca

Mn

Al

Li

Ni

Cs

Sr

Ba

Ce

Nd

Eu

Mo

Zr

Others

38.2

12.40

9.89

9.37

2.0

0.28

0.52

0.38

0.40

1.11

4.21

0.08

1.7

4.5

8.7

40.0

12.50

9.6

9.5

2.0

0.2

1.0

0.4

0.6

1.2

1.7

0.07

2.2

5.0

4.9

38.3

15.1

11.9

7.8

4.9

3.5

6.5

2.8

0.9

43.8

15.2

12.4

8.0

5.1

3.6

4.2

2.9

1.0

80.5

3.8

12.9

2.2

0.2

0.07

0.2

0.7

0.6

0.003

0.2 0.4

aAnalysis reported by PNL using KOH fission/ICP solution analysis.

bNominal composition.

cAnalysis by acid dissolution/ICP solution analysis.

[BATES], has been completed, and the test conditions are summarized in Table11. The PNL 76-68 glass was exhaustively studied because MCC data on a similarglass were available for comparison [MCC]. Less extensive tests were donewith the other glasses.

Identical leach tests were done on both activated and nonactivatedsamples. For the nonactivated samples, each leachate was divided into two

equal aliquots, one of which was analyzed in-house for silicon (colorimetry)and cesium (atomic absorption), the other of which was analyzed commerciallyfor silicon and other elements (inductively coupled plasma spectroscopy, ICP)

Table 11. Leaching Test Conditions: MCC-1 Procedure.

Normalized Elemental Loss (NL)i, g/m2

Glass Type

LeachingTime,days

Siliconnonactivated

colorimetry

sample

ICP

Siliconactivated sample

colorimetry

Cesiumnonactivated sample

AA-1 AA-2

Cesiumactivated sample

AA-1 NAA

PNL 76-68

SRL 211*

SRL 211 c

2

7

14

22

40

2

714

40

14

40

Pyrex No. 7740 c 14

4

8

15 0.122

30 1

6

78

9 0.4

4 3

5

10

16 1

23

34 4

7

10

10 1

10 2

12 2

10

1 0.5

6

814 1

32 2

4

4

7 2

7

15

26 0.4

36

48 5

5

8

10

10 2

aa

30 2

40

53 9

a

a

8 7

15 12

25 1 20 0.8

35 30

54 2 41 3

4

6

5

6

a 8 2 9 1

a b b

1

aBelow detection limit.

bAnalyses incomplete.

cThese glasses contain no cesium.

90*C, Deionized Water

40

and cesium (atomic absorption). The dual analysis provided a reliabilitycheck for the methods of solution analyses, in addition to providing data bnseveral elements present in the solution (ICP). For the activated samples,each leachate was also divided into two aliquots. One aliquot was analyzedfor silicon and cesium in-house; the other was analyzed for 13 4 Cs and otherradionuclides using NAA.

By comparing the silicon release from activated and nonactivatedsamples in these tests, it is possible to determine whether the glass networkhas been affected by neutron irradiation. Similarly, the release of highlymobile cesium can be studied. Finally, the overall leaching behavior of non-activated 76-68 glass can be confirmed by comparing the present results withMCC data on similar glass [MCC].

The leach results for silicon and cesium (as determined for eachglass) are given in Table 11. Leach data for PNL 76-68 glass at 900C areillustrated in Figs. 7, 8, and 9. The overall leaching behavior of the PNL76-68 glass (Fig. 7) indicates that the alkali metals, Na and Cs, togetherwith Mo and B, have the highest normalized elemental losses, (NL)i--these

70 T 10

60 1-

E

U)U)0JJ1-z

w-JW

N-J4

0Z

50 F

401

8B -No

Mo

Cs

0 Mo

o BA Co ICP

O Na- Si" Cs) AA

Co

30 F-

20I

I0

0 1~ IfI

0 10 20 30 40TIME, days

Fig. 7. Normalized Element Losses fromPNL 76-68 Glass Leached at 90*Cin Deionized Water

i

i

I

--

i

T T

41

50

E 40

0J

30

z90

U) U

0

W 20NJ

0

O COLORIMETRY (NONACTIVATED)" O ICP (NONACTIVATED)

e o f COLORIMETRY (ACTIVATED)0

0 1- -0 10 20 30 40

TIME , days

Fig. 8. Normalized Silicon Loss fromActivated and NonactivatedSamples of PNL 76-68 Glassat 900C

levels being slightly greater than that for Si. Of the alkaline earths, onlyCa could be detected; a constant amount of this element was released intosolution with time. These general leaching trends are similar to those re-ported previously for PNL 76-68 glass [MCC]. For each element detected, theratio between the present values of (NL)i and the previous data [MCC] is 0.7.This establishes the consistency of the two sets of experimental results andindicates that no unexpected leaching behavior occurred in the glass studiedhere.

The comparison of silicon releases, (NL)5i, for activated and non-activated samples (Figs. 8) indicates that neutron irradiation has no signif-icant effect on (NL)5i. The precision for the replicate samples done at 14and 40 days is 10% and all the (NL)si values are within this deviation.

The values of (NL)Cs as determined by NAA are about 20% lower thanthose determined spectroscopically (Fig. 9). This trend is observed for alltime periods and may be due to a systematic error in the solution-handling orcounting processes or may be due to the different manner whereby (NL)i iscalculated when using NAA.

Generally, to calculate (NL)i, it is necessary to know the value of

fi, the mass fraction of element "i" in the specimen when the experimentstarts. This value is determined for a typical sample and not for each indi-vidual sample. However, when NAA is used, (NL)i is calculated from

42

A

50

N

o~40 0

20 00J

230 0UWw

AW

NI

~20

o 0 AA (NONACTIVATED)f 0 FE (NONACTIVATED)

10 A" AA (ACTIVATED)f NAA (ACTIVATED)

0 10 20 30 40TIME, days

Fig. 9. Normalized Cesium Loss fromActivated and NonactivatedSamples of PNL 76-68 Glassat 90 C

Ai W

(NL)i = (AX S (1)0

where Ai is the activity of isotope "i" in the leachate and AO is the originalradioactivity of this isotope in the specimen. W and SA are the specimenweight and geometric surface area, respectively. Instead of an average f,an accurately measured value for A0 and Ai are determined for each sample.This fundamental difference in the method of calculating (NL)i could resultin a systematic difference.

A similar set of experiments was done for two Savannah River Labo-ratory glasses, SRL 211 and SRL 211* (Table 10). For these tests, only theSRL 211* was irradiated since it contained the elements in concentrationssuitable for leach testing by activation. To make the SRL 211 glass serve asa control sample, it was leached without being activated. Both glasses wereleach-tested according to the MCC-1 procedure. The results are given inTable 11.

43

The initial values of (NL)Si and (NL)Cs obtained for the SRL glassesare similar to that observed with the PNL 76-68 glass. However, in the SRLglasses, the normalized releases of both elements becomes identical and con-stant after about 7 days of leaching. After 40 days, the releases are con-siderably lower than those observed for PNL 76-68 glass. To determine whetherneutron irradiation affects the leaching process for silicon and cesium, theresults for activated and nonactivated samples can be compared (Table 11).There is no significant difference ( 15%) in the normalized silicon and cesiumreleases of activated and nonactivated samples. This indicates that there wasno observable effect in SRL 211* glass due to neutron irradiation.

Experiments to detect neutron irradiation damage effects were alsodone for No. 7740 Pyrex glass. For this glass, only silicon release for acti-vated and nonactivated samples could be compared. The release of severalother trace elements present in Pyrex could be monitored with NAA, but not byany other methods of analysis. The test procedure varied somewhat from thatfor MCC-1 in that 200 beads (6 mm) were used to facilitate silicon detectionand the only sampling period used was 14 days. Silicon loss for the nonacti-vated samples was detected using ICP analyses and colorimetry and was detectedby colorimetry in the activated samples. The results (Table 11) contain un-certainties reaching factors of 3, but the data indicates that the siliconrelease of activated and nonactivated samples varies by no more than a factorof 2. On the basis of the uncertainty in these data, no large effect due toneutron irradiation is observed.

A description of the elemental and waste form utility of the method,together with associated sensitivity limits has been presented previously[STEINDLER]. With this report, the validation of NAA for the measurement ofleaching from glass waste forms is complete. NAA has been shown to be aviable method of detecting elements released from solid glass waste formsduring the leaching process. No damage due to neutron irradiation during theactivation process could be detected for silicon or cesium, and associatedneutron attenuation effects are minimized by proper control of the waste formdimensions. The method is able to simultaneously detect many elements thatare representative of a wide range of leaching behavior and at sensitivitylevels necessary to monitor the very low element release associated withnuclear waste forms.

The validation of NAA for crystalline waste forms (SYNROC) is con-tinuing and is discussed in Section Il.E, Service Role Studies.

D. Weathering Experiments

Weathering studies to measure the leach rates of weathered glass[STEINDLER] are continuing. The weathering of glass samples at 80 C and 90%relative humidity has been in progress for over 200 days. Significant newinformation concerning methods of detecting hydrogen in glass was obtained atthe Materials Research Society meeting in Boston (November 1980). Furtherdiscussions indicate that in-house secondary ion mass spectroscopy (SIMS)analysis of the hydration (weathered) layer should be possible and reliable.

44

E. Service-Role Studies

ANL has taken part in the round-robin sponsored by the MCC, which incor-porates 18 sites and is designed to establish the reliability and precisionof MCC-1 as the leach test procedure. Hopefully, each participant will alsocomment on how the procedure could be improved. The round-robin provides foreach laboratory doing a series of leach tests on at least two waste forms pro-vided by the MCC. ANL has tested PNL 76-68 glass and a Pomona basalt; thesolution analysis has been done by the Analytical Chemistry Laboratory at theChemical Engineering Division.

The matrix of leach tests have been completed, and the solutions analyzed.The test conditions and some auxiliary results are given in Table 12. Testsof up to 28-day duration were completed using deionized water, silicate water,and brine leachants. The glass leachates were analyzed for silicon, cesium,and strontium. The basalt leachates were analyzed for silicon, calcium,sodium, and iron. In addition, some of the initial solids were also analyzedalong with a standard leachate solution (SL-1) provided by the MCC. The so-lution analyses are given in Table 13.

Several aspects of the data are worth mentioning. Dual blanks of eachleachant were to be run. This was done. However, for only one blank of eachleachant was enough solution retained for analysis. The remaining threeblanks were repeated and analyzed. Analyses of the initial blanks showed thatsignificant quantities of cesium and sodium were present. The source of thiscontamination was the water used for AA analyses.

The present data for PNL 76-68 glass can be compared with the resultsreported above. The present data are about 15% lower than the previous data,but since the glasses were not from the same batch, exactly identical resultswere not expected.

The final report, including suggested changes in the procedure, will besent to the MCC before January 31, 1981.

Leach tests are also being conducted on three types of Westinghousealkoxide glass. These glasses are rich in aluminum. Supposedly, they havevery low leach rates. A full matrix of MCC-1 tests was initiated, and testsof up to 14-day duration have been completed. Initially, NAA and solutionanalysis will be done on only one 28-day sample for each glass type.

Weight loss data will be available for all samples. Weight loss datafor samples leached up to 14 days indicate these glasses to be about fivetimes more durable than PNL 76-68 glass; however, conclusions await completesample analysis.

The validation of NAA for crystalline waste forms is being done on SYNROC.These tests, in conjunction with Materials Science Division of Argonne NationalLaboratory, will indicate any advantages of using NAA on SYNROC and will doc-ument any neutron irradiation effects. A complete series of tests will bedone on a batch of SYNROC scheduled to be prepared in January 1981. InitialSYNROC leaching tests are in progress to establish what other analyticaltechniques may be used for comparison with NAA and to identify the major

45

Table 12. 1CC-1 Round Robin Leach Test Conditions and Results

Leaching Initial Surface Mass,Sample Time, Mass, A Mass Area, SA Final

Sample No. Leachant days g g x 10-4 cm2 g/m 2 pH

PNL 76-68 glass

Pomona Basalt

DIWa Blank

828384858687888990919293949596979899

100101102103104105106107108109110111112113114115116117

118119149

DIWaDIWDIWDIWDIWDIWDIWDIWSWbSWSWSWSWBC

BBBB

DIWDIWDIWDIWDIWDIWDIWDIWSWSWSWSWSWBBBBB

DIWDIWSW

37

1428.728.728.728.728.728282828282828282828

37

1428.728.728.728.728.728282828282828282828

282828

0.36360.34450.34410.34820.41010.38080.32590.34190.28380.30830.35330.34460.38820.32060.43200.35860.31690.3059

0.29180.35760.29960.31790.33860.34220.36770.22370.23180.25880.32560.32450.32600.35070.42640.31480.40960.2901

172540555653535045464446474652494649

192630292731292319202422191621161617

4.184.234.294.224.314.264.194.244.154.204.254.264.314.254.344.234.204.22

4.134.224.164.164.214.234.234.044.074.114.204.194.194.224.284.184.234.12

4.15.99.313.013.012.412.611.810.911.010.310.810.910.812.011.611.011.6

4.66.27.27.06.47.36.95.74.74.95.75.34.53.84.93.83.84.1

9.39.39.49.39.39.39.39.49.59.79.79.46.96.96.96.97.0

8.18.87.98.08.07.97.89.49.49.29.29.25.85.75.85.75.7

4.154.069.01

aDeionized water, initial pH - 6.3.

bSilicate water, initial pH - 7.8.

CBrine, initial pH - 6.5.

matrix components leached. Two 28-day tests, by MCC procedures, have been com-pleted and the leachates have been submitted for analysis by various methods atthe Chemical Engineering Division Analytical Chemistry Laboratory and by dcplasma spectrometry at Trace Elements, Inc., Park Ridge, IL. Two samples havebeen sent for irradiation and subsequent NAA analysis.

Table 13. Leachate Analyses for the MCC-1 Round Robin

Sample Si ,Sample No. mg/L

Cs,

(NL)i mg/LSr,

(NL)Cs mg /LCa,

(NL)Sr mg/LNa,

(NL)Ca mg/LFe ,

(NL)Na n¬/L

PNL 76-68 glass

Pomona Basalt

DIW BlankSW BlankB BlankMCC Blank

2.83.76.1

0.220.841.393.433.333.333.173.263.262.852.672.62

0.170.180.180.0770.0770.0780.0570.0490.049

3.312.921.352.751.251.248.850.250.243.841.140.3

5.45.65.62.42.42.41.71.51.5

(NL)Fe

12.318.724.135.934.436.630.131.131.322.323.122.9

3.87.08.9

12.212.311.98.98.58.75.55.75.6

6.39.6

12.418.517.218.915.516.016.111.511.911.8

1.62.93.85.15.25.03.73.63.72.32.42.4

828384878889929394979899

100101102105106107110111112115116117

118150152SL-1

0.50.671.10

C'

47

III. BRITTLE FRACTURE STUDIES(L. J. Jardine, W. J. Mecham,and R. H. Pelto)

A. Introduction

To demonstrate safe methods of disposal of radioactive wastes, a propertyof materials that must be known is their resistance to environmental disper-sion. Part of this effort is developing a methodology for characterizing thefracture behavior of representative brittle waste materials under conditionsof mechanical impacts that could occur in the event of accidents in thehandling and transportation of such waste forms. The impact-fracture resultsare to be characterized in terms of (1) surface area and (2) the amount ofparticles of respirable size. These measures of potential dispersibility fromimpacts are to be related to other properties of the materials and to theconditions of impact.

Measurements of fracture particulates formed in impact tests of smallspecimens are being used to provide data to establish quantitative correla-tions of the fracture properties of the materials with the conditions of im-pact. Such data are useful both for evaluating alternative materials and forpredicting the consequences of accidents in realistic situations. Becausetheories of brittle fracture do not yield quantitative predictions, the va-lidity of such correlations must be verified by specific empirical data. Themethodology is being developed to accomplish this iii an efficient, reliable,and economical way--that is, by correlating (1) simplified models of dynamicstress distribution and (2) small-particle statistics with reproducible dataobtained in standardized measurements of fracture particulates formed underknown conditions of impact.

Crack propagation is nearly unpredictable in detail, because it is acatastrophic rate process accompanied by complex patterns of reflected stresswaves [BRADT, AVERBACH]. However, in our studies, a useful model for approx-imating the results of impact fracture has been developed by combining threeengineering methodologies: dimensional analysis applied to dynamic stressdistributions [LANGHAAR], the two-parameter lognormal probability functionapplied to the size distribution of fracture particulates [HERDAN], and thestatistical correlation of the surface area of the fracture particulate withthe input energy dissipated in the fracture process [ZELENY]. The combinedbrittle-fracture model identifies and relates material properties, geometricalparameters, and impact energy input in such a way that the surface area andsize distribution of the fracture particulate are defined for a wide range ofbrittle waste forms and impact conditions.

B. Status of Results Previously Reported on Measurements of FractureParticulates from Impacted Brittle Materials

When plotted on lognormal graph paper, the particle-size distributionsof fracture particulates in our work always follow a straight line for allparticle sizes smaller than a certain upper-limit size, DL, which is about5 mm for a typical glass specimen. These graphical size distributions areplots of the mass fraction of particles smaller than size D, as determinedby sieving and by Coulter counter analyses measured for particles of any size

48

down to several micrometers. These cumulative mass fractions are equal tocumulative volume fractions for materials of a given density, and the coor-dinates of the graph are marked so that lognormal size distributions plot asstraight lines. (Examples of such graphs and the mathematical definitionsare given in Section III.C. below.) Such straight-line plots define the twoparameters of the lognormal probability function: the geometric mean sizeof the lognormal size distribution, Dg, and the geometric standard deviation,ag. From data obtained thus far [STEINDLER], the fracture particulates canbe characterized by Dg and 0g for all impact tests over a wide range of im-pact velocities and energies, including drop-weight* and free-fallT tests ofbare and canistered specimens having diameters from about 1 to 60 cm. Thelognormal-distribution particulate described by Dg, Gg, and DL includes therespirable particles (D less than 10 pm) and typically includes more than 90%of the total fracture surface area.

The mathematical form of the lognormal size distribution is such that Dgand 0g describe not only the cumulative volume fraction of material as a func-tion of size, Pv(D), but also the cumulative surface fraction, Ps(D) and thecumulative number fraction Pn(D). Thus, sieving data provides the parametersfor a complete geometric description of the fracture particulate. The meansurface/volume shape factor, a , is defined by the ratio of surface area tovolume of the fracture particulate.

Absolute surface areas of fracture particulates can be measured repro-ducibly by the BET gas-adsorption technique. This surface measurement wasused by Piret and associates [ZELENY] in impact-calorimetry tests to establishthe surface fracture strength property, Yf, for Pyrex and crystalline quartz.In the final phase of this work at the University of Minnesota [ZELENY], yfwas found to be nearly constant at 78 J/m2 of fracture surface area, indepen-dent of input energy, Wi, over a range of 0.6 to 12 J/cm3 of material. Ourpreliminary tests have shown results consistent with Piret's results. Thissurface/energy correlation has its physical basis in the conversion of kineticenergy into stored work of elastic compression in the material, followed bythe dissipation of this energy into heat when the brittle material fracturesinto a particulate under the high stresses; there appear to be multiple stagesof three-dimensional binary cleavage.

In the preceding quarterly report [STEINDLER], size distributions of thefracture particulates of seven different materials were analyzed lognormally,and relative fracture-surface areas (for shape factors assumed to be equal)were used to calculate relative impact strengths (based on the Yf for Pyrexand crystalline quartz). Characteristic values of Dg, ag, and DL were found.

Work done in the present report period was as follows: (1) correlationsof lognormal particle-size parameters with input energy density were developedfurther, (2) grain sizes of materials previously impacted were measured and

*In a drop-weight test, a weight is dropped onto the sample from a

selected height, giving essentially a two-sided impact.

tIn a free-fall test, a test specimen is dropped from a selected height

onto a steel plate resting on a concrete floor, giving a one-sided impact.

49

were compared with particle size distributions, (3) microscopic measurementswere initiated to obtain further statistical data on particle shape factors,(4) work has been continued on the calibration of the BET surface area mea-surement, and (5) free-fall impact tests of bare and canistered specimenswere initiated to clarify time-dependent behavior. Results are describedbelow.

C. Correlation of Lognormal Particle Size Parameters with Input EnergyDensity in Drop-Weight Experiments

In [STEINDLER], particle size distributions were plotted for the frac-ture particulates from drop-weight side impacts of 38-mm-diameter Pyrex glasscylinders at energy densities of 2.5, 1.2, 0.42, and 0.21 J/cm3. In thegraphical analyses used, the masses measured for the various size incrementsseparated by sieving were normalized into mass fractions by dividing by themass (i.e., volume) of the original specimen. That is, if the cumulativemass of particles smaller than sieve size D is M(D), then the cumulative mass

fractions are given by MN), where Mo is the original mass of specimen. Of

course, the cumulative volume fractions are equivalent to mass fractions:V(D) = - .M(D) By extrapolation of the straight line of the plot of V(D)

Vo V0the lognormal parameters, Dg and ag, are determined. In general, the mathe-matical description of the fracture particulate up to size DL is defined by

Pv (D < DL), where Pv is the cumulative lognormal probability function and,

by definition, Pv (D; Dg, og) = V(D), where V. is the volume of a "complete"

lognormal distribution, that is, for 0 < D < w. There is no actual physicalmass available to be measured as the normalization volume, V. Note that fora given empirical V(D), the values of Dg and ag will depend on V.. In prac-tice, fracture particulates are lognormal only for particle sizes smaller thansome upper limit size, DL.

As described above, the normalization volume, V,, was assumed in ourstudies to be equal to Vo, the original total volume of the impacted body.This normalization is our preferred and reference method. As described in[STEINDLER] in relation to edge and corner impacts, a possbile interpretationfor the sole purposes of correlation is to arbitrarily define V. = 2V(8 mm)--that is, as twice the volume of all material passing through an 8-mm sieveopening. A discussion is presented below based on the normalization, but nofurther use of it is anticipated at this time.

When this alternative and arbitrary normalization was applied to the re-sults of side-impact experiments in which Pyrex was impacted at four differentenergies, the particle size distribution obtained and lognormal parameters arethose shown in Fig. 10 for Pyrex impacted at four different energy densities.(The original volume, Vo, was uced in calculating energy density.) The cumu-lative volume fraction is the ordinate in Fig. 10, Pv(D) = V(D)/2V(8 mm). Eachstraight line represents lognormally size particulate for the given D and 0g.The respirable fraction (particles smaller than 10 um) are shown as Vl0 im)V0.In this normalization, the ratio V(8 mm)/Vo is a parameter describing impacteffects.

The impact parameters tabulated in Fig. 10 arr lotted against energydensity Wi/Vo in Fig. 11, where Wi is the input energy of impact and Vo is theoriginal volume of the impact d body.

50

W

J 97

w 90

50000

V 10Ez

10~-o> 10 - 03-

S 10-5-

103 1 3 10-2 10,

SIEVE SIZE, D, m

Symbol Glass Wi/V0 (J/cm3) V(8mm) D (m)cr V(IOpm)

o Pyrex 2.5 0.72 0.013 8.7 56 x10-5o Pyrex 1.24 0.50 0.014 7.8 14 x 10 5

V Pyrex 0.43 0.28 0.016 7.3 4 x 10-5A Pyrex 0.2 I 0.22 0.021 7.0 1.3 x 10-5

Fig. 10. Pyrex Fracture Particulates as a Functionof Energy Density in Drop-Weight Experi-ments, Using an Arbitrary Normalization

The error bands shown for the data points plotted in Fig. 11 are estimatedapparent variabilities of the values of Dg and ag based on various attemptsto draw the "best straight line by eye." More systematic error analysis byan approach described in Section III.D is planned.

As is evident from Fig. 11, the lognormal parameters Dg and ag are eachnearly constant (independent of energy density) over the range of measuredenergy densities; the values of V(8 mm)/Vo are approximately directly pro-portional to energy density. As observed in our previous work, for a givenmaterial and impact configuration, ag tends to be constant over a range ofenergy densities, giving rise to a family of straight lines having the sameslope. As the energy input increases in tests with the same specimen volume,that fraction of the volume stressed above the threshold for particulate frac-ture increases, resulting in larger values of V(8 mm). These parameters andthe shape factor, ag, determine the total surface area of the particulate.No direct measurements of surface area were made in these particular tests,but such measurements are discussed below in Section III.E.

The values of the respirable fraction, V(10 m)/V 0 , obtained from datareported in Fig. 10, are plotted in Fig. 12. On this log-log plot, the values

of V(10 m)/Vo form an approximately straight-line function of energy density

51

1.0

o - o x 101

S 0.8-

> 0.6-

- - V(8mm)/Vo

b 0.4-

E D (m) x 10S02 9

o _p p--

00 0.5 1 .0 1.5 2.0 2.5 3.0

ENERGY DENSITY, Wi/Wo, J/cm3

Fig. 11. Particle Parameters as a Function of Energy

Density Based on the Arbitrary 2V(8 nun)

Normalization. Drop-weight side impacts of

Pyrex cylinders, 3.8-cm diameter by 6.4-cm

leng th . An energ y density of 1 J/cm3 is

equivalent to impact at 30 m/s or to free-

fall from a distance of 45 m.

Wi/Vo. In these tests, no Coulter counter measurements were made, and the

amount of 10-um particles was obtained by extrapolation of the sieving data.

However, the general validity of such extrapolations has been verified in other

tests at this site. Further tests, such as those in Section D, are in progress

to measure the fragments to sizes as small as 10 um.

From the above data, correlations of particle parameters for drop-weight

side impacts of Pyrex cylinders can be expressed as follows:

Dg = 0.016 m

ag = 7.-7

V(8 MM 2.5V + 0.17(0 )0

( 1.47

E D (m)- 13 x10 O

0 Vo

where 0.2 < < 2.5, .SCm

52

o

0

z0I-V

LL.

W-J

a.U)W

10~2

10- 4 h

0.1 10

ENERGY DENSITY, Wi/Vo, J/cm3

V(0an)Fig. 12. Respirable Fraction V( )

v

as a Function of Energy

Density for Pyrex Side

Impacts.

Preliminary values of the same parameters have been obtained for initialdrop-weight (DW) edge and corner impacts and for free-fall (FF) side impactsof Pyrex specimens. These initial test data were plotted, using the arbitrary8-mm normalization. From these plots, the particle parameters are summarizedas follows:

Impact TestWi J

Vo ' cm 3

V(8 mm)V0

Dg, m V(10 um)

V0

DW 900 cone 0.42 0.014 0.017 10.3 2 x 10-5

DW corner slice 0.42 0.022 0.066 11.0 0.6 x 10-5

DW cyl side 0.42 0.28 0.016 7.3 4 x 10-5

FF cyl side 0.42 0.14 0.016 7.3 1.4 x 10-5

For the side impacts, values of Dg and ag are similar, but lower values ofV(8 tea)/Vo and V(10 um)/Vo indicate that fracture efficiency was lower for

0

I

53

the free-fall test. A possible explanation of this effect is presented inSection III.G below. Further tests are required to establish definitive cor-relations of particle size parameters with energy density for free-fall testsand for corner impacts.

D. Effect of Grain Size on Size Distribution of Fracture Particulates

Particle size distributions of fracture particulates for seven differentmaterials were presented in [STEINDLER]. These materials were Pyrex, fusedquartz, crystalline quartz, a "machinable" glass ceramic ("MACOR"), microcrys-talline U0 2 (reactor-grade ceramic), and two natural minerals, (1) sandstone(a conglomerate) and (2) nepheline syenite (a coarse-grained aluminosilicate).Coulter counter measurements were made down to particle sizes of about 5 um;the approximate lognormality of all particles smaller than 100 pm was shown.During the present report period, grain sizes were determined for U02, neph-eline syenite, and sandstone. These grain sizes are compared in more detailwith measured particle sizes .

Sandstone is a conglomerate of hard silica grains weakly cemented togetherwith a second phase. The grain size was determined by sieve analysis afterthe sandstone was disaggregated by immersion in a 1% sulfuric acid solutionin an ultrasonic bath. In Fig. 13, the lognormal plot of grain size is super-imposed on the size distribution of the fracture particulate resulting from adrop-weight side impact of two cylindrical specimens (both with a 22.1-mm dia-meter; 44.9- and 17.6-mm lengths). Evidently, a lognormal distribution ofgrain size (mass median, Dg = 200 pm) occurs along with a lognormal fractureparticulate which has an upper limit size DL = 120 pm. This bimodal lognormalpattern is consistent with the conglomerate nature of sandstone.

% 99.9V W/V 0= 1.2 J/cm3 V(8mm)/V=0.30Z 99.0-0

U GRAIN SIZE490.0a:

Ud

50.0-

S10.0 - - FRACTURE PARTICULATE

E

N 1.0

-5 -04 I-3 1-2 10-10-S 10 10 10,D

MESH SIZE, D, m

Fig. 13. Particle Size Distribution of Sandstone Side-Impacted at1.2 J/cm3 . Collective sample from two specimens

54

Grain sizes of both U02 and nepheline syenite were determined bymetallographic analyses. Samples of U0 2 were prepared* by mounting in aglass-reinforced epoxy resin, rough polishing with a 600-grit silicon carbidepaper and fine polishing on a 1-pm diamond wheel. Samples were etched for twominutes in dilute acid. Specimens were photographed* at 200X and 50OX magni-fication. The intercept method [HILLIARD] was used to estimate grain size.This method uses a circle overlay and involves counting the number of grainintersections over the circumferential distance. An average grain size isdetermined that is reproducible, but there may be a slight systematic under-estimate of grain size. The nepheline syenite grain sizes were determined ina similar way, but with a magnification of 6.5X.

The mean grain sizes so determined were 705 100 jm for nepheline syenitand 8.8 2.5 Pm for U0 2. These grain sizes are shown in the graphs of com-plete particle size distribution in Figs. 14 and 15. In both figures, normal-ization of the particle size data was based on total specimen volume, V0. Inthe case of nepheline syenite, the original specimen was a cylinder (31.9-mmdiameter by 39 mm) obtained by core drilling. The U02 specimen consisted ofthree cylindrical pellets (each 13.7-mm diameter by 13.6 mm). Overall impactenergy density was 1.2 J/cm3 for both specimens.

In Fig. 14, the mean grain size of 705 im corresponds to an inflectionpoint of the two straight-line sections of the particle-size-distribution

97

oc 90--j 80-a 70 O DATA POINT OBTAINED BY SIEVE ANALYSIS

t 50 O DATA POINT OBTAINED BY COULTER COUNTER40 ANALYSIS

J 30- AVERAGE GRAIN0 20-SIZE705pm

W 10--

9w 5a 2

00 Iz 0.5--

z= a 0.2w' 0.1

a 0.01

0.001- .

0.0001

I 10 100 1000 10,000 100,000PARTICLE DIAMETER, pm

Fig. 14. Particle Size Distribution of Ne heline Syenite Rock,Drop-Weight-Impacted at 1.2 J/cm3

*Preparation and photography were done by C. Steves of the Materials

Science Division.

55

97-

W 90-J 80 - DATA POINT OBTAINED BY SIEVE ANALYSIS

60 -0 DATA POINT OBTAINED BY COULTER COUNTER50 ANALYSIS

w 40 -W 30--Si 20-

W 1o0

52

oc |

0.5--z a 0.2w' 0.1

IL 0.01 --

N 0.001

0.0001

I 10 100 1000 10,000PARTICLE DIAMETER, jim

Fig. 15. Particle Size Distribution of U02, Drop-Weight-Impactedat 1.2 J/cm3 . Collective sample from three pellets.

curve. The deviation of the 5 and 6 um points from the straight line is anartifact of the Coulter counter method and the pulse electronics. These twoas plotted should not be included in the fits of the data. Such a bimodalsize distribution indicates that fracture at the grain boundaries is somewhateasier than across the grains, but the grain-boundary fracture is much lesspronounced than the case with sandstone.

In Fig. 15, the mean grain size of U02 (8.8 um) is within the particlesize range analyzed by the Coulter counter. From the plotted points, it ispossible to construct an inflection point at about the 9-jm grain size. How-ever, a noncrystalline specimen, fused quartz (Fig. 16), also shows such aninflection. Probably, the slight inflection is not due to crystal grain size,but to the insensitivity of the Coulter counter to particles smaller than about5 jm and the fact that this is a cumulative plot. To allow comparison the log-normal particle-size distribution for crystalline quartz is shown in Fig. 17.These specimens were each cylinders- of about 30 mm diameter by 70 mm.

Quantitative evaluations of the "goodness of fit" of lognormal parametersare needed to make more detailed characterization. A computer-assisted linearregression analysis to accomplish this is being developed that uses an approachdescribed later in this report section.

From tests made thus far (e. g., crystalline quartz and fused silica), itappears that there is no significant difference in the fracture behavior ofvitreous and crystalline forms of the same or similar material. Both sand-stone and nepheline syenite have gross lattice discontinuities at the grainboundaries, whereas U02 ceramic does not. Further tests are planned that areto be performed on samples of simulated waste forms (glass and crystalline)requested from several laboratories. The tests described above indicate thatthe general methods being developed are capable of characterizing diversematerials.

56

97 ~ ,*'ii ---- ui

w 90 - DATA POINT OBTAINED BY SIEVE ANALYSIS-J 80-

70 -0 DATA POINT OBTAINED BY COULTER COUNTER60 ANALYSIS

Co 50-Wc, 40-0 -30--

20

10-5

4

00 I AVERAGE GRAIN

i z 0.2 SIZE 8.8 pm-zc4 0.2-W 0.1

0.001-

Co 0.001-

S0.0001-

I 10 100 1000 10,000PARTICLE DIAMETER, pm

Fig. 16. Particle Size Distribution of Fused Quartz,Drop-Weight-Impacted at 1.2 J/cm3

97

ac 90-w 80 - DATA POINT OBTAINED BY SIEVE ANALYSIS

4 70 0 DATA POINT OBTAINED BY COULTER COUNTER

o50 ANALYSISWu 40--1. 30-0cs W 20-

1- 10-5-2-

00 IZ z 0.5-Z 4 0.2W J 0.1

W 0.01a. - -N 0.001

0.0001-

I 10 100 1000 10,000

PARTICLE DIAMETER, pm

Fig. 17. Particle Size Distribution of Crystalline Quartz,Drop-Weight-Impacted at 1.20 J/cm3

57

Ordinary graphic analysis provides no measures of errors or of the "good-ness of fit." However, linear regression (analysis of variance) can readilybe applied to empirical data plotted in straight lines. The cumulative volumefraction of the lognormally distributed particulate, PV(D), as defined abovein terms of empirical values of the cumulative volume function V(D) obtainedby sieving, can be related to the standard cumulative normal probabilityfunction, P(u):

V(D) = Pv(D) = P(u),

where u is the standarized normal variate, here defined as

logD log Dgu = loo - gglogag logag g

If we define X as a function of D: X = log D, then the linear relation be-tween cumulative mass fraction and size D is given by u = AX + B, where

1 log DgA = and B =- .

logoa logoag g

Computer programs are being developed for linear regression and graphicalpresentation of sieving data.

E. Microscopic Measurements of Particle Size to DetermineShape Factors

A contract has been placed with North Carolina State University (NCSU)for the determination of particle sizes and shapes of fracture particulatesby computer-based, statistical stereometric methods, using a scanning electronmicroscope and an ellipsoidal model of fragment particle shape. A report onthese micromeasurements and the statistical shape factors is expected duringthe next report period, when these measurements will be compared with oursieving data and surface area measurements on lognormal fracture particulates.

Spherical particles have volumes and surfaces unambiguously representedby linear size, D, which is the single dimension: diameter. The volume, Vi,

and surface area, Si, of a single spherical particle are given by Vi = Di

2 6and Si = wDi, and the surface/volume shape factor is Si/Vi -D . In general,

however, irregular particles formed by brittle fractu*.e must have their geo-metric relations defined by two shape factors, av anv as. Thus, by defini-tion,

i aV - 3a D;S D ai v i iaDs i'Vi aD. (1)

i v i

58

The surface/volume shape factor, as/av, can be defined for the BET surfacearea, S(D), and the cumulative volume, V(D), of a particulate of upper-limitsize D. In terms previously defined for a lognormal particulate,

DV(D) =Za D3 = V Pv(D) (2)

v i Go v

DS(D) = asD.= S PS(D) (3)

The normalization quantities, Vi,, and S, are related by the shape factors andby the moments of the lognormal probability function. That is,

Sm_ as 1 a 0.5 In og (4)V- a D gv g

a

As described in [STEINDLER], mean values of a lognormal shape factor, ag = a,v

can be found for the range of empirical values of Dg and ag.

Preliminary results from both the microscopic measurements at NCSU andour BET and sieving measurements indicate that the mean shape factor, ag, isconstant, i.e., independent of size, for Pyrex fracture particulate in thelognormal size range. Shape factor evaluations are continuing.

F. Calibration of BET Measurements and Sieves

Work has continued on the calibration of BET surface area and sievingmeasurements [STEINDLER]. Absolute standards of several types have been ob-tained, including glass microspheres of an approximate lognormal size distri-bution and also monosize spheres. In addition, our measurements of fractureparticulates are being checked against those of other laboratories.

As described in Section E, statistical analysis of microscopic particlemeasurements together with calibrated measurements of surface areas is expectedto establish the shape factors and size parameters of fracture particulates.Particle-size distributions determined by sieving are described as functionsof D, the nominal size of the sieve opening. Sieve sizes can be determinedboth by direct microscopic measurements and by weighing or measuring the sizesof sphere that just pass through or "stick" in a particular sieve frame. Theshape factor, av (defined in the previous section), can be determined by thesame method [HERDAN, LESCHONSKI]. This work is continuing.

G. Preliminary Consideration of Free-Fall Tests

As described in [STEINDLER], fracture particulates of Pyrex specimensimpacted in preliminary free-fall tests show generally similar lognormal sizedistribution, but with less extensive fracture than do particulates fromdrop-weight tests with the same input energy.

59

In the present period, 38-mm-dia cylindrical Pyrex specimens with onehemispheric end (to receive the impact) were prepared for free fall tests.This hemispheric shape makes results less dependent on precise impact orien-tation, which is difficult to achieve in free-fall tests and drop-weight im-pacts.

Annealed specimens of about 146 g each were impacted in the drop-weightdevice, using the 10-kg deceleration mass (i.e., falling weight) with energyinputs equivalent to 18.2-m (60-ft) and 27.4-m (90-ft) free-fall impacts. Theparticle size distributions agree reasonably well with those for drop-weightside impacts of Pyrex cylinders shown in Fig. 10.

In two free-fall tests of this specimen from the 18.2-m height, this typeof specimen bounced about 4.6 m, remaining unfractured except for a slight(less than 1 g) spallation fracture at the contact point. The calculatedmaximum elastic stress in this test was well above the fracture threshold,but the calculated time to achieve maximum stress (and to start the rebound)is only about 40 is. Since the glass is subjected to stresses above thresholdvalue for less than 40 us, it is believed that the specimen failed to fracturebecause of the short time available for cracks to initiate and propagatethrough the specimen. (The equivalent calculated stress duration for thedrop-weight test was 240 us.) This difference in time may be critical, basedon reported rates of crack propagation [DOREMUS, AVERBACH].

Our simplified dynamic stress analysis indicates that the above compres-sion time is proportional to the square root of the deceleration mass. There-fore, the time-dependent fracture behavior is such that there may be a minimumsize (mass) for meaningful free-fall tests specimens. Preparations are beingmade for free-fall tests of 1-kg canistered specimens of waste glass. Largerspecimens will also be required in order to establish the scaling laws forimpact fracture. Scaling laws are well established for elastic and plasticdeformation [LANGHAAR].

A preliminary test of a canistered glass specimen was made in the drop-weight device. An annealed Pyrex cylinder of 38 mm was used; the canisterwas staiiuiess steel type 304 of 1.4-mm wall thickness. The canister hada welded bottom of 0.132-mm thickness but no top. The glass cylinders fitrather loosely into the canister with a 0.30-mm clearance. The specimen wasimpacted on the side with an energy of 180 J. The particle size distributionwas compared with that for a similar bare specimen. Both sieving and Coultercounter measurements were made so that size fractions were measured down toabout 5 in. The particle size distributions for the bare and canistered spec-imens agreed closely, implying that little impact energy was absorbed by thecanister material. Additional tests are planned.

60

IV. FLUIDS IN ROCK(M. G. Seitz, D. L. Bowers,*and J. Dunningt)

A. Development of Dynamic Stream Simulation Methodology

The work reported here included the evaluation of approaches for ground-water stream experiments to simulate a hydraulic breach of a nuclear wasterepository located in bedded salt. This work is in support of the Waste Iso-lation Pilot Plant (WIPP) program conducted at the Sandia National Laboratory.The work involved the completion of an experiment and the analyses of mate-rials in a study of the penetration of water into rock salt.

1. Penetration of Rock Silt by Water

An experiment was set up to investigate the penetration of waterinto rock salt. Because of the low permeability exhibited by the salt inprevious experiments [STEINDLER], we used in this experiment tritiated waterthat that could be monitored radiochemicaily to provide us with a sensitivemeasure of the penetration of water into the salt.

An experiment consisted of placing four cylindrical McNutt rocksalt cores (8.89 cm long and 2.20-cm dia.) in core holders having a designschematically shown in Fig. 18. These cores and holders were assembled, andpressurized water (14 MPa, 2000 psi) was applied to the outsides of the Tefloncore sleeves with a Haskel MS 188 pump. This pressure was used to confinethe Teflon sleeve around the sides of each core, making a nearly perfect seal.Two milliliters of a solution consisting of 80% solution saturated with rocksalt from the McNutt formation, 20% distilled water, and 40 1Ci tritium as

VALVE

MONEL400 4 2.

SPACER T4 AND

McNUTT ARGONROCKULTRA-S A LTPURE

SLEEVE-- ---

EPRE SSURE SL EEVE -CONFINING WA TE R

T RANSDUCE R

AI - DR IVE NHYDRAULIC

PUMP

BLICKMAN HOOD

Fig. 18. Schematic of the ExperimentalSetup to Study Water Penetra-tion into McNutt Salt Cores

*Member of the Analytical Group of the Chemical Engineering Division.

tConsultant, Indiana University.

61

water was added to the top of each core to detect the penetration of water.The atmosphere above the solution was purged with argon, and argon was appliedthere at a pressure of 760 kPa (110 psi). The experiment was performed at anaverage ambient temperature of 21*C. The four cores were pressurized for du-rations of 31, 61, 103, and 137 days.

The sleeve confining pressure and the argon pressure were monitoredby pressure transducers and were recorded on a strip chart for the 137-dayduration of the experiment. During the experiment, the argon pressure abovethe cores was either stable or varied slowly [ranging from 710 kPa (103 psi)to 785 kPa (114 psi)].

The sleeve confining pressure varied considerably during the exper-iment, as depicted in Fig. 19. The pressure oscillated at first, with aperiod of 20 to 30 h. No correlation was found between the confining pressureand (1) ambient temperature, (2) time of day, or (3) the line air pressure thatmaintained the confining pressure. After the second core was removed (after61 days) from the pressurized lines, the confining pressure became consider-ably more stable, suggesting that the pressure oscillations were caused by apeculiarity of the second core holder.

A rock core was removed from the pressurized lines after isolatinga core holder with valves, releasing the argon pressure, then removing thetritiated water from on top of the rock core. The sleeve confining pressurewas then decreased, and the core was removed from the core holder. The bottomexit tube was inspected for water, then rinsed to produce a sample for tritiumanalyses by liquid scintillation counting. Tritium detected in the rinses of

25

a -OSCILLATIONS HAVE2 A 20-TO-30 HOURj 20 PERIOD THIRD

-FIRST SECOND CORECORE CORE REMOVED

a. REMOVED REMOVED FOURTH

Z REMOVED_

0

zz

> 5WWJ

0 20 40 60 80 100 120 140

TIME, days

Fig. 19. Depiction of the Sleeve Confining Pressure duringthe Water-Penetration Experiment. (Water infil-tration of rock salt.)

62

the first two rock cores was attributed to solution that had not been com-pletely removed from the top of the rock core and had passed along the out-side of the core after the confining pressure was lowered. The bottom of thelast two rock cores was rinsed before the confining pressure was removed, andthese rinses did not contain detectable tritium.

The rock cores were sectioned, using a diamond saw with Freoncoolant. Sectioning of all cores (Table 14; lowest segment number, 1, was atthe top of the core) was very similar to that reported last quarter for thefirst core [STEINDLER]. Then the sections were dissolved in water and ali-quots of the solution were analyzed for tritium by scintillation detection.The tritium activity per gram of rock for the various rock segments is givenin Table 14. The analyses are plotted in Fig. 20 against depth of segment inthe core.

The results indicate that although the water had penetrated to onlya depth of about 1 cm in the first 31 days, the water completely penetratedthe 8.85-cm-long core after 61 days. Nonetheless, the concentrations of tri-tiated water are always less at the bottom than at the top of the cores, andthe gradient that was established in 61 days in core No. 2 remained fixedthereafter, as indicated by similar tritium distributions in cores 3 and 4.

I 2 3 4 5DEPTH WITHIN THE

6CORE, cm

7 8 9

Fig. 20. Specific Activity of the Segments of theRock Core Relative to their Depths inthe Core

-Q-4 CORE I'-0-4 CORE 2'-a-CORE 3

CORE 4

STEADY STATE

FIRST CORE

aC

00

0

104E

0ccw

0 10-}H-

S101LUW.

0

63

Weights, Activities, andFour Rock Cores

Lengths of Segments from the

Core Segment Segment Activity SegmentNo. No. Weight, g d/(min-g) Length, cm

11111

22222222

33333333

444444444

la2a3a4a5b

la2a3a4a5b6b

7b8b

la2a

3a4a56b7b8b

la2a

3a

4a

5

6

7

8

9

1.242.334.103.977.39

1.412.523.914.119.758.80

11.4317.51

1.492.392.694.34

15.547.33

10.3614.18

1.581.973.217.034.716.287.499.41

12.89

1.4 x 1042.0 x 103

<10<10<10

6.42.11.71.32.2

1.26.31.76.0

x.

x.

x.

x.

x

510430360

104

103

103

103

10 3

0.300.561.000.960.89

0.340.610.940.991.181.061.382.11

0.360.580.651.051.880.881.251.71

0.380.480.771.700.580.760.901.141.56

x 105x 103x 104x 103

51016026

145

2.2 x 1047.1 x 1033.4 x 1031.8 x 1031.3 x 103

670410310320

aThese segments each consist of one-half of the circular shapeof the cylinder. Other segments consisted of cylinders sampledalong the length of a rock core.

bThese segments of the cores were sectioned and analyzed atthe same time (after 137 days) as the segments of core No. 4.

These results suggest that after 61 days, a steady-state transfer process isestablished in which water moves from the reservoir at the top of the rockcore down through the core to the bottom, where it may evaporate.

Table 14.

64

Using this interpretation, we fitted the tritium data in Fig. 20with a solution to Fick's law of diffusio%. For steady state conditions,with a source of water at one end of the core and a sink at the other (evap-oration), the concentration of water will vary linearly along the length ofthe core. This distribution, fitted to the data in the center of the rockcore (2 to 6 cm depth), is plotted as a continuous line in Fig. 20. (Alsodrawn is the steeper curve of tritium distribution after 31 days.) Thesteady state distribution is consistent with the data from segments of cores2, 3, and 4, excepting the top-surface segments. A probable cause of thehigh tritium concentrations of the top segments is recrystallization, inwhich tritium is incorporated as the hydrated fraction of the newly formedminerals. This recrystallization effect would not be represented in a tritumdistribution expected from diffusion alone.

The results shown in Fig. 20 are not consistent with water perme-ation caused by a hydraulic gradient imposed on the rock cores because if thiswere the case, we would expect that the water would saturate the cores uni-formly from top to bottom.

From the steady-state diffusion curve, we estimate that the maximumconcentration of water subject to diffusion in rock salt in contact with wateris 7.5 x 10-5 g of migrating water per gram of rock salt or 75 ppm. This es-timate was obtained by using the intercept of (1) the postulated steady-statediffusion curve [of 1.4 x 103 tritium distegrations/(min-g)], and (2) the ac-tivity of the brine [of 1.89 x 107 d/(min-g)]. The activity of the brine wascalculated from its activity by volume [of 2.27 x 107 d/(min-mL)], using anestimated density for the brine of 1.2 g/mL.

The average concentrations of water incorporated by recrystalliza-tion at the top of the rock cores were estimated (by a calculation similar tothat described above) to be 0.07, 0.33, 0.63, and 0.11% by weight for cores1, 2, 3, and 4, respectively. We would expect that most of the recrystallizedwater would occur very near the surface of the rock cores and would have aconcentration at the surface greater than the calculated average values forthe whole segments.

The amount of water in recrystallized minerals appears to be readilymodified. Prior to sectioning and analyses, rock core No. 4 was exposed tolighting for about 10 min for photographs. The low amount of recrystallizedwater on the surface of this core in comparison to cores 2 and 3 may have beendue to evaporation losses.

Considering further the diffusion-controlled process, we can esti-mate the diffusion rate of water in the rock salt from the tritium profileseen in core 1 (Fig. 20). The amount of water entering the rock cores wouldincrease with additional time, then remain constant at steady state conditions.

At steady state conditions, the amount of water, Ma,, entering a coreis given by the formula

Mm = 1/2 1Co

65

where Co is the water concentration at the upper surface and 1 is the lengthof the rock core [CRANK, p. 50]. Based on the data reported here,Co = 7.5 x 10-5 g/g of core and 1 = 8.89 cm, and so

M = 3.33 x 10-4 cm00

or 7.27 x 10-4 g/(cm2 of rock core surface), using 2.18 g/cm3 as the densityof rock salt. The diffusion rate of water through the rock salt can be esti-mated from the amount of water that penetrated rock core No. 1 in the first31 days of the experiment. The amount of water entering each square centimeterof the top of rock core No. 1 in 31 days is given approximately by

M31 days = 2 Co(W1+2)/A

where Co is the maximum concentration of migrating water (equal to7.50 x 10-5 g/g at the top surfaces of the rock cores), W1+2 is the weight ofthe two top segments of rock core No. 1, and A is the cross-sectional area ofthe core (equal to 3.80 cm2). From this formula,

M31 days = 1.41 x 10 g water/cm2

The ratio of the amount of water that entered the core after 31 days to theamount that had entered at steady state is

M /M = 0.194.31 days

This ratio can be used to calculate the diffusion coefficient for water inthe rock salt core. From [CRANK],

M 8= 1 - 8 1 2 exp -D(2n + 1)2 ,2t/12

W 2 n=0 (2n + 1)

where D is the diffusion coefficient, t is the duration of water diffusion,and 1 is the length of the core. The ratio can be approximated by

t = 1 - 2exp -D2t/ -2 8exp -9Di 2 t/12

M o 729,r2

By using Mt/Me, = 0.194 t = 2.68 x 106 s (31 days), and 1 = 8.89 cm, we findthat D = 2.0 x 10-7 cmI/s.

Now, the transport rate of water through the rock salt cores, F, can bereadily calculated by Ficks law of diffusion

F = -D dCdx

66

where C is the concentration of diffusing water at a core length

Co = 7.50 x 10-5 g/g x 2.18 g/cm3

= 1.64 x 10-4 g/cm3

-4

= Co/i = 1.648910 = 1.84 x 10-5 (g/cm4)

Then

F = 2.0 x 10~7 (cm 2 /s) x 1.84 x 10-5 (g/cm 4 ).

= 3.68 x 10-12 g/cm2 -s

Using the cross-sectional area of a core (3.80 cm2) for the core that waspressurized the longest (137 days), we would expect that no more than 170 iLof water would traverse the 8.89-cm-long rock core. This volume of solutioncould easily have avoided detection because of evaporation.

2. Conclusions

Tritiated water appears to be incorporated into rock salt cores byrecrystallization (including hydration) at the surface in contact with brineand by permeation due to diffusion. Tritiated water appears to be removed byevaporative loss at the core surface in contact with air. Water is incorpo-rated by recrystallization at concentrations up to 0.7% by weight, but readilyevaporates when the salt is exposed to air. Water subject to migration bydiffusion reaches a maximum concentration of 7.5 x 10- g/g (75 ppm) in saltthat is in contact with aqueous brine solution. The observed behavior ofthe water is consistent with its having a diffusion coefficient of2.0 x 10-7 cm2/g. Permeation of the water through the rock due to the imposedpressure gradient does not appear to be significant, based on the strong andsustained gradient in tritiated water concentration observed experimentally.This interpretation of the results could be tested in additional experimentsto study fluid flow through rock salt.

B. Determination of Residual Oil Saturation in Depleted OilReservoirs

A log-inject-log technique has been examined that uses gamma-emittingradiochemicals and a detector to determine the residual oil saturation (SOR)of a depleted oil well. This is the most favored technique and consists ofinjection into the reservoir of a solution containing a noninteracting radio-chemical, logging the bore hole with a gamma detector, injecting a solutioncontaining another radiochemical into the reservior, and then logging the borehole again. The residual oil saturation of the reservoir can be calculatedfrom the difference in the count rates of the two logs. This procedure isfavored over the procedure of using the natural radiation in the reservoir tomake one log because the artificial gamma-emitting radiochemicals and theiractivities are known unequivocally.

67

Factors affecting the accuracy of SOR determination are (1) the extentof radiochemical adsorption on reservoir rock, (2) drilling mud that is onthe bore hole wall or in the (perforated) casing, (3) heterogeneities ofdensity and porosity of the reservoir material, (4) aqueous fluid in thereservoir, that is not displaced by injection, (5) residual radiochemicalin the bore hole, (6) reservoir fluid drift, and (7) detector anisotropy.

An experimental program to examine the accuracy of the proposed loggingmethod uses core flooding with radiochemical injections. In the plan, onesandstone core is saturated with oil, then depleted of oil by being floodedwith aqueous reservoir fluid. Another core (similar to the first) is simplysaturated with aqueous reservoir fluid. Both cores are then injected withan aqueous fluid containing a radiochemical that does not react with thesandstone. The gamma signals from both cores are measured. An attempt willbe made to relate the difference in the two signals to the residual oil inthe first core.

1. Geochemical Characterization Relevant to Nuclear Logging

An effort was made to identify possible characteristic elements orratios of elements in aqueous reservoir fluids that could be used in nuclearlog-inject-log techniques. For this effort, several hundred chemical analysesof aqueous reservoir fluids from producing petroleum reservoirs in limestone,quartz, sandstone, arkosic and lithic sandstone, dolomites, chalk, and shalewere solicited from a variety of sources in the petroleum industry, the gov-ernment, and the geologic literature. Seventy of these analyses were studied,using element and element ratio diagrams. If the different aqueous fluids canbe chemically classified, they may be monitored by means of a distinguishingchemical or chemical ratio detectable by the nuclear log response.

Most of the seventy analyses used in the study contained concentra-tions of the following solute species: Na, Ca, Cl, Sr, Br, HCO3, C03, Mg,Fe, I, K, and SO4. The concentration ratios of all chemical pairs were cal-culated and plotted against the other chemical ratios. Also, the concentra-tion of each solute species was plotted against the concentration of eachother species. Nine of the elemental and element-ratio plots displayed sig-nificant correlation of aqueous reservoir fluid composition with reservoirrock type. These plots are of Ca vs. Mg, Sr vs. K, Br vs. I, Mg vs. K,Br vs. SO4, Br/S04 vs. Ca/Mg, K/Na vs. Ca/Cl, Sr/S04 vs. Na/Fe, andNa/Fe vs. Mg/Cl. The fourteen elements or element ratios used in the plotscorrelate with reservoir rock type.

The Br vs. I plot gave the clearest grouping according to reservoirrock type. This suggests that a nuclear log for bromine or iodine would beof sufficient accuracy to quantify the amount of indigenous aqueous fluid re-maining in a reservoir after flooding with a fluid extraneous to the reser-voir. This work will be concluded after a few representative correlationdiagrams are generated by computer using a large (>70) number of analyses.

2. A Proposed Method of Determining Residual Oil

A precise knowledge of the amount of oil remaining in a reservoiris crucial to the selection and economics of a tertiary oil recovery method.

68

Therefore, considerable effort has been expended by the oil industry to accu-rately measure the residual oil in a reservoir (SOR). Methods of determiningSOR are summarized in two recent publications [BOND, WYMAN]. SOR is measuredusing (1) cores of reservoir material, (2) production data and material balanceequations, (3) logs of carbon/oxygen ratio, hydrogen content, etc., (4) welltests (drawdown test, buildup test, etc.) and (5) injections of radio orchemical tracers followed by pumping out of the well.

D. S. Webster (Argonne National Laboratory) suggested that SOR couldbe determined by injecting a reservoir with a fluid in which petroleum wassoluble, then extracting the fluid to determine its petroleum content (i.e.,by reservoir stripping followed by production, or the solvent-extraction-assaymethod). Although reservoir stripping in conjunction with well logging tech-niques is discussed in both reviews cited above, in neither review is theremention of the approach of pumping the solvent out of the well for analysisof SOR. Because the solvent-extraction approach is potentially very accurate,a development of this approach is reported here.

In this approach, a reservoir is injected with a fluid in whichpetroleum is soluble. The immobile petroleum in a reservoir is allowed todissolve in the fluid, then the reservoir is produced (i.e., the solvent ispumped out of the well). The ratio of injected fluid to petroleum in theproduced fluid is indicative of the quantity of residual oil in the reservoir.If mixing of the injected fluid with the petroleum is slow and if the timebetween injection and production is sufficient to allow the injected fluidand petroleum to mix, the ratios of injected fluid to petroleum in successiveproducts may be as in Fig. 21. A high ratio in the initial product (from 0-to 10-gal production) would be expected in the absence of mixing of (1) thefluid injected into the piping and into the bore hole with (2) petroleum inthe reservoir. A constant fluid to petroleum ratio at 10- to 40-gal produc-tion would be expected because of nonreversible flow of the injected fluid,since fingering, injected fluid buoyancy, etc. would allow petroleum-richfluid to be drawn into the bore hole before all of the injected fluid is re-covered. The uniform ratio at 10- to 40-gal production can be used to deter-mine the residual petroleum in the reservoir if the porosity of the reservoiris known and if the aqueous fluid is completely displaced by the injectedfluid. In this case, the relative volume of petroleum, Vp, is given by

V = "/(1 + R)

where $ is the porosity of the reservoir (in cm3 per cm3 of reservoir) and Ris the ratio of volume of injected solvent to volume of petroleum at the pla-teau of the production curve. Since the plateau of the curve in Fig. 21 isat a ratio of 7.8, this would indicate that the volume of petroleum in thereservoir is only 11% of the reservoir pore space.

Initially, refined petroleum -ight be used as the injection solvent.A minor element that occurs in the ur-arined petroleum (such as sulfur) couldthen be used to monitor the ratio of injected fluid to petroleum in the pro-duced fluid. The properties of the injected fluid and the petroleum wouldbe very similar, which might be beneficial in some cases. The refined petro-leum would not adversely modify the physico-chemical state of the reservoir.

69

Z

J J00ci)or0 Wy

0 Hox_ __ _ __ _ __ _ _

0 20 40 60 80 100

TOTAL PETROLEUM SOLVENT PRODUCED, gal

Fig. 21. Anticipated Ratios of Injected Solvent to ExtractedPetroleum as the Well is Pumped (after injecting areservoir with solvent in which petroleum is soluble)

The amount of water extracted together with the petroleum and injected solventwould be indicative of the departure from simple plug flow that is desirableduring injection and pumping. Various volatile fractions from petroleum re-fining could be tested as possible injection solvents. The most suitablefluid may have a volatility between those of heavy oil and gasoline. Such afluid would not be prohibitively expensive.

Alternatively, alcohol could be used as the injection fluid. Alco-hol is often used to strip oil from a depleted oil reservoir. The alcoholwould have properties (e.g., density and viscosity) more similar to those ofwater than to those of petroleum, which may be beneficial since the reservoirwould have been swept previously with water. However, mixing of the alcoholand water might cause increased analytical difficulties and complications indata interpretation.

A third possible injection fluid is Triton X100, which is a nonionicsurfactant used for tertiary oil production. It is believed that oil is sol-uble in this surfactant. Triton X100 is currently delivered to oil fields andis relatively inexpensive; its interactions with the reservoir liquids androcks are already known to some extent and are tolerated.

To determine if this approach to measuring SOR has been developedpreviously, a review of the petroleum-production literature would be required.Laboratory support required for the development of this approach includes (1)selection of an injection fluid that has desirable properties, (2) studies ofthe mixing kinetics of the injected fluid with petroleum, and (3) developmentof on-line equipment to analyze the ratio of injected fluid to petroleumduring production.

70

V. TRACE ELEMENT TRANSPORT IN LITHIC MATERIAL BY FLUID FLOW ATHIGH TEMPERATURE(A. G. Seitz and R. A. Couture)

A. Introduction

This report covers the final planned experimental work on Cs+-Na+ ionexchange on kaolinite, along with some new work on rock hydrology, includingwork on the permeability of limestone from the Notch Peak, Utah, area and de-velopment of a new core holder for further studies of rock hydrology.

B. Ion Exchange on Kaolinite

Transport of ions by fluid flow through porous media is of great interestfor the disposal of radioactive waste and in certain areas of geology. Thereis currently great interest in the use of clay minerals as backfill in radio-active waste repositories. Their impermeability, swelling properties, andion-exchange properties are highly desirable. Consequently, information isneeded on their high-temperature properties and mass transport propertiesduring fluid flow. There is also current theoretical interest in the mecha-nisms of ion exchange and in a-priori calculations of exchange equilibria bymeans of Born-Haber cycles [EISENMAN, EBERL]. Unfortunately, there are re-ports that the ratios of transport of ions through beds of minerals do notalways agree with the rates calculated from equilibrium considerations [SEITZ-1980A] and that ion-exchange equilibria depend on such parameters as thesolid/liquid ratio [GRIM] and trace element concentrations [SEITZ-1980A].

In our work [SEITZ-1980B] on Cs+-Na+ exchange on kaolinite as a functionof temperature, it was found that transport rates of Cs+ during flow throughpacked columns of kaolinite are about four times lower at room temperaturethan predicted from equilibration experiments on slurries in test tubes.(Kaolinite is a non-swelling clay mineral, [Al2Si2O 5(OH)4j, which owes itsion-exchange capacity primarily to broken bonds at the surface.) Consequently,an effort was made to identify all the important factors which affect the ionexchange. It was determined that the difference between observed and calcu-lated transport rates was not due to experimental error and apparently was notdue to retardation by the streaming potential, but to an actual difference ofthe distribution coefficient (or ratio) for the column and the test tube ex-periments. It was also determined that the distribution ratio observed in thetest tube experiments is independent of the amount of 134Cs tracer added, atall concentrations used in the column experiments. Therefore, the differencein tracer concentration cannot explain the difference between the predictedand observed transport rates. It was discovered that the distribution ratioin test tubes depends greatly on time and on solid/solution ratio. The timefactor alone does not explain the difference; also, since the distributionratio decreases with increasing solid/solution ratio of the slurry [SEITZ-1980B], the solid/solution ratio does not explain the low transport rates.

Several reports in the literature have shown that sorption of Cs+ by ka-olinite is time-dependent; the cause is thought to be a slow change in thestate of aggregation (flocculation) of the clay. It was established by Tamersand Thomas [TAMERS] that the Cs+ exchange capacity of kaolinite changes slowlyover a period of weeks, whereas isotopic equilibrium is established quickly.

71

This means that a secular change in the clay itself is involved. However,there is little or no direct evidence that the state of aggregation changeswith time. Our discovery [SEITZ-1980B] that the distribution ratio dependson the solid/liquid ratio supports the view that sorption of Cs+ depends onthe state of aggregation of the clay, provided that there is no alternativeexplanation for the dependence on solid/liquid ratio.

A possible alternative explanation is initial disequilibrium between solidand liquid. If the solid is initially out of equilibrium with the liquid withrespect to exchangeable ions, the extent of reaction during an experiment willdepend on the solid/liquid ratio. At equilibrium in 0.1M NaHCO3 , the exchangeions will consist overwhelmingly of Na+, regardless of the solid/liquid ratio.However, if there are two or more exchange sites with very different charac-teristics, with one site present in very low concentration, the distributioncoefficient might be very sensitive to sorbed trace element concentration.(Measurements of the sorption isotherm will show that this is an actual pos-sibility.)

Consequently, we did several experiments (results are reported below) toinfer what effect trace elements might have on the distribution coefficientfor cesium. First, we determined the isotherm for Cs+ sorption by kaolinitefrom 0.1M NaHCO3 as a function of concentration of Cs+. Then we determinedthe effect of preequilibration of the clay with NaHCO3 solution before reac-tion. Next, we determined the rate of isotopic equilibration by adding traceronly near the end of an experiment. Finally, we compared unwashed clay andwater-washed clay with clay which had been removed from the top (entrance) ofa column and which had therefore been washed thoroughly with NaHCO3.

1. Procedures

The method for sorption measurement was the same as was used previ-ously [SEITZ-1980B] and consisted of rotating suspensions in stoppered testtubes. One batch of clay was first prewashed with freshly prepared 0.1M NaHCO3solution in the ratio 0.0375 g clay/mL solution. The mixture was agitatedovernight and then centrifuged lightly; the procedure was repeated, and thenthe clay was given a final wash with H20 and centrifuged lightly. The claywas dried at 50*C, and the (soft) lumps were crushed and mixed gently to re-duce inhomogeneities. The procedure was duplicated simultaneously on anotherbatch, with deionized water instead of NaHCO3 solution, to obtain a control.Care was taken to ensure equal and simultaneous treatment. Previous measure-ments of distribution coefficients indicated that if the dependence of thedistribution ratio on the solid/liquid ratio were entirely due to cesium nat-urally present as an exchangeable ion in the clay and if there were no Cs+ inthe solution, washing with NaHCO3 solution would remove about 50% of the ex-changeable cesium from the clay. This would increase the distribution ratio(at 24 h) from about 65 to 90 mL/g. Whether it does so is a test of the as-sumption that the dependence of the distribution ratio on the solid/liquidratio is due to removal of cesium from the clay.

The kinetics of the Cs+ sorption was studied by adding Cs+ tracerto slurries of NaHCO3 solution and clay four hours before the conclusion ofan experiment (which ran for a total of 116 h). The data were compared withthose in which the tracer was added at the beginning.

72

The Cs+ sorption isotherm was determined without pretreatment ofthe clay. Various amounts of stable CsCl were added initially in addition to13 4Cs. The total amount of cesium remaining in solution was calculated fromthe fraction of tracer remaining in solution times the amount initially pre-sent. At the lowest cesium concentrations, the amount of cesium initiallypresent had to be estimated by using the supplier's specification for thetracer of 20-100 Ci/g Cs.

2. Results

The effects of isotopic equilibration time and of prewashing areshown in Fig. 22. The counting errors are fairly large in this experimentbecause of low activity used, but it is clear that prewashing with NaHCO 3slightly decreased the amount of sorption by kaolinite in comparison withprewashing with deionized water. The effect of washing the kaolinite with0.1M NaHCO3 in a column is also small and negative. Thus, Kd apparently did

Ho

100

90

E

80

70

60

50 -

40_0.1 I0t

% KAOLINITE IN SUSPENSION

100

Fig. 22. Effects of Prewashing and Equilibration Timeon Distribution of Cs+ between Kaolinite and0.1M NaHCO3. 116 h, pH of 8.8.

1 - Clay prewashed with deionized water2 - Clay prewashed with 0.1M NaHCO33 - Clay prewashed with 0.1M NaHCO 3 , 1 3 4 Cs

tracer added 4 h before end of experiment4 - Clay taken from top of column washed

with 0.1M NaHCO3

2

73

not increase due to net transfer of trace elements to the solution during pre-washing. The dependence of Kd on solid/liquid ratio observed in previous ex-periments [SEITZ-1980B] is such that it cannot be due to net transfer of traceelements from the solution to the clay or from the test tubes to the clay.Therefore, we believe that the new evidence strengthens the conclusion thatthe distribution coefficient depends on the state of aggregation of the clay.Evidently, prewashing in NaHCO3 solution has some effect on the clay, but theeffect is not that expected on the basis of ion exchange.

The results of short isotopic equilibration are also shown in Fig.22. Comparison of curves 2 and 3 shows that equilibrium between Cs+ and theclay-water suspension is not reached after four hours. Evidently, we are notstudying an instantaneous surface phenomenon.

The isotherm for sorption of Cs+ by kaolinite in 0.1M NaHCO3 isshown in Table 15 and Fig. 23. The solid/liquid ratio was 0.0375 g/mL. Theresults are presented in Fig. 23 in terms of fractional loading of the ex-change sites, calculated as the mole fraction of the exchange capacity (about9.8 x 10-5 mol/g). Loading of only a tiny fraction of the sorption sites byCs+ changes the distribution coefficient markedly: an increase in the loading

60 'FROM COLUMN

DIFFERENT BATCH OF NoHCO 3

50

40-

E

30

20

1010 10' 0.100~7 10~ 10 104

FRACTIONAL LOADING

Fig. 23. Isotherm for Sorption of Cs+ by Kaolinitein 0.1M NaHCO3 , Expressed as Fractional

Loading of Exchange Capacity. Also shownare Kd for kaolinite taken from top of acolumn that had been used with NaHCO3, andtwo points determined with a differentbatch of NaHCO3. Amount of cesium presentin NaHCO3 and kaolinite assumed to be zero.24-hour equilibration.

74

Table 15. Distribution Coefficient of Cs+ between Kaolinite andFresh 0. IM NaHCO3 . 24-Hour Equilibration;Solid/Liquid = 0.0375 g/mL

Comnent [Cs+], Ma [Cs+], Mb Loading Fract5.onb Kd,mL/g

8.4 x 10-5 8.4 x 10-5 4.3 x 10-5 20

7.6 x 10-6 7.6 x 10-6 3.0 x 10-6 256.9 x 10~7 6.9 x 10-7 - 7.0 x 10-7 2.3 x 10-7 31

5.8 x 10-8 6.0 x 10-8 - 6.8 x 10-8 1.5 x 10-8 - 1.7 x 10-8 424.9 x 10-10 2.5 x 10-9 - 1.0 x 10-8 4.8 x 10-10 - 1.9 x 10-9 54

0 2.0 x 10-9 - 1.0 x 10-8 3.8 x 10-10 - 1.9 x 10- 9 53

From top of columnsecond batch of NaHCO 3 0 2.0 x 10- 9 - 1.0 x 10-8 3.8 x 10-10 - 1.9 x 10-9 58

0 2.0 x 10-9 - 1.0 x 10-8 3.8 x 10- 1 0 - 1.9 x 10- 9 59From top of columnsecond batch of NaHCO 3 0 2.0 x 10-9 - 1.0 x 10-8 3.8 x 10-10 - 1.9 x 10- 9 57

atncludes only stable CsCl added; does not include carrier present with tracer, or cesium present in

starting materials, or contamination.bBest estimate. Includes stable CsCl added and estimate of carrier; does not include cesium fromstarting materials or from contamination.

fraction by 2 x 10-8 gives a 20% decrease in Kd. It is again observed thatthe Kd determined for kaolinite removed from the top of a column is only about5% larger than that for fresh, untreated kaolinite. Use of another batch ofNaHCO 3 also gives a Kd about 5% larger.

3. Discussion

The sorption of Cs+ by kaolinite is exceedingly complex. Sorptionis relatively slow: isotopic equilibrium is not attained after four hours;sorption continues for more than 100 h, possibly much longer.

The distribution coefficient decreases with increasing solid/liquidratio in a rotating test tube but increases if the kaolinite is packed in acolumn. Previous use in a column (either the top or bottom) does not affectthe distribution coefficient significantly, but prewashing the kaolinite inNaHCO3 solution decreases Kd slightly; the method used to make the Kd mea-surement is much more important than the history of the kaolinite.

We attribute the behavior of the kaolinite to its state of aggrega-tion in the suspensions. It was proposed in a previous report [SEITZ-1980B]that the degree of edge-to-face flocculation increases at high solid/solutionratio and that this flocculation decreases the number of charged sites avail-able for sorption of ions. It was also proposed that compaction in a columndisrupts flocculation and thereby increases the effective exchange capacity.Schofield and Samson [SCHOFIELD] too have proposed disruption of the structureby compaction, but it still is difficult to verify our hypothesis directly.

A somewhat greater change in Kd with cesium concentration was ob-served previously [BO], although it was thought to be related to possible im-purities in the clay. In our work, the clay was found to be extremely pure,with a small amount of quartz the only impurity detectable by X-ray diffraction.

75

C. Permeability of Big Horse Limestone

In the preceding report [SEITZ-1980C], the geology of the Notch Peak,Utah, area was described. Briefly, a granitic pluton has intruded sedimentsconsisting of limestone, and interbedded limestone and siltstone. Migrationof rare earths from the pluton over several kilometers through the limestonehas been inferred (J. C. Laul, oral communication). Skarns are present inthe area, but they are restricted to areas near the contact. The limestoneis rather impermeable, and it is questionable how much fluid could have beentransported through the limestone in the postulated manner. It is felt,rather, that outward fluid migration took place through fractures, whereskarns are now located. Fractures would have opened during the intrusions.

In order to resolve the question, a sample of the unmetamorphosed lime-stone was obtained and its permeability was measured. (Evidence for lack ofmetamorphism is (1) the lack of stratigraphic thinning (which would have beencaused by loss of volatiles) and (2) distance from the intrusion.) In thepreceding report (SEITZ-1980C), we deduced on the basis of preliminary mea-surements that substantial amounts of fluid could have been forced throughsmall cracks in the 10,000 y after intrusion. Assuming the maximum possiblepressure at 4.4-km depth, up to 10 L of water could have flowed through asquare centimeter of rock at 200*C. However, it seems likely that much ofthe flow would have been diverted through large fractures.

On the basis of further measurements reported below, it is not necessaryto modify those conclusions. The purpose of this report is to describe themethod and details of the results of the permeability measurements. Consider-able emphasis has been placed on determining whether fluid flows through therock or around it during measurement.

1. Method

Sample number 7/12/80/12 from the Big Horse member of the Orr Forma-tion was selected and cored. The unmetamorphosed limestone is quite compact,and appears to be of very low permeability, although a little water soaks rap-idly into the rock. There are a few interconnecting joints which appear tofollow grain boundaries. The scale of jointing is such that interconnectingjoints can be sampled in a core 2.2 cm in diameter, which we hav used forpermeability measurements.

A cylinder 6.74 cm in length and 2.18 cm in diameter was cut in thelaboratory with a core drill and saw. The core was enclosed in a core holderwhich encloses the rock in a pressurized Teflo sleeve, and fluid was thenpumped through the core. The core holder was Immersed in a water bath. Theconfining pressure on the Teflon must exceed the maximum pressure of thefluid to be forced through the rock, plus the Fl:astic rebound pressure of theTeflon. In practice, an excess confining pressure of around 3.4 MPa is ade-quate, although much lower values allow leak.ge. At the beginning of theexperiment the confining pressure was held at over 14 MPa overnight to makethe surface of the Teflon conform to the rock.

An aqueous dye solution was used as the confining fluid; pure waterwas used for the permeability measurements. A 0.2% fluorescein solution waspumped through the rock in a final experiment.

76

Several methods of measurement were tried. The method that workedbest was gravimetric measurement of the flow rate at constant pressure dif-ference. The pressure at the inlet of the core was measured by both a trans-ducer and an inexpensive Bourdon gauge. The two devices agreed to within0.2 MPa; data from the transducer were used for the interpretation. Thepressure at the outlet was ambient.

The permeability was determined from the formula

k = VP,

where k = permeability, m2

p = viscosity, Pa-s

q = flux, m3 -m 2 -s-1

and VP = pressure gradient, Pa/m

2. Results

The results, shown in Table 16 in the original experimental sequence,are highly reproducible and are believed to be reliable. The core assemblywas completely disassembled and the core removed. After reassembly, the mea-sured permeability was observed to be unchanged, to within 1%. By the useof 0.2% fluorescein solution as the confining fluid, it was shown that no con-fining fluid leaked into the limestone during the flow measurements. Thus,the measured flow rates cannot be erroneously high because of the addition ofconfining fluid. It is also certain that any creep of the seals which sepa-rated the fluids did not significantly distort the flow rate measurements.Pumping continuously to increase the confining pressure prevented the flowrate from changing by more than 0.0001 g/min.

There is good evidence that the observed flow was through the coreand was not due to leaks around the core. After the experiment, the insidesurface of the Teflon was found to correspond closely to the rock surface.Where there were cracks in the rock, ridges were visible in the Teflon; areasof the Teflon that contacted smooth metal spacers were polished. Furthermore,increasing the confining pressure by 50% at 25*C reduced the flow rate by only3-5%. This slight reduction in flow rate is easily attributable to actualcompaction of the rock pores [COLLINS, p. 13]. Finally, the excellent repro-ducibility after the core was disassembled and reassembled suggests that therewere no major leaks around the core.

A dye experiment was also performed to test penetration of the coreby the solution. A 0.2% solution of fluorescein was pumped through the core,and the core was disassembled and examined under ultraviolet light. Solutionin the major cracks fluoresced brightly. Possibly, fluorescein was introducedinto the cracks by external contamination when the assembly was opened. Afew drops did contact the surface but were wiped off immediately. However,

77

Permeability Measurements ofthe Notch Peak, Utah, Area.asterisk indicates there wasflow rate when the confiningmeasurement.

Limestone Sample 7/12/80/12 fromErrors shown are ranges. Anno significant difference inpressure was increased during

Confining Pressure, Flow Rate, Permeability,T, C AP, MPa MPa g/min m2

25.0 6.9 17.7 0.00527 0.00003 2.00 x 10-123.4 17.8 0.00248 0.00002 1.93 x 10-12

10.3 17.7 0.00838 0.00001 2.18 x 10-12

13.8 19.1 0.01177 0.00008 2.30 x 10-12

13.8 27.4 0.01122 0.00002 2.19 x 10-12

10.3 27.1-27.2 0.00815 0.00017 2.12 x 10-1217.2 26.5 0.01416 0.00018 2.21 x 10-12

20.7 26.1-27.6* 0.01778 0.00004* 2.31 x 10-12

Core assembly dismantled and reassembled.

25.0 3.4 13.7 0.00247 0.00004 1.92 x 10-12

50.0 3.4 14.4 0.00357 1.73 x 10-12

6.9 14.5 0.00850 0.00008 2.05 x 10-123.4 14.5 0.00362 0.00004 1.75 x 10-12

an attempt to duplicate the results by deliberate contamination failed. One-half of the core was rinsed and blotted dry. Then a few drops of dye wereintroduced onto the surface and immediately blotted off. The whole texturedsurface of the rock fluoresced where the dye had touched; no such surfacefluorescence had been observed where the core had been uncontaminated. Thus,there is no evidence that the dye in the cracks came from the surface of thecore; rather, it must have come from interconnecting cracks or from the rockpores. Finally, a fresh surface of the core was broken and was found tofluoresce. The water must have moved throught the rock, not around it.

D. Discussion

Permeability appears to increase with increasing pressure gradient. Figure24 shows a plot of flow rate vs. AP. The increase in flow rate is unlikelyto be due to errors in the pressure gauges since all gauges tested agree withnarrow limits, which are much too small to account for the observed changes.The increased permeability may be due to the effect of air in the rock. Nearthe top of the core, air would be highly compressed; near the bottom of the

Table 16.

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0.02

25*C f17.9 MPa confining pressurex ~ 27.6 MPa confining pressure

50 C A 13.8 MPa confining pressure

wQ 0.01

0

0 6.9 13.8 20.7LP, MPa

Fig. 24. Flow of Water through a Limestone Core as a Functionof Pressure Difference, AP, and Temperature

core, air would not be very compressed and would occlude part of the porosity.The result would be a small error which would be relatively most importantat low pressure gradients and high temperatures.

Alternatively, the nonlinearity might be due to leakage around the rockcore, or to actual changes in the rock with increasing pore pressure.

E. Hydrologic Properties and Ground Water Composition ofNorthern Illinois Granite

Work is continuing on the measurement of the permeability and trace.-element transport properties of granite from the North American Shield,obtained from Northern Illinois drill hole UPH-3. Recent work [FRITZ-1979, 1980] has shown the existence of old CaCl2 brine as groundwater inPrecambrian granitic rocks of the Canadian Shield. This groundwater hasbeen studied extensively and is thought to be the reaction product of normalgroundwater and granitic rocks during a very long contact period [FRITZ-1979,1980]. Other work has suggested that diffusion into dead-end rock pores maybe important factors in retarding transport of dissolved materials [GRISAK-1980A, 1980B, 1980C, 1981; NERETNIEKS].

Our experience suggests that elution experiments on granite could be de-signed to sample tightly held pore water and to study transport of dissolvedsubstances through the rock matrix. Cores of Precambrian granite from NorthernIllinois drill hole UPH-3 were obtained in order to test these ideas. Afairly impermeable rock was selected and prepared, and a special core holderwas built for the experiments.

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1. The Core Holder

The core holder consists of two cylindrical end plates with zero-dead-volume fittings (i.e., so that there was no dead volume to act as mixingchambers in the fluid stream). Teflon tape is wrapped around the cylinderof rock, and a tight Neoprene sleeve (bicycle tire inner tube) is placed overthe rock and the end plates. The sleeve is cemented to the end plates, tubingis fitted to the end plates, and the assembly is placed in a pressure vessel.The result is a totally enclosed rock cylinder with capillary tubes leadingto the ends.

The end plates are constructed of stainless steel (type 316); theinlet tubing is stainless steel with an inner diameter of 0.25 mm; and theexit tubing is glass-lined steel with an inner diameter of 0.3 mm.

The core holder has been tested and found not to leak, although itremains to be established whether fluid can leak between the outside of therock cylinder and the Neoprene sleeve.

2. Design Problems

The core holder was originally designed with a 3-mm-thick Teflonsleeve instead of a Neoprene sleeve. Assembly was accomplished by firstheating the Teflon, since the inner diameter of the Teflon is 0.5% smallerthan the diameter of the end plates. However, the fit was not tight enough,and the device leaked, even after it was pressurized to 135 bars (13.5 MPa)at 35*C overnight. Clamps and rubber cement did not stop the leak.

The reader is referred here to three works which describe similarcore holders [DONALDSON, POTTER, BRACE]. Donaldson described elution experi-ments on sedimentary rocks; Potter measured permeabilities in the nanodarcyrange with a Teflon core holder. However, to our knowledge, no one has doneelution experiments on igneous (low-permeability) rocks.

One bizarre and surprising observation poses a serious experimentaldifficulty. Methyl violet (pKi = 1) solution was used as the pressurizing so-lution in leak tests, in order to differentiate between it and the solutioninside the core holder. In addition to being intensely colored, it stainsthe Teflon and the steel, which is useful for the post-experimental examina-tion of failed core holders. Unfortunately, in leaking past the Teflon, themethyl violet (0.001% solution) is removed completely from solution, so thatpure water emerges from the end plates! This casts doubts on the use of dyesas tracers to check for leaks in our experiment. Dyes and other tracers thatstain the apparatus should be avoided!

3. Results

At the time of writing, halide-rich brine is being eluted from acore of Precambrian granite (permeability about 0.35 ydarcy) from NorthernIllinois. The experiment is not complete, but a method appears to be estab-lished for sampling groundwater in very impermeable rock formations. Theresult also appears to generalize the occurrence of brine in granitic rocksof the North American Shield.

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VI. LIGHT WATER BREEDER REACTOR PROOF-OF-BREEDING (LWBR POB)ANALYTICAL SUPPORT PROJECT(N. M. Levitz, T. F. Cannon, P. G. Deeken, R. E. Nelson,J. E. Parks, L. E. Trevorrow, C. G. Wach, I. 0. Winsch,J. E. Fagan, Henry Lautermilch, R. E. Brock andR. J. Meyer)

This project includes responsibility for the destructive analysis (sched-uled for FY 1986) of full-length irradiated fuel rods from the LWBR. Theresults will be used by the Bettis Atomic Power Laboratory (BAPL) in supportof their nondestructive assay of the end-of-life (EOL) core to determine theextent of breeding.

A. Full-Scale Shear (FSS)(P. G. Deeken, J. E. Fagan, and R. E. Brock)

Testing and development of the full-scale shear (FSS) continued in threemain areas: test shearing with the shear head-end assembly in the mockup areaat Chemical Engineering Division (Bldg. 205); assembly, testing, and trouble-shooting of the shear feed system at Chemistry Division (Bldg. 200); and in-stallation work in Cell M-3 (Bldg. 200).

Final testing of the shear head-end assembly in the mockup area was com-pleted. Testing comprised shearing of short Th0 2-loaded Zircaloy-clad rodsections to determine the error associated with locating the cutting (orshear) plane--particularly as it relates to the apparent cut through the fuelaloa:. This work is also closely allied to investigating cross-contaminationlevels at intersegment boundary cuts. Previous work had assumed that therewould be no error inherent in the actual cut, and so this issue was not ad-dressed during earlier pilot-scale shear studies. A total of 22 short rods,representing each of the four sizes of fuel rod to be assayed at end-of-life(EOL), were prepared and sheared in the present test program. The effects ofcutting speed (about 5 and 100 cm/s) and blade shape (flat-bottom and pointed)were studied.

The results of the tests indicated that the error associated with a seg-ment boundary cut due to shearing, considering the ceramic fuel apart fromthe cladding, is well in excess of the limits established for rod segmenting;the allowable error is 0.025 cm. Errors, converted to equivalent lengths offuel in inches, ranged from 0.074 cm for the seed rod specimens, to 0.13 cmfor the 2.11-cm-dia reflector rod. A higher speed and the pointed blade gavethe best results (i.e., the smallest error values).

These results are considered preliminary because the setup used in themockup work for holding the fuel rod samples in the shear was not as rigid asthat in the FSS assembly, i.e., slippage may have occurred. Further work onthis problem (done with the shear installed in Cell M-3) will be aimed atreducing the error by controlling variables such as shear speed, cut length,and blade configuration.

The shear feed/measurement system in its present out-of-cell locationunderwent further laser calibration to determine the error associated withpositioning the carriage that feeds the fuel rod into the shear; the carriage

81

is driven by a precision ball screw. Evaluation of the data obtained during45 traverses of the carriage over its 2.3-m (90-in.) length of travel alongthe shear bed indicates that the feed system has the ability to position arod for shearing with an error no greater than 0.005 cm (0.002 in.) Further-more, excellent agreement (maximum difference, 0.005 cm) was achieved betweenthe two independent carriage travel-measuring systems, the Index-Syn,* whichfeeds (and counts) electronic pulses to the ball screw drive motor (the numberof pulses is correlated with motor revolutions), and the NPL,t which indepen-dently counts ball screw revolutions (through a geared-down transducer). Allcritical components were calibrated, including spares.

Final assembly of the shear is in progress. Electrical connections ofthe shear and the computer for automatic operation were made and tested.Additional sensors such as a shear blade speed indicator were added to providefurther insight into the shear process, as well as for automatic control.

The fuel rod weighing system which meets the requirements for fuel rodweight prior to segmenting has been redesigned; the system is being fabricated.The new design involves replacement of the two existing load cell weighingelements with two higher sensitivity electronic balances; the two balancesare used simultaneously.

In-cell work continued, with the fabrication and installation of ashielded rack for the storage of three end-of-life (EOL) fuel rods in IN-40(secondary) shipping containers. The rack provides surge capacity in theevent rod shipments are delayed. Other modifications to the alpha enclosurewere completed that improve remote capabilities.

Shearing of dummy fuel rods in the out-of-cell mockup is expected tocommence in the first quarter of 1981, followed by in-cell installation ofthe shear and final shear testing and qualification.

B. Single-Unit Dissolver (SUD)(I. 0. Winsch, T. F. Cannon, C. G. Wach, and Henry Lautermilch)

Installation of equipment for the single-unit or prototype dissolver (SUD)system in the shielded cell (K-3) is considered complete as of December 30,1980. However, since the work is developmental, changes of equipment may beneeded in the course of dissolution studies.

Boiling water and boiling nitric acid tests were performed to determinethe characteristics of the SUD and SUD system with the external heaters onthe stainless steel (SS) secondary vessel, rather than in the annular regionbetween the primary tantalum vessel and the SS secondary vessel. These testsare part of a continuing investigation aimed at establishing a practical de-sign for the EOL multiple dissolvers. Tests were completed with minimum andmaximum liquid volumes, i.e., 500 and 3000 mL. Results showed that heatup and

*Series 70, product of Control Systems Research, Division of Contraves

Goerz Corp., Pittsburg, PA.

tNorthern Precision Laboratories Inc., Fairfield, NJ.

82

cooldown rates were excessive due to the large annular space between the inner(tantalum) vessel and outer vessel. As described below, a contoured SS outervessel which minimizes this annular space has been fabricated for the nextseries of tests.

Three SUD dissolution tests were completed in which sheared lengths ofTh02-filled simulated seed, standard blanket, and reflector rods were dis-solved in nitric acid-HF mixture. Good audio evidence of boiling during thedissolutions was obtained with an accelerometer attached to the top portionof the dissolver. Visual inspection of the dissolver, dissolver basket,hulls, and filters following each test indicated that complete dissolution ofthe Th02 had been achieved. Analytical results for samples taken during thetests are yet to be received. Experience was also gained with liquid andoff-gas sampling in these tests.

A decontamination procedure for the blend tank was tested in conjunctionwith the dissolution of a simulated standard blanket rod, using acid ref lux(two 200-mL batches of 0.5N HNO3); silver was added to the dissolver solutionas a tracer to verify decontamination. Sample analyses for silver are notavailable; however, smears of the blend tank surface following decontaminationfailed to indicate any thorium activity. This suggests that the decontami-nation factor was relatively high.

The MODICON* programmable controller is in use as a part of the SUD con-trol system; this system contains a configuration of 56 valves and pumps. TheMODICON provides for semi-automatic switching of this configuration from onestate to another as required by the operating procedure. The controller under-went initial testing during water transfer tests; some reprogramming needs wereimplemented. It was then used successfully in the thoria dissolution tests.

C. Multiple Dissolver System(I. 0. Winsch, T. F. Cannon, C. G. Wach, and Henry Lautermilch)

Four dissolvers similar to that developed for the SUD are to be installedin Cell M-1. This equipment will be used for the dissolution of 33 LWBR-POBfuel rods (actually, about 350 segments) during the end-of-life (EOL) cam-paign. Major activity for this subtask is centered in the Engineering Divi-sion, with equipment in the design stage.

A Work Plan for FY 1981 for the design and fabrication of the MultipleDissolver System, prepared by ANL-ENG, is under review by the ChemicalEngineering Division. The work has been divided into eight subsystems forconvenience of monitoring/control. Subsystem titles and status follow.

1. Dissolver Heater Mockup. A stainless steel outer vessel contouredto the shape of the tantalum dissolver has been fabricated such that the gapbetween vessels is minimized. External heaters are to be tested with thisarrangement, rather than the heaters being in the annular region clamped tothe tantalum vessel. The heater control console has been completed, but the

*Gould, Inc., MODICON Division, Arlington Heights, IL.

83

work is slightly delayed since heater delivery has been delayed to January1981. Furthermore, the stainless steel vessel is temporarily at the tantalumvessel fabricator* being used as a gage for the four tantalum dissolvers(i.e., the primary vessels). Initial testing of the system, scheduled forMarch 1981, will be with the SUD dissolver.

2. Dissolver Vessel Assembly. Fabrication of the four tantalum dis-solvers is complete, and the units have successfully passed hydrostatic pres-sure and helium-leak tests. Radiographs of longitudinal welds are undergoingreview by the ANL Quality Assurance Division (QAD).

3. Dissolver Cooling. Detailing has started.

4. Off-Gas Processing. Detailing has started. The off-gas systemoriginally conceived was considerably simplified in terms of the degree ofautomation (e.g., switching of valves on off-gas traps) and the reductionand elimination of traps, without impacting safety considerations. Thechanges should result in significant cost saving.

5. Acid Transfer. Detailing has started.

6. Blend Tank Assembly. Several blend tank features remain to befirmed before design is finalized. Among them are type of mixing, manner ofdraining (which can impact cross-contamination), method of decontamination,and types of connectors. A blend tank management plan must be developed inwhich the number and size(s) of blend tanks, frequency of change-out, etc.are considered.

7. Electrical Power/Controls/Instrumentation. Design work has started

on the portable control console, main power (electrical) distribution panel,and emergency power distribution panel.

8. Cell Modifications and Equipment. Layout drawings for the worktable and alpha containment structure have been completed and are being re-viewed by the Engineering Division. Piping and instrumentation (P&I) drawingsfor off-gas processing, dissolver cooling, and acid transfer have been updated.

D. Scrap and Waste

(L. E. Trevorrow and R. E. Nelson)

1. Spray Calcination of Dissolver Solutions

Although a number of waste types will emerge from the LWBR-POB Pro-ject, priority is being given to the development of a method of disposing ofthe dissolver solution because of the difficulties associated with disposalof highly radioactive liquids. It is planned to convert the dissolver solu-tion to solids by spray calcination.

Concepts and drawings of the proposed operations, equipment, andarrangement in the hot cell were subjected to a two-day concept design review

*Fansteel Corp., Torrence, CA.

84

by members of the ANL Engineering, Chemical Engineering, and OccupationalHealth and Safety Divisions. Many of the layout drawings of large compo-nents that were prepared for the review are considered capable of beingreadily converted into fabrication drawings. From the review, it was con-cluded that the design should generally proceed as originally conceived. Thereport of the concept design review, containing a summary of proceedings,action items, and recommendations, will serve as a guide for directing thefuture course of the work.

Revisions of the calcination system concept involving both vesselsand the P&I arrangement have been made by the Chemical Engineering Division.The revisions are in response to recommendations made in the design reviewand also are aimed at reducing the total cost of the system. Substantialdecreases in overall size and cost by the consolidation of vessels and reduc-tion of instrumentation are anticipated.

A simple apparatus for the testing of spray nozzles under nonradio-active conditions was assembled. This apparatus permits air and water inputto a nozzle a: controlled pressure. The nozzle is mounted in a glass chamberto permit visual observation of the spray.

The program for this subproject includes assembly and nonradioactivetesting (mockup) of the calciner and off-gas system at the Chemical EngineeringDivision (Bldg. 205) prior to installation in Cell M-4 of Chemistry Division(Bldg. 200), the intended site of EOL operations. Mockup plans were alteredwhen an area became available in Bldg. 205 that will permit the equipment tobe assembled in the same configuration that it will have in EOL operations.Current plans, therefore, are to assemble the calciner and off-gas system onskids, one or two skids at most, that can be transported with the assembledequipment intact to its location in Cell M-4. This procedure is expected tosave considerable time and cost in comparison with the previously intendedprocedure; because of a limitation in the area, the earlier procedure wouldhave required disassembly of the mockup system and reassembly in a differentconfiguration in the cell. The impact of this change in plan on the overallsubproject schedule is being evaluated. Scheduling itself continues to be animportant phase of the overall task; Gantt charts (bar charts) are being used.

The first draft of a procedure for operating the calciner and theoff-gas system has been prepared. This conceptual operating procedure is tobe used to guide analyses of safety, needs for instrumentation, and the de-velopment of a software program for operation, control, and data acquisition.It is expected that development of the final operating procedure will be anongoing, iterative process.

2. Waste Shipment

Waste calcine package design and procedures for packaging mustcomply with ANL safety rules, DOE shipping rules, and the acceptance criteriaof the waste-acceptor contractor. The Department of Transportation (DOT)Spec 6M package has been used as a guideline for package design. Our adapta-tion of this design provides for calcined waste to be contained in primary andsecondary vessels of stainless steel, lead shielding, and centering media,

85

nested in a 55-gal drum. The intended waste-acceptor contractor, Rockwell-Hanford, requires that the radiation dose be limited to 200 mr/h at the sur-face of the drum. Additional work on the package design was done, with theaid of dose rate calculations by computer code (DOT-IV), to increase the totalweight of shielding to the DOT limit of a total package weight of 640 lb forthe Spec 6M package. DOE permission to use an even greater weight of shieldingper package, up to a total package weight of 840 lb, is being sought. Addi-tional shielding would permit the shipment of larger amounts of radioactivity

per package and would reduce the total number of packages shipped and there-fore the total cost of shipping.

Several organizations have been asked for advice on package designand shipping procedures. Information has been exchanged with DOE-CORO, Nu-clear Engineering Corporation (NECO), Mound Laboratory, and Rockwell-Hanford.Acceptance of LWBR wastes by Rockwell was discussed in a meeting at ANL withJ. Anderson of Rockwell.

A document describing the procedures and packaging to be used forshipping the solidified dissolver waste from the ANL site was reviewed byproject personnel, then distributed to Rockwell-Hanford, the intended waste-acceptor contractor, and also to DOE-CH. Informal opinions from both of theseorganizations based on this document are that the intended procedures and hard-ware seem to be acceptable. Such a document will now be formally transmitted.

E. Computer System(J. E. Parks)

The computer system presently being developed for use with the LWBR-POBAnalytical Support Project was described in [STEINDLER]. Considerable pro-gress was made during this report period in the areas of basic hardware ac-quisition and checkout, software development, and interfacing with the FSSand SUD, as reported below.

1. Hardware

Major hardware items are a central processing unit (CPU), a VAX-11/780,* which is shared with other Chemical Engineering Division users, adedicated computer, a PDP-11 /23 ,* and CAMAC modules. The CPU was shipped bythe vendor in late December and is due at ANL in early January 1981. It willbe installed in Bldg. 205 (Chemical Engineering Division) in a specially builtroom now under construction; the room is scheduled to be ready by February1981. Installation and checkout of the CPU has a high priority.

All components of the PDP-11/23 computer were received, assembledby the Chemical Engineering Division Computer Section in Bldg. 205, and tested.Malfunctions were traced to minor manufacturing flaws which were easily cor-rected. The software for testing the unit was "customized" to our hardwareconfiguration. The computer was subsequently moved to Bldg. 200 (ChemistryDivision) and installed in M-Wing adjacent to the shielded cells. It has alsobeen connected to the phone lines leading to Bldg. 205, which will provide theconnection to the VAX-11l/780.

*Products of Digital Equipment Corp. (DEC), Maynard, Mass.

86

CAMAC modules for both the FSS and SUD systems have been installed.A manual "crate controller" has also been installed for the FSS system, whichallows simulation of computer/operator interactions.

Temporary wiring of the controls and sensors to the CAMAC systemfor the FSS, nearly complete, will permit full testing of the automation fea-tures of the FSS. Testing of the interconnections by use of the manual "cratecontroller" has paralleled this work. Reconnection of the shear system to thecomputer after its installation in the shielded cell (Cell M-3) will be expe-dited by the use of detailed cross-connect tables which are in preparation.

Interfacing of the SUD with the computer system has proceeded in asomewhat different manner, since the SUD is already mounted in the shieldedcell (Cell K-3). The CAMAC system for the SUD has been assembled, tested, andinstalled adjacent to Cell K-3. All connections between the main SUD controlconsole and the CAMAC system have been made. Functional testing of theseconnections will be done directly with the PDP-11/23 computer as soon as itbecomes operational. Two interface modules remain to be assembled, the SUD-computer "communication console" and the balance-computer interface. Workhas been initiated on documentation of the SUD-computer interface via thepreparation of cross-connect tables.

2. Software

Software development is required for both the VAX-11/780 and thePDP-11/23 computer. The status of the four major programs follows.

a. The monitoring program (MONITR), which continuously reads ap-proximately 130 process sensors (temperature, pressure, flow) has been writtenand debugged. This program is used with the PDP-11/23. In addition to sensingdata, the program stores data, compares sensed values with set points andactuates alarms, etc.

b. Preparation of the Executive Operating Program is about 75%complete; this program is for the PDP-l1/23. This work involves incorpora-tion of the FSS and SUD operating procedures, for later call-up by the opera-tor at a CRT terminal; actual programming of the procedures is scheduled tobe started in late January. Yet another phase, formatting of the informationat the CRT, is also scheduled for early definition.

c. Work has been started on the Data Management Program for thePDP-11/23 and the Report Generating Program for the VAX-11/780, which is in-tended to prepare summary report sheets on fuel segments and/or entire rods.

Parts of all of these programs will be in use during the next sev-eral months.

F. Blend Tanks (BTs)(J. E. Parks)

The development of a suitable blend tank (BT) system for EOL has severalmajor facets. Among them are (a) actual design(s), including size (to accom-modate different solution volumes); (b) overall strategy for uaage, which

87

includes management issues such as the number of BTs required, reuse vs. dis-card, and decontamination; and (c) the problem of sampling with negligiblesampling error (part of a qualification plan). Results of recent activitiesfollow.

1. Design

Except for gasketing and the type of mixing that will be used, basicdesign features have been set by the ANL Engineering Division. Mixing optionsinclude magnetic stirring, mechanical mixing, and ultrasonic mixing; a decisionabout mixing awaits testing by Chemical Engineering Division, planned for nextquarter.

The concept of disposable plastic BT liners was reviewed. Onlyexploratory testing of "welded" plastic samples is planned for next quarter.

A "Preliminary BT Design/Management Summary" by the ANL EngineeringDivision showed EOL BT costs ranging from $37 K to $107 K, depending on thedesign and the strategy chosen. This document is expected to guide BT devel-opment work.

2. Blend Tank Decontamination

Since decontamination and reuse of BTs might impact analyticalresults via cross-contamination, requirements for a BT decontamination proce-dure were drafted. A BT Decontamination Test Plan was issued, and BT decon-tamination tests are in progress. Initial results with a 12-L BT indicatethat following solution withdrawal via a dip tube, the volume of solutionremaining in the vessel is higher than desired. If this solution retentioncan be minimized, it should be possible to reduce the contaminant level to anacceptable value (on the order of 0.01% of the original amount present) byperforming two or three rinses (either by spraying or by refluxing) with200-mL batches of dilute acid. Further testing with uranium-containingsolution to verify this is planned.

3. Sampling Error

ANL is required to demonstrate that sampling of the BT results innegligible sampling error. A statistically based experimental plan for thispurpose was prepared and submitted to the New Brunswick Laboratory at ANL forreview. The plan will be incorporated into the Single Unit Dissolver SystemQualification Plan.

G. Analytical(R. J. Meyer)

Two new automated analytical instruments will be used for the EOL analy-ses. They are a mass spectrometer and a multichannel analyzer (gamma-counting)system. The status of these instruments is as follows: Design drawings forthe mass spectrometer have been approved, and the manufacturer has been autho-rized to begin construction. The laboratory that will house this instrumentis being readied, and a layout sketch of the floor arrangement has been made.Work has started on the specifications for the gamma-counting system; Thesespecifications are scheduled to be complete in the next quarter.

88

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93

Distribution for ANL-81-13

Internal:

M. Ader T. J. Gerding J. SimmonsJ. K. Bates D. R. Hamrin M. J. Steindler (10)R. L. Breyne J. E. Harmon L. E. Trevorrow

R. E. Brock L. J. Jardine S. VoglerL. Burris J. H. Kittel (2) C. G. WachF. A. Cafasso M. Krumpelt D. S. WebsterT. F. Cannon H. Lautermilch A. A. ZieglerR. A. Couture R. A. Leonard A. B. KrisciunasE. J. Croke W. J. Mecham ANL Patent Dept.J. E. Fagan K. M. Myles ANL Contract FileP. R. Fields R. E. Nelson ANL Libraries (4)B. R. T. Frost W. B. Seefeldt TIS Files (6)

M. G. Seitz

External:

DOE-TIC, for distribution per UC-70 (325)Manager, Chicago Operations Office, DOES. A. Mann, DOE-CHArgonne Universities Association:

President

C. B. Alcock, U. TorontoJ. T. Banchero, U. Notre DameP. W. Gilles, U. KansasR. I. Newman, Fripp Island, S. C.

S. W. Ahrends, Oak Ridge Operations Office, USDOET. W. Ambrose, Battelle Pacific Northwest Lab.C. K. Anderson, Combustion EngineeringR. E. Barletta, Brookhaven National Lab.G. S. Barney, Rockwell Hanford OperationsBattelle-Columbus Labs.R. C. Baxter, Allied-General Nuclear ServicesB. C. Blanke, USDOE-DA, Miamisburg, 0.E. Bondietti, Oak Ridge National Lab.D. Bowersox, Los Alamos National Lab.M. C. Britton, Corning Glass WorksR. Brown, Allied Chemical Corp., Idaho FallsL. L. Burger, Battelle Pacific Northwest Lab.D. Camp, Lawrence Livermore National Lab.D. 0. Campbell, Oak Ridge National Lab.W. Carbiener, Battelle-Columbus Labs.W. T. Cave, Mound Lab.B. H. Cherry, GPU Services Corp.J. M. Cleveland, U. S. Geological Survey, Lakewood, Colo.F. E. Coffman, Office of Fusion Energy, USDOEJ. J. Cohen, Lawrence Livermore National Lab.Commonwealth Edison, Vice Chairman, ChicagoC. R. Cooley, Office of Nuclear Waste Management, USDOEJ. L. Crandall, Savannah River Lab.M. C. Cullingford, Nuclear Regulatory CommissionR. Cunningham, USNRC, Nuclear Materials Safety & Safeguards

94

G. H. Daly, Office of Nuclear Energy, USDOE (2)B. R. Dickey, Allied Chemical Corp., Idaho FallsJ. E. Dieckhoner, Div. Waste Products, USDOEJ. Dietz, Los Alamos National Lab.R. L. Dillon, Battelle Pacific Northwest Lab.R. G. Dosch, Sandia National Labs., AlbuquerqueG. H. Dyer, Bechtel Corp.G. Eichholz, Georgia Inst. Technology0. J. Elgert, Richland Operations Office, USDOEB. Erdal, Los Alamos National Lab.E. D. Erickson, Rocky Flats Plant

D. Ferguson, Oak Ridge National Lab.Foster Wheeler Corporation, Library (HQAP)R. G. Geier, Rockwell Hanford OperationsE. S. Goldberg, Savannah River Operations Office, USDOE (2)S. Goldsmith, Battelle-Columbus Labs.D. Gordon, Savannah River Lab.

J. P. Hamric, Idaho Operations Office, USDOES. G. Harbinsen, San Francisco Operations Office, USDOEM. Harwell, Battelle Pacific Northwest Lab.C. A. Heath, Office of Nuclear Energy, USDOEL. L. Hench, U. FloridaT. B. Hindman, Jr., USDOE-SRB. F. Judson, General Electric Co., San JoseR. G. Kepler, Sandia National Labs., AlbuquerqueC. J. Kershner, Mound Lab.

F. J. Kiernan, Aerojet Energy Conversion Co., WashingtonJ. F. Kircher, Battelle-Columbus Labs.S. J. Lambert, Sandia National Labs., AlbuquerqueG. Lehmkul, Rocky Flats PlantW. H. Lewis, Nuclear Fuel Services, RockvilleR. C. Liikala, Battelle Pacific Northwest Lab.Los Alamos National Lab., DirectorA. L. Lotts, Oak Ridge National Lab.R. Y. Lowrey, Albuquerque Operations Office, USDOEB. M. Ma, Iowa State U.L. Machta, NOAA, Silver SpringR. Maher, Savannah River Plant

J. C. Mailen, Oak Ridge National Lab.W. J. Maraman, Los Alamos National Lab.A. B. Martin, Rockwell International, Canoga ParkM. L. Matthews, Nuclear Power Development, USDOED. J. McGoff, Waste Technology Branch, USDOED. L. McIntosh, Savannah River Lab.W. H. McVey, Office of Nuclear Energy, USDOER. E. Meyer, Oak Ridge National Lab.S. Meyers, Office of Nuclear Energy, USDOENASA, John F. Kennedy Space CenterR. D. Nelson, Battelle Pacific Northwest Lab. (2)G. K. Oertel, Office of Nuclear Energy, USDOED. A. Orth, Savannah River Lab.B. Paige, Allied Chemical Corp., Idaho FallsH. Palmour III, North Carolina State U.J. H. Pashley, Oak Ridge Gaseous Diffusion Plant

95

A. M. Platt, Battelle Pacific Northwest Lab.H. Postina, Oak Ridge National Lab.C. A. Preskitt, IRT Corp., San DiegoJ. J. Reilly, Brookhaven National Lab.J. F. Relyea, Battelle Pacific Northwest Lab.G. L. Ritter, Exxon Nuclear Co., Inc., Idaho FallsH. E. Roser, Office of Asst. Secretary for Defense Programs, USDOER. Roy, Pennsylvania State U.K. J. Schneider, Battelle Pacific Northwest Lab.R. L. Seale, U. ArizonaT. A. Sellers, Sandia National Labs., AlbuquerqueJ. Serne, Battelle Pacific Northwest Lab.B. Silva, Lawrence Berkeley National Lab.A. Squire, Hanford Engineering Development Lab.J. A. Stiegler, Sandia National Labs., AlbuquerqueS. Stoller, The S. M. Stoller Corp.G. Stukenbroeker, NL Industries, Wilmington, Del.J. L. Swanson, Battelle Pacific Northwest Lab.USDOE Div. of Nuclear Power Development, Nuclear Fuel Cycle Programs Br.USDOE Office of Basic Energy SciencesUSDOE Office of RRT, EngineeringUSDOE Office of RRT, TechnologyUSDOE Idaho Operations OfficeUSDOE New Brunswick LaboratoryUSDOE San Francisco Operations OfficeUSDOE Southern California Energy OfficeH. H. Van Tuyl, Battelle Pacific Northwest Lab.V. C. A. Vaughn, Oak Ridge National Lab.E. E. Voiland, General Electric Co., Morris, Ill.R. D. Walton, Jr., Office of Nuclear Waste Mgmt., USDOEC. D. Watson, Oak Ridge National Lab.L. L. Wendell, Battelle Pacific Northwest Lab.J. B. Whitsett, Idaho Operations Office, USDOEW. J. Wilcox, Oak Ridge Gaseous Diffusion PlantA. K. Williams, Allied-General Nuclear Services, BarnwellR. 0. Williams, Rocky Fiats PlantD. D. Wodrich, Rockwell Hanford OperationsJ. E. Yanoski, Office of Nuclear Energy, USDOED. Zeigler, Rocky Flats PlantArizona, U. of, Dept. of Nuclear EngineeringMichigan Technological U., LibraryD. W. Moeller, Kresge Ctr. for Environmental Health, BostonG. Murphy, Iowa State U.H. Rosson, U. KansasE. R. Stansberry, Purdue U.W. E. Wilson, Washington State U.W. F. Witzig, Pennsylvania State U.I. Neretnieks, Royal Inst. Technology, Stockholm, Sweden

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