NBS SPECIAL PUBLICATION

U.S. DEPARTMENT OF COMMERCE/National Bureau of Standards

Nuclear Waste

I Management

Program

Summary

Report
NATIONAL BUREAU OF STANDARDS

The National Bureau of Standards' was established by an act of Congress on March 3, 1901. The Bureau's overall goal is to strengthen and advance the Nation's science and technology

and facilitate their effective application for public benefit. To this end, the Bureau conducts

research and provides: (1) a basis for the Nation's physical measurement system, (2) scientific and technological services for industry and government, (3) a technical basis for equity in trade, and (4) technical services to promote public safety. The Bureau's technical work is per- formed by the National Measurement Laboratory, the National Engineering Laboratory, and

the Institute for Computer Sciences and Technology.

THE NATIONAL MEASUREMENT LABORATORY provides the national system of

physical and chemical and materials measurement; coordinates the system with measurement

systems of other nations and furnishes essential services leading to accurate and uniform

physical and chemical measurement throughout the Nation's scientific community, industry,

and commerce; conducts materials research leading to improved methods of measurement,

standards, and data on the properties of materials needed by industry, commerce, educational

institutions, and Government; provides advisory and research services to other Government

agencies; develops, produces, and distributes Standard Reference Materials; and provides

calibration services. The Laboratory consists of the following centers:

Absolute Physical Quantities 2 — Radiation Research — Chemical Physics —

Analytical Chemistry — Materials Science

THE NATIONAL ENGINEERING LABORATORY provides technology and technical ser- vices to the public and private sectors to address national needs and to solve national problems; conducts research in engineering and applied science in support of these efforts; builds and maintains competence in the necessary disciplines required to carry out this research and technical service; develops engineering data and measurement capabilities;

provides engineering measurement traceability services; develops test methods and proposes

engineering standards and code changes; develops and proposes new engineering practices;

and develops and improves mechanisms to transfer results of its research to the ultimate user. The Laboratory consists of the following centers:

Applied Mathematics — Electronics and Electrical Engineering2 — Manufacturing

Engineering — Building Technology — Fire Research — Chemical Engineering2

THE INSTITUTE FOR COMPUTER SCIENCES AND TECHNOLOGY conducts

research and provides scientific and technical services to aid Federal agencies in the selection, acquisition, application, and use of computer technology to improve effectiveness and

economy in Government operations in accordance with Public Law 89-306 (40 U.S.C. 759), relevant Executive Orders, and other directives; carries out this mission by managing the Federal Information Processing Standards Program, developing Federal ADP standards

guidelines, and managing Federal participation in ADP voluntary standardization activities; provides scientific and technological advisory services and assistance to Federal agencies; and

provides the technical foundation for computer-related policies of the Federal Government.

The Institute consists of the following centers:

Programming Science and Technology — Computer Systems Engineering.

'Headquarters and Laboratories at Gaithersburg, MD, unless otherwise noted; mailing address Washington, DC 20234.

: Some divisions within the center are located at Boulder, CO 80303.
CLtV < Nuclear Waste Management Program

Summary Report

(j c Workshop on Research and Development Needs ' ^

in Backfill for Long Term Storage of Nuclear Waste

in Deep Geologic Repositories, held at the National

Bureau of Standards, Washington, DC, April 13-14, 1981

Riley M. Chung, Editor

Center for Building Technology

National Engineering Laboratory

National Bureau of Standards

Washington, DC 20234

Prepared for:

Office of Nuclear Waste Management

Department of Energy

Washington, DC 20585

U.S. DEPARTMENT OF COMMERCE, Malcolm Baldrige, Secretary

NATIONAL BUREAU OF STANDARDS, Ernest Ambler, Director

Issued December 1983
Library of Congress Catalog Card Number: 83-600607

National Bureau of Standards Special Publication 668

Natl. Bur. Stand. (U.S.), Spec. Publ. 668, 228 pages (Dec. 1983)

CODEN: XNBSAV

U S GOVERNMENT PRINTING OFFICE

WASHINGTON: 1983

For sale by the Superintendent of Documents, U.S. Government Printing Office, Washington, DC 20402
Table of Contents

Page

Abbreviations of Organizations v Agenda of Workshop vi

Abstract ix

Introduction, Riley M. Chung, National Bureau of Standards 1

Papers presented to the Workshop and Published in this Report

PNL Backfill Development Project,

E.J. Wheelwright, F. N. Hodges, J. H. Westsik, Jr., and L. A. Bray

Pacific Northwest Laboratory 5

Design of Backfill (= Overpack) Materials, Rustum Roy and S. Komarneni

,

Pennyslvania State University 21

Backfill Barrier Research and Development for Radioactive Waste

Isolation in Salt (Abstract Only), E.J. Nowak, Sandia National

Laboratory 38

The Development and Testing of Waste Package Backfill Materials for a Nuclear Waste Repository Located in Basalt, M.I. Wood, Rockwell

Hanford Operation 39

Conditions Affecting the Backfill in a Waste Package for Repository

in Tuff (Abstract Only), J.K. Johnstone, Sandia National Laboratory:

and E.N. Vine, Los Alamos National Laboratory 57

Critical Factors Affecting the Use of Montmorillonite as a Nuclear

Waste Backfill Material, G.W. Beall, A.G. Lamkin, and K.L. Kelly,

Radian Corp 59

Selection and Evaluation of Buffer and Backfill Materials for Use in

a Nuclear Fuel Waste Disposal Vault: The Canadian Program,

G.W. Bird (Presented by D. Cameron), Atomic Energy of Canada Limited .. 66

iii
Table of Contents (Continued)

Page

Evaluation of Thermal Properties of Buffer Materials for a Deep

Underground Nuclear Waste Disposal Vault, H.S. Radhakrishna , Ontario

Hydro, Canada 87

Bentonite Shielding of Rock-Deposited Radioactive Wastes, Roland Pusch,

University of Lulea, Sweden 112

Paper Presented to the Workshop but withdrawn from Publication

Effect of Hydrothermal and Cyclic Wetting - Drying Treatment on Properties and Characteristics of Montmorillonitic Clay, A.J. Selti,

R.N. Yong, McGill University; and G.W. Bird, Atomic Energy of Canada

Limited, Canada

Paper not Presented to the Workshop but Included in this Report

Evaluation of Backfill as a Barrier to Radionuclide Migration in a High Level Waste Repository, T.M. Ahn, R. Dayal , and R.J. Wilke,

Brookhaven National Laboratory 125

Presentation of Results of Workshop Questionnaire, Riley M. Chung, NBS ... 149

Summary of Future Research and Development Needs, Riley M. Chung, NBS .... 190

Appendix A. Questionnaire Distributed During the Workshop A-2

Appendix B. Organizing Committee B-2

List of Participants B-3

iv
ABBREVIATIONS OF ORGANIZATIONS

rVLUllllL LjLlt-.Lgy Ul V_/d Lldtdd JjlUil LcU

ANT rtXgOLlLLfc; A r~ rr a n n o MQt"iAnQl INaLlUllal TXidUOLdLOLy o r\f"\ i"* o trvrxT RrnnVVia A7Pn Wot"! Anal T QK^rafnyu

crt/nrs VjCLIUCL 1 Ul UUJ.XUJ.Llg 1 CLllllUXUg Y ^ llaLlUllal UULCaU L/J_ i) LallUa 1 Uo

U W1—

i

\J t- IJ d JL L. lilt- L1L. OX IjUcL

LjrVO Li LiUo rxXdulO o OLlcULll 1L LdUUL a LU L y

T RT T a t»t LjaWlcLlLC t* o n r* o RDela y1/Nclcy d 1 qt.7 TLidq Katq UU L a fLULnr\f

y

T T T TLdwi 0 t.t fori pq T n T7 /^i v~ mr\ t~ a To r\ a v 0 t~ /*"\ t*tt cliCc Li X V c L lllO I tr Lid DO L d L O L y riaic jl ldi undLdCLcrizaLiori i<eriLci

LN D O iNd LXOIlcl X DU JVtidU OX jldllUdlub

MRP iNUC-LeaL Keguiacory ^ouimissiori

OMTJT U1N W J_ ui nee or iNucicdL waste lboidLion

ORWIYNTIN Li OWdrV 0 IV 1MU5C Pi' n cr d Molid fiL1ULnnaLd X1 TLiaUULaLUiy a Knra h attt PNL Battelle Pacific Northwest Laboratory

RHO Rockwell Hanford Operation

SNL Sandia National Laboratory

SRL Savannah River Laboratory

V
AGENDA

Monday, April 13, 1981

8:30 - 9:00 Arrival at NBS, Lecture Room B Administration Building

9:05 - 9:15 Welcome - Riley M. Chung, NBS

9:15 - 9:25 NBS Nuclear Waste Program - William P. Reed, NBS

9:25 - 9:35 Structures and Materials Programs of CBT/NBS,

Edward 0. Pfrang, NBS

9:45 - 10:10 Pacific Northwest Laboratory's Backfill Development Project,

E.J. Wheelwright, PNL

10:10 - 10:35 Design of Backfill (Overpack) Materials, Rustum Roy and

S. Kamarneni , Pennsylvania State University

10:35 - 11:00 Backfill Barrier Research and Development for Radioactive

Waste Isolation in Salt, E.J. Nowak , SNL

11:00 - 11:25 The Development and Testing of Waste Package Backfill Materials

for a Nuclear Waste Repository Located in Basalt, M.I. Wood,

RHO

11:25 - 1:00 Lunch at NBS Senior Lunch Club

1:00 - 1:15 Conditions Affecting the Backfill in a Waste Package for

Repository in Tuff, Part I, J.K. Johnstone, SNL

1:15 - 1:40 Conditions Affecting the Backfill in a Waste Package for

Repository in Tuff, Part II, E.N. Vine, LASL

v i
1:40 - 2:05 Critical Factors Affecting the Use of Montmorillonite as a Nuclear Waste Backfill Material, G.W. Beall , A.G. Lamkin,

and K.L. Kelly, Radian Corp.

2:05 - 2:20 Selection and Evaluation of Buffer and Backfill Materials for

Use in a Nuclear Fuel Waste Disposal Vault : The Canadian

Research program, G.W. Bird, AECL

2:20 - 2:35 Coffee Break

2:35 - 2:50 Evaluation of Thermal Properties of Buffer Materials for a Deep Underground Nuclear Waste Disposal Vault,

H.S. Radhakrishna, Ontario Hydro

2:50 - 3:15 Effect of Hydrothermal and Cyclic Wetting - Drying Treatment

on Properties and Characteristics of Montmorillonitic Clay,

A.J. Sethi, R.M. Yong , McGill University, and G.W. Bird, AECL

3:15 - 3:40 Bentonite Shielding of Rock-Deposited Radioactive Wastes,

R. Pusch, University of Lulea, Sweden

3:40 - 3:55 Coffee Break

3:55 - 4:50 Open Discussion and Adjournment, R.M. Chung, F.Y. Yokel, and

L.A. Salomone, NBS

Tuesday, April 14, 1981

9:00 Arrival at NBS, Building 226, CBT Conference Room

9:00 - 12:00 Informal Discussions on Backfill with Geotechnical Engineering

Staff

12:00 - 1:30 Lunch at NBS Senior Lunch Room

vii
1:30 - 3:00 Return to CBT Conference Room to continue discussion

3:00 - 4:00 Preparation of Written Recommendations, E.J. Nowak and

F.Y. Yokel

4 : 00 Adj ournment

viii
ABSTRACT

An international workshop on the backfill component of a multi-barrier

underground isolation system for nuclear waste disposal/storage was conducted

on April 13 and 14, 1981, at the National Bureau of Standards, Washington, D.C.

This report includes the papers presented and/or submitted to the workshop for

publication. It also contains the surveyed results of a questionnaire distributed to the participants at the workshop. The workshop held a one-day informal

in-depth discussion on the subject and a summary of these discussions is

included in this report, which states areas where further research and development needs are required to better understand the fundamental mechanisms of the

backfill in the waste repository.

Key words: backfill; compressibility; geology; nuclear waste; permeability

(hydraulic conductivity); research and development; shear strength;

sorption; swelling; thermal conductivity.

ix

INTRODUCTION

In 1979, the Interagency Review Group (IRG) Report, presented to then President

Carter, stated that there were about 76 million cubic feet of long lived radioactive nuclear wastes stored at various interim sites around the United States.

In addition to these wastes generated by military and civilian programs , there

are large quantities of uranium mill tailings, transuranic wastes, and spent

fuel discharged from commercial nuclear power reactors. Nuclear wastes will be

continuously generated by the military weapons and propulsion programs and by

the operation of over 65 nuclear power reactors and associated fuel service

facilities. Therefore, no matter how uncertain the future that the nuclear

industry is facing, nuclear wastes do exist and their environmentally safe disposal will remain as a critical public concern in both the States and overseas

.

Major classes of nuclear wastes include high level wastes (HLW), transuranic

wastes (TRU) , low level wastes (LLW), uranium mine and mill tailings, and gaseous effluents. The NBS program was directed toward high-level and transuranic

wastes which pose the greatest hazard to society, and thus require the greatest

technical input for their safe management and disposal.

A number of alternative methods for disposing of high-level wastes have been

proposed such as placing them into an orbit around the sun, transmutation by

neutron bombardment, or deposition into subseabed. In the IRG report these

wastes are recommended to be put into a stable form and disposed of by burial

in deep geologic sites using a multi-barrier system. As a simplified example,

the multi-barrier system consists of a glass waste form, a metal canister

containing the waste form which is lowered into a borehole drilled from the

floor of a tunnel in the repository, backfill placed between the canister and

the borehole wall, and the repository rock.

The glass is only one of the possible waste forms. Others including ceramics

and composites (e.g., synthetic rocks) have been considered. Metal canisters

made of stainless steel, copper, and titanium alloys were included in the

study; It is probable that the selections between waste package and geologic

1
settings are inter-related and the choice of one could influence the optimum

choice of the other. Thus, it is logical to tailor a waste package to the

requirements of the geohydrologic environment of a geologic site.

Up to the present (1981), geologists have not been able to reach a consensus on

the ability to adequately predict long-term geologic stability. For this

reason, the Nuclear Regulatory Commission (NRC) is currently (as of 1981)

proposing regulations which rely heavily on the selection of waste form and the

design of a total waste package for providing a safe geologic storage/disposal

system. The proposed NRC regulations focus on four different time periods for

the management of both high-level and TRU wastes. The first time period

includes waste processing, transportation, and repository operation. The

second time period, which is 50 to 100 years after waste emplacement, is the

interval proposed by NRC for possible retrieval of waste from the repository

if difficulties are detected. The third (up to 1000 years after emplacement)

and fourth (beyond 1000 years) time periods follow from ingestion toxicity of

fission products as a function of decay. For the third time period, the proposed NRC regulations would require that there be reasonable assurance that

there will be no release of radionuclides from the waste package for the 1000

years time period. The proposed regulations would require the release rate

for radionuclides not to exceed lO -

-* of its radioactivity per year for the

fourth time period beyond 1000 years.

In view of the above consideration, a great deal of new scientific and

engineering information is needed to arrive at a technical public consensus.

A large technical program has been developed by the Department of Energy (DoE)

to provide the necessary information for waste containment and the program

ranges from scientific research to building processing plants and studying

candidate repositories. It is conceivable that the credibility of the properties measurements program of DoE to the regulatory agencies of the government,

the scientific community, and the general public will be based upon a referenceable basis for test methods and long term predictive capability to determine

materials and systems performance. And it is in this capacity that NBS was invited by DoE in 1979 to participate in the national nuclear waste storage/

disposal program.

2
The NBS program covers all the components of the multi-barrier system except

the repository rock. The study of the backfill component was assigned to the

Geotechnical Engineering Group of the Structures Division in Center for Building

Technology. The objectives of the NBS backfill program can be summarized as follows

• To formulate performance criteria against which backfill materials can

be evaluated.

• To develop or adopt methods for measuring the performance of backfill

materials

.

• To evaluate the consistency and accuracy of available data on backfill

materials

.

• To select and characterize research materials.

• To determine, and to the extent possible, standardize the tests by which

backfill materials are characterized and evaluated.

One major activity in our first year program was to conduct an international

workshop with respect to the specific needs of the backfill in a waste package

geohydrologic environment. This led to the gathering of thirty-eight people on April 13th and 14th, 1981 at the NBS Campus. Experts from national laboratories

and universities as well as from Canada and Sweden participated in the program.

On the first day, eleven papers were presented. The second day was devoted to

informal in-depth discussions on the subject. Discussions were focused on the

current state of backfill evaluation and the research and development needs in

research materials, test variables, test methods and test equipment, and

fundamental mechanisms of the backfill in a repository environment.

This report includes the papers presented to the workshop and also one paper

which was submitted later for publication. Unfortunately, three papers were withdrawn by the authors because of the lack of time for completing the manuscripts. The report also presents the surveyed results of a questionnaire

distributed to the participants at the workshop. Finally, a summary of the

3
discussions generated during the workshop states areas where further research

and development needs are proposed to better understand the fundamental mechanisms of the backfill in a waste repository; thus leading to a better prediction

of the long term behaviors of the backfill material.

4
PNL BACKFILL DEVELOPMENT PROJECT^ 1 )

E. J. Wheelwright

F. N. Hodges

J. H. Westsik, Jr.

L. A. Bray

Pacific Northwest Laboratory

Richland, Washington 99352

ABSTRACT

Preliminary results of the first year backfill development project of the

Pacific Northwest Laboratory (PNL) were presented. A consultant committee

which consists of five members was formulated at the inception of the project.

All of them are from outside the nuclear community with complementary areas of

expertise. The desired material properties of backfill were evaluated and its

functions were divided into two categories: the primary and supporting attributes. Primary attributes are those which can have a direct effect upon one

or more nuclide transport mechanisms. The supporting attributes do not directly

affect a nuclide transport mechanism but provide support to the primary

attributes. Of course, there is some overlap and interdependency

.

A test program has been formulated. The materials selected for study include

Nabentonite, Ca-bentonite , and illite with quartz sand mixtures. Some small

amount of other materials will also be added to the mixture to study the sorption properties of anions or for Eh/Ph control. The test program will address

several areas including compaction of materials, measurement of hydraulic con- ductivity, measurement of ion retardation, measurement of thermal conductivity,

determination of mechanical properties, and determination of chemical stability

under repository conditions. Only the test results of compaction and hydraulic

conductivity are presented in this paper. The compaction density has been

shown to be a function of compaction pressure and the water content of the

U Work Sponsored by Office of Nuclear Waste Isolation.

5
sample during compaction. Proper degree of hydration of the clay materials is

important in order to obtain maximum density. Hydraulic conductivity measurements are being carried out in a test set-up developed at PNL. Preliminary

test results indicate that values of hydraulic conductivity of less than

10~12 cm/sec can be expected from pure Na-bentonite clay. It should be pointed

out that all hydration conductivity tests were conducted under rather high

hydraulic gradients which were in the range of 3.5 x 10^ to 2.1 x 10^. These

hydraulic gradients are considered much higher than those the backfill will be

actually subjected to in the field. Extrapolation of experimental data to the

actual gradient region will thus overestimate the flow. Therefore, the

laboratory results on hydraulic conductivity are conservative.

INTRODUCTION

Concepts currently under investigation for the disposal of spent nuclear fuel

or separated high-level nuclear waste involve deep emplacement of the waste in

continental geological formations. In such disposal concepts, primary canisters

or primary canisters contained within one or more overpacks are placed in an oversized hole bored into the geologic medium, and the space between the surface of the canister or overpack and geologic medium is backfilled with a suitable material or combination of materials. In such a concept, the waste form,

the canister, the overpack, the backfill, and the geology itself each constitute

a barrier to transport of the radioactive materials to the biosphere. The use

of multiple, well-engineered, independent barriers significantly increases the

probability of containment for the required time. We recognize that PNL is not

the first to recognize the value of using an engineered backfill and we do want

to make maximum use of the work of others.

The Backfill Development Project at PNL is sponsored by the Office of Nuclear

Waste Isolation (ONWI). The project has been funded for just over 12 months.

UinWI has directed PNL to develop backfill materials for use in repository systems in basalt and tuff. We will do a very limited number of tests of material

in salt. The current PNL program is limited to materials development and

testing.

6
SCOPE

The project scope is given below. The first two items are largely completed

and our attention is now focused on the third item.

o A definition and prioritization of the desired attributes of the

backfill material-quantitative specifications where possible.

o A screening of potential backfill materials and the selection of a manageable number of candidates, i.e., 6 to 12, for laboratory evaluation. This will include multi-component material to satisfy multifunctional requirements.

o Laboratory evaluation of the candidate backfill

measuring their performance against the defined

COMMITTEE OF CONSULTANTS

At the inception of the project, a consultant committee was formed by selecting

people of recognized stature. A deliberate effort was made to secure experts

from outside the nuclear community with complementary areas of expertise. The

committee consists of five members and their names and affiliations are given

in table 1. Dr. Lindsay's interests are in the application of electrochemistry

to soils, minerals and groundwater. Dr. Low is preeminent in the area of claywater interactions. Dr. Berner's areas of interest include geochemical processes and rock-water interactions. Dr. Lerman's field of interest is geochemical modeling. Dr. Rollins was selected because of his engineering experience

in the structural aspects of clays.

The committee has functioned to review our research plans, and to review and

comment on published work from other backfill programs. They will critique

our experimental results and conclusions and we hope that the work will result

in some jointly authored publications.

materials aimed at

requirements

.

7
FUNCTIONS OF BACKFILL

On the basis of a literature search and a brainstorming session with the

consultant committee, the list of backfill attributes was selected as shown in

table 2. The list is divided into primary and secondary attributes. Primary

attributes are those which can have a direct effect upon one or more nuclide

transport mechanisms. The supporting attributes do not, by themselves, directly

effect a nuclide transport mechanism but provide support to the primary

attributes. There is, of course, some overlap and interdependency

.

MATERIALS

The materials selected for study are shown in table 3. The principal materials

are Na-bentonite-Quartz sand mixtures. Tests are planned in which small percentages of other materials will be added in an attempt to enhance specific

additional properties, i.e., absorption of 129j or Eh/pH control, without

seriously detracting from the primary function of providing a barrier to groundwater migration. Calcium-bentonite and illite are included in the list so we

can study the effects upon hydraulic conductivity of the conversion of

Na-bentonite to one of these phases.

TESTING PROGRAM

Experimental work has been initiated in the four areas shown in table 4. Most

comments will be made about work in the first two areas.

Compaction of Materials

Clays, quartz sand, zeolites, minerals, and metal powders are being mixed in

various ratios and pressed in an air-hydraulic press to study the relationship

of compaction density to material composition and pressing conditions. The

objective is to obtain a backfill material composition that will maintain its

physical shape during placement, and will then prevent water intrusion, provide

enhanced radionuclide sorptive capacity, provide adequate thermal transport,

and buffer the Eh-pH of the backfill environment.

8
The results of compaction experiments with undiluted Na-bentonite (Volclay CS502/ ) are shown in figures 1 and 2. The solid lines in figure 1 show the relationship between applied pressure and final density for Na-bentonite containing

differing amounts of absorbed water. A reference point of 0.0 wt percent

absorbed water was assumed for clay that was heated at 110°C for 16 hours.

Vacuum drying at the same temperature for the same time gave comparable dehydration. On this basis, the as-received clay contained 7.5 wt percent absorbed

water. Clays of higher hydration were obtained by equilibration with water

vapor in a sealed system. After the pressed pellets were weighed and measured,

each was dehydrated by heating in an oven at 110°C for 16 hours. The dehydration results are shown by the dashed lines in figure 1. The four pellets

pressed from previously dehydrated clay powder did not change during the pellet

dehydration step.

In figure 2, pressed pellet density is shown as a function of the water content

of the unpressed clay for compaction pressed at four different pressures. The

importance of using clay with the proper degree of hydration is clearly shown.

Measurement of Hydraulic Conductivity

A schematic of the equipment assembled for measuring hydraulic conductivity of

compacted materials is shown in figure 3.

Details of the constant-volume flow through cells are shown in figure 4. The

backfill material to be tested is compressed inside the heavy-wall cylinder by

a hydraulic press in a separate system. When assembled as shown, the tie bolts

hold the components together and resist expansion of the clay as it is hydrated.

The difference in pressure on the load cell before and after pumping gives the

clay swelling pressure.

Preliminary results from experiments completed and still in progress are shown

in table 5. In these tests, the materials were compacted at 138 MPa. Synthetic

Registered trademark of American Colloid Company, Skokie, Illinois.

9
Grande Ronde (Hanford basalt) groundwater was used as the permeating fluid. As

the ratio of clay to sand is increased, the conductivity decreases by several

orders of magnitude to less than 10 - ^ cm/sec for pure Na-bentonite clay. The

hydraulic conductivity of the 25 wt percent clay-75 percent sand mixture shows

a decrease by a factor of two over the 189 hour period following the initial

breakthrough.

Because of the low hydraulic conductivities of these backfill materials,

experimental hydraulic gradients must be significantly higher than those

expected in a geologic repository in order to obtain data in a reasonable time

period. The data given in table 6 shows that for a 50 percent clay-50 percent

sand mixture, the hydraulic conductivity is reasonably constant over a hydraulic

gradient range 3.5 x 10 4 to 2.1 x 105. For these gradients the backfill materials appear to follow Darcy's Law. Deviations from Darcy's Law are not

expected until the hydraulic gradient drops below 100. Extrapolations of

experimental data to this region will usually overestimate the flow. Therefore,

the laboratory results are conservative.

10
Table 1

DBM Consultants

Dr. W. L. Lindsay Colorado State University

Dr. P. F. Low Purdue University

Dr. R. A. Berner Yale University

Dr. A. Lerman Northwestern University

Dr. R. Rollins Brigham Young University

11
Table 2

Backfill Attributes

Primary

o Exclude groundwater from the waste canister

o Retard migration (immobilization) of selected radionuclides

o Control groundwater pH and Eh

Supporting

o Be compatible with geology and other engineered barrier materials

o Self-seal minor cracks or fractures in backfill and interfacing geology

o Retain performance criteria at all repository temperatures

o Retain performance criteria during and after gamma radiation

o Maintain mechanical stability

o Provide adequate thermal conductivity

12
Table 3

Clay

Sand

Zeolites

Metal powder or fibers

Minerals /Rocks

Backfill Materials

Na-bentonite

Ca-bentonite

Illite

Treated Na-bentonite

Si0 2 (10-230 mesh)

Clinoptilolite

13X

Zelon 900

Iron

Aluminum

Lead Oxide

Pyrite (FeS2)

Ferrosand

Basalt

Tuff

13
Table 4

Current Materials Testing Program

o Compaction of materials

o Measurement of hydraulic conductivity

o Measurement of ion retardation

o Measurement of thermal conductivity

14
Table 5

Hydraulic Conductivity of Clay/Sand Mixtures

Percent

Na-Bentonite

Initial Hydraulic Time to Migrate

Percent Sand Density Conductivity One Centimeter

(80-120 Mesh) (g/cm3) (cm/sec) (years)

25 75 2.1 2 x 10-9* 15

50 50 2.1 3 x 10" 12 11,000

100 2.1 7 x 10-13 45,000

Hydraulic conductivity decreasing with time

15
Table 6

Pressure Effects on Hydraulic Conductivity of 50-50 Clay/Sand Mixture

Hydraulic Equivalent Time at

Gradient Pressure Drop Pressure Total Time

(psi) (hours) (hours)

Hydraulic

Conductivity

(cm/ sec)

1.55 x 105

7.03 x 104

3.52 x 10 4

7.03 x 10 4

2.11 x 10 5

1.58 x 10 5

2200

1000

500

1000

3000

2250

169*

72

62

72

97

334

169

241

309

381

478

812

4.1 x 10~ 12

3.5 x 10~ 12

3.1 x 10 -12

3.1 x 10 -12

2.9 x 10" 12

* Includes 64 hours before initial water breakthrough

16
2.3

2.2

2.1

COE

o

>-

2.0

1.9

1.8

DRY SHRINKAGE = 1.3%

DENSITY CHANGE = 4%

5.8 wt% H 2 0

7.5 wt% H 2 0

DRY SHRINKAGE = 0.2%

DENSITY CHANGE = 5%

13.5 wt% H 2 0

'0 wt% H 2 0

DRY SHRINKAGE = 3%

DENSITY CHANGE = 5.5%

VOLCLAY, CS-50

2.86 cm DIA

15 g CLAY — HELD AT COMPACTION PRESSURE

FOR 1 MINUTE

AFTER MEASURING & WEIGHING,

THE PELLETS WERE PLACED IN AN OVEN AT 110°C FOR 16 HOURS

100 200

COMPACTION PRESSURE (MPa)

300

Figure 1. Density of Na-Bentonite as a function of compaction pressure and

water content

17
0 5 10 15

WEIGHT PERCENT WATER

Figure 2. Density as a function of compaction pressure and H2O content

18
RESERVOIR

BACK PRESSURE

RELIEF VALVE

PUMP

5000 psi

0.01 ml/min

2

CAPILLARY

TUBE

PERMEABILITY

CELL

PRESSURE

TRANSDUCER

Figure 3. System set-up for hydraulic conductivity testing

19
Figure 4. Schematic drawing of equipment for hydraulic conductivity

measurement

20
DESIGN OF OVERPACK (= BACKFILL) MATERIALS

Rustum Roy and S. Komarneni

Materials Research Laboratory

The Pennsylvania State University

University Park, PA 16802

ABSTRACT

Research on backfill (= overpack) design at The Pennsylvania State University

has focussed on the stability of such materials under repository conditions,

the interactions of these materials and waste under repository conditions, and

the sorption and fixation properties of backfill materials. Montmorillonite

among clay minerals and mordenite and clinoptilolite among zeolites have been

identified as the best backfills based on the present studies. Areas for

further research include thermal stability studies of backfill materials under

repository conditions in conjunction with natural analogs, waste-backfill

interactions and design of backfill materials for purposes such as maintaining

a reduced environment and keeping the repository dry.

INTRODUCTION

For most of the life of the research and development program for nuclear waste

management, the role of the waste form—the source of radionuclides—was considered secondary if of any significance at all. As late as in the WIPP documents,

absurd "assumptions" were made such as the total 'dissolution' of the waste

forms. It does not, therefore, come as any surprise that the concept of the

entire waste package itself as a subsystem of redundant barriers has had a difficult time gaining acceptance and understanding. Since 1973 we have advocated this approach (7), and figure 1 shows a late model of our concept of the

waste package.

We have consistently used the term "overpack" for the silicate-oxide material

to be packed around the canister; however, since different terminology has

21
been adopted in this conference, we shall henceforth use the term "backfill"

for what in all our referenced work is called "overpack."

The roles of a backfill barrier in any repository can be envisaged as follows:

(a) to selectively adsorb and fix (retain) radioactive ions, in the case of a leakage; (b) to interact with waste to form new crystalline phases under repository conditions; (c) to maintain a reducing environment; (d) to keep the

water from reaching the canister. The composition and structure of backfill

materials should vary depending upon whether salt or other geological formations

are used. There is, however, very little published work on the design of backfill. The Swedish KBS study in 1977 (1) was the first to propose a specific

tailor-made set of materials. Research on backfill design at The Pennsylvania

State University has examined several facets of the materials problem: (a) hydro

thermal stability of the minerals under repository conditions (3); (b) interactions between waste and backfill under repository conditions (4, 6); (c) high

sorption and fixation capacity for radwaste ions by materials treated in different ways (2). In the following we present first a summary account of the

several preliminary research projects on backfill conducted at The Pennsylvania

State University. Then we analyze the role of the backfill in the evaluation

of waste forms.

TRANSFORMATION OF CANDIDATE BACKFILL MATERIALS UNDER RADWASTE REPOSITORY

CONDITIONS

Although there will no doubt eventually be a wide variety of repositories for

different wastes (defense and commercial), of different ages, in different

countries, there is no doubt that mild "hydrothermal" conditions will exist in

a nuclear waste repository during the thermal period. Such exposure to T, pH.20,

etc. , can significantly alter the minerals in a backfill prior to leakage of

radioactive ions because the backfill is in the immediate vicinity of the

nuclear waste containing canisters. If the backfill materials are easily

altered (i.e., thermally unstable), the very purpose of using them may be lost.

Therefore, several candidate backfill materials have been screened for their

stability (3). Clay minerals such as montmorillonites and vermiculites and

zeolites such as clinoptilolite , chabazite, phillipsite, mordenite, and erionite

22
which are candidate backfill materials were treated under hydrothermal

conditions of 200 and 300°C with a pressure of 30 MPa for 28 days. Results of

X-ray diffraction (XRD) analysis revealed that montmorillonites did not seem

to alter while vermiculites were partially chlorotized by hydrothermal treatment at 200 and 300°C. Natural zeolites excepting phillipsite did not seem to

have transformed at 200°C but transformed to various extents at 300°C. Selective sorption of Cs and Sr decreased to different degrees in all the above

minerals excepting two montmorillonites and mordenites after hydrothermal

treatment because of the nature and extent of mineral transformation under

these hydrothermal conditions as revealed by XRD and Cs and Sr sorption measurements. For example, Cs sorption (ml/g) values decreased from 8100 to 1000

and 4600 to 2800 in erionite and chabazite respectively, while they decreased

drastically from 2600 to 90 and 3900 to 70 in vermiculite and phillipsite

respectively after hydrothermal treatment at 300°C (table 1). Montmorillonites

among clay minerals and mordenites among the zeolites studied here seem to be

very resistant to hydrothermal alteration and therefore may be preferred in

a backfill on this account . These results point out that the effects of temperature and water pressure on mineral transformations must be experimentally

determined before the selection of proper backfill materials is made. A study

of mineral alteration under natural conditions of burial diagenesis or metamorphism may give additional clues in the selection of proper backfill materials. While we have examined several candidate materials in silicate rock

environments in a preliminary way, the stability of backfill materials in the

brine of a salt repository has yet to be so investigated. No detailed studies

have appeared.

SORPTION OF IONS BY CLAYS, ZEOLITES, AND THEIR MIXTURES

Any material that is to be used as a backfill should have a maximum useful

capacity for sorption and fixation of the specific sets of radwaste ions during

and after the hydrothermal period. This presents the clay and zeolite mineralogist with an intriguing challenge. What is the nature of the adsorption equilibria between a clay + zeolite when several ions are present together? Reality

is even more complicated; and we need to study such equilibria with both multiple ions and also with multiple adsorbent phases present. In our study along

23
these lines (2), we measured on several clays, zeolites, and oxides for

their sorption and fixation properties, individually and in various mixtures,

using a simulated waste solution (table 2). These studies show that mixtures

of such phases cannot be treated as simply additives; both synergistic and

negative interactions are observed. Among those studied mixtures, zeolites

such as clinoptilolite plus mordenite or clinoptilolite plus phillipsite are

the best for sorption of ions. Phillipsite, however, is very unstable (table 1)

and therefore might not be suitable where fixation is important and the material

is heated in the repository prior to adsorption.

Backfill materials in a salt repository should have high specific sorption in

brine in order to be effective. Zeolites were found to be the most effective

for Cs uptake from saturated NaCl solutions as shown in table 3.

THE QUESTION OF "FIXATION" OF IONS

These same backfill materials also serve as adsorbents in waste water clean-up

and possibly as precursors to a final waste form. Hence it is of considerable

interest to determine how strongly ions which are adsorbed on these phases are actually held on the phase. For example, certain zeolites used in clean-up at

TMI would release 100 percent of their Cs to groundwater, while others may not.

Furthermore, simple heat treatments could radically alter the fixation of the

radionuclides by the adsorbent. Fixation of Cs was measured by extraction in

0.1N KC1 because K is a similar ion to Cs and is effective in replacing Cs.

Our study of these phenomena shows that there exists an enormous range in the

fixation by rather similar zeolites. Table 4 summarizes these results. It

shows, for example, that while analcite has a low exchange capacity, it will

fix virtually 94 percent of the Cs by heating to 300°C. On the other hand, the

zeolite actually being used in Harrisburg, is one of the poorest for fixation,

since even after heating at 800°C almost 89 percent of the Cs washes out. Concentration of the radionuclide is also an important variable. At the maximum

loading level of zeolites at TMI, the Cs can be adsorbed on phillipsite and the

latter heated to (a maximum of) 1000°C to make an excellent radiophase with

nearly 100 percent fixation. This, if incorporated into a concrete matrix,

could make an effective waste form having the same range of leachability as glasses.

24
INTERACTIONS OF WASTE ELEMENTS WITH POTENTIAL BACKFILL MATERIALS UNDER

REPOSITORY CONDITIONS

When we first pointed out the absolute necessity for evaluating waste package

performance by simulating repository conditions and studying what we called

"waste-rock interaction" (8) , we used this shorthand expression principally to

stress the fact that the simple leach tests in water were of no significance

whatsoever in determining the fate of the waste form in a repository environment after container failure. It was essential we argued as we do now, to take

the first order approximation (that given the range of volumes of fluid flow

projected) that the compositional system which will be brought to near- equilibrium will be waste-container-overpack (= backf ill)-rock-fluid (WCORF).

Thus while the waste-rock-fluid research studies now in full swing in many

institutions, including our own laboratory, are ultimately much more informative

regarding the limits of what can happen at the waste package edge, they do not

represent the best approximation to reality at all.

Since the volume and mass of the backfill can be at least as large as that of

the waste form, and the canister may be a reasonable fraction thereof, it is

quite certain that the WCORF system should be studied as a whole. White et

al. (10) have indeed launched a major systematic study of several waste forms,

several containers and several rock types. If one considers silicate host

rocks and corrosion resistant ceramic or metallic canisters, then the ternary

systems of greatest interest are certainly those consisting of combinations of

various overpacks (= backfills), various waste forms and water. Thus a series

of waste-container-overpack interaction studies are obviously of vastly greater

significance for waste package evaluation than "leach tests." We have begun in

a preliminary way to study ternary waste-overpack-f luid interaction under simulated repository conditions for various waste forms. Thus Cs , Mo and U ions

in CS2M0O4 and 3-CS2U2O7 phases of spent fuel elements have been shown to react

with potential backfill materials to form pollucite, powellite, and uraninite

respectively (5, 6). Figure 2 shows the concentration and time dependence of

the reaction. Figure 3 shows the reaction of clinoptilolite with sintered

ceramic to form analcime containing Cs (9). The reaction of waste elements

with backfill barriers will of course change with the nature of groundwater

25
(table 5) if the latter is a concentrated solution. Backfill materials fix the

largest percentages of Cs in deionized water by forming pollucite or analcime

(figures 2 and 3), but. not in bittern brine. However, montmorillonites , vermiculites, and zeolites are good fixers of Cs, both in deionized water and bittern

brine. Since montmorillonites and some zeolites are quite stable under

repository conditions (table 1) they may be preferred for a backfill.

The various lines of the authors' preliminary research reported herein support

the following general theses:

o That compositional tailoring of the backfill (= overpack) materials is

without doubt the most cost effective way of improving the waste package performance. Compared to the complexity of remote and high temperature processing required for making glass or ceramics, the normal cold manipulations of

composition and emplacement of backfill are of trivial cost.

o That zeolite and clay minerals and their interactions in the mild

hydrothermal regime with the principal radionuclide threats in radwaste will

provide the data to assess the limits of what the backfill can do to immobilize the radionuclides in new radiophases not present in the original

waste form.

o That waste package evaluation will only be meaningfully simulated from

geochemical-petrological views when the waste-canister-overpack-rock-f luid

interactions have been studied. These kinds of data should shortly eliminate

the entire family of "leach tests" as totally meaningless.

ACKNOWLEDGEMENTS

The studies of waste-rock interaction and the waste-canister-rock stability

referred to in this work, have been supported by the U.S. Department of Energy

through sub-contract E512-3400 with the Office of Nuclear Isolation, Battelle.

Overpack research has been supported by the Pennsylvania Research Corporation,

The Pennsylvania State University.

26
REFERENCES

1. Jacobson, A., and R. Pusch, "Deposition of high-level radioactive waste

products in bore-holes with buffer substance," KBS Teknish Rapport 03,

Stockholm (1977).

2. Komarneni, S., and R. Roy, "Superoverpack : Tailor-made mixtures of

zeolites and clays," in Scientific Basis for Nuclear Waste Management,

Vol. 2, C. J. Northrup, Jr. (Ed.), pp. 411-418 (1980).

3. Komarneni, S., and R. Roy, "Hydrothermal transformations in candidate

overpack materials and their effects on Cs and Sr sorption," Nucl. Tech .

,

54: pp. 118-122 (1981).

4. Komarneni, S., and R. Roy, "Interactions of Backfill Materials with

Cesium in a Bittern Brine Under Repository Conditions," In Preparation

for Nucl. Tech. , 56: pp. 575-579 (1982).

5. Komarneni, S., and B. E. Scheetz, "Hydrothermal Interactions of Basalts

with Cs and Sr of Spent Fuel Elements: Implications to Basalt as a Nuclear Waste Repository," J. Inorg. Nucl. Chem. , 43: pp. 1967-1975

(1981).

6. Komarneni, S., and W. B. White, "Hydrothermal Interactions of Clay

Minerals and Shales with Cesium Phases from Spent Fuel Elements," Clays

& Clay Minerals , 29: pp. 299-308 (1981).

7. Mendel, J. E., J. L. McElroy and A. M. Piatt, "High-Level Radioactive

Waste Management Research and Development Program at Battelle Pacific,"

Adv. in Chem. Ser. 153, M. H. Campbell (Ed.), Amer. Chem. Soc, Washington,

DC, pp. 106-107 (1976).

27
8. Roy, D. M. , et al., "Borehole Cement and Rock Properties Studies," Annual

Progress Report to Office of Nuclear Waste Isolation, Union Carbide

Corporation, Oak Ridge, TN 37830, Y/OWI/SUB-4091 /6 (1977).

9. Sasaki, N. , S. Komarneni , B. E. Scheetz, and R. Roy, "Backfill-Waste

Interactions in Repository Simulating Tests," Scientific Basis for Nuclear

Waste Management, Vol. 6, S. V. Topp (Ed.), North-Holland, New York, pp.

397-404 (1982).

10. White, W. B. , W. P. Freeborn, S. Komarneni, D. S. Minser and G. D. Cox,

"Chemical Interactions of Shale Rock, Prototype Waste Forms, and Prototype

Canister Metals in a Simulated Waste Repository Environment," Proceedings

of the 1980 National Waste Terminal Storage Program Information Meeting,

ONWI, Columbus, OH (1980).

28
Table 1

Effect of Hydrothermal Treatment on Cs Sorption by Candidate Backfill Materials

Sample name, state

Cs sorption, (ml/g)

200°C, 30 MPa 300°C, 30 MPa

Untreated 28 days 28 days

Montmorillonites

Na - montmorillonite , WY

Ca - montmorillonite, TX

Ca - montmorillonite, AZ

150+20

190+20

590+60

140+10

200+20

510+50

170+20

160+20

280+30

Vermiculites

Poole verraiculite, SC

Williams vermiculite, SC

2600+300

2900+300

60+10

100+10

90+10

60+10

Zeolites

Phillipsite, CA

Chabazite, AZ

Erionite, CA

Erionite, OR

Clinoptilolite , CA

Clinoptilolite , ID

Mordenite, AZ

Mordenite, NV

3900+400

4600+500

8100+800

4300+400

2900+300

3800+400

4600+500

4100+400

3200+300

4800+500

6900+700

3200+300

3500+400

4200+400

5100+500

70+10

2800+300

1000+100

3600+400

2600+300

3200+300

4500+500

4800+500

+ Denotes 90% confidence limit on the mean.

29
Table 2

Sorption* and Fixation* of Ions from Simulated Waste Solution by

Individual and Mixtures of Minerals

Total mg % Cs Total mg % Rb Total mg % Sr

Sample of Cs fixed of Rb fixed of Sr fixed

sorbed/

g

sorbed/

g

sorbed/

g

Conasauga shale, TN 1. 56 23 0. 168 45 0.330 0

Montmorillonite

,

AZ 4. 24 33 0.378 21 0.913 0

Vermiculite, SC 4. 76 65 0.218 55 1.213 0

Gibbsite, heated 0. 00 — 0.000 — 0.675 0

Clinoptilolite

,

AZ 5. 35 81 0.643 77 1.490 26

Clinoptilolite

,

ID 5. 35 82 0.648 82 1.590 75

Mordenite, AZ 5. 43 92 0.558 61 0.618 53

Mordenite, NV 5. 37 89 0.558 69 0.410 70

Erionite, OR 5. 28 89 0.650 63 0.550 7 Erionite, CA 5. 51 91 0.648 75 1.346 9

Chabazite, AZ 5. 47 88 0.598 68 0.945 0

Phillipsite, CA 5. 44 89 0.663 94 0.775 0

Linde, 3A 2. 40 38 0.073 0 1.778 87

Linde , 4A 4. 27 70 0.323 40 1.783 70

Linde, 5A 5. 06 69 0.345 41 1.670 30

Mixtures of

Natural Zeolites

Clinoptilolite, ID +

Mordenite, AZ 5.34 85 0.598

Clinoptilolite, ID +

Mordenite, NV 5.32 83 0.623

Clinoptilolite ID +

Erionite, OR 5.39 82 0.605

63

66

66

1.428

1.541

1.445

80

80

68

30
Table 2 (Continued)

Total mg % Cs Total mg % Rb Total mg % Sr

Mixtures of of Cs fixed of Rb fixed of Sr fixed

Natural Zeolites sorbed/

g

sorbed/

g

sorbed/g

Clinoptilolite, ID +

Erionite, Ca 5.42 82 0.635 67 1.590 68

Clinoptilolite, ID +

Chabazite, AZ 5.33 79 0.620 63 1.482 62

Clinoptilolite, ID +

Phillipsite, CA 5.34 83 0.655 84 1.532 72

Erionite, CA +

Mordenite, AZ 5.39 89 0.605 62 1.242 47

Erionite, CA +

Mordenite, NV 5.44 86 0.628 67 1.330 52

* Sorptions of Cs , Rb, and Sr were measured from a simulated waste solution

prepared by dissolving PW-4b waste (Komarneni and Roy, 1980). Fixation here

is simply the percent remaining in solid after 1 extraction for 24 hours with

15 ml of 0.1N NaCl + 0.1N CaCl2 solution.

31
Table 3

Cs Sorption by Backfill Materials from Saturated NaCl*

Cs sorption

Sample name Kd (ml/g)

Montmorillonite , WY 6

Montmorillonite , TX 6

Montmorillonite, AZ 12

Vermiculite, Poole, SC 0

Vermiculite, Williams, SC 2 Clinoptilolite, ID 14

Clinoptilolite, CA 18

Phillipsite, AZ 24

Chabazite, AZ 28

Mordenite, AZ 30

Contained 8 ppm Cs, 4 ppm La, 4 ppm Nd and 2 ppm Sr.

Sorption of Sr, La, and Nd could not be detected.

32
Table 4

Fixation of Cs in Cs-Loaded Zeolites After Heat Treatment as Determined by One 0.1N KC1 Extraction

Exchangeable Cs from Cs Fixation, %

unheated zeloites with

Zeolite one 0.1N KC1 extraction,

meq/100 g 350°C 600°C 800°C

Clinoptilolite , ID 143 0.0 0.0 0.0

Mordenite, AZ 114 0.0 0.0 7.9

Chabazite, AZ (similar

to Ionsiv IE 95

proposed for TMI) 137 0.0 0.0 11.3

Erionite, CA 131 6.6 11.4 16.8

Phillipsite, CA 171 0.0 85.0 89.9

Phillipsite, NV 197 0.0 69.1 87.2

Analcite, AZ 6 94.0* 99.5 99.8

* Heated to 300°C instead of 350°C.

33
Table 5

flnaiyoco 01 i>3 in Solution and XRD Characterization or Reaction

Products* in Water vs Brine

*

Total

mg of Cs Cs uptake Pollucite Cs uptake Pollucite

Backfill Barriers added/ in deionized formation in bittern formation

100 mg water, % by XRD brine, % by XRD

sample

Shales

Conasauga 15.2 82.1 V.Stl 6.8

Salona 13.1 55.7 M 11.0 ND

Basalt Phase

Labradorite 14.8 99.7 V.St 9.7 Kin

Clay Minerals

Illite 17.9 31.7 SI 12.0

Mica 12.8 22.3 ND 9 . O ND

Montmorillonite , AZ 17.4 44 . 4 ND 46.7 ND

Vermiculite, Poole 1 7 /, 1 / . H 93.6 ND

Vermiculite, Williams 17.4 97.2 Kin

Zeolites

Phillipsite, CA 17.4 50.0 Tr

Mordenite, AZ 17.4 39.6 ND

Clinoptilolite, ID 17.4 30.6 Km

* Hydrothermally treated at 200'C/300b for 2 months.

V.St. = Very strong; ND = None Detected; M = Medium; Tr = Trace; SI = Slight.

34
Figure 1. A concept of the waste package

35
i—i—i—i—i—i—I

1 r~

COLLAPSED MONTMORILLONITE

^] POWELLITE

^

POLLUCITE

i—i—i—i—i—

r

200°C

300 bors

MONTMORILLONITE + Cs 2 Mo04 , Si/Cs = 2. 60 doys

MONTMORILLONITE + Cs 2 Mo04

, Si/Cs = 8. 60 days

MONTMORILLONITE + Cs2 Mo04 , Si/Cs = 2. 10 days

J I I L

UNTREATED MONTMORILLONITE

J I

l I J L 40 36 32 28 24 20 16 DEGREES TWO THETA (CuK«)

12 8

Figure 2. X-ray dif f ractograms of the products of hydrothermal interactions

between CS2M0O4 and Ca-montmorillonite

36
Figure 3. Scanning electron micrograph of analcime (containing Cs ) formed

from interaction of clinoptilolite with sintered ceramic at

300°C under a confining pressure of 30 MPa for 28 days

37
BACKFILL BARRIER RESEARCH AND DEVELOPMENT FOR

RADIOACTIVE WASTE ISOLATION IN SALTl/

E . J . Nowak

Sandia National Laboratory—^

Albuquerque, New Mexico 87185

ABSTRACT

Tailored backfill material eraplaced around waste forms and other waste package

components can provide a hydrologic barrier to brine flow and a sorptive barrier

to radionculide migration. The backfill can also chemically modify brine (e.g.,

modify Eh and pH) to minimize its corrosivity and to maximize the sorption of

radionuclides. Mixtures of bentonite (K^ = —2000 ml/g for Pu and --1000 ml/g

for Am, permeability = <1 microdarcy, swelling pressure = 18 MPa, all in con- centrated brine; thermal conductivity = —1.0 W/mk for dry bentonite-sand

mixtures) with other sorptive materials ( mordernite with = —30 ml/g for Cs

,

and sodium titanate with = -130 ml/g for Sr, all in concentrated brine) are promising candidate materials for a repository in salt. Measurements of radionuclide migration rates in backfill mixtures, permeabilities to brine, swelling

pressures in brine, thermal conductivities, hydrothermal reactions in brine,

and conditions for pre-compaction of backfill solids by isostatic pressing are underway. Among research and development needs are measurements of migration

rates of species in leachates from waste forms and of particulates; measurements

of radiation effects; engineering assessment of backfill manufacturing processes, emplacement techniques, and in situ physical behavior; and measurements

of chemical interactions among backfill components, other waste package

components and salt at elevated temperatures in a radiation environment.

}J This work was supported by the U.S. Department of Energy under Contract

DE-AC04-76-DP00789.

U A U.S. Department of Energy facility.

38
THE DEVELOPMENT AND TESTING OF

WASTE PACKAGE BACKFILL MATERIALS

FOR A NUCLEAR WASTE REPOSITORY

LOCATED IN BASALT

M. I. Wood

Rockwell Hanford Operations

Richland, Washington 99352

ABSTRACT

A deep repository located in Hanford basalt has been proposed for the disposal

of high-level nuclear waste. Backfill materials are required to make up the

backfill component of a multiple barrier waste package and to fill the rooms,

tunnels, and shafts of the repository sealing system. The major emphasis of

this paper is the backfill component of the waste package. However, much of

the discussion is applicable to the repository sealing system. The functions,

material properties, candidate materials, and testing methods for backfill are discussed. Because of the variety of backfill functions, physical, mechanical,

and chemical properties are required of an adequate backfill. Therefore, a multicomponent backfill will probably be needed. The geochemical environment

of the basalt site and the physical and chemical reactions that are likely to occur in the near-field environment have been used to derive the essential

testing needs.

INTRODUCTION

The Basalt Waste Isolation Project (BWIP) is currently assessing the feasibility

of developing the techi.ology for a deep geologic repository for the disposal of

high-level nuclear waste in the basalt flows of the Columbia River Plateau.

One important aspect of the development of a qualified repository in basalt is

the development of optimum backfill materials. Backfill materials are required

to make up the backfill component of the waste package and to fill the rooms,

tunnels, and shafts excavated prior to the emplacement of the waste. This

39
discussion is focused primarily on the backfill component of the waste package.

However, the tests being performed now and those planned in the future for the

waste package component are applicable to the repository sealing system as well.

The development of backfill materials is being achieved through the following

logical sequence of steps :

1. The functions of a backfill component are defined.

2. The chemical, physical, and mechanical properties required to perform

these functions are defined.

3. Candidate materials are chosen which are characterized by these properties.

4. Tests are designed and implemented to select the best backfill materials.

Two key elements which are needed to design useful experiments are (a) an understanding of the near-field geochemical repository conditions throughout

the service life of the repository, and (b) identification of the chemical

and physical properties which will involve the backfill in this environment

and may adversely affect its functional capacity.

5. Suitable backfill materials are chosen on the basis of these data and data

collection from other National Waste Terminal Storage (NWTS) and international sources.

BACKFILL FUNCTIONS, PROPERTIES, AND CANDIDATE MATERIALS

A fundamental premise of the multiple barrier concept is the development of a barrier system in which one barrier not only promotes the physical and chemical

stability of another barrier or set of barriers, but also provides redundant

functional ability in the event of the failure of one or more barriers. The

current waste package concept designed by the Basalt Waste Isolation Project is

given in figure 1. In practical terms, the outer component protects and

provides redundant functional support for the inner components. In this case,

the backfill component supports and protects the overpack and canister. The

40
backfill has two major functions: (a) to minimize contact between these

components and the host environment; and (b) to control the release of radionuclides into the host environment within the recommended guidelines in the

event of canister and overpack failure. In table 1, the functions of the

backfill are correlated along with the chemical, mechanical, and physical

properties of candidate materials which will be used to determine their

relative acceptability.

The first function of the backfill can be divided into two parts. First, the

backfill component is required to minimize contact between the waste package

and groundwater because of the corrosive reaction which will take place,

perhaps to a significant degree, over the first 1,000-year period. The most

important parameter here is the permeability of the material. Swelling

pressure measures the ability of the backfill to fill voids in the space

between the waste package and the bedrock. Second, the backfill must act as

a shield to protect the canister from the impact of falling rock and undue

stress resulting from minor bedrock movement. The important parameters are bearing capacity, plasticity, and shear strength. These properties measure the

ability of the backfill to keep the waste package in its emplacement position

and to retain its physical integrity. Thermal conductivity is an important

parameter since chemical alteration of backfill materials could result if the

temperature buildup is excessive within the backfill. A change in mineral

phase assemblage could affect both the physical integrity and water-excluding

ability of the backfill.

Retardation of radionuclide migration can be accomplished by the backfill via

three principal mechanisms: (1) sorption; (2) reaction; and (3) diffusion

control. Sorption and reaction capacities are strongly dependent on chemical

properties of the backfill (backfill composition, reactivity of backfill

materials to specific radionuclides, temperatures). Diffusion controls the

rate of radionuclide transport only under certain physical conditions. These

conditions are most likely to occur in a dry impermeable backfill or in a saturated backfill where the rate of groundwater flow is essentially zero. Due to the small thickness of the proposed backfill component around the waste

package, diffusion controlled transport may prove to be a most effective means

41
of radionuclide retardation. Consequently, the water-excluding properties of

backfill components (permeability, swelling pressure, etc.) are significant

parameters to consider in terms of radionuclide containment within the

backfill.

PRELIMINARY BACKFILL MATERIAL CANDIDATES

An optimum backfill material is likely to be made up of several materials since

no single material has been found in the course of a literature search which

can satisfy all of the functional requirements listed in table 1. It is

unlikely that a satisfactory single component will be found.

Two component types of candidate backfill materials have evolved on the basis

of the functional requirements, a major component and a minor component. The

major component makes up the bulk of the backfill and can be composed of more

than one material. The best major component will satisfy a maximum number of

the functional requirements. In addition to the functional properties listed

above, two general constraints were applied to make up a preliminary candidate

list. First, since backfill will be in direct contact with the geological

environment (basalt, secondary minerals, groundwater), it must remain in its

functional state in that environment under ambient conditions during the time

period of geologic control. Furthermore, the backfill must be compatible with

host rock and groundwater during the thermal period when various chemical

parameters are disturbed by the introduction of waste into the system (e.g.,

temperature, pH, oxygen fugacity). The consequences of this constraint are

that synthetic materials are eliminated as potential major components. Further

preference is given to those natural materials presently found in the basalt

environment. Second, potential major component materials must be readily

available in quantities sufficient to fill the bulk of the repository volume.

Three potential major component candidates satisfy these two criteria and are presently the subject of extensive investigation. They include: (1) bentonite,

(2) crushed basalt, and (3) zeolites. Bentonite has been studied in depth

primarily by Swedish scientists and shown to have a variety of useful properties

In particular, the lower permeability and swelling capacity of this clay makes

42
it an attractive material for minimizing contact between the groundwater and

the overpack or canister metal. In addition to its water excluding potential,

bentonite has a very high sorption capacity relative to other clays and

geologic materials. Crushed basalt is an obvious candidate in a repository

located in basalt, because of its availability following the excavation of the

repository and the need to dispose of this material. Preliminary hydrothermal

and sorption tests indicate that altered basalt is highly reactive with uranium,

plutonium, americium, cesium, and strontium (2). Zeolites are considered as a potential backfill component because of their high cation exchange capacity.

Also, minor amounts of zeolites are found in the fracture mineralization in the

Columbia Plateau Basalts.

There is also a need for a minor backfill component which is specifically

tailored for transport retardation (either by sorption or reaction) of key

long-lived and potentially mobile radionuclides (e.g., 129j and 99tc) . Attention is focused on single phase and the minor backfill component may be

composed of several phases.

PHYSICAL AND MECHANICAL TESTING

Perhaps the most critical physical parameter of a backfill component is its

permeability. Ideally speaking, if a backfill component existed which would

remain perfectly dry, then the inner components of the waste package would

remain intact and contain the waste completely for an indefinite period of

time. This material probably does not exist. However, if the time required

for saturation is as long as several hundred years, the design of a waste

package will be affected significantly. Therefore, it is important to examine

the sequence of events which lead up to the saturation of a backfill.

The physical state in which groundwater exists in the near-field environment of

the waste package depends on the temperature and pressure conditions of the

system. These parameters will change with time as the three major periods of

repository service life evolve. During the operational period, the waste

packages are loaded into boreholes and the rooms above the boreholes will be

left open. The backfill volume will be characterized by high temperature

43
(250° to 300°C) and low pressure (nearly atmospheric). During the thermal

period after the repository system is sealed, temperature will decrease slowly

and pressure will increase slowly because the room backfill has been emplaced

and is slowly saturating and swelling. During the period of geologic control,

the temperature eventually will return to the ambient value of 65°C and pressure

will rise to its maximum value (100 to 300 bars). Thus, the overall trend is

that of decreasing temperature and increasing pressure in the backfill component

of the waste package. This trend is consistent with water first being in the

vapor phase, then in the vapor and liquid phases, and finally in the liquid

phase. Further, it is reasonable to assume that the backfill will be dehydrated

when water is in the vapor state, partially saturated when water is present as both liquid and vapor, and fully saturated when water is liquid.

The length of time in which a backfill will remain dehydrated is most sensitive

to the pressure increase of the waste package system. Bentonite is probably the

only major backfill component which will contain significant amounts of water

when it is emplaced (10 to 15 wt%). This interlayer or intergranular water

should be the only source of water in the backfill during the operational

period because the rock will be pumped dry. Water should also vaporize and

escape into the empty room above unless the rate of vapor diffusion through the

backfill is slow enough to allow a buildup in effective pore pressure that is

sufficient to induce the stability of the liquid phase. Saturation would then

begin as soon as groundwater resaturates the host rock and contacts the

backfill

.

A far more likely scenario is that vapor diffuses easily through the backfill

up to the room and the effective pore pressure is not increased. This mechanism

may also continue after the room is backfilled and pumping of the repository

ceases. The room backfill will be relatively porous and dry, thus presenting

avenues of escape for vaporized groundwater. A stable liquid water phase will

not exist until the backfill temperature decreases to low enough temperatures

(100° to 150°C) where small pressures ( 1 to 5 bars) are sufficient to promote

the stability of liquid water. At this point, a wetting front will be

established and the saturation process will begin.

44
Estimation of the time required to saturate the backfill after water is stable

in the liquid phase is a difficult task. Both heat and mass transfer (liquid

and vapor) will be occurring simultaneously. Parameters which will affect the

saturation process include the temperature of the backfill, the absence or presence of a temperature gradient, the degree of compaction of the backfill,

the availability of groundwater from the surrounding fracture system, and the

swelling of bentonite as it comes into contact with the groundwater.

If the, total time involving dehydration and rehydration of the backfill is on

the order of hundreds of years, it is likely that the waste package can be

simplified considerably. For example, if corrosion of canister metal through

interaction with groundwater is delayed for this period of time, the reduction

of oxygen content and tempeature should significantly reduce the corrosion

rate. Also, if a metal is used which is thermodynamically stable in the

ambient basalt environment (such as copper), the canister is likely to remain

physically intact indefinitely. Thus, a waste package consisting of a canister

and a backfill may be sufficient to contain the waste within recommended

guidelines

.

After the backfill is saturated, contact between groundwater and the overpack

or canister must be kept to a minimum. This goal is achieved by engineering a backfill which is of low permeability when it is finally saturated. The

permeability must be low enough for diffusion to be the rate controlling

mechanism of transport. In this manner, the concentration of corrosive agents

in the groundwater (oxygen, chlorine, etc.) coming in contact with the metal

surfaces are minimized. Also, if the inner components of the waste package are breached, and radionuclides are transported through the backfill, diffusional

movement allows the maximum time for precipitation and/or sorption to occur.

Given these considerations, a number of tests must be conducted to determine

the behavior of a bentonite-based backfill. They include the following:

45
1. Determination of Backfill Compaction and Porosity

Prior to experimentation, backfill materials would be compacted and analyzed to

determine the porosity of the material. A three-dimensional picture of porosity

can be achieved by characterizing thin sections parallel and perpendicular to

the direction of compaction. This work would be completed for all types of

proposed backfill materials (bentonite, bentonite-basalt , bentonite-basaltzeolite)

.

2. Dehydration of Bentonite

Compacted bentonite is dehydrated at 300°C for varying lengths of time. Four

factors should be measured: (a) the rate of vapor discharge, (b) pore pressure,

(c) the degree of fracturing the steam channeling that may result from vapor

discharge, and (d) analysis of the solid phases following discharge. These

data will indicate the effects of dehydration on the chemical and physical

integrity of* bentonite.

3. Rehydration of Bentonite

Compacted bentonite which has been dehydrated is rehydrated over an appropriate

temperature range. Simulated Hanford groundwater is used in the test. Data to

be collected include (a) the permeability as a function of different temperatures, the absence or presence of a thermal gradient, and the degree of compaction; (b) analyses of the groundwater at the completion of the experiment; and

(c) analyses of the bentonite following the experiment. These data will be

used to estimate the time required to saturate the backfill, the effect of

altered groundwater chemistry on the corrosion rates of the canister or overpack,

and the effect of chemical alteration on the sorptive and reactive capacity of

the backfill.

4. Field Tests of Backfill Dehydration and Rehydration

On the basis of laboratory data, full scale tests of backfill dehydration and

rehydration processes will be done in the field at various temperature and

46
pressure conditions expected to occur during the service life of the repository.

The performance of these tests will require development of backfill fabrication

and emplacement techniques as well as instrumentation to measure dehydration

and rehydration processes. These tests are advantageous because variations in

the groundwater backfill interaction can be determined which may occur due to

the spatial arrangement of the backfill component of the waste package.

CHEMICAL TESTING - SORPTION AND HYDROTHERMAL TESTING

If the inner components of the waste packages are breached, the backfill

component must be capable of reducing rate of radionuclide release to the host

rock. This is accomplished through sorption processes or through precipitation

of radionuclides in low solubility phases as transport through the backfill

occurs. In order to design an accurate and efficient testing program to

evaluate the chemical reactivity of backfill materials with radionuclides, two

factors must be considered:

1. The tests must simulate the geochemical environment at the repository site.

The range of conditions expected to occur in a repository located in the Hanford

basalt is summarized in table 2. A reference site groundwater is listed in

table 3. The most critical parameters are temperature, groundwater composition,

Eh and pH. Changes in temperature affect reaction kinetics and phase solubilities. The pH and Eh values are directly related to the oxygen content of the

system which dictates the stable oxidation state of polyvalent radionuclides.

This is an important factor for radionuclides such as uranium, which is highly

soluble in the hexavalent state (oxidizing environment) and nearly insoluble in

the quadrivalent state (reducing environment).

2. A choice of which radionuclides to use in these experiments must be made

because an attempt to investigate all of the isotopes in the nuclear waste

inventory is expensive, time-consuming, and unnecessary. Simple one-dimensional

transport modeling has been used to determine which radionuclides are most

likely to reach the biosphere in hazardous amounts (3). The results of this

analysis indicate that the choice of which radionuclides are most hazardous is

somewhat dependent on the time of release. If the canister is breached in the

47
first thousand years, then the short-lived fission products, i.e., cesium and

strontium, present a hazard. If the canister remains intact for the thousand

year period, then these elements no longer exist in hazardous amounts as a result of radioactive decay. The alternate group of radionuclides which are most hazardous include both long-lived cations (uranium, neptium, plutonium,

and americium) and long-lived anions (iodine, selenium, and possibly technetium)

Because of their long half-lives, these radionuclides are hazardous regardless

of their release time. The anionic radionuclides, iodine, and selenium, are given the most immediate priority because few geologic phases are capable of

reacting with them. Cesium and strontium are given the lowest priority because

experiments completed on the sorptive and reactive interaction between basalt

and these elements suggest that they can be adequately retarded with the

addition of crushed basalt to the backfill.

Two areas of chemical testing have been defined on the basis of temperature.

Backfill sorption studies are being completed in the range of 65° to 100°C. A

test matrix has been designed to account for the variety of geochemical

conditions discussed above. Potential sorptive and reactive backfill materials

and the corresponding radionuclides to be sorbed are listed in table 4. The

following elements make up the sorption test plan:

1. All experiments are being run with simulated repository site groundwater

(table 3).

Parallel experiments are being completed under oxidizing and reducing conditions

In order to create the reducing environment, hydrazine (N2H4) is added as an Eh

poising agent.

2. Static tests are currently being conducted at 65°C. The reactants are placed in a polyethylene bottle which is agitated and kept in a heated water

bath (65°C). Similar tests will be completed at 90°C. On the basis of static

test data, flow-through experiments will be initiated to simulate groundwater

flowing through a backfill material surrounding a breached waste package.

48
3. Isotherms are conducted by doing experiments at several concentrations for

each element in order to determine the effects of radionuclide concentration on sorption coefficients.

4. One element and one potential backfill material are tested in one experiment. The experiments are run for up to 60 days and replicated. This

experimental procedure is necessary not only to determine the most effective

sorptive materials, but also to isolate specific sorptive reactions which

will occur in a multi-component backfill and involve many radionuclides. On

the basis of single radionuclide-single backfill material tests, multiple

radionuclide-single material tests will be completed to determine the degree

of competition for sorption sites among various nuclides.

Hydrothermal testing of backfill materials must satisfy two major goals.

First, the chemical stability (i.e., solubility) of these materials in the

basalt environment must be determined. Second, the ability of these materials

to reduce radionuclide release rates (primarily through precipitation of

radionuclide-bearing solids) under near-field repository conditions must be

determined. As with the sorption experiments, the variety of geochemical

conditions that will occur during repository life must be included in the

test matrix. The following parameters make up the test matrix:

1. The experimental reactants include (a) groundwater-backf ill materials,

(b) waste f orm-groundwater-backf ill materials, or (c) groundwater-backf ill

materials-basalt. The progression of experiments moves from single component

systems to multicomponent systems. This sequence of testing enables the

investigator to determine which backfill materials are most reactive with key

radionuclides. Assuming that the reactions occurring in the single component

systems will also occur in the multi-component systems, the data will be used

to project the degree of fixation of hazardous radionuclides in the backfill

over long periods of time.

2. Temperatures of the experiments are 150°C and 300°C and the pressure is

300 bars. These values represent the upper temperature and pressure limits

expected in the near-field repository environment.

49
3. Parallel experiments will be run under oxidizing and reducing conditions.

Experimental methods for controlling Eh at hydrothermal temperatures are currently being developed.

4. Experimental methods for controlling the pH are currently being developed.

A pH buffer must be developed in those experiments where a reactant (i.e.,

basalt) is not present to buffer the solution (a pH value of approximately 9.4

at 65°C and decreasing with increased temperature).

5. Parallel backfill experiments will be conducted with simulated and actual

waste forms. If experiments . involving simulated wastes are duplicated by

experiments with actual wastes, then fewer hot cell experiments need to be

done and extra time, experimental difficulty, and expense can be avoided.

6. The experimental apparatus that is being used for the backfill hydrothermal

tests is a Dickson-type rocking autoclave. With this pressure vessel, liquids

can be periodically withdrawn and analyzed without interrupting the experiment.

Continuous sampling at temperature and pressure enables the investigator to

determine when steady state conditions are reached. Solubilities of backfill

materials and their capability to react with key radionuclides will be

determined

.

7. As with the sorption experiments, flow-through experiments will be

conducted on the basis of the data gathered from the Dickson apparatus and

modified to permit flow-through measurements.

Most of the backfill chemical testing completed to data has been done to

evaluate basalt as a reactive material capable of precipitating solids. These

data indicate that basalt has an excellent potential for sorbing or precipitating stable solids of the radionuclides, cesium, strontium, rubidium,

uranium, plutonium, and americium. Thus, the data indicate that basalt is an excellent backfill material and host medium for the disposal of nuclear waste.

50
SUMMARY

Analysis of the functions of a backfill component within a multiple barrier

waste package system indicates that a wide range of physical, mechanical, and

chemical properties must characterize an acceptable backfill. Consequently, a variety of tests on several materials must be completed. The types of sitespecific tests which need to be designed and completed have been discussed

above. Our current data base indicates that an adequate backfill material can

be specified for use in a repository located in basalt.

REFERENCES

1. Coons, W. E., Moore, E. L. , Smith, M. J., Kaser, J. D. , The Functions of an Engineered Barrier System for a Nuclear Waste Repository in Basalt , RHO-BWI-LD-23 , Rockwell Hanford Operations, Richland, Washington

(January 1980).

2. Smith, M. J., G. J. Anttonen, G. S. Barney, W. E. Coons, F. N. Hodges,

R. G. Johnston, J. D. Kaser, R. M. Manabe , S. C. McCarel , E. L. Moore,

A. F. Noonan, J. E. O'Rourke, W. W. Schulz , C. L. Taylor, B. J. Wood,

and M. I. Wood, Engineered Barrier Development for a Nuclear Waste

Repository Located in Basalt , RHO-BWI-ST-7 , Rockwell Hanford Operations,

Richland, Washington (May 1980).

3. Wood, B. J., Estimation of Waste Package Performance Requirements for a Nuclear Waste Repository in Basalt , RHO-BWI-ST-10 , Rockwell Hanford

Operations, Richland, Washington (July 1980).

51
0 O'X) 1—1

CO

i-l CD uo

3. u

•HB

4-1 3

CPco

ex0o

CJ

rH1 1 •H

4H

i-H

a

ctl 33 X)

QJ

H X!

4-J

4HO

CO

QJ •H

4_)

CUGXo

in a,i—i

COCo

•H

4-J CJc

X)c

03

COcoH

4-1 oc

hi

no

cx

COc

COMH 4-J 3

cu cu T3 Q

•H 3 H o

CJ )-l 3 "H C > o a

•h

w

CO 4-1 ad coo m cn o

cu 3 X3 O 4-1 H

4-1 O

CO 4-1 T3

5-i

cfl 4-1 cu C4

CDSO

CO ^ 4-J o 3

Cfl 0) (X 0co

(-1 •H>cw

cu 4J

CO

cd

4—1o

CO 3 o o pa

•H

4J S

cfl o H iH O

cn co 4-1 •H •rH 0> CO 0 >

4-1 3 i-i •rl

Cfl <4_| 4-J 4-» Qi M-l CO •H O

•H CO CJ 4-1 4-J Q >-> OJ 03 SO u 4-1 i-i Cu 3 c o >> •H cx 03 cu >^ O

a, x> rHU i-i 4-J CJ

co •H bO 4-J •H 3 H XI 3 CUJ co CJ

H

cflO Cfl •r-l 3 •rH cfl

M Sh 0) 1—|

•rH i-i 4-)

g

H 4-1 0 rH i-l CO co in 3 S-i CJ cfl CU cO 0) O cu 3 CU JZ rH jt; cj Cm CO 33 co (X tH * c c >^ o o 4J H •rl -H

4-J 4-1 CJ cx Cu 03 M i-l CX O o rtj CO CO U 3 c >> o o 4-J •H •H H

4-1 4-)

u

O CJ rd

cfl cfl CX

CU a) CO od O

4-) o >^

cfl 4-J Cu •r-l j3 e c CJ 4-1 M O Cfl bO >,

•H CU c .p X) 4-J cfl p*. cu rl 3 o CJ 4-1 M >

CO CU •rl 4-J •H

4-1 bO CJ CO

H 4-J CO o 3 H CO CJ CO >H •rl 4-J i-l e 3

CU Cu in co cO i-l Tj

in cfl cfl OJ CU c

4-J ai j3 X! o

CO PQ (X CO

H CJ

cuu3 4-1 CO H 3 CO CJ o >, 0) CO >^

iu -H 4J >-i cx 4-1 0) CO •H Cu CO >» •H

4-J 3 i-H cj 4-J >

cfl rH •H bO •H •H

i-S CJ xi 3 bOO 1 1 4-J

X cfl

H c •rl CO CJ

W CU

H •H 4-J

B 3 e rH i-l CO u -a

CU CO cO cu c

QJ cu rH o

Cm co pq Pi CJ

CO3Q

•rH

4-J CJ33

IX

H co

Cfl OJ 3 -H O 4J •H M

4J at

CJ o, 3 O 3 in Pn Cm

CO uHHO

CO

-a3

(0

co T3

•H3

cr H3

cu

cu3

4-1 CU XI

CO

cu T3

CJ3co

•HT3

COu

S-i o

CMC

CU •HoH

cw

4-iCUo

CJ3O3

X>

•HM

X>T3

OJ

!-i 3

co

COCU S

52
Table 2

Range of Geochemical Conditions in the Near-Field Repository

Environment (at 1,000 m in depth)

Ambient Conditions Thermal Period

Temperature 65°C 65°C-250°C (canister)

Pressure 30 MPa (lithostatic)

10 MPa (hydrostatic)

Atmospheric to 30 MPa

Groundwater pH

(R^SxO^ control)

Groundwater Eh

(QFM to

Ni - NiO control)

9.4 - 10.0

-0.3 to -0.5 volts

5.5 - 9.4

-0.5 to +0.54

53
Table 3

Measured Grande Ronde Groundwater Composition (Smith and Others, 1980)

Anionic Concentrations mg/£ epm*

CO

~2 30 1.003

HCOo"3

36 .589

0H

_

3 .182

H

3

Si0

4

~

137 1.438

Cl

_ 148 4.174

-2 so4 108 2.249

F~ 37 1.947

TOTAL epm = 11.582

Cation Concentrations mg/ % epm

Na+ 250 10.875

K+ 1.9 .049

Ca+2 1.3 .065

Mg+2 .04 .033

Si02 121

TOTAL epm = 11.022

*epm = equivalents per million

54
Table 4

Candidate Backfill Materials for Reducing the Rate of Release

of Key Radionuclides in a Basalt Repository

Potential

Radionuclide( s) Backfill Material

I, Se Bornite : Cu^FeS^

Chalcopyrite : CuFeSo

Ultramarine: Na

g ( AlSiC^)^

Graphite: C

Copper : Cu

Np Apatite: Ca5 (F, CI, OH)(P0 4

) 3

Azurite: O^CCC^^COH^

Basalt

Tc, U, Pu, Am Basalt

Bentonite

55
Figure 1. BWIP conceptual waste package in storage hole

(Coons and Others, 1980)

56
CONDITIONS AFFECTING THE BACKFILL IN A WASTE PACKAGE

FOR A REPOSITORY IN TUFF

J. K. Johnstone, Sandia National Laboratories, Div . 4537,

Albuquerque, N.M. 87185

E. N. Vine, Los Alamos National Laboratories, Mail Stop 514,

Los Alamos, N.M. 87545

ABSTRACT

We have initiated a program to support waste package development and design for

a repository in tuff. Initial efforts have focused on definition of a reference waste package, determination of the repository thermal and geocheraical

environment and publication of a research plan. The components of the reference

waste package are the waste form, canister, overpack and backfill. For commercial high level waste (CHLW), the backfill thickness is 41.3 cm. The reference

backfill composition is 70 to 90 percent crushed, welded tuff and 30 to 10

percent Na-montmorillonite clay.

A parametric study has been completed to evaluate the time-dependent thermal

environment in a waste package with most emphasis placed on its effect on the

backfill. The study included both CHLW and spent fuel. The parameters that were varied included canister power (2.16 and 1.08 kW/canister for CHLW and

0.55 kW/canister for spent fuel), areal power density (100 and 50 kW/acre),

backfill thermal conductivity (0.30, 0.75 and 1.1 W/m°C) , tuff thermal conductivity (1.55 and 2.40 W/m°C), and time (100 to 1000 years). The emplacement

hole diameter was 122 cm or 61 cm. Thermal properties of the rock and waste

package were assumed temperature independent. We used a finite element, 2-

dimensional, axisymmetric heat conduction code to calculate the radial temperature gradient along the canister midplane. The calculations show that, for a given heat loading, the thermal profiles are most sensitive to the thermal con- ductivity of the backfill. In the most severe environment (CHLW, 2.16 kW/

canister, 100 kW/acre, kbackfill = 0.3 W/m°C, and ktuff = 1*55 W/m°C)

temperatures in the backfill may exceed 535°C. The Na-montmorillonite and

57
some components of the crushed tuff in the backfill would be unstable in the

potential hydrothermal environment and ground water surrounding the waste

package. The conditions imposed by the spent fuel are much less severe and do

not appear to affect the stability of the backfill components.

Geocheraical studies include investigation of the mineralogy, major and minor

trace element rock composition and ground water conditions in a welded tuff

horizon. The reference composition for the ground water suggests that it is

chemically benign under ambient conditions. The pH is neutral to slightly

basic. The Eh remains uncertain. The potassium to sodium weight ratio is

about 0.1 which classifies the system as potassium rich. The high potassium

content accounts for the potential instability of the Na-montmorillonite at

temperatures in excess of 100°C.

A research program has been developed to support the waste package development.

The heart of this program is a long-term series of "chemical vector" experiments

in which the interactions between the various waste package components is

evaluated in a sequence reflecting the flow of water through the waste package.

The experiments provide for control of temperature, pressure, radiation, water

flow rate, and the number of components in the system.

58
CRITICAL FACTORS AFFECTING THE USE OF MONTMORILLONITE

AS A NUCLEAR WASTE BACKFILL MATERIAL

G. W. Beall, A. G. Lamkin, and K. L. Kelly

Radian Corporation

8500 Shoal Creek Blvd.

Austin, Texas 78758

ABSTRACT

The use of montmorillonite as a component in nuclear waste backfill mixtures

was first suggested by Swedish KBS (1) and later in the United States (2). The

advantages of montmorillonite in such nuclear waste backfill largely stem from

its large cation exchange capacity and from its swelling properties, which

result in extremely low permeabilities. This paper details the effects of

compaction and common groundwater cations such as calcium, magnesium, potassium,

and sodium on the properties of montmorillonite, and what consequences these

effects have for nuclear waste disposal.

INTRODUCTION

The suggestion that montmorillonite would make an ideal component in a nuclear

waste backfill mixture was made in Sweden (1) and later in the United States

(2). The characteristics that make montmorillonite suited for a nuclear waste

backfill component are its large cation exchange capacity and its swelling

properties. The cation exchange reactions for a large number of radionuclides

have been studied (3). It has been shown that in the pH range of interest in

most groundwater systems these cation exchange reactions are overshadowed by

sorption reactions involving hydrolysis and carbonate complexation. These

facts have been incorporated into a backfill mixture that still retains montmorillonite as a component to act as a water barrier, but that contains additional minerals which undergo specific irreversible sorption of radionuclides

of importance in nuclear waste (4). The fact that the swelling properties of

montmorillonite are critical to its inclusion as a backfill component makes

it imperative that a thorough understanding of these properties be obtained.

59
The majority of work in the past on the physical properties of montmorillonite

have centered on the use of the sodium form of montmorillonite. That emphasis

has prompted this paper to address some factors that are critical to the proper

application of montmorillonite as a water barrier in a backfill material for

nuclear waste disposal. The critical factors are cation effects and clay

orientation due to montmorillonite morphology upon swelling.

EXPERIMENTAL PROGRAM

Weighed samples of Wyoming montmorillonite obtained from the Source Clay Mineral

repository, The University of Missouri at Rolla, were equilibrated with various

molar concentrations of Ca, Mg , K, and Na. In all cases the counter ion was chloride. The molar concentration employed in each set of equilibrations were

1, 0.5, 0.25, 0.125, 0.0625, and 0.03125. The samples were equilibrated for

four hours with continuous agitation. The samples were then centrifuged for ten

minutes in a clinical centrifuge at approximately 3000 rpm. The supernatent was then decanted and the volume determined accurately.

Another equilibration study was conducted using a synthetic groundwater solution

containing the following molar concentrations: Ca —.005, Mg —.001, K—.00025,

and Na—.004. Again the anion was chloride. This was compared with a solution

of only .004 M Na. Weighed samples of the Wyoming montmorillonite were equilibrated with these solutions for 30 minutes with continuous agitation and then

centrifuged for 30 minutes. The supernatent was then decanted and the volume

determined

.

RESULTS AND DISCUSSION

The majority of work on the use of montmorillonite in nuclear waste backfill

mixtures has been predicated on the cation exchange and swelling properties of

the sodium form of montmorillonite. These studies have not dealt with the

effect of various cations common in groundwaters or with how various physical

processes affect montmorillonite cation exchange or swelling properties.

60
The concern with cation effects on montmorillonite stems from the ubiquitous

nature of several cations in groundwaters. Table 1 gives the general range of

concentrations for major ions in undisturbed groundwater in contact with igneous

rocks. It can be seen that Ca"

1-*", Mg"*""*", and K"

1" are of the same magnitude of

concentration as sodium. In the case of many brines the concentration of Ca++,

Mg"1-1", and K+ can reach 1 molar range. The fact that Ca^-1" and Mg"1-1" are of

approximately the same concentration as sodium does not in itself present problems with the cation exchange and swelling properties of montmorillonite. The

problem arises from the fact divalent Ca and Mg are preferentially taken up

by montmorillonite versus sodium. In general the exchanging power of cations

on montmorillonite follow Na § K § Ca § Mg § NH4. This fact again would not

in itself cause problems with the properties of montmorillonite. The problem

arises from the fact that both the Ca and Mg strongly affect the swelling of

montmorillonite. This arises largely from the hydration behavior of the cations

involved, Ca and Mg being more highly hydrated than sodium. It has been calculated (5) that the ordered water layers on the Na montmorillonite extend approo o ximately 7.5 A and for the calcium form 10 A. It is further calculated that

the transition from highly ordered water at the surface out to completely random

o water is a gradual one extending to 100 A in the case of the sodium form in

contrast to the calcium form, where the transition appears to occur abruptly

o at about 15 A. This, difference strongly affects the swelling properties of

montmorillonite as illustrated in figure 1, where the effect of concentrations

of Ca"*

-1", Mg"

1-1", Na+

, and K+ upon swelling of the Na form of montmorillonite is

presented. It can easily be seen that both Ca and Mg strongly affect swelling

of Na montmorillonite. As an example, the swelling volume of the Ca form is a factor of 2.5 less than the Na form at 0.1 M. This appears to be less of a problem with brines since all the curves coalesce at close to one molar. To

illustrate the magnitude of the effect on swelling that calcium and magnesium

can have at levels common in groundwaters, the sodium form of montmorillonite

was equilibrated in a single equilibration, with a synthetic groundwater con- taining 60 ppm Ca, 25 ppm Mg, 10 ppm K, and 100 ppm Na. This groundwater

caused 4 percent less swelling compared with a solution containing only sodium.

It must be emphasized that this effect was found after only one equilibration.

In a real groundwater situation the effect would be much greater since there

would be an infinite supply of equilibration groundwater compared to a finite

61
amount of backfill. These facts then open the possibility that in a real

nuclear waste repository, the montmorillonite would initially expand as primarily a Na montmorillonite and then, as it comes to equilibrium with the groundwater, could shrink and actually create fissures through which groundwater

migration could be quite rapid. This is what occurs quite frequently to ponds

utilized for watering cattle where the ponds were lined with montmorillonite.

These ponds with time begin to leak because of the Ca, Mg , and Na equilibration

This effect will most likely be worsened because the groundwater will tend to

have higher concentrations of cations than those used in our experiments. This

is largely due to the higher temperature around the nuclear waste canisters.

The second factor that has been largely overlooked is the anisotropic nature of

montmorillonite. Montmorillonite ' s morphology can be described as sheet-like.

The sheets are quite thin with indistinct and irregular edges. This morphology

tends to cause montmorillonite to orient strongly parallel to the C-axis. This

orientation can be caused by physical handling such as compacting and extruding

or by simple gravity settling. This tendency to orient must be strongly consid

ered in any backfill design to ensure that the swelling properties observed in

the laboratory can indeed be obtained in the field.

CONCLUSIONS

The fact that calcium and magnesium have such a pronounced influence upon

swelling properties of montmorillonite hints that it would be prudent either to

employ the Ca form of montmorillonite in nuclear waste backfill, or to tailor

the ratio of Ca , Mg, and Na to fit the groundwater composition to which it will

be exposed. If indeed tailoring of montmorillonite occurs, the possibility of

_i i_ adding small amounts of Fe would eliminate the need to add ferrous-containing

minerals to the backfill as an oxygen getter. This tailoring might lend itself

to other substitutions that could be of benefit in a backfill.

The second factor of importance is that the tendency for montmorillonite to

orient and thus swell anisotropically must be considered in field emplacement

of montmorillonite as a backfill material.

62
REFERENCES

1. Pusch, R. Highly Compacted Bentonite as a Buffer Substance , KBS Technical

Report 74, Karnsbranslesaherhet , Stockholm (1978).

2. Allard, B. and G. W. Beall. "Actinide Retaining Backfill Materials in

Underground Repositories for Alpha-Contaminated Waste." Proceedings of the

IAEA meeting, Vienna (1981), p. 667-677.

3. Serne, R. J. "Waste Isolation Safety Assessment Program," Progress Reports

1977 and 1978.

4. Beall, G. W. and B. Allard. "Chemical Aspects Governing the Choice of

Backfill Materials for Nuclear Waste Repository," Nucl. Tech. (in press).

5. Grim, R. E. and F. L. Cuthbert. "The Bonding Action of Clays, I, Clays in

Green Molding Sands," Illinois State Geological Survey. Rept. Invest. 12

(1945).

63
Table 1

General Composition of Granitic Groundwater

Species Concentration in Nature, mg/L_L

HCO3 60-400

Si02 (total) 5-60

SO?

- 3-40

4

CI" 5-50

Ca 2+ 10-60

Mg2+ 2-25

K+ 1-10

Na+ 10-100

F~ 0.01-5

HP0^~ 0.01-0.5

Fe2+ 0.5-20

pH 7.2-8.5

1 Probable concentration range for undisturbed deep groundwater in

contact with igneous rocks.

64
10

1.0

< 0.1

A SODIUM

POTASSIUM

v MAGNESIUM

O CALCIUM

0.01

0.001 1

10 15

SWELLED VOLUME OF THE CLAY (ml)

20

Figure 1. The effect of various cations on the swelling of montmorillonite

65
SELECTION AND EVALUATION OF BUFFER AND BACKFILL MATERIALS

FOR USE IN A NUCLEAR FUEL WASTE DISPOSAL VAULT:

THE CANADIAN PROGRAM

G. W. Bird

Fuel Waste Technology Branch

Atomic Energy of Canada Limited

Whiteshell Nuclear Research Establishment

Pinawa, Manitoba, ROE 1L0

ABSTRACT

The selection and evaluation of backfill materials for use in a nuclear fuel

waste disposal vault have been divided into two parts based on the physical

location of the materials in the vault. Materials in the immediate vicinity

of, and in contact with, the waste containers are termed buffer. Backfill

includes all materials which fill the remainder of the vault after the waste

and buffer have been emplaced.

Potential buffer materials are being evaluated by:

1. Chemical studies which include (a) ion exchange capacities and

selectivities for various radionuclides, (b) hydrothermal stability in

groundwater at the pressure and temperature conditions in the vault,

(c) reactivity to the waste container, waste form and host rock (d) radiolysis effects, (e) assessing the role of organic materials and biological pro- cesses and (f) a study of possible additives which could be used to control

redox potentials or to trap particular radionuclides.

2. Physical studies which include thermal conductivity and diffusivity,

swelling pressures, compressibility and shrinkage, mechanical strength,

density, hydraulic conductivity, and diffusivity for water and specific

radionuclides.

66
3. Conceptual engineering studies where equipment and techniques for

preparation, emplacement and quality control are being developed.

4. Development of mathematical models to describe the physical and chemical

behavior of the buffer within the vault and to provide an analysis of

potential pathways and distances of radionuclide transport.

The backfill will occupy a much larger volume than the buffer and will require

a different technology for emplacement. These differences lead to additional

constraints on material selection and evaluation. As much of the required

physical and chemical data will come from the buffer development program,

backfill research will be concentrated on engineering aspects such as handling,

emplacement and quality control technologies, and on the development of mathematical models for water transport in the vicinity of and within the vault.

The schedule for the buffer and backfill development program calls for

preliminary evaluations of the materials to be completed by the end of 1983.

From these evaluations, materials will be selected for engineering-scale

testing in an underground laboratory.

INTRODUCTION

Safe permanent disposal of radioactive waste requires the isolation of a number

of diverse chemical elements from the environment for a long period of time.

The Canadian approach to nuclear fuel waste management is disposal in vaults

mined deep into crystalline rocks for the Precambrian Shield (4). A vault will

consist of a number of mined chambers some 500-1000 meters below the surface,

the shafts leading to it and any other underground facilities. These facilities

are described in general terms in reports AECL-6415 (2) and AECL-6416 (3).

After disposal operations have been completed, the vault must be backfilled and

sealed in such a manner as to return the land to a natural condition where no further maintenance or supervision is necessary.

In the post-closure period, the vault and surrounding rocks will become heated

from the decay heat of the radioactive waste. The maximum temperature reached

67
will depend on the amount of radioactivity emplaced per unit area. Currently

the maximum temperature is assumed to be not more than 150°C (1). The elevated

temperature period will last from tens to several thousands of years depending

on the waste form.

After closure, the only significant mechanism of radionuclide release which can

be identified is that deeply circulating ground water may penetrate to the

waste, leach out the radionuclides, and carry them back to the surface. A

number of protective barriers, including the waste form itself, its container,

the buffer material surrounding the container, the backfill and sealing materials, the massive geological formation and, finally, dilution, dispersion and

retention in the geosphere, minimize the probability of radionuclide escape to

the biosphere.

The Canadian Nuclear Fuel Waste Management Program has separate research

programs on each barrier. This paper describes the research programs on the

buffer and backfill barriers.

The term buffer refers to that material which immediately surrounds the waste

containers and isolates them from the surrounding rock and backfill (see figure

1). The ideal buffer would have the following chemical and physical properties

1. Low hydraulic conductivity to groundwater and very low porosity for ionic

diffusion.

2. Thermal conductivity equal to or greater than that of the surrounding rocks

3. Long-term chemical and physical stability at the temperatures, pressures

and groundwater conditions of the vault.

4. No unfavorable chemical reactions between the buffer, the waste container,

waste or the surrounding rock.

5. Good sorption properties to retard the movement of radionuclides, should

the containers be breached.

68
6. Ability to condition incoming groundwaters so that they are benign with

respect to the container materials.

The term backfill refers to that material used to fill any part of the nuclear

waste vault other than filled by the waste form or buffer. By this definition

(see figure 1), the backfill never comes into direct contact with the waste

containers and hence cannot react directly. The remaining physical and chemical

specifications for the backfill remain similar to those for buffer. The volume

of the backfill is much greater than that of the buffer, and this somewhat

restricts the choice of materials on grounds of effective resource utilization

and costs.

The buffer and backfill development programs are described separately below.

BUFFER DEVELOPMENT PROGRAM

A number of possible buffer materials have been identified, and a composition

of 10 - 20% of bentonite clay in crushed granitic rock and fine aggregate has

been suggested in our conceptual design studies (4). Additional materials

which will be investigated include kaolinite- and illite-based mixtures and

natural clays, which are mixtures of illite, montmorillonite and kaolinite. A

short list of potential clay-based buffer materials is being prepared on the

basis of existing literature and on our preliminary research results.

Bauxite, carbonaceous shales, fly ash, mine tailings and finely crushed rock

such as peridotite and dunite (which will hydrate readily and increase in

volume) are also being tested, although these materials are of more interest

as backfill.

Two alternative concepts suggested for waste form and buffer emplacement are shown schematically in figure 2. Each concept requires different handling

technology, and different physical properties would be expected for each type

of emplacement. For these reasons, the buffer development program has been

divided into an investigation of: (1) the physical properties, (2) the chemical

69
properties and (3) engineering technology. The logic of the buffer development

program is described and shown schematically in figure 3.

Physical Properties of Buffers

The physical properties of greatest concern in selecting a buffer material are thermal conductivity and diffusivity, hydraulic conductivity to groundwater,

load bearing capacity, thermal expansivity, thermal stability (5), and shrinkage

by drying.

Ideally, the thermal conductivity of the buffer should be similar to or greater

than that of the rock; otherwise heat dissipation from the irradiated fuel con- tainers to the surrounding rock may not be adequate and high temperature gradients could result. Maintaining a thermal conductivity equal to or greater

than its surrounding rock may not be achievable in practice since thermal

conductivity is highly dependent on mineralogy and water content.

Moreover, the thermal gradient within the buffer could induce water and salt

movement from the hotter zone to the cooler and vice versa. Each buffer material will be stable with respect to moisture migration for certain ranges of

water saturation and in this condition it is described as "thermally stable"

(5). Currently an experimental program is underway to test the thermal properties of various buffers. The main thrust of the initial work is to characterize

materials in terms of thermal conductivity and moisture migration as a function

of temperature, temperature gradient, etc. The moisture content of a thermally

stable buffer will depend on water pressure, grain size, hydraulic conductivity,

suction pressure of the buffer, density and groundwater chemistry. The overall

buffer program must address all these parameters and assess some of the

interactive effects.

The load bearing capacity of any potential buffer must be adequate to support

the mass of the containers so that they do not sink through the buffer and make

direct contact with the rocks. At the same time the buffer material should not

be brittle or susceptible to shrinkage cracking by water loss. The bentonitebased buffers recommended in the Swedish KBS-concepts (6, 7, 8) appear to meet

70
these physical requirements, although considerable research is required to

confirm and supplement the existing data and develop engineering-scale

techniques for handling and emplacement.

Thermal expansion and swelling pressures, generated as the buffer becomes

heated and takes up water, are also of concern. The total pressure from thermal

expansion and swelling must not exceed the crushing strength of the containers.

The current design for the waste container requires it to withstand crushing at

the hydrostatic head of water for the vault depth with a safety factor of two

(9). If pressures are generated in the buffer or backfill which exceed this

value, the waste containers could be deformed and premature failure could occur.

One of the research goals of the buffer and backfilling program is to ensure

that such a high pressure transient does not occur.

The rate and mechanism of water, salt, radionuclide and oxidant transport

through the buffer to/from the waste containers are two of the most important

parameters in buffer selection as they control both the physical and chemical

behavior of the system by contributing to both heat and mass transport. Empirical and theoretical relationships have been suggested for calculating mass

transport rates in soils at room temperature and below, at atmospheric pressure

(10, 11, 12). An experimental program is planned to test the usefulness of

these relationships at higher temperatures and pressures, and for temperature

gradient conditions. Salt and water migration will be measured for a range of

buffer densitities, water chemistries, pressures, temperatures and temperature

gradients in order to develop empirical or theoretical models which can be used

to simulate underground disposal conditions. This program will be integrated

with the buffer-chemistry research program and will provide experimental input

to the near-field mass transport modelling (see figures 3 and 4).

Chemical Properties of Buffers

The chemical parameters which are of concern in choosing a buffer composition

are: (a) total ion exchange capacity, (b) hydrothermal stability under the temperature and pressure conditions in the vault, (c) rates and types of chemical

reactions between the buffer, groundwater, waste container, waste form and host

71
rock, (d) the effects of the radiation field on the chemistry, (e) potential

for organic and biological reactions to alter the chemistry (f) the use of

buffer additives for redox control or selective sorption of radionuclides. For

the buffer-chemistry studies, other research groups will provide data on the

chemistry of the groundwater and how this chemistry changes as the water reacts

at elevated temperatures with the rock surrounding the disposal area (13, 14).

Ion exchange capacity of a number of potential buffer materials is being studied

and work is being carried out on diffusion of non-reactive solutes through

buffers. The results of these programs should give us a preliminary estimate

of the rate of movement of solutes through the buffer and will be supplemented

by elevated temperature and pressure experiments. The data will supplement the

results obtained by Nowak (10) and Neretnieks (11) and will be useful for

transport modelling.

The hydrothermal stability of the buffer in groundwater at elevated temperatures

will be studied for various groundwater compositions. The information existing

to 1979 has been summarized by Weaver (15) who notes that clay-mineral stabilities are highly dependent on composition of the interstitial water. With unfavorable water compositions, hydrothermal alteration can begin at temperatures

as low as 40°C for smectite-type clays. Hydrothermal reactions of the buffer

are of most concern if they lead to increased hydraulic conductivity and

increased mass transport of radionuclides. Hydrothermal reaction between the

buffer, waste container and/or the waste form must not lead to degradation of

the waste and enhanced release of the radionuclides.

An assessment of the effect of radiolysis is included in the study of buffer/

waste f orm/container/water interactions. Some preliminary studies have been

reported on radiolysis effects (16) and further work is underway, but it is

unclear whether the experimental conditions are relevant to the conditions of

disposal. Currently there is little evidence to suggest that radiolysis effects

will be significant.

Another aspect of the buffer chemistry which is being investigated is the role

of trace amounts of organic material which are present in practically all clay

72
deposits. This material could prove detrimental through the effects of

biological activity. Some organic materials can act as nutrients for bacteria,

and the bacteriological reactions can alter the pH and redox conditions in the

groundwater. This could accelerate container corrosion or increase reaction

rates between various buffer components. In addition, some organic molecules

can complex a variety of radionuclide anions and cations by increasing their

solubility and hence transportation rates in groundwater (17).

In addition to using natural materials for buffers, it may be possible to add

components which would create desirable chemical conditions. For example, it

is suggested in the KBS review (5, 6) that 0.5 percent f errophosphate (vivianite) be added to the buffer as an oxygen getter. Other suggestions have

included adding materials which could trap radionuclide anions (18). In evaluating these proposals, it is necessary to demonstrate that not only do the

chemicals function as predicted but also that no unfavorable reactions result.

For example, if f errophosphate is used, we must show that it will not serve as

a nutrient for biological activity or buffer the pH of the water-backfill system at unfavorable values, e.g., highly alkaline or highly acidic. In addition,

the material must not result in complexing and enhanced transport of the radionuclides should the waste containers be breached. Our research on these topics

is continuing.

Engineering Aspects of Buffering

The two irradiated fuel disposal concepts under consideration are illustrated

in figure 2. Each concept puts different restraints on the engineering and

emplacement of the buffer. In the first concept (figure 2a), the waste containers are placed on a layer of compacted buffer covering the disposal room

floor. Additional buffer is spread over and around the containers, using

radiation-shielded equipment, and compacted. The remaining free volume of the

disposal room would then be backfilled.

In the second concept (figure 2b), the radioactive waste containers are placed

in holes drilled into the floor of the disposal room. Two methods have been

suggested for emplacing the buffer around these waste containers. In one

73
method the container is placed on a pad of buffer at the bottom of the hole,

and powdered buffer is poured around the container and compacted by vibration.

An alternative method for buffer emplacement in boreholes has been proposed in

the Swedish KBS-project (8). They propose to compress blocks and rings of 90

percent bentonite - 10 percent water to a density of 2.1 - 2.2 Mg/m^, and place

these blocks and rings around the waste containers.

Each waste emplacement concept has a number of advantages and disadvantages.

For example, borehole disposal is particularly appealing from the point of view

of radiation safety. Moreover, a highly compressed buffer such as bentonite

has the advantage of creating a zone with very low hydraulic conductivity (8) as the buffer will swell by taking up water with which it comes in contact.

Swelling pressure built up should lead to the extrusion of some bentonite into

any pores and fractures in the rock, further inhibiting access of water. However, additional rock mechanics research effort beyond that required for the

backfilled drift may be necessary to establish the stability of the boreholes

and room floors.

If laboratory testing shows a buffer material and engineering concept to be

suitable, a number of practical aspects must be addressed. These include

development of methods of material preparation, handling, and quality control

on a suitable scale of operation, development of emplacement techniques for the

proposed disposal option, and inspection and testing of the emplacement method.

Some engineering feasibility studies are planned including a review of material

and transportation costs, ease of material handling, storage and quality control.

These are not particularly difficult engineering operations and hence are not

priority items in the research program. However, the results must be available

in order to plan the testing program for the underground research laboratory

proposed for construction in the Lac du Bonnet batholith near the Whiteshell

Nuclear Research Establishment. This program, which will be described in more detail by Simmons and Soonawala (19), is scheduled to begin around 1985/86.

74
Mathematical Modelling

Mathematical models are being developed for both buffer and backfill. The

models are designed to describe the physical and chemical behavior of the com- ponents within the vault. The chemical and physical models are being developed

separately.

a. Physical Modelling

The physical modelling has three objectives: (1) to assess the significance of

each component of the vault in terms of mass transport of radionuclides from

the vault, (2) to assist in developing materials performance specifications and

(3) to assist with developing environmental assessment models for the entire

vault

.

To develop a mass transport model for each component requires knowledge of:

1. The vault thermal-mechanical response, including short-term and steady-state

temperature and stress distributions.

2. The rate of water transport into and out of the vault.

3. The major pathways taken by water moving into, through and out of the vault.

4. The mechanism of water movement (e.g., by diffusive or advective flow)

within each vault compartment.

Sketches of three geometries we are trying to model are shown in figure 5. The

mathematical analysis has been developed for a two-dimensional diffusion model

for the borehole emplacement concept (see figure 5a and reference 20). The

analysis will be developed initially for a saturated buffer, and will later be

extended to account for the unsaturated-f low regime which would be present

during vault flooding. As data become available on the diffusion coefficients

for various solutes in the groundwater, mass transport of these will be

incorporated into the model.

75
During the later stages of the modelling work, it will be necessary to

supplement the diffusion model to include any advective component of mass transport in the regions where advection dominates the flow pattern. To do this, it

will be necessary to have better estimates than are currently available of the

hydraulic conductivities of the buffer, backfill, host rock and damaged zones surrounding the vault (figure 5b and c). Preliminary estimates of the hydraulic

conductivities of potential buffers and backfills will be available from this

program, and we are relying on the hydrogeology and rock mechanics research to

supply the information on the host rock and the damaged zone. The program

schedules require that a preliminary mass transport model be available for the

vault by the end of 1983.

b. Geochemical Modelling

Both thermodynamic (equilibrium) and kinetic data must be incorporated into the

geochemical modelling for the vault.

In the chemical treatment, the vault could be divided into a series of

compartments. The components of a particular compartment could be computed to come to a thermodynamic or kinetic "equilibrium". Water of "equilibrium" com- position could then be transported by advection or diffusion, using the physical

mass transport model described in the last paragraph, to another compartment,

and a computation of a new "equilibrium" composition could be carried out.

Using this procedure to transport water through a number of vault compartments

would give a time-dependent chemical transport model.

The buffer/backfill development program will provide chemical equilibrium and/

or kinetic data, mass transport mechanisms, pathways and rates in support of

geochemical modelling. In some cases, sub-models may be developed for

individual vault components.

BACKFILL DEVELOPMENT

Much of the physical and chemical characterization of probably backfill

components is being carried out through the buffer development program, and

7b
backfill research will concentrate on engineering and mathematical modelling.

Supply and quality control of the backfill material will be more difficult

because of the large volume requirements. The materials will have to be mined,

crushed, dried and transported to the repository site; crushed rock or aggregate

will be added and the mixture homogenized and tested before removal to the backfilling site. The type of equipment and techniques required for backfilling

will depend on whether the area being backfilled is a horizontal room, a drift

or a vertical shaft.

Room and Tunnel Backfilling

Several aspects of backfilling emplacement will require research and development

once the actual backfill composition has been selected. Emplacement techniques

and quality control will probably be the most important considerations. For

example, if fine-grained material is used as a part of the backfill, then some water may be required to minimize dust hazard and to ensure good compaction.

It will be necessary to know the optimum percentage saturation and how to control this percentage within the required range during emplacement. If the backfill can be emplaced with a suitable water content, it may be feasible to

utilize the swelling properties of materials such as bentonite to ensure that

the backfill penetrates voids and small fractures in the rock.

Control of porosity and hydraulic conductivity during emplacement will be

important as these factors could control the rate of water penetration to the

disposal zone. Moreover, if the backfill is emplaced wet and if the hydraulic

conductivity of the surrounding rock is low, the volumetric expansion of the

trapped water during the heating period following emplacement may control the

pressure regime. Because the irradiated-fuel waste containers are designed to

withstand a pressure of 10 MPa (9), care must be taken to ensure that pressure

does not exceed this value.

Regardless of the method backfilling, we will need to be able to predict the

physical response of the backfilling system and of the vault. A mathematical

model of vault behavior will have to be developed and be available prior to

full-scale testing of backfill emplacement techniques in the URL. This model

77
can then be used to compare predictions and observations in a carefully

instrumented full-scale backfilling operation in an appropriate mined room (19)

Shaft Backfilling

Shaft backfilling will differ from room and tunnel backfilling in several

significant ways. A different emplacement technology can be developed as the

shaft will be accessible to machinery at all levels and the work will be done

on a horizontal surface with adequate headroom at all times. As the shafts

will not be close to the waste disposal rooms radiation hazards will not be

present and temperatures will not rise significantly above ambient. This means

that there will be no thermal restrictions on the choice of materials. Ease of

compaction, long-term settling properties, load bearing capacity and hydraulic

conductivity of the compacted mass will be the most important parameters. Low

hydraulic conductivity will be required, not only through the backfill itself

but also along the contact between the backfill and shaft liner, or the sur- rounding rock if the liner has been removed. If it is not possible to obtain

a sufficiently low hydraulic conductivity along this interface, separate seals

may be required in some sections of the shaft.

Emplacement of the shaft backfill will require sufficient compaction to prevent

long-term settling which could leave voids in the shaft.

SUMMARY

The Canadian nuclear fuel waste management research program on buffer and

backfill development has four principal objectives:

1. To develop materials and technology for backfilling and sealing a waste

vault

.

2. To develop and execute a research program to provide experimental data and

theoretical support for objective 1.

3. To develop detailed mathematical models of buffer, backfill and sealing

barriers and use these models to assess the significance of each barrier.

78
4. To contribute to the development of a mathematical model for the long-term

safety assessment of the entire disposal vault.

The research program has been divided into evaluation of the physical, chemical

and engineering properties and mathematical modelling. Logic diagrams for the

programs are shown in figures 3 and 4. Initially the work is concentrating on buffer development because the buffer will control the source terms for any

radionuclide transport from the waste. The studies described in this paper will

lead into a large-scale testing program in an underground research laboratory.

REFERENCES

1. Boulton, J., (Editor) "Management of Radioactive Fuel Wastes: The Canadian

Disposal Program", Atomic Energy of Canada Limited Report, AECL-6314

(1978) .

2. Acres Consulting Services Ltd. and Associates, "A Disposal Centre for

Irradiated Nuclear Waste: Conceptual Design Study", Atomic Energy of

Canada Limited Report, AECL-6415 (1980).

3. Acres Consulting Services Ltd. and Associates, "A Disposal Centre for

Immobilized Nuclear Waste: Conceptual Design Study", Atomic Energy of

Canada Limited Report, AECL-6416 (1980).

4. Acres Consulting Services Limited, "Irradiated Nuclear Fuel and

Immobilized Waste Vaults: BAckfilling System", Atomic Energy of Canada

Limited Technical Record, TR-66 (1980).

5. Radhakrishna , H.S., "Heat Flow and Moisture Migration in Cable Backfills",

Ontario Hydro Research Quarterly _20, 10 (1968).

6. Karnbranslesakerhet , "Handling and Final Storage of Unreprocessed

Spent Nuclear Fuel, Volume II, Technical", Stockholm (1979).

79
7. Karnbranslesakerhet , "Handling of Spent Nuclear Fuel and

Final Storage of Vitrified High Level Reprocessing Waste", Stockholm

(1977).

8. Pusch, H. , "Highly Compacted Sodium Bentonite for Isolating Rock-Deposited

Radioactive Waste Products", Nuclear Technology 45, 153 (1979).

9. Cameron, D.J., "Fuel Isolation Research for the Canadian Nuclear Fuel

Waste Management Program", Atomic Energy of Canada Limited Report,

AECL-6834 (1982).

10. Nowak, E.J., "The Backfill Barrier as a Component in a Multiple Barrier

Nuclear Waste Isolation System", Sandia National Laboratories Publication,

SAND 79-1109 (1980).

11. Neretnieks, I., "Transport of Oxidants and Radionuclides Through a Clay

Barrier", KBS-Teknisk Rapport 79 (1978).

12. Taylor, S.A. and Cary, J.W. , "Linear Equations for the Simultaneous Flow

of Energy and Matter in a Continuous Soil System" , Proceedings of Soil

Science Society of America, 28 (1964).

13. Betcher, R.N. and Pearson, R. , "Hydrogeological Research for the Canadian

Nuclear Fuel Waste Management Program", Atomic Energy of Canada Limited

Technical Record, TR-151 (1982).

14. Sargent, F.P., "Applied Chemistry and Geochemistry Research for the

Canadian Nuclear Fuel Waste Management Program" , Atomic Energy of Canada

Limited Technical Record TR-149 (1982).

15. Weaver, Charles E., "Geothermal Alteration of Clay Minerals and Shales:

Diagensis" , Office of Nuclear Waste Isolation Technical Record, ONWI-21

(1979).

80
16. Allard , B., Kipatsi, H. , Rydberg, J., "Radiolysis Effects in Packing

Material", KBS-Teknisk Rapport 56 (1977), (AECL miscellaneous translation

464).

17. Mean, J.L. and Hastings, D.W. , "Status Report on the Importance of Natural

Organic Compounds in Groundwater as Radionuclide-Mobilizing Agents", Office

of Nuclear Waste Isolation Technical Report, ONWI-84 (1979).

18. Bird, G.W. , "Geochemistry of Radioactive Waste Disposal", Geoscience Canada

6, 199 (1979).

19. Simmons, G.R. and Soonawala, N.M. , (Editor) "Underground Research

Laboratory: Underground Experimental Program" , Atomic Energy of Canada

Limited Technical Record, TR-153 (1982).

20. Cheung, S., Bird, G.W. , and So, C.B., "Diffusion Modeling of the Borehole

Emplacement Concept for a Nuclear Fuel Waste Disposal Vault," Proceedings

of the NEA Workshop on Near Field Phenomena in Geological Repositories for

Radioactive Waste, NEW/OECP, Paris, pp. 375-388 (1981).

81
PLUGGED & SEALED

BOREHOLE

SHAFT LINER

GROUTED ZONE

OF WEAK ROCK

ROOM BACKFILLING

\ GROUTED ZONE ^ AROUND SHAFT

SHAFT SEAL

SHAFT > BACKFILLING

I

BUFFERED WASTE CONTAINERS

Figure 1. Illustration of components of a backfilled nuclear fuel waste disposal

vault (not to scale). See figure 2 for illustrations of two possible

waste container/buffer/backfill layouts

82
IN-ROOM EMPLACEMENT BOREHOLE EMPLACEMENT

Figure 2. Schematic illustration of the two disposal concepts (not to scale)

83
84
REVIEW "STATE-OF- ART" BACKFILLING TECHNOLOGY

LIST COMBINATIONS

OF MAJOR & MINOR COMPONENTS

PRELIMINARY COST & ENGINEERING

FEASIBILITY STUDIES

SHORT LIST PROMISING COMBINATIONS —> —f> —

PHYSICAL PROP. EVALUATION (PERMEABILITY, COMPACTION,

SWELLING, ETC.]

CHEMICAL PROP. EVALUATION (WATER CHEMISTRY, ION EXCH. REDOX VALUES, ETC.)

DEFINE RANGES OF CHEMICAL & PHYSICAL PARAMETERS

TEST WITH NEAR-FIELD MASS TRANSPORT MODEL

HYDROLOGY & ROCK MECHANICS i PROGRAM INPUT /

REFERENCE CONCEPT

FOR VAULT ASSESSMENT

GENERIC RESEARCH

CONTINUES

CONTINUE ENGINEERING DEVELOPMENT IN URL

Figure 4. Logic diagram for the backfill development program

* Refers to activities which are part of buffer development program

(see figure 3)

85
DIFFUSIVE TRANSPORT ADVECTIVE TRANSPORT

z

Figure 5. Illustration of the geometry of waste, buffer and backfill emplacement

concepts being modelled. In (A) all mass transport is by diffusion

and flux gradients can be calculated for the vertical (Z) and radial

(R) directions. In (B) and (C) both diffusion and advection are con- sidered. The relative contribution of diffusion advection can be

calculated by varying the hydraulic conductivities of the buffer,

backfill, host rock, and the thickness and hydraulic conductivity of

the damaged zone which exists around the room and borehole

86
EVALUATION OF THERMAL PROPERTIES OF

BUFFER MATERIALS FOR A DEEP UNDERGROUND

NUCLEAR WASTE DISPOSAL VAULT

H. S. Radhakrishna

Research Division

Ontario Hydro,

Toronto, Canada

ABSTRACT

Thermal properties of four types of sand/clay mixtures were investigated to

evaluate their suitability for buffer material. Tests were done on mixtures of

sand/clay over a range of mix proportions, moisture content and inter-particle

density and temperature, using a transient heat probe method. Sealbond (illite)

and kaolin exhibited better thermal-conducting properties than the swelling

clays, namely bentonite and zeolite. Graded silica and crushed granite with

about 15 to 25 percent clay added to it appeared to produce a workable material

with good compactability , strength and thermal properties. Tests for evaluating

the temperature and hydrostatic pressure on the permeability, strength and

volume change characteristics of the materials are underway.

INTRODUCTION

Safe and permanent disposal of radioactive waste requires isolation of a number

of diverse chemical elements from the environment for a long time. The Canadian

approach to nuclear waste management is disposal in vaults mined deep into

crystalline rocks of the Precambrian shield (1). This underground disposal

vault will consist of a number of chambers, some 1000 meters below the surface,

interconnected by a system of tunnels and access shafts as described in the

Conceptual Design Reports by Acres Consulting Services Ltd. (2, 3). In the

Canadian concept the nuclear waste, packaged in canisters, will be emplaced in

boreholes drilled from the floor of the chamber about 4.7 meters deep, as shown

in figure 1. The annular space between the host rock and the canister will be

filled with a suitable material called the 'buffer'. The repository vault itself

87
is filled with a material called 'backfill'. Thus the primary difference

between the buffer and the backfill is in the fact that the buffer is in direct

contact with the canister while the backfill is not.

An alternative concept to the borehole emplacement is the room emplacement in

which the canisters will be embedded in a buffer mass within the rooms also

shown in figure 1.

After disposal operations have been completed, the design criteria requires

backfilling and sealing of the entire vault in such a manner as to isolate any

harmful chemical elements that may be released from the waste either by

accidental or natural breakdown of the canister.

In the post-closure period, the only plausible failure mechanism which can be

identified is the prospect that the deeply circulating ground water may penetrate

to the waste, leach out radionuclides and carry them back to the surface. As a defense mechanism against such a transport, a series of engineered barriers are proposed in the vault design concept. These are the waste form itself, the

canister, the buffer, backfills and the seals or plugs for the mined shafts and

boreholes. These, in consort with the geological barrier, should perform the

job of isolating the waste from the human environment.

The buffer and backfill, as engineered barriers, are required to perform the

following functions: buffer the chemistry of ground water to enhance the life

of the canister; capture and retard the migration of radionuclides if they are released from the waste form; provide mechanical support to the canister and

also absorb any excessive stresses induced by movements of the rock; and

dissipate the heat generated by the waste form, in the process of its decay, to

the surrounding rock mass.

The physical and chemical attributes that the buffer material should possess in

order to perform the above mentioned functions effectively over long periods of

time are discussed in detail by Bird (4). A research program to develop buffer

materials has been planned (5).

88
The three phases in the buffer/backfill development program are: (i) preliminary

selection and evaluation of materials, (ii) research into their physical and

chemical properties, (iii) engineering development.

The study described in this paper deals with the heat transport aspects of the

buffer/backfill materials. The objectives of this study are to evaluate the

thermal properties of the candidate materials and to assess the heat transport

mechanism and the associated moisture migration phenomenon in the buffer/backfill

media. The results presented in here will supplement the results from other

studies in the buffer development program to enable a selection of appropriate

materials and installation techniques.

CANDIDATE MATERIALS

For reasons of longevity and durability, geological materials such as rock

flours and clays are the most favored candidates for buffer/backfill material.

Swedish programs have investigated the bentonite-based materials for buffer,

backfill and other seals for the deep underground disposal vault. Laboratoryscale experiments by Pusch (6) indicated several positive attributes of highly

compacted bentonite, namely its low permeability, self-sealing ability by

swelling upon wetting and good ion exchange capacity.

Some of the disadvantages of using bentonite are its low thermal conductivity,

high swelling potential and chemical instability at temperatures of 100°C and

above. In the Swedish conceptual design, the maximum temperature in the buffer

mass is limited to 100°C, however, in the Canadian reference design, the temperature at the canister surface could exceed 100°C. The limitations on maximum

permissible temperatures in the buffer mass and host rock (for reasons of

thermal spalling) will obviously limit the amount of radioactive waste that can

be emplaced per unit floor area.

Research programs in the United States are currently investigating a range of

materials (including cement, chemical resins and geological products) for various

vault seals. The chemical behavior of clays and crushed rock under the anticipated vault environment is being evaluated (7). Specially formulated backfills

89
designed to adsorb actinides are suggested by Bell and Allard (8). Ground

basalt and salt beds are also being examined for the backfill material (9).

THERMAL PROPERTIES OF BUFFER/ BACKFILL MATERIALS

Since the decay process of the emplaced radioactive waste generates heat, the

buffer backfill and host rock surrounding the canister will act as both heat

sink and heat transfer media. The heat stored in the buffer will be relatively

insignificant in comparison to that in the backfill and host rock because of

its small volume. However, the buffer/backfill will be the only channel for

heat flow between the canister and the host rock. A buffer and backfill with

good thermal conductivity would prevent excessive temperature build up in the

canisters and would also help distribute the heat more uniformly in the rock

mass

.

Since the buffer and backfill substance is likely to be a particulate material

placed in a moist condition, its effective thermal conductivity will be a function of parameters such as composition, thermal properties of the components

and their geometrical arrangements. Typical thermal conductivities of component

materials in a soil-like medium are summarized in table 1

.

From this table it is apparent that air has the lowest thermal conductivity of

all the components in a soil, water has a thermal conductivity about 25 times

that of air, and solids have thermal conductivities varying from one to 15

times that of water. Because of this wide variation in the thermal conductivities of the component materials, the effective thermal conductivity of the

medium is dependent on the volumetric proportions of each component and their

arrangement

.

Heat transport through soil media has been a subject of investigation for more

than a decade. The basic quantitative physics involved in the prediction of

the effective thermal conductivity of granular material is contained in the

fundamental theoretical work of Maxwell (10), which was extended and modified

by DeVries (11). Attempts have also been made by a number of investigators,

notably Kersten (12), Knutsson (13) and Jackson (14), to calculate the thermal

90
conductivity of soil-like materials, with different degrees of success. Work

carried out by these investigators has resulted in a better understanding of

the factors affecting the heat transport qualities of soil. Some of the highlights of the current understanding on this subject that are pertinent to the

evaluation of thermal properties of the candidate buffers and backfills are summarized below:

1. At normal temperatures the convective heat flux in soil media is negligible

for pore spaces smaller than 0.15 mm. However, the radiative flux is a function of the cube of the mean ambient temperature and, therefore, it may

not be negligible at very high temperatures (15).

2. The effective thermal conductivity increases with the size of particles;

thus coarse sands have higher thermal conductivities than silts and clays

(16).

3. The thermal contacts or bridges created by the liquid film at the solid

particle interfaces are the principal contributors to the significant

dependence of the effective thermal conductivity of soil to its liquid (or

water) content (17).

4. Effective thermal conductivity of the composite soils will be greatest when

the continuous matrix is the component with the highest thermal conductivity,

and volume fractions of the discrete, lower thermal conductivity components

are lowest. To increase the thermal conductivity of sand-like materials it

is necessary to couple adjacent particles with a binder whose thermal

conductivity is high with respect to that of air. This is most effectively

done by using a fluid binder with a heat-conducting filler which remains

adsorbed to the particles even under high thermal gradients (14).

5. In practical terms the effective thermal conductivity of granular fill

material can be adequately improved by selecting quartz-based solid particles

as the major component with a clay binder such as kaolin or calcium carbonate

(limestone dust) in proportion of eight to ten percent by weight, and

compacting to a high degree of interparticle density (19).

91
In summary no analytical models are available that can predict the effective

thermal conductivity of a wide range of moisture content and temperature for the

composite materials including clays. As a result, experimental techniques to

measure the thermal conductivity of the candidate buffers have been employed for

this study.

EXPERIMENTAL TECHNIQUES

Thermal conductivity of solids and particulate materials is determined by either

steady-state heat flow measurements or transient heat dissipation measurements.

The steady-state methods have the disadvantages of producing errors in measurements due to moisture migration in the direction of thermal gradients in moist

soils and creating convection currents in saturated soils. Furthermore, these

steady-state experiments are time consuming.

Transient heat dissipation methods are generally free of the above disadvantages

and the measurement techniques are relatively simple and fast. Thus the

transient heat dissipating method is favored for routine measurement of the

thermal conductivity of particulate materials. In the transient heat dissipating

method, which was used in this study, a slender probe simulating a line heat source is installed into the test sample and the temperature rise of the probe

under a constant heat input is monitored over a period of 15 minutes. The

slope of the slight line portion of the temperature vs log time plot is used

for calculating the thermal conductivity of the material. Thermal conductivity,

k, of the test material is calculated as:

t

k

2.303 W log

10 V E L(T2 -Ti)

where W = input thermal power for the probe in watts

L = length of the probe (m)

T\ and T2 = temperatures of the probe (°C) at times t^ and t 2 respectively

92
A microprocessor-based instrument was used to control the constant rate of heat

input to the sample, to collect the temperature data at specified time steps

and to fit a straight line to the data points by using the least-squares technique.

Figure 2 illustrates the logic diagram for this thermal property analyzer.

MATERIALS TESTED

The buffer materials tested consisted of mixtures of coarse-grained and finegrained materials in different proportions. The coarse-grained materials were:

- uniform graded Wedron silica sand

- graded silica sand

crushed granite

The fine grained materials (clays) used:

Blackhills bentonite

air-floated Georgia kaolin from Harrison and Crossfield (Can) Ltd, Montreal

Sealbond (air-floated shale) from Domtar Inc , Toronto

zeolite from Occidental Minerals, Wyoming

The materials were mixed in different proportions to form the test specimens.

The physical properties of these materials are summarized in table 2. The

chemical composition of these materials is summarized in table 3.

The silica sand used in this study was procured from local suppliers and was blended to produce a particle size gradation similar to the one used in Swedish

studies (20). The crushed rock was produced from the granite of the Lac du

Bonnet batholith near Pinawa, Manitoba. This rock outcrop is considered to be

similar in geochemical composition to the rock at depth. Results of detailed

chemical analyses of this rock are described in reference (21).

93
The following mixes were tested:

l Wedron sand + Avonlee bentonite (WSAB)

9 Graded silica + Blackhills bentonite (GSBB)

J o Graded silica + sealbond (GSSB)

A \J L a. U t: Li o X X J. v_ d r

t; 3 • Graded silica + zeolite (GSZ)

D o Crushed granite + Blackhills bentonite (CGBB)

7. Crushed granite + sealbond (CGSB)

8. Crushed granite + kaolin (CGK)

Clay proportions in the above mixes were varied from 15% to 50%. The grain size

distribution curves for individual components are shown in figures 3 and 4.

SAMPLE PREPARATION AND TESTING

The sand/clay mixtures were blended dry in proper proportions by dry weight and

then mixed with water in a Hobart mixer to produce a uniform and homogeneous

material

.

Several methods of mixing and different periods of curing after mixing were

tried to select a procedure could produce homogenous material and consistency

in test results. Mixtures with clay proportions greater than 25% by weight,

especially those with bentonite, were generally difficult to blend. They

also produced the same degree of inhomogenity in the moisture content.

The mixed materials were compacted into standard Proctor moulds (102 mm diameter

by 115 mm high) using standard Proctor compaction energy conforming to the ASTM

procedure (D-1557).

THERMAL CONDUCTIVITY OF CANDIDATE MATERIALS

Thermal conductivity measurements were made on each of the Proctor-compacted

samples, both in its as-packed condition and over-dried (at 105°C) condition.

The results of these measurements are summarized in table 4. For mixes 2, 3

94
and 4 the thermal conductivity as a function of moisture content at a constant

dry density was established by making thermal conductivity measurements on large

samples (200 mm by 350 mm) subjected to stage drying. The results of these

tests are presented in figures 5 and 6. Thermal conductivity measurements were

made on bentonite and sealbond based mixes at elevated temperature (100°C) to

examine the temperature effects on thermal properties of candidate materials.

The following observations can be made from the test data:

1. Use of graded silica instead of uniform sand improves both the compaction

and the thermal conductivity of the mix materials. However, the crushed

granite mixes produce' slightly lower thermal conductivity values than the

graded silica mixes.

2. Sample density and thermal conductivity decreases significantly with the

increase in clay content. This effect is more pronounced in the case of

bentonite based mixes (figure 7).

3. The thermal conductivity of the bentonite based mixes varied from 1.0 W/

(°C m) to 3.0 W/(°C m) depending on the percentage of bentonite and the

type of sand used. The lowest thermal conductivity values occurred in the

dry state of materials which ranged from 0.5 W (°C m) to 1.0 W/(°C m)

.

4. Of the four sand/clay mixtures tested, the mixes containing 15% kolin or

15% seal bond had the best thermal conductivities. These were in the range

of 1.5 W/(°C m) in totally dry state to 4.0 W/(°C m) in moist state.

5. Zeolite based mixes had somewhat better thermal conductivities than the

bentonite based mixes but the samples appeared to be somewhat porous due to

the large percentage of silt size particles in zeolite.

6. A marked reduction in thermal conductivity occurred upon drying for all of

the mixes. The thermal conductivity appeared to be far more dependent on

the soil moisture than the ambient temperatures of up to 100°C (figure 8).

95
7. For all of the mixes, there appeared to be a critical moisture level below

which a marked change in the thermal conductivity would occur, and for the

mixtures with 15 percent clay the critical moisture content was about 2 percent. In earlier studies of backfills for underground cables, it has

been demonstrated that the critical moisture content at which a sharp change

in thermal conductivity occurs also corresponds to the condition of highest

moisture migration in vapor phase and thermal instability (22). Moisture

migration experiments to be performed next will examine the potential for

thermal instability in the candidate mixtures.

CONCLUSIONS

Buffer/backfill materials made up of sand/clay mixtures will have adequate heat

transport qualities with sealbond based mix having the best thermal properties

of the four sand/clay mixes tested. None of the mixes will, in dry state, be

as conductive as the host rock. Good compaction and moisture control will be

necessary to achieve the desired thermal properties. Preliminary tests suggest

that a high ambient temperature does not alter the thermal properties significantly. The high degree of moisture dependency of the heat conduction properties

of the buffer/backfill should be considered in modelling the thermal transients

in the repository vault. A preliminary thermal analysis with a thermal conductivity value of 0.5 W/°C m for buffer should give the upper limit for canister

temperature.

REFERENCES

1. J. Boulton (Editor), "Management of Radioactive Fuel Wastes" The

Canadian Disposal Program", Atomic Energy of Canada Limited Report,

AECL-6314 (1978).

2. Acres Consulting Services Ltd. and Associates, "A Disposal Centre for

Irradiated Nuclear Fuel" Conceptual Design Study" , Atomic Energy of

Canada Limited Report, AECL-6415 (1979).

96
3. Acres Consulting Services Ltd. and Associates, "A Disposal Centre for

Immobilized Nuclear Waste" Conceptual Design Study", Atomic Energy of

Canada Limited Report, AECL-6416 (1979).

4. G. W. Bird, "Possible Buffer Materials for Use in a Nuclear Waste Vault",

Atomic Energy of Canada Technical Record, TR-72 (1979).

5. G. W. Bird, "Vault Sealing for the Canadian Nuclear Fuel Waste Management

Program", Atomic Energy of Canada Limited Technical Record, TR-145 (in

preparation)

.

6. R. Pusch, "Highly Compacted Sodium Bentonite for Isolating Rock-Deposited

Radioactive Waste Products", Nuclear Technology 45, 153 (1979).

7. E. J. Nowak , "Radionuclide Sorption and Migration Studies of Getters for

Backfill Barriers", Sandia National Laboratories, Sand 79-1110 (1980).

8. G. B. Bell and B. Allard , "Chemical Aspects Governing the Choice of Backfill Materials for Nuclear Waste Repositories", Proceedings of Symposium

on Backfill as an Engineered Barrier for Radioactive Waste Management,

American Chemical Society, 1980, August 24-29.

9. M. I. Wood and W. E. Coons, "UMTANUM Basalt as a Potential Backfill

Component in a Repository Located Within the Columbia River Basalt",

Proceedings of Symposium on Backfill as an Engineered Barrier for Radioactive Waste Management, American Chemical Society, 1980, August 24-29.

10. J. C. Maxwell. A Treatise on Electricity and Magnetism. Oxford

University Press, London, 1892.

11. D. A. De Vries. Thermal Conductivity of Soil, Mededel, Landbouuhogeschool , Wageningen, Netherlands, 1952.

12. M. S. Kersten, "Thermal Properties of Soils", Engineering Experimental

Station, University of Minnesota, Tech., Bulletin 29 (1949).

97
13. S. Knutsson, "Thermal Conductivity of Highly Compacted Bentonite",

Karn-Bransle-Sakerhet Teknisk Rapport 11 (1977).

14. K. W. Jackson, "Enhancement of Thermal Energy Transport Through Granular

Media", Ph.D Thesis, Georgia Tech, 1980.

15. H. W. Russell, "Principles of Heat Flow in Porous Insulators", J. Amer.

Ceram. Soc. 1_8, 1 (1935).

16. W. Woodside and J. H. Messmer, "Thermal Conductivity of Porous Media.

I. Unconsolidated Sands", J. Appl. Phys. 32^, 1688 (1961).

17. A. Gemant , "The Thermal Conductivity of Soils", J. Appl. Phys. 2_1_, 750

(1950).

18. L. D. Bauer, W. H. Gardner and W. R. Gardner, Soils Physics, Wiley, 1972

19. J. I. Adams and A. F. Baljet, "Thermal Behavior of Cable Backfill

Materials", IEEE Transactions on Power Apparatus and Systems, PAS-87

,

1149 (1968).

20. S. Knutsson, "Heat Conduction Experiments with Buffer Substance of

Bentonite/Pite Silt", Karn-Bransle-Sakheret Teknisk Rapport 28 (1977).

(Atomic Energy of Canada Limited Miscellaneous Translation, AECL-trMISC-611).

21. Analytical Science Branch, Whiteshell Nuclear Research Establishment,

"Analysis of Surface Outcrop Rocks in the Lac du Bonnet Batholith, Field

Samples 1977", Atomic Energy of Canada.

22. H. S. Radhakrishna , "Heat Flow and Moisture Migration in Cable Backfills'

Ontario Hydro Research Quarterly 20, #2 (1968).

23. S. P. Clark, Jr., "Handbook of Physica 1 Constants, revised edition",

Geological Society of America, Momoir 97 (1966).

98
24. S. Knutsson, "Thermal Conductivity Tests on Highly Compacted Bentonite"

,

Karn-Bransle-Sakerhet Teknisk Rapport 72 (1977). (Atomic Energy of Canada

Limited Miscellaneous Translation, AECL-tr-MISC-468 (1978).

25. R. C. Weast (Editor), CRC Handbook of Chemistry of Physics, 55th edition,

Cleveland, Ohio, CRC Press, 1975.

99
Table 1 Thermal Conductivity of Geological Materials

Dry Thermal

Material State Moisture Density Conductivity Information

Content W/(°C m) Source

(%) (kg/in 3

) Moist Dry

Quartzite Rock natural 1.5 2400-2700 9.0 8.0 23

Granite Rock natural 2.0 2600 4.0 3.3 23

Limestone natural 4.0 2400-2600 3.5 2.5 23

Sandstone natural 2.0 2300-2700 4.0 3.0 23

Shale natural 4.5 2000-2300 1.8 1.0 23

Crushed Quartz compacted 6.0 2200 3.0 1.6 * Crushed Limestone compacted 6.0 2100 2.2 0.8 A

Crushed Shale compacted 10.0 2100 1.3 0.6

Silty Clay compacted 14.0 1800 1.3 0.6 A

Ottawa Sand compacted 5.0 1700 2.0 0.5 A Glacial Till compacted 12.0 1900 2.3 0.8 A Bentonite Clay compacted 25.0 2020 0 9 0 .

8

24

Coal Ash compacted 15.0 1500 1.0 0.3 A

Water non- convective 0.63 25

Air non- convective 0.025 25

* Ontario Hydro, unpublished data

100
Table 2 Physical Properties of Components in Candidate Buffers

Air Dry

Size Fractions Moisture

Specific (%) Content

Component Gravity Sand Silt Clay (%)

Wedron Sand 2.65 100 0 0 0.3

Graded Silica 2.70 90 10 0 0.5

Boundary Dam Fly Ash 2.60 11 87 2 — Blackballs Tower Bond Bentonite 2. 18 0 18 82 8.2

Avonlea Bentonite 2.45 0 39 61 8.8

Avongel Bentonite 2.21 0 12 88 9.3

Sealbond

- 2.76 0 65 35 1.4

Kaolin 2.60 0 30 70 1.1

Zeolite 2.33 35 60 5 4.4

101
0 O <r 1 cx r~~ CO

t—i c o

<u '1—1 00

4-) 0 O ~. LO 1—1

CO

C s

* Q sj CO

O tiQ 1

i>

—1 r—

0 1 1 CO CO 00

CN] 11 |I rH CO

OJ >

1-7—

n i O 1 i X- ^j- CO 1 11 * nwi

-J t—

—J | ) • > t 1

CO rH Lji

rrt /—1 0 1 —(—

CCm ,~r* 'rH (—1 ( 1 \—1 (—-\ -

' Q

j_l Kn * *

rH

IU

Lrj

,1 I /—1 . —\ V—

M—I

/—

M cO *

/—i

<«J

<-t

rUS O —> •H

*j

•rH CO

QQ CNJ 1 s '—> /-\ 1—M c—1 1 —| Q

CO P •H

"~® | <—1

•rH CN

•rH rrl /—s ( J

Q ' 1

S-l

U I 1 11 •rH —1 P

d)

i—1 rCr -\

CO en O CN CN

O 0 1—1 o CM O O ro O m

CD • rH Tt . O t

* rsj CN] , ^ cOoHB ro

cu

OO 0 r- CNJ 1 1 1 1 1 s ' 

* CO

r -\ - X * JU rH

CN OO 00 ' ' O * 0 •H —

C/} 0^ *

CO •K 0

13

H XI

G 1—1

CO c CD CD *

CO

H rH 0 tJ -X co co rH pp •H •H T3

CO •H c H c C *

•HG XI jG 0 0) 0 O c

S-iO 0) ^ OJ

u 4-1 •H

0)

M T3O

dd

d rH rH

•u TJ CO COO 0) 0 0) CO 0

CO CU

H rH

H pp > pq CD CO

a 3:

O CQ

m

u

CUu

4-1 O

CO 4H cO4J d CO T3 S

TJ

>^ CU -Q

Cfl T3 -H

OJ rH

•H X) rH 3

Q*

CX c 3 3

Cfl r-J CO O

4J

CO Q* -X

102
Table 4

Thermal Conductivity of Candidate Buffer/Backfills

Thermal Thermal

Moisture Conductivity Conductivity

Mix Content Dry Density in Moist in Dry

Composition.!/ Range Range Condition Condition

(%) (kg/ml) W/°C m W/°C m

WSAB-15 rD

i "7 1 / 11UU •- 1 / / u 1 . J Z . o 0. / 1 .

0

WSAB-25 5 16 1200 • - 1760 1.2 2.7 0.6 0.9

WbAB JU 1 D iX JjU - i add 1 9 1 * Z 9 n U.J n a

GbBB-lU £.D

i n1U i O jU zuuu Z • U j . U U . H U. / GSBB-25

/^CTJD bMD jU

GbbB- 1

J

A Q ZUjU Zi/J i n J . U1 c

1 . j 1 . / GSK-15 4 9 2000 - 2200 2.5 3.5 1.8 2.0

GSZ-15 5 12 1725 - 1850 0.8 3.5 1.8 2.0

CGBB-15 5 19 1750 - 2000 1.3 2.6 0.8 1.0

CGBB-25 6 12 1620 - 1710 1.0 1.5 0.6 0.8

CGBB-50 12 17 1500 - 1550 0.8 1.0 0.5 0.6

CGSB-15 6 10 1890 - 2200 2.4 3.4 1.4 1.8

CGSB-25 7 12 1600 - 2050 1.4 2.4 0.8 1.0

CGSB-50 9 15 1850 - 1980 1.8 2.4 0.9 1.0

CGK-15 6 10 1900 - 2050 2.2 3.0 1.7 2.0

CGK-25 9 12 1850 - 1980 1.7 2.5 0.9 1.2

CGK-50 9 15 1770 - 1860 2.

1

2.2 1.0 1.2

]J Notations for Mix Compositions:

WS = Wedron Sand

GS = Graded Silica

CG = Crushed Granite

AB = Avonlea Bentonite

BB = Blackhills Bentonite

SB = Seal Bond

K = Kaolin

103
V

104
Program

Microprocessor

Output

Temp vs Time TR, TD at Low Power

TR, TD at High Power

T hermocouple

Reader

High Power

Power

TR Probe Low Power

Time

Figure 2. Logic diagram of the thermal properties analyzer

105
ure 3. Grain size distribution curves for the coarse fraction of buffer

mixes

106
#140 #50 #30 #10 3/4" 1-1/2" 6"

0.001 0.002 0.006 0.02 0.06 0.20 0.60 2.0 6.0 20.0 60.0 100 GRAIN DIAMETER (mmj

Figure 4. Grain size distribution curves for the fine fraction of

buffer mixes

107
GSS-15

8" x 14" SAMPLES

DRY DENSITY = 2055 kg/m3

MOISTURE CONTENT = 6.0% dry wt

3 4 5 6 MOISTURE CONTENT

10

E

CJ

</> 3

i—

r

GSBB-10

8" * 14" SAMPLE

DRY DENSITY = 1836 kg/nr*

MOISTURE CONTENT = 6.9% dry wt

J I I . I

i

I 3 4 5 6 7 iOSSTURE CONTENT [%)

10

Figure 5. Thermal conductivity vs moisture content relationship at

constant density from stage drying tests for

A) 85% Graded Silica + 15% Sealbond Mix

B) 90% Graded Silica + 10% Black Hills Bentonite Mix

108
GSK-15

8" x 14" samples

Dry density = 2068 kg/m3 Moisture content = 6.9% dry wt

MOISTURE CONTENT (%)

Thermal conductivity vs moisture content relationship at

constant density from stage drying tests for 85% graded

silica + 15% kaolin mix

109
20 40 60

CLAY CONTENT (% by dry wt)

CLAY CONTENT (% by dry wt)

Figure 7. Thermal conductivity and density as a function of clay content in

crushed granite and clay mixes

110
o

I

4.0

3.0

« 2.0

1.0

600

GSSB-25

J 1600 1700 1800

DRY DENSITY (kg/m*)

GSBB-25 -

2000

o0

25 50 75

TEMPERATURE [°C]

• GSSB-25

oGSBB-25

100 125

Figure 8. Influence of density and temperature on thermal conductivity

of crushed granite - clay mixtures

111
BENTONITE SHIELDING OF ROCK- DEPOSITED RADIOACTIVE WASTES

Roland Pusch

Division of Soil Mechanics

University of Lulea

Lulea, Sweden

INTRODUCTION

The Swedish KBS concepts for final deposition of canisters with nuclear wastes

from reactors, imply that the canisters be placed in deposition holes which are bored from the floors of tunnels at about 500 m depth in Swedish crystalline

bedrock. The canisters are isolated from the rock by a "buffer mass" consisting

of pure bentonite in the case of unreprocessed wastes, and the deposition plant

is finally sealed by backfilling shafts and tunnels with bentonite/sand mixtures

(figure 1). The design, involved techniques, and expected barrier effects have

been described in various KBS reports. A short description of the properties

of highly compacted bentonite has been published by Pusch in 1979 (1).

The bentonite, which surrounds the canisters, is applied in the form of blocks

of highly compacted bentonite. The blocks are produced by isostatic compaction

of commercial bentonite powder under pressures in the range of 50-100 MPa, which

yields block densities of 2.1-2.2 t/m^ when using Wyoming Na bentonite MX-80

powder. The compacted bodies are cylindrical and they can be manufactured by

sawing, drilling etc. so that cylindrical discs, rods, annuli, and practically

any shape can be obtained (figure 2). This means that the canisters can be

enclosed in a tightly fitting stack of such blocks which are easily applied in

each deposition hole.

The particularly attractive properties of the highly compacted Na bentonite,

some of which will be commented in this report, are:

o Extremely low permeability

o Low diffusivity

o Considerable cation adsorption capacity

112
o Sufficient bearing capacity to support canisters

o Sufficient ductility to sustain minor rock displacements without

transferring very high stresses to canisters

o High swelling and self-sealing capacities

THE BENTONITE MATERIAL

Bentonite is the term for smectite-rich clay formed by devitrification of the

natural glass component of volcanic ash deposited in prehistoric lakes and

estuaries. The crystal lattice of montmorillonite_l/ is characterized by a variable c-axis spacing (figure 3) which provides the large internal mineral surface

area for cation adsorption, and the strong swelling ability.

The commercial bentonite powder for the production of compacted blocks is obtained

by simple grinding and drying of crushed raw material. The grains are silt-sized

and consist of face-to-face-grouped stacks of lamellae, i.e., the microstructural

order of the natural bentonite beds (figure 4).

Compaction of the powder under high pressure welds the aggregates together and

yields a strongly coherent, talk-like product. A convenient water content of

the bentonite powder is 10% for the production of blocks with a bulk density of

about 2.15 t/m3. This water content corresponds to a degree of water saturation

of 50-60%.

MICROSTRUCTURAL PROCESSES ASSOCIATED WITH WATER UPTAKE AND SWELLING

The water uptake and swelling are associated with particle rearrangement meaning

that properties like permeability and diffusion should only be determined after

the establishment of internal microstructural equilibrium. The driving forces

when partially saturated highly compacted bentonite takes up water are due to

the strong affinity of the montmorillonite to water, and to capillary action.

The aggregates start to take up water in their outer parts and thereby expand by

1/ The major smectite clay mineral is Wyoming bentonite, the KBS reference

clay.

113
which the interaggregate voids are successively filled with a clay gel. As long

as there are differences in water content in various parts of the system, there

is a tendency of adjustment so that a homogeneous microstructural state is finally

achieved.

The migration of water leading eventually to water saturation can be considered

as a diffusion process governed by a "concentration" gradient. The "concentration" equals to the water content, which has a fairly low value (approximately

10 percent at a bulk density of 2.1-2.2 t/m^) at the start of the water uptake,

and which finally reaches a certain maximum value (20-25 percent) under completely

confined conditions. The isotropic character of the highly compacted bentonite

suggests a simple form of the diffusion equation for uniaxial migration:

3w = d 9_ ( 3w)

3t 3x 3x

where w = water content

D ~ 3.10" 10 m2 /s

This finding is of great practical importance since it offers a simple way of

predicting the rate of water intake and the distribution of the water that has

been taken up by the clay blocks.

SWELLING PRESSURE

If a partly water saturated, compacted bentonite body has access to external

water and is confined so that its total volume remains constant, i.e. in principle the conditions in figure 1, the individual crystallites take up water in

connection with an expansion in the c-axis direction. When this expansion can

no longer proceed as a consequence of the limited space, a "swelling pressure"

is exerted on the confinement. Figure 5 illustrates a generalized relationship

between the swelling pressure ps and the bulk density MX-80 bentonite at complete

water saturation.

114
A close examination of the influence of pore water chemistry on the swelling

pressure has shown that the concentration and composition of dissolved salts

(maximum concentration corresponding to that of the oceans) did not have any

o noticeable effect on p g when the bulk density was higher than about 2 t/m . This is because electrical double layers are not, or incompletely developed at

o interlamellar distances less than 10-15 A. Instead, the high swelling pressure

at such high densities is related to the pressure of strongly fixed, "steric"-

type interlamellar water molecules. The matter is dealt with in various KBS

reports (2).

PERMEABILITY

The presence of largely immobilized interlamellar water leaves only narrow,

tortuous passages for water migration between stacks of montmorillonite sheets.

The resulting permeability is therefore extremely low (figure 6).

The interaction of mineral lattices and water molecules - mainly through

hydrogen bonding - is strong in the narrow passages, such as interlammellar

space, and less intense in wider passages, the variation in interaction being

stochastic. While water flow through clays of low density can be considered

as the motion of a fluid with a definite and constant viscosity, it should

rather be regarded as a shear-induced displacement of a "structured" medium in

the case of dense bentonites. This suggests a close resemblance to the creep

of clay. It is in fact illustrated by the obvious retardation of the flow rate

after changing the hydraulic gradient (figure 7), by the non-Darcy behavior,

and by the very obvious influence of temperature (3).

SELF-HEALING EFFECTS

The unique relationship between bulk density and swelling pressure suggests

that homogeneous conditions and uniform water distribution are attained in confined, sufficiently dense bentonite bodies if given enough time for the establishment of internal equilibrium. The strong ability to swell and become

homogeneous is illustrated by figure 8 which shows how 3 loose pieces of precompacted bentonite in a steel cylinder absorbed water from one end and turned

115
into one, coherent, swollen, and fairly homogeneous piece of clay after only 3

weeks. Such self-healing requires, however, that the average density is of the

order of 1.5 t/m3. Lower densities imply non-associated water and thus a possibility of the formation of permanent local zones with "excess" water.

The property of very dense bentonite to swell out and fill voids and form,

ultimately, a uniform clay body is the main principle of a new technique of

sealing boreholes (4). Perforated, closely fitting copper tubes filled with

highly compacted bentonite cylinders are inserted in the hole, its outermost

part then being sealed with cement. Water is taken up from joints which intersect the hole and a completely homogeneous state of the bentonite is expected

after a sufficiently long time. Wider joints require sealing by pretreatment

with bentonite, cement, or some other suitable substance. Pilot tests in 6 m

long holes in granite are presently being run in the Stripa mine.

REFERENCES

1. Pusch, R. , 1979, Highly Compacted Sodium Bentonite for Isolating RockDeposited Radioactive Waste Products, Nuclear Technology Vol. 45, Sep. 1979,

pp. 153-157.

2. Pusch, R. , 1980a, Swelling Pressure of Highly Compacted Bentonite, KBS

Technical Report 80-13.

3. Pusch, R. , 1980b, Permeability of Highly Compacted Bentonite, KBS Technical

Report 80-16.

4. Pusch, R. and Bergstrom, A., Highly Compacted Bentonite - A Self-Healing

Substance for Nuclear Waste Isolation, Proc . Mat. Res. Soc. Ann. Meet.

Nov. 16-21 1980 Boston, Mass.

116
£

Figure 1. Schematic view of tunnel and deposition hole containing canister

with unreprocessed reactor wastes. I) bentonite/sand , II) highly

compacted bentonite, C) copper canister

117
Figure 2. Example of precompacted ring of bentonite

118
LEGEND

® OH

O O

° Mg, Al

• Si, Al

Figure 3. Probable crystal structures of montmorillonite ; left model

according to Hofmann, Endell and Wilm; right model according to Edelman and Favejee. The intrastructural space contains water molecules and cations

119
Figure 4. Upper picture shows scanning micrograph of MX-80 aggregates.

Notice the tapered shape of the aggregates (lOOx). (Photo

D.M. Anderson, New York State University at Buffalo)

Lower picture illustrates, schematically, the

microstructural pattern of bentonite powder

120
2.5

2.0

1.5

1.0

0.1 10 100

P

s

, MPa

Figure 5. Approximate relationship between swelling pressure, P s /density

of MX-80 bentonite at complete saturation and 20°/C

121
Z5

2.0

m

3 1.5

1.0

V

> r44

10

43

10

k, m/s

-12

10

=11

10

Figure 6. The coefficient of permeability versus bulk density. Where two

values are plotted for one and the same density, the left repre

sents i = and the right 10^ (i = hydraulic gradient).

Broken curves are upper and lower boundaries of literaturederived values

122
GRADIENT

10

4 510

3

210

3

. 10

3

-13

3-10

-13

10 J 1 1 1 1 0 10 20 30 40

TIME, days.

Figure 7. Example of peremability test with smoothened curve for k versus time after the test start

123
3C

p ~13

3t

70 mm

4i 43

i

i tO

36 33(33; 32

| l_Ll__L_

I I

1

1 25)25-251 25 I n '

i —

i--

t22j22

, 22j 23 I

Figure 8. Swelling test with precompacted bentonite. The right picture

shows the average water contents of symmetrically situated

elements after 3 weeks

124
EVALUATION OF BACKFILL AS A BARRIER TO RADIONUCLIDE MIGRATION

IN A HIGH LEVEL WASTE REPOSITORY 1 ^

T. M. Ahn, R. Dayal , and R. J. Wilke

Department of Nuclear Energy

Brookhaven National Laboratory

Upton, NY 11973

ABSTRACT

The feasibility of using highly sorptive synthetic minerals such as zeolites or titanates as backfill in a HLW repository has been evaluated in terms of the NRC

1,000 yr containment and 10~5/yr controlled release criteria. The results indicate that for groundwater velocities below 1 ft/yr, diffusion and sorption are

the dominant processes controlling radionuclide migration in backfill systems.

A 3-ft-thick bed of synthetic zeolite backfill can provide total containment of

activity released continuously from a 55-gallon glass monolith for a period of

up to 1,000 years. For longer time periods of up to 10^ years, the controlled

release rate of 10~5/yr can be met by a >10-ft-thick bed of zeolite backfill,

assuming the glass monolith to be the source term. In the absence of the waste

form as an engineered barrier, a bed thickness of 3 to 10 ft is required to

Support for this work was provided by the U.S. Nuclear Regulatory Commission

Contract No. DE-AC02-76CH00016

.

This report was prepared as an account for work sponsored by an agency of the

United States Government. Neither the U.S. Government, nor any agency thereof,

or any of their employees, makes any warranty, expressed or implied, or assumes

any legal liability or responsibility for any third party's use, or the results

of such use, of any information, apparatus, product or process disclosed in

this report, or represents that its use by such third party would not infringe

privately owned rights. The views expressed in this paper are not necessarily

those of the U.S. Nuclear Regulatory Commission.

125
satisfy the controlled release criterion for a period of up to 10^ years. With

a bed thickness of 200 ft, synthetic zeolites can be used as backfill to provide

containment of the total U.S. waste inventory (109 Ci), if it were placed in one repository, for periods up to 10& years. Zeolites are known to exhibit

radionuclide-specif ic sorption properties. We believe that the range of sorption

coefficients (lO^-lO^ ml/g) considered in the calculations will cover the various

zeolite-radionuclides systems. We wish to emphasize that the above estimates

are conservative since diffusion occurs radially and we have only considered

uni-dimensional transport in our models. In addition, radioactive decay of the

nuclides has not been considered in the calculations.

We also wish to note that in this report we have not dealt with radiation and

thermal stability of backfill materials. Zeolites have been used to clean

radioactivity from water systems and have been loaded to as high as 75,000

curies per MTHM. Although we believe that zeolites should be structurally

stable to low radiation loadings (approx. 1,000 curies per cubic foot) and that

radiolysis efforts should not be important at these levels of radiation, very

little work exists in this area. In addition, we have not addressed the reten2-

tion of long-lived anionic species such as SeO ^, I-, TcO^ . It appears

that other materials will have to be added to backfill to maximize its

effectiveness for retention of all radionuclides of interest.

INTRODUCTION

The operational life of a geologic high level waste (HLW) repository after

decommissioning can be divided into two time periods: (a) the period (<1,000

yr) during which the short lived fission products dominate the hazard posed by

the waste, and (b) the long term period (>1,000 yr) during which the hazard is

dominated by the very long lived isotopes including the actinides [1]. In view

of these considerations, the Nuclear Regulatory Commission (NRC) had adopted a strategy for regulatory and licensing the disposal of high level radioactive

wastes in geologic repositories which requires that the engineered systems of

the underground facility meet the following criteria [2]:

126
a. Containment of all radionuclides in HLW for at least the first 1,000 years

after decommissioning of the geologic repository, assuming expected events

and processes. This containment shall result from properties of the waste

package.

b. Starting 1,000 years after decommissioning of the geologic repository, the

radionuclides present in HLW will be released from the engineered system to

the geologic setting at an annual rate that in no case greater than one part in one hundred thousand of the total activity present within the

underground facility at that time, assuming expected processes and events.

c. For transuranic waste (TRU) , the engineered system shall be designed so that following decommissioning of the geologic repository the annual

release rate from the underground facility into the geologic setting is at

most one part in one hundred thousand of the total activity present in the

underground facility at any time following decommissioning.

Most of the current Department of Energy (DoE) programs in high levels waste

research are investigating the waste form and container as potential components

of a waste package in meeting the containment requirements outlined above. In

this report, we have attempted to evaluate the effectiveness of using synthetic

zeolites and titanates as potential backfill materials in an HLW repository.

It should be pointed out that a DoE waste package is defined to include everything that is placed in the waste repository emplacement hole, i.e., the waste

form, filled container, overpack, sleeve, and backfill. The NRC waste package

includes discrete backfill [3] in contrast to non-discrete backfill in DoE

definition. Since the 1,000 yr criterion applies specifically to the waste

package and the 10~-Vyr controlled release criterion refers to the engineered

system, we have evaluated the effectiveness of backfill in meeting both performance requirements: discrete backfill for 1,000 yr containment and non-discrete

backfill for controlled release.

127
Zeolites and Titanates as Backfill

Although backfill in an HLW repository has been identified to provide several

functions, the most important role is to retard migration of radionuclides in

groundwater that has reached the waste in the event of a container breach. It is primarily this property that we have considered in our evaluation of synthetic

zeolites and titanates for backfill. In addition to providing a high potential

for required retention of radionuclides, synthetic backfill materials of regularly shaped small spheres can be used, through prior selection of the mass,

volume, shape, and packing densities, to predetermine and control the groundwater flow rates and patterns into and through the engineered underground

facility for varying aquifer pressures and flow rates. Such a modification of

the hydrodynamics of an engineered facility could be advantageous in that rapid

radionuclide transport by advection would be minimized, resulting in diffusioncontrolled transport.

Existing technology, widely used in purification systems for water reactors,

has demonstrated that synthetic materials including a wide range of zeolites

can be used to quantitatively hold up fission products and actinides. A most

prominent example of zeolite application to waste management is the use of

IONSIV-IE95 and Zeolon 900 in the Submerged Demineralized System (SDS) to be

used to decontaminate water from the Three Mile Island Unit 2 Nuclear Power

Plant [4], Another such example is the storage of krypton-85 in sodalite zeolite [5J. Recently, Nowak [6] and Winslow [7] at Sandia Laboratories and

Komarneni and Roy [8] at Pennsylvania State University have examined the sorption properties of zeolites as potential backfill material for use in a HLW

repository. Pennsylvania State University is also investigating the use of

mixtures of zeolites and clay minerals for backfill material around a waste

package and the fixation properties of sorbed radionuclides of various zeolite

minerals [9].

Sodium titanate materials have been developed at Sandia Laboratories for

quantitative removal of fission products and actinides from commercial liquid

wastes [10, 11]. Subsequent pressure sintering of the loaded material results

128
in a dense ceramic waste form resistant to leaching. Based on the high decontamination factors observed in HLW studies, the Sandia titanate materials are also being investigated for the decontamination of defense liquid wastes stored

at the Hanford site in Washington.

In this report, we have attempted to evaluate the feasibility of using highly

sorptive materials such as synthetic zeolites and titanates as backfill in a HLW repository. The effectiveness of such a barrier with respect to radionuclide retention has been assessed in terms of the NRC release criteria

discussed above.

Radionuclide Transport Processes

To evaluate the migration of radionuclides in an ion exchange backfill barrier,

we have to first consider the dominate transport mechanisms. The principal

mechanisms controlling the nature and extent of radionuclide transport in sorptive materials are ion exchange, sorption, diffusion, precipitation, advection,

and other irreversible reactions. Ion exchange and sorption processes are essentially reversible (at least over short time periods), resulting in a net

retardation in the diffusion-controlled ion migration. The extent of retardation will be primarily determined by the sorption kinetics and equilibria of

solid-fluid interactions. Non-reactive ions such as tritium, however, do not

exhibit interactions with particle surfaces. In this case, ion exchange and

sorption processes are not significant and, consequently, the migration is

primarily controlled by diffusion and advection.

Precipitation of dissolved radionuclides as authigenic minerals will decrease

the rate of radionuclide migration. Formation of new minerals, immobilizing

radionuclides release from the waste form, as a result of quaternary interactions involving waste/backfill/host rock with hydrothermal fluids has been

documented [12]. However, such data are not available for the various backfill

materials. Therefore, precipitation, as a retardation process, will not be

considered in this report.

129
Similarly, irreversible reactions such as fixation of sorbed ions with aging of

the sorbent can also decrease the rate of radionuclide migration. Recently, it was reported that synthetic zeolites, used for decontamination purposes at

Hanford, were found to fix certain radionuclides irreversibly in a period of

less than five years [13]. Although this mechanism could be potentially effective in retarding radionuclides in a bed of synthetic zeolites, the available

data are insufficient at this time for this mechanism to be considered in our transport model.

At a recent Waste Rock Interaction Technology Meeting in Seattle, a paper was presented describing the role of fine suspended particulates in transporting

radionuclides [14]. This process could serve as a potentially important mechanism for radionuclide migration in argillaceous backfill materials which commonly

contain extremely small particulates. However, in the case of synthetic materials, the particle size distribution would be very narrow, giving rise to a relatively uniform particle size. Therefore, we believe that this mechanism

will not contribute significantly to an increase in the rate of radionuclide

migration in backfill systems. However, the significance of this process as a transport mechanism for radionuclides should be investigated.

The principal transport mechanisms which we have considered for evaluating

zeolite backfill are diffusion, sorption, and advection. Since groundwater

flow rate is relatively low in a realistic repository situation, dispersion

associated with advection is considered to be negligible. An example of radionuclide transport by advection, diffusion, and a combination of advection, diffusion, and sorption processes is displayed in figure 1 [15]. As shown in

figure 1, pure diffusion corresponds to radionuclide migration as a result of

interstitial concentration gradient for non-reactive ions such as tritium. For

reactive ions such as Cs+ and Sr2+ and assuming bulk fluid flow to be negligible, the transport is primarily determined by diffusion and sorption, where the

sorption equilibrium coefficient determines the extent of ion retardation as compared with diffusional transport for non-reactive ions. For systems having

a high groundwater flow velocity, advection becomes the dominant transport process as compared with diffusion (figure 1). Dispersion effects, resulting from

130
the variation in bulk fluid flow velocities between the particles, could give

rise to a broadening of the advective front.

MODEL CALCULATIONS

We have developed one-dimensional models that describe quantitatively the

transport of radionuclides in a backfill bed. A range of values are considered

for groundwater flow rate, sorption coefficient, bed thickness, and transit

time.

Phenomenological Model 1

In model 1, the details of which are given in appendix 1, we have assumed a waste glass, having a constant leach rate over a longer period of time, to be

the source term. A constant concentration is maintained at x=0 , the interface

between the waste glass and backfill. In other words, it is assumed that the

radionuclide source is non-depleting, time invariant. The transport mechanisms

considered are diffusion, sorption, and advection. The groundwater velocity is

assumed to be constant with time and uniform throughout the bed. The dispersion

term, resulting from localized velocity variations around the particles, is

assumed to be negligible. Finally, radioactive decay of the source term and

precipitation and fixation of the radionuclide are not considered.

Nowak [6] simplified a material balance equation around a differential volume

element in the backfill bed. His formulation is directly applicable to the

present analysis. It consists of a linear partial differential equation with

three terms: (1) the rates of accumulation (or depletion) of sorbing species

in the liquid and on the solid, (2) the net transport of liquid phase species

by advection, and (3) the net transport of liquid phase species by ion diffusion. The groundwater flow in a repository must be considered and could serve

as a potentially important transport process, especially for non-reactive

radionuclides

.

Using model 1, the results were calculated for a range of conditions and the

concentration-bed thickness profiles for various transit times are displayed in

131
figure 2. In figure 2(a), for a sorption coefficient of 1,000 ml/g and a water

velocity of 1 ft/yr, the 1,000 year and 10,000 year profiles show that for

relatively short time periods diffusion plus sorption are the rate controlling

processes, with the initial concentration dropping by a factor of >10^ within

20 feet of backfill thickness. However, for longer time periods of up to

100,000 years and greater, transport associated with bulk fluid flow becomes

dominant, even for reactive ions having a sorption coefficient of 1,000 ml/g.

In the absence of bulk fluid flow, the advective term becomes zero. As shown

in figure 2(b), the transport is dominated by diffusion and sorption. These

findings are in general agreement with those reported recently by Nowak [21].

For materials having negligible sorptive capacity such as quartz, and in a zero groundwater flow stationary system, radionuclide transport is determined

by pure diffusion via interstitial water (figure 2(c)), emphasizing the importance of sorption as a retardation process in the transport of radionuclides

through a backfill barrier.

Phenomenological Model 2

In this model, we exclude the waste form as an engineered barrier which

maintained constant concentration at the interface of the glass and backfill by

continuous congruent leaching over a period of time. Instead, at zero time

(t=0), the total concentration of the radioactive waste is hypothesized to enter

into the backfill instantaneously along the x direction as a sharp step increase

(more precisely as a delta function) in its liquid phase concentration c from

zero to a constant value CQ

. We derived a mathematical expression using the

appropriate initial and boundary conditions in appendix 2. The calculated

results were obtained using the same range of parameters given in table 1.

The assumption of an instantaneous pulse at time t=0 and continuously depleting

over a period of time for the source term in model 2 means that no reliance is

placed on the waste form as an engineered barrier. Concentration-bed thickness

profiles for various transit times, based on model 2 using the parameters given

in table 1, are given in figure 3.

132
Evaluation of NRC Criteria

Although the concentration-bed thickness profiles generated using models 1 and

2 (figure 2 and 3) are useful in terms of identifying the processes that control

radionuclide migration in a sorptive backfill system and in determining their

relative magnitudes, quantitative information is required to address the

performance objectives set by NRC. For this purpose, we assumed the source

term to be a glass monolith having a volume of 55 gallons. Assuming 30 percent

waste loading in the glass, an activity of approximately 4.2 Ci/cm^ of glass

monolith is calculated. The age of the waste is assumed to be approximately

1.5 years old [21]. The average glass leach rate is considered to be 10"^

g/ cm^-yr

.

Based on assumed leach rates for the waste glass, surface areas of the glass

monolith and the total amounts of radioactivity released from the source for

the periods lCp and 106 years as estimated from the profiles shown in figure 2, we estimated the total amount released from a 3-ft-thick backfill, having a range of sorptive capacities, in a period of 1,000 years at the groundwater

flow velocity of 1 ft/yr. As shown in table 2, a 3-ft-thick bed of backfill

materials, such as zeolites having a sorption coefficient of 10^ ml/g, will con- tain practically all the inventory released from the waste for a period of up

to 1,000 years. This implies that a discrete zeolite backfill will satisfy the

NRC 1,000 year containment criterion. Table 3 shows calculated results of fractional controlled release rate from a 10- and a 100-ft-thick backfill bed over

a period 10 6 years. In these calculations, the assumptions made for the source

term and backfill are the same as those described in table 2. Assuming a leach

rate of less than 10"^ g/cm^-yr and a surface area of the glass monolith to be

3.6 x 10^ cm^ , the NRC controlled release criteria can be met using a >10-ftthick non-discrete zeolite backfill bed for periods of up to 106 years. These

calculations show the significance of a waste form having a low leach rate with

respect to meeting the controlled release criterion. In such a situation, the

sole purpose of the backfill would be to control water ingress in order to

minimize the probability of container corrosion. It must be emphasized that

bulk fluid flow, based on a groundwater velocity of 1 ft/yr, also contributes

133
to radionuclide transport. In a realistic repository situation, the flow velocities are much smaller than that assumed in our calculations. In that case,

the transport would be predominantly diffusion-controlled, resulting in a more effective containment by the backfill barrier.

Using model 2 and making the same assumptions for the source term and backfill

barrier, we obtained estimates of the total amount of activity released from a 3-ft-thick backfill barrier in a period of 10^ years. As shown in table 4, the

calculated values indicate that the 1,000 year containment criterion will be

met by a 3-ft-thick bed of backfill with sorption coefficients in the range of

10^ ml/g. Also included in table 4 are values for fractional amounts of activity

released from the backfill barrier in 1,000 years.

Using the same model and considering a time period of 10 5 years and bed

thicknesses of 3 ft and 10 ft, we obtained estimates of fractional release per

year from the backfill bed (table 5). The results show that for backfill materials having sorption coefficients in the range of 10 ^ ml/g, the annual fractional release from 3-ft-and 10-ft-thick beds are 10~5/yr and 10~6/yr, respectively. This indicates that the controlled release criterion of 10~5/yr will

be met by highly sorptive backfill materials such as synthetic zeolites and/or

titanates for long period of time.

Model Calculation for Total U.S. Waste Inventory

In order to demonstrate the effectiveness of backfill as a barrier to

radionuclide migration, we assumed the source term to be the total U.S. waste

inventory and calculated the total thickness of backfill that would be required

to provide controlled release rate over a period of 10^ years.

Using models 1 and 2, we used a value of 109 Ci [23] for the total waste

inventory in the U.S. [12] and 100- and 200-ft for bed thicknesses. The calculated estimates of radionuclide release are compiled in tables 6 and 7. Table

6, based on model 1, shows that the annual release from 200-ft-thick backfill,

having a sorption coefficient of 10^ ml/g, is on the order of 10~80 Ci/yr.

This indicates that the total containment will be achieved during this period.

134
Table 7, based on model 2, also shows that a 200-ft-thick bed of backfill

(1^=10^ ml/g) will provide total containment. A bed thickness of 100 ft,

however, is not sufficient to obtain the desired containment.

CONCLUSIONS

For groundwater velocities below 1 ft/yr, diffusion and sorption are the

dominant processes controlling radionuclide migration in highly sorptive backfill materials such as synthetic zeolites. A 3-ft-thick bed of synthetic zeolite backfill can provide total containment of activity released continuously

from a 55-gallon glass monolith for a period of up to 1,000 years. For longer

time periods of up to 10^ years, the controlled release rate of 10~^/yr can be

met by a >10-ft-thick bed of zeolite backfill, assuming the glass monolith to

be the source term. In the absence of the waste form as an engineered barrier,

a bed thickness of 3 to 10 ft is required to satisfy the controlled release

criterion for a period of up to 10^ years. With a bed thickness of 200 ft,

synthetic zeolites can be used as backfill to provide containment of total U.S.

waste inventory (10^ Ci) , if it were placed in one repository, for periods of

up to 106 years.

ACKNOWLEDGMENTS

We thank D. G. Schweitzer for initiating this study and C. Sastre for his

assistance in computer work. E. J. Nowak , A. Okubo , and G. W. Beall reviewed

the manuscript. We thank them for their constructive comments. We are also

grateful to Grace Searles, Katherine Becker, Sharon Moore, and Cathy Van Noy

for typing and preparation of the manuscript. We also would like to thank

Ulesia Gray of the Word Processing Center for final typing of the manuscript.

REFERENCES

1. R. Dayal , K. J. Swyler, and P. Soo, NRC Nuclear Waste Management Technical

Support in the Development of Nuclear Waste Form Criteria , BNL Report,

BNL-NUREG-27961 (1980), Chapters 4 and 5.

135
2. Code of Federal Regulations, 10 CFR 60, "Disposal of High-Level Radioactive

Wastes in Geologic Repositories," (Proposed Rule), Advance Notice of

Proposed Rule Making, Federal Register, May 13, 1980.

3. Code of Federal Regulations, 10 CFR 60, "Disposal of High Level Radioactive

Wastes in Geologic Repositories," (Proposed Rule), Advance Notice of

Proposed Rule Making, Federal Register, July 8, 1981.

4. E. D. Collins, J. E. Bigelow, D. 0. Campbell, L. J. King, and J. B. Knauer

,

"Evaluation of the Submerged Demineralizer System (SDS) Flowsheet for

Decontamination of High-Activity-Level Water at the Three Mile Island Unit

2 Nuclear Power Station," ORNL Report, 0RNL-TM-7448 (1980).

5. R. W. Benedict, A. B. Christensen, J. A. Del Debbio , J. H. Keller, and

D. A. Knecht, "Technical Feasibility of Krypton-85 Storage in Sodalite,"

in The Scientific Basic for Nuclear Waste Management - Vol. 2, Plenum

Press, N. Y. (1980).

6. E. J. Nowak, "The Backfill as an Engineered Barrier for Nuclear Waste

Management," Sandia Report, SAND 79-0990C (1979).

7. CD. Winslow, "The Sorption of Cesium and Strontium From Concentrated

Brines by Backfill Barrier Matreials , " Sandia Report, SAND 80-2046 (1981).

8. S. Komarneni and R. Roy, "SUPEROVERPACK: Tailor-made Mixtures of Zeolites

and Clays," Presented at Materials Research Societ Annual Meeting,

Massachusetts, 1979.

9. R. Roy and S. Komarneni, "Design of Overpack (Backfill) Materials," Paper

presented at the National Bureau of Standards Sponsored Workshop on Research and Development Needs Relating to Backfill, Gaithersburg , Maryland,

April 1981.

10. R. G. Dosch, "The Uses of Titanates in Decontamination of Defense Waste,"

Sandia Report, SAND 78-0710 (1978).

136
11. D. G. Dosch, "Application of Titanates, Niobates, and Tantalates to

Neutralized Defense Waste Decontamination-Materials Properties, Physical

Forms, and Regeneration Techniques," Draft Sandia Report, SAND 80-1212

(1980).

12. B. E. Scheetz, S. Komarneni , D. K. Smith, and C. A. F. Anderson,

"Hydrothermal Interaction of a Ceramic Waste Form With Basalt," in Alternate Nuclear Water Forms and Interactions in Geologic Media , Proceedings

of the Workshop held in Gatlinburg, Tennessee, May 1980.

13. R. E. Barletta, Personnal Communication.

14. G. G. Eichholz, "Subsurface Migration of Radioactive Waste Materials by

Particulate Transport," Abstrate of paper presented at Waste Rock

Interation Technology Meeting, 13-15 October (1980), Seattle, Washington.

15. I. Neretnieks, "Retardation of Escaping Nuclides From a Final Depository,"

KBS-TR-30, (1977).

16. R. K. Kibbe and A. L. Boch, Technical Support for GEIS: "Radioactive

Waste Isolation in Geologic Formations, Vol. 21, Groundwater Movement and

Nuclide Transport," Y 10 WI/FM-36/21 (1978).

17. L. D. Ramspott, "Waste Isolation Project: FY-78" , UCRL-50050-78

.

18. I. Neretnieks, "Transport of Oxidants and Radionuclides Through a Clay

Barrer," KBS-TR-79, (1978).

19. D. W. Breck, Zeolite Molecular Sieves , J. Wiley & Sons, N. Y. (1974),

Chapter 9.

20. R. C. Weast, Handbook of Chemistry and Physics, CRC Press, 56th Ed. (1975).

137
21. E. J. Nowak, "The Backfill Barrier as a Component in a Multiplier Barrier

Nuclear Waste Isolation System," Sandia National Laboratories, SAND

79-1109, 1980.

22. Management of Commercially Generated Radioactive Waste, DOE Report,

DOE/EIS-0046-D, Volume 2, Appendices (1979).

23. D. Majumdar and J. P. Indusi, "Inventory of HLW in the U.S.," BNL-NUREG27817 (1980).

138
Table 1

Parameters Considered in Radionuclide Transport Calculations

Backfill Bed Thickness (ft): 1, 3, 10, 100

Transit Time (years): 1,000, 10,000, 100,000, 1,000,000

Sorption Coefficient (ml/g): 1, 10, 100, 1,00, 10,000

Ion Diffusion Coefficient ( cm2/sec) [ 6] : 10~5

Groundwater Velocity ( f t/yr )[ 16 , 17 , 18 ] : 0, 1

Bed Porosity [19] : 0.34

Bed Density (g/cm3)[20]: 0.423

139
Table 2

Total Release of Radioactivity from a 3-ft-thick

Backfill in 10 3 Years (Model 1)

Backfill Sorption Coefficient Total Assumed Released from Backfill (Ci)

Kd (ml/g) SA = 3.6 x 10 4 (cm2 ) SA = 3.6 x 106(cm2

)

100

1,000

10,000

4 x 10 4

4 x 10°

5 x 10-46

4 x 106a

4 x 10 2

5 x 10~44

a Artificial number indicating total inventory release.

V = 1 ft/yr; backfill thickness = 3 f t ; leach rate (LR) = 10~3 g/cm2 -yr;

C

Q = LR x SA x transit time/area under the concentration curves;

SA = surface area of the glass monolith.

Table 3

Fractional Release Rate from a 10-ft-and a 100-ft-Thick

Backfill in 10^ Years (Model 1)

Fractional Release Rate From Backfill (/yr)

Backfill Sorption Coefficient SA = 3.6 x 10 4 (cm2 ) SA = 3.6 x 10 6 (cm2

) Kd (ml/g) 10-f

t

100-f

t

10-f

t

100-f

t

100 6.8 x 10 -5 6.7 x 10 -3 6.8 x 10"3 6.7 x 10~3

1,000 6.7 x 10-5 6.0 x 10"5 6.7 x 10"3 6.0 x 10~3

10,000 6.0 x 10~5 3.6 x 10 -9 6.0 x 10 -3 3.6 x 10"7

V =1 ft/yr; leach rate (LR) = 10-3 g/cm2 -yr;

SA = surface area of the glass monolith.

140
Table 4

Fractional and Total Release from a 3-ft-Thick Backfill

in 103 Years (Model 2)

Backfill Sorption Coefficient

Kd (ml/g) Fractional Release Total Release (Ci)

100 9.97 x 10-1 8.80 x 105

1,000 3.25 x 10"4 2.90 x 10 2

10,000 5.93 x 10"50 5.20 x 10~44

Vg

= 1 ft/yr, backfill thickness = 3 ft.

Table 5

Fractional Release Rate from a 3-ft-Thick Backfill and a 10-ft-Thick Backfill in 10 5 Years (Model 2)

Backfill Sorption Coefficient

Kd (ml/g)

100

1,000

10,000

Fractional Release Per Year (/yr)

3-ft 10-ft

>10"5 >10"5

>10"5 >10-5

9.97 x 10-6 1.85 x 10~6

Vs = 1 ft/yr.

141
Table 6

Annual Release From a 200-f t-Thick Backfill for 1()6 Years, Assuming

the Source of Term to be Total U.S. HLW Inventory (Model 1)

Backfill Sorption Coefficient Annual Release From Backfill (Ci/yr)

Kd (ml/g) SA = 2.4 x 10j ( cm2) SA = 2.4 x ip7 ( Cm2)

100 4 x 102 4 x 104

1,000 3 x 102 3 x 104

10,000 3 x 10-80 3 x iq-78

V =1 ft/yr; backfill thickness = 200 ft; leach rate (LR) = 10"3 g/cm2 yr

;

SA = surface area of all the glass monolith considered.

Table 7

Annual Release From a 200-f t-Thick Backfill for 106 Years, Assuming

the Source Term to be Total U.S. HLW Inventory (Model 2)

Backfill Sorption Coefficient Annual Release From Backfill

Kd (ml/g) (Ci/yr)

100 103

1,000 103

10,000 10-79

Ve = 1 ft/yr.

142
DISTANCE, TIME

Figure 1. Principal radionuclide transport processes in sorptive backfill

material [after Neretnieks [15]]

143
Figure 2. Concentration vs. bed thickness profiles for various transit times

based on model 1: (a) Kd = 1000 ml/g, V

g

= 1 ft/yr; (b) kd = 1000

ml/g, V

g

= 0 ft/yr; (c) k

d = 1 ml/g, V

g

= 0 ft/yr; A = 10 3 yr,

B = 104 yr, C = 105 yr, D = 106 yr

144
Figure 3. Concentration vs. bed thickness profiles for various transit times

based on model 2: (a) Kd = 1000 ml/g, Vg

= 0 ft/yr; (b) kd = 100

ml/yr, V = 1 ft/yr; A = 10 3 yr , B = 10 4 yr , C = 10 5

, D = 10 6 yr

145
APPENDIX 1

The simplified equation of the differential mass balance is

6c(x,t) 6c(x,t) ?f 6 2 c(x,t) =

'* 5t Rf 5x /J Rf 6x2

where

R

f = [1 + _S—<LJ, the retardation factor

e

x = the distance from the interface of the glass and the zeolite

along the direction of flow and longitudinal diffusion

t = time

c(x,t) = liquid phase concentration, quantity of sorbing species per unit

volume of liquid at distance along flow direction x and time t

e = effective porosity of bed (fraction of bed volume containing

flowing liquid)

Vg = average interstitial velocity of flowing water

Df//o - coefficient of longitudinal fluid diffusion with a tortuosity

factor to account for the tortuous diffusion path through the

porous bed

Pb = the bulk packing density of solid sorbent, mass of solid per

unit bed volume

Kd = the distribution coefficient for a linear sorption isotherm. The

ratio of quantity of sorbed species per unit mass of solids to quantity of mobile species in the liquid phase per unit volume of liquid,

146
The boundary condition is

c(x = 0, t > 0) = C 0

,

The initial condition is

c(x > 0, t = 0) = 0,

Define

U(x,t) = c(x,t) • exp

-V V

2 • t

• exp

/l Df 2/2 • Df

•

Then, Equation (1) is reduced to a heat equation of U(x,t) as function of x and

t. With the rearrangement of boundary and initial conditions, the Fourier transform is applied to the heat equation, and the subsequent convolution gives the

following solution

c(x,t)/C = 1 -

\ erf c V * ft

_g

^R

f

. x

2 3/4 /

D

f

• Rf

2 3/4 /D

f

• t J

- exp < ^2 V • x

S erf c V • ft

S , 3/4 j— . + 3/4

R

f

• x

D

f R

f

2 D

f

• t j

(2)

where

erfc(y) = 1 - erf(y) and erf(y) is the error function of y.

The integration from x = backfill thickness to infinity is done numerically for

the total release from backfill.

147
APPENDIX 2

The same differential equation is used as in model 1. The new constraint

instead of the boundary condition is

oo

/ C(x>0, t>0)dx = C total

o

and the new initial condition is

C(x > 0, t = 0) = C total 6(x)

where 6(x) is a delta function of x and C tota i is a total quantity of sorbing

species.

The transformation technique may not be plausible in this case for the closed

form solution. However, since we know the response of step function from

appendix 1 , the response of the delta function should be the derivative of the

response of the step function with respect to x: the derivative of equation

(2) of appendix 1. Further, we need the integral from x equals the backfill

thickness of infinity, which is equivalent to the response function of

appendix 1 (equation 2).

148
PRESENTATION OF SURVEYED RESULTS OF WORKSHOP QUESTIONNAIRE

A questionnaire containing ten questions covering a wide range of problems

relating to the use of backfill was handed out to the participants during the

workshop for them to fill out and return to NBS. A copy of the questionnaire

is included in Appendix A of this report. Out of the twenty-eight non-NBS participants, eleven responded the request. It should be noted that the Canadian

representatives only felt able to respond to the questions dealing with experimental techniques and equipment. They felt that all other questions are premature based on the current state of research and development. This point of

view was echoed by a few other researchers at the national laboratories, as will be seen later in the report.

The surveyed results of all questions are given in this section. Names of the

respondents are not spelled out in the presentation, but are identified by

letter designation. Before the first question was asked, it was thought appropriate to ask the respondents to indicate their areas of expertise and the percentage of time which they have devoted to the nuclear waste problem. Table 1 summarizes their responses to this question. For those answered (7 out 11), the

percent of time worked on this problem varies from 30 to 100 with 5 respondents

working on full time basis. The areas of their expertises reflect a wide range

of disciplines including chemistry, geology, civil engineering, seismicity, and

metallurgy.

Question 1 asked "What do we expect from the placed backfill?" and proceeded to

list seven functions that the organizing committee of the workshop believe the

backfill should be able to perform. Four of the respondents did not answer

this question. Three respondents completely accepted what was presented and

made no comments to either modify the statements or add additional functional

requirements for the emplaced backfill. The answers to Question 1 can be

summarized as follows:

• It should serve as a hydrologic barrier to groundwater intrusion to the

canister-waste form. It should also provide strong retardation of

groundwater flow after intrusion.

149
• Its swelling and plasticity characteristics should enable it to seal

cracks within the backfill and localized cracks in the surrounding rock

to minimize the potential of groundwater intrusion.

» It should serve to buffer mechanical stresses and deformations.

• It should serve as a chemical conditioner of intruding groundwater by

modifying pH, Eh, and chemical composition to minimize the corrosion

potential of canisters.

9 It should sorb and fix radionuclides to the extent possible.

» It should also be tailored specifically to retard or fix anionic

chemical species.

• The backfill should be expected to be stable in the repository

environment, i.e., it should be compatible with host-rock-imposed geochemical environment, with other barriers and waste form, and should

have long term stability in expected radiation field.

• Finally, the backfill should be expected to perform all the necessary

functions across a broad range of pressure, temperature and saturation

conditions.

Question 2 requested the participants to list the criteria for the material

properties they consider important for selection of the backfill and then

distribute 100 points among these criteria based on their relative importance.

In Question 3, eight (8) pertinent material properties were listed for the

participants to distribute 100 points based on the relative importance.

Furthermore, both questions asked the respondents to choose not more than four

(4) material properties for preliminary screening purposes.

Nine out of eleven respondents answered Questions 2 and 3 and the results are presented in table 2. The numbers listed under the respondent are the points

out of 100 which he/she has assigned to each material property and the numbers

150
in the brackets are the priority of the material properties which the respondent

chose for the preliminary screening process. In addition, a scoring system

was devised as shown at the bottom of table 2 to tally the total points in the

brackets scored on each material property. Consequently, the material properties can be ranked as shown on the last column in table 2. Thus, for preliminary screening purposes, the researchers would choose the backfill properties in the following order: sorption, hydraulic conductivity, thermal

conductivity, chemical stability, swelling/shrinkage characteristics and plasticity, bearing capacity/shear strength, and diffusivity. Diffusivity was not on the original list but was brought up by Respondent B. The respondents also

provided criticism of the method used in selecting material properties. One

comment was that chemical stability does not belong to a list as given in Question 3 because if the material does not survive in a repository environment, all

the other properties will then be meaningless. The issue of the interdependent

nature of some of these material properties was also raised, e.g., the hydraulic

conductivity of clay materials is closely related to swelling and plasticity

and obviously the bearing capacity is related to shear strength of the material.

Question 4 was asked in two parts. The first part addressed the advantages

and disadvantages of manufactured backfill with artificial additives versus various mixtures of natural materials. The second part asked the participants

to indicate whether the location and type of parent material selected for the

repository affect the answer to the first part of Question 4 and to the backfill evaluation in general. Nine out of eleven respondents responded to this

question and their opinions are summarized in table 3. In short, artificial

backfill has the advantages of ease in achieving better quality control and

being available to be readily tailored for specific needs. However, it is disfavored for its comparably expensive nature, and most critically, for its lack

of proof in providing long term stability in a repository environment for the

required time period.

From the information revealed in table 3, it can be concluded that the natural

materials are favored as the backfill mainly because they offer a much better

chance to achieve long term chemical and thermodynamic stability, for the reasons that these materials have already been seriously tested under a geologic

151
time frame. The answer of part II question to part I is very much divided.

First of all, only 5 respondents responded to this part of the question.

Among the five, two indicated affirmative and three negative, and none offered

reasons to back up their responses. As to the part II question' for the backfill

evaluation in general, the answer is that the choice of backfill has to be

repository dependent, particularly with respect to the overall chemical stability. One respondent indicated that this part of Question 4 was too broad

and that the answer could go either way.

Question 5 was also asked in two parts. In the first part (Question 5(a)), the

participants were asked to construct the ideal backfill using material such

as montmorillionite or bentonite, zeolite, sand, crushed host rock materials,

additives for sorption of selected radionuclides, and others that the participants may add to the list. Question 5(b) asked whether they think the design

concept should change if factors such as availability, quality control, and

cost are considered. Table 4 showed the constituents of the backfill envisioned by the eight respondents who answered this part of the question. Two of

the respondents believe that it is too early to make such a selection. For

those who made the selection, the consensus is that the backfill has to be sitespecific and also to be dependent on design loading conditions. A question on possible thermal instability of zeolites during the thermal period of the

repository was brought up. To summarize, it is clear that certain amount of

clay materials will be included. Depending upon where the backfill will be

placed, the amount of clays may vary from 5 to 100 percent. For example, the

highly compacted pure bentonites have been considered for placement around the

canister as a buffer, whereas backfills to be used to fill in the tunnels may

only need 5 to 10 percent of the fine-grain materials. Because of the inexpensive nature of crushed host rocks and their many inherited favorable properties

that they should be blended together with other ingredients to form the backfill.

Only 5 out of the eight respondents went further to answer Question 5(b) and

their responses are given in table 5. Apparently, at the present thinking

the major components of the desired backfill, i.e., clay, sand, and crushed hot

rocks, are abundant in their natural existence and are relatively uniform in

their composition, thus will not impose serious problems on availability, cost,

152
and quality control. Of course, in the final design stage all these factors

will have to be considered which means that some minor modifications of the

backfill composition would be anticipated.

As commented by one respondent, Question 6 may well outline one of the most

difficult problems needing to be resolved in order to use the backfill as an effective barrier. The question requested the participants to describe in

their vision the placement procedures and anticipated difficulties in the

installation of the backfill. One respondent disqualified himself /herself to assess the problem and another believed that the problem has to be site-specific

without elaborating his/her comments. For those four who answered the question,

their answers can be summarized separately as follows

o Respondent B: - There is no identified major problem of backfilling in

relatively dry rock.

- If inlet of water is high, there may be some difficulties

of containing swelling bentonite in the deposition hole

before backfilling the tunnel.

- Sealing water-bearing rock joints and drainage facilities

may be necessary.

o Respondent E: - Possible procedures are briefly described below but not in

the order of probability:

• Pressed billets (donuts or hemi-cylinders ) are inserted

around the canisters or positioned in the hole prior to

insertion of the canisters, or

• Backfill is poured in under dry condition and be tamped

around the canister, or

• Backfill is poured in around canister in a slurry state,

or

153
• Backfill will be dry poured into borehole, tamped, and

then core-drilled to accept canister.

- All of the above four approaches have serious limitations,

most of which affect the bulk properties of the backfill,

especially homogeneity. Selection and evaluation of an emplacement procedure is one of the most important actions

that could be taken at the present time.

o Respondent G: - Preformed blocks of compressed bentonite mixer, probably

fabricated at the repository site, will be placed in the

emplacement hole.

- The canister/overpack assembly will be emplaced in the

backfill line hole.

- Gaps resulting from necessary tolerances will be filled

with powdered bentonite.

- A section of preformed bentonite mix will be placed over

the top of the canister/overpack.

- A ram, attached to the bottom loading transporter, will be

braced against the tunnel roof and will compress the backfill into one solid piece surrounding the canister/overpack.

- A temporary radiation shield, possibly a block of magnetite

concrete, will be placed over the top of the emplacement

hole.

o Respondent I: - Backfill must be pre-installed due to radiation protection

and quality control requirements.

154
- The backfill also can not be pre-installed too long in

advance in order to assure that size (opening for canister/

overpack) is maintained, unless

• If "overpack" becomes a sleeve around which backfill is

packed, there could be a longer lag between backfill and

waste emplacement.

• Drilling 48" diameter holes of 20 ft deep that are both

plumb and straight is difficult and expensive.

• Dry, highly compacted backfill as in the Swedish concept

will have to be kept away from water - a non-trivial task

in an operating mine, even a "dry" one.

• The simplest field placement would involve backfill outside

a sleeve. The backfill would be rodded in place.

Question 7 requested the participants to indicate for a number of test variables

the desired range/type as well as those actually used for tests already completed

in their projects. These variables included temperature; hydraulic gradient

and liquid/ solid ratio of groundwater/leachant ; the pH, Eh, species and concen- tration of ionic composition; corrosion products; and radiolysis products of the

solution composition; mechanical stresses such as compression, swelling, shear,

cyclic wetting/drying; particle size distribution and density of the backfill;

and radiation dose rate and total dose. Answers from the respondents were summarized with respect to each variable as presented in tables 6 through 11.

Part (a) of Question 8 asked the participants about the test equipment used in

evaluating the backfill properties including swelling, permeability, thermal

conductivity, strength, sorption, and chemical analysis. Part (b) of the same question had the participants to describe the test methods and the availability

of test procedures for evaluation of the above mentioned backfill properties.

Tables 12 through 22 summarize the responses from those who answered the

question.

155
Some general observations can be made from the answers to Questions 7 and 8:

(a) there exists a wide range as well as many types of test variables which the

respondents considered important considerations in their programs. This is

another example of why the backfill probably has to be site-specific to best

suit the waste loading as well as the hydrogeologic condition of the repository

sites. It was also recommended by the researchers that at this stage of the

work, ASTM Standards should be used wherever they are applicable. Some test

methods, such as those used for hydraulic conductivity and swelling determinations, will require consensus among the researchers as to how they can be

adopted for the nuclear waste standard work. Only a few organizations have

started the material testing under high pressure and elevated temperature

conditions. The test equipment is unique in each case and the method of

testing will have to be developed during the execution of the preliminary test

program.

Question 9 requested the participants to briefly discuss the change in backfill

properties as a function of time and the participant's vision of the sequence of

events happening to the backfill after its placement. They were also asked to

identify major problem areas associated with the development of the backfill.

Descriptions from the five respondents were quite similar and the focus was to

address the mechanism of a hydrothermal interaction region within the emplaced

borehole. Detailed description of the nature of the problem will be delayed to

the next section where further research and development needs will be discussed.

Finally, table 23 summarizes the reference sources where information with regard

to the characteristics of backfill materials are located. It also gives the

most commonly used references in their research work. It can be seen that for

each researcher who responded has preferences and their preferences are very

much dependent upon his/her training, background, and familiarity with the

reference sources and their accessibility.

156
Table 1

Area of Expertise and Percent of Devoted Time of the Respondents

Respondent Area of Expertise Percent of Time

A o Backfill 30

o Primary Waste Form

o Radionuclide Migration

B o Performance of Backfill

C o Development of Test Procedures 85

o Corrosion of Engineered Barriers

D o Geochemistry 100

100

o Geology and Geochemistry 100

o Geochemistry 100

o Rock Mechanics

o Site Exploration

o Seismology

K o Geotechnical Engineering 50

o Heat Transfer

157
158
o o o o o

159
o

0) H

eg H

01

4_>

cu t3

f3O

to, CO

0)

cu

4->

>,

42

-C

CU

(3o>

Ed

14-1

CJ

ce

CO

cu T3V

43

CO I I oo

inI o o

CO

43 i o 3o 3I o

4Jo

cj CO

43

O

CM o M

cu a>a

4-1 4-1 o

•H i-4 C! C

rHO o 4J

CtJ 4J rH CO •H c rH o M a;

a; 43 o a) T3 l e u CU

4-> •H 42 c rH CO eo 0 o 3

rH e cu CO- t-i IK) CO U

in

03

CI) CD rH

4J cfl o •H 0 u

CVJ iJ 1h CD CO r4 c

a. CU •H

Cfl 14-4 e

<+-! O

exMO

03

(JO

mh

cna)>

jj •HX)

-a

CO

CO -aodo

•H

CO

J-i

-d

a)

.U) CJCU rH

CU

CO

03M

CU 43

•uO

COco

*T-| 1 1

4-1 o >-> cu rH t i 1—1 rH CO

QJ TO CO CO •rHo CO rS

Ml CU C cu 4-1 iH •H

ki J£ #-* t- c*>— CD o O 0) g a 4_> P

c3 a)

M o cu y-t O cu

4-1 OS

<D >. Tj rH i

j CU -O

!-J 4-) CO

cC CO rH o

CU cuW) cO aj o M) o 3 fHcu

CO

cu ' —/

ri—1 lOi •i-i t | i—5

CD cO Cu •iH

i-i rH CU CO 4—> ca cOP

CU 43

4-) 1—1 3

CU 43

4J 0)>

,3

M) Cfl C3 O CUCuX

JJ cu

U3 CU 4-1

4-) rH CO

cO Xj cO

l-j CO cu cu 4-1 rH

a, co e 0)

cu 4-1 4J o 4J

(3 c >, o >. rH

rH c

T3 43 •H c cO cO

cu 43 a O S-J 0) CU T3 (3. CJ

CO 43 o

03

CU

4-> o

160
Table 5

Will the Design Concept Change if The Following

Factors are Considered?

spondent Availability Cost Quality Control

A • No No No

E Possibly No No

F Current materials being considered are available in sufficient quantity and not

prohibitively expensive

G The final design will, if necessary,

consider all of these factors

t

1 No No No

161
Table 6

Test Variable - Temperature

Respondent Desired Actually Used

A 25° - 200°C

B - 20° - 90°C

C 60° - 250°C

D

E

F 65° - 300°C 65 - 300°C

G 25° - 350°C 25°

H best obtained from reference repository

conditions published by OWWI

I Not adequately defined at present in -

the U.S. program

J 25° - 150° 25° - 100°

K 100° - 15Q°C 100°C (in progress)

162
Table 7a

Respondent

A

B

C

D

E

FG

H

IJ

Test Variable - Groundwater/Leachant (Site-Specific)

Desired Actually Used

Granitic, Tuff, Basaltic groundwater,

and Brine

'Allard" synthetic groundwater

Grando Ronde (proposed site groundwater) Same as left

1. Synthetic granitic groundwater

2. Basaltic groundwater

3. WN-1 Saline solution

4. Standard Canadian sheild

saline solution

Granitic water tap H2O, deionized H2O

163
Table 7b

Test Variable, Groundwater /Leachant , Hydraulic Gradient

Respondent Desired Actually Used

A

B

C

D

E

FG

H

I

JK

Very low

10"2 to 10 6

Not defined yet

50 - 500

None when in diffusivity measurement

10 4 to 10 6

50 - 100

164
Table 7c

Test Variable, Groundwater /Leachant , Liquid/Solid Ratio

Respondent Desired Actually Used

A

-

B - Realistic for density of highly compacted

bentonite = 2.1 t/m-^ and mixture with 10

to 30 percent bentonite

C Low under expected -

condition

D

E

F

G

H

I -

J Not defined yet

K

165
Table 8a

Test Variable, Solution Composition (site specific)

Respondent Desired Actually Used

A

-

B

C

° For certain repositories, specific solutions

are available, e.g., Hanford GW, Brine A, Brine

B, Tuffs related GW, and Granite GW

° Must test ambient oxygen pressure to very

low oxygen to simulate open repository and

conditons often sealing

D

E

FG

11

IJ

K

166
Table 8b

Test Variable, Solution Composition, pH

Respondent Desired Actually Used

A 4-10

B - 8-9

C

D

E

F 6-10 6-10

G 4-10 9.8

H

I -

J 5.5 - 8.5 5.5 - 8.5

K

-

167
Table 8c

Test Variable, Solution Composition, Eh

Respondent Desired Actually Used

A

BC

DE

F

G

II

1

JK

Aerated to Fe+2/ Fe+3

-0.5 - + 0.5

-0.2 - atmospheric

Fe304 - Fe203 Buffered

-0.5 - +0.5

atmospheric

Air saturated H2O

168
H

IJ

K

Table 8d

Test Variable, Solution Composition, Ionic Composition

Respondent

Species Concentration

Desired Actually Used Desired Actually Used

Ca+2

, Mg+2

, Na+, K+, co

3=

, so 4=

Site-specific ambient

groundwater for

basalt and tuff plus

solutions resulting

from rock groundwater

reaction at elevated

temperature

Ambient groundwater

from Hartford Basalt

(Grande Ronde)

169
Table 8e

Test Variable, Solution Composition,

Corrosion Products

spondent Desired Actually Used

A

B

C

D -

E

F

-

G To be determined

II -

IJ Work, is undefined

K

170
Table 8r

Test Variable, Solution Composition,

Radiolysis Product

spondent Desired Actually Used

A

B - Hydrogen

C This can be answered only by

radiation tests

D

E

F

G To be determined

H

IJ Work is undefined

K

171
Table 9a

Test Variable, Mechanical Stresses, Compression

(Lithostatic) Pressure

Respondent Desired Actually Used

A

B

C

D

E

F 300 bars (30 MPa)

G 0-30 MPa

H

I

J ^25 MPa 100 kPa - 20 MPa

K up to 15 Mpa None

172
Table 9b

Test Variable, Mechanical Stresses, Swelling

Respondent Desired Actually Used

A

B - Confined w/access to H2O

C -

D

E

F

G 0 - > 10 MPa

H

I

J Undefined

K

173
Table 9c

Respondent

A

BC

D

F

G

H

IJ

K

Test Variable, Mechanical Stresses, Shear & Cyclic

Wet ting/ Drying

Shear Cyclic Wetting/Drying

Desired Actually Used Desired Actually Used

To be determined on site specific basis

Undefined

174
Table 10a

Test Variable, Material Physical Properties, Particle

Size Distribution

Respondent Desired Actually Used

A

-

3

C

D

-

E

-

F

-

G To be determined

H

I

J

K Clays to sands Mixtures of crushed rock

and clay

175
Table 10b

Test Variable, Material Physical Properties, Density

Respondent Desired Actually Used

A

B - 2.1 t/m 3 (highly compacted

bentonite)

C

D

E

F

G To be determined

H.

I -

J

K. 1.6 to 2.2 g/cm3

176
Table 11

Test Variable, Radiation Dose Rate for Total Dose

Dose Rate Total Dose

Respondent Desired Actually Used Desired Actually Used

A

B - 450 krad/hr - 7 Mrad

C ~ 10 7 R/hr ( no radiation tests done on backfill)

D

E

F

-

G 4 x 10^ R/hr - cumulative ~

1010 R

H - - - -

I - - - 10 9 R

J Undefined Undefined

K - - - -

177
Table 12

Test Equipment Used to Evaluate Backfill Swelling

Respondent Equipment

A

B Permeameter described by Pusch

C Dilatometry

D

E

F Permeameter

G Load cell attached to the PNL permeability

cell

H

I

J Applicable ASTM standards

K o One dimensional consolidometer

o Swelling chamber (Yong & Warkentin, 1975)

o Automatic swelling pressure apparatus

(Holland, 1968)

178
Table 13

Test Equipment Used to Evaluate Backfill Permeability

Respondent

A

B Perraeameter described by Pusch

C o Water migration - rigid volume flow cell

o Radionuclide - modified version of above

migration

D

E

F Perraeameter

G PNL permeability cell

H

IJ Applicable ASTM standards

K o Constant head permeameter

o Pressure decay permeameter

179
Table 14

Test Equipment Used to Evaluate Backfill Thermal Conductivity

Respondent Equipment

A

B "Transient hot strip" method by Prof. Lunden

C Cut - Bar thermal conductivity apparatus

D

E

F Thermal probe

G Under study to determine best equipment and

techniques for application to backfill

materials

H

I

J o Transient heat probe

o Guarded hot plate (steady state method)

K o Transient hot probe

o Guarded hot plate (steady state method)

180
Table 15

Test Equipment Used to Evaluate Backfill Strength

Respondent Equipment

A

B Described by Pusch

C Mechanical compression test

D

E

F Triaxial test

G Standard uniaxial and triaxial test equipment

H

I

JK Triaxial compression apparatus

181
Table 16

Test Equipment Used to Evaluate Backfill Sorption

Respondent Equipment

A

B

C

D

E

F Teflon tubes, autoclaves

G Column (retardation) techniques are being

developed for use in PNL permeability cell

H

IJ

K

182
Table 17

Test Equipment Used to Perform Backfill Chemical Analysis

Respondent Equipment

A

B

C

D

F XRD, STEM, SEM, ICP , Chromatograph , AA

G ICP, AA, XRF

H

J

K

183
Table 18

Test Methods and Procedures for Backfill Swelling Evaluation

Respondent Test Method Test Procedures

A

B

C

D

EF

G

H

I

JK

Described by Prof. Pusch

o Measured by mechanical testing

o Creep can also be measured by

standard method

Use of a load cell attached to

the PNL permearaeter

Applicable ASTM standards

o Stress-controlled method

o No volume change method

Described by Prof. Pusch

Under development

Applicable ASTM Standards

o To be developed

o Available

184
Table 19

Respondent

A

B

C

D

E

F

G

H

IJ

K

Test Methods and Procedures for Backfill Hydraulic

Conductivity Evaluation

Test Method Test Procedures

Described Dy Prof. Pusch Described by Prof. Pusch

Hydraulic conductivity is a "stan- Variety of techniques is

dard" method of assessing; extra- available in open

polation to low pressure is literature

overly conservative

PNL permeability cell Under development

Applicable ASTM standards

Steady-state flow method

Flow velocities using tracers

Applicable ASTM standards

o Available

o Available

185
Table 20

Test Methods and Procedures for Backfill Thermal

Conductivity Evaluation

Respondent Test Method Test Procedure

A - -

B "Transient hot strip" method by Prof. Lunden See left

C Relatively straight forward compared to

other properties (importance diminishes with

emplacement time, i.e., long term assessment

is not that critical)

D - -

E - "

F

G Most test method are relatively standard

H

I

J - "

K Transient line source method Available

186
Table 21

Test Methods and Procedures for Backfill

Strength Evaluation

Respondent Test Method Test Procedure

A

B Method described by Prof. Pusch See left

C Mechanical testing -

D

E

F

G Most test methods are relatively standard

H

I

J Applicable ASTM standards Applicable ASTM

standards

K Triaxial testing Available

187
Table 22

Test Methods and Procedures for Backfill

Sorption Evaluation

Respondent Test Method Test Procedures

A

B

C

D

EF

G

H

IJ

K

measurement; its primary difficulty

is the specification of proper, relevant boundary conditions, e.g., Eh, pH,

Tests will have to be conducted over a range of both variables.

Column (retardation) technique Under development

188
Table 23

Reference Sources for Backfill Properties

Respondent

Reference Sources Regarding

Backfill Characteristics

References Used

Most Often

A

B

C

D

E

FG

H

IJ

K

No comprehensive data compilation

is available

Proc. of Material Research

Society Annual Meetings

American Chemical Society

Meetings

Nuclear and Chemical Waste

Management

o Swedish Documents

o KBS reports

o AECL documents

o ASTM Standards

o ASTM Geotechnical Testing

Journal

o Technical Reports from

ANSL, SNL, ORNL , LLL , SRL

,

and PNL

o Swedish Reports

o Proc. of American Society

of Soil Sciences

o International Journal of

Heat and Mass Transfer

o Soil Sciences

o Transactions, Geophysical

Union

o ASCE Geotechnical Engineering

Division

189
SUMMARY OF FUTURE RESEARCH AND DEVELOPMENT NEEDS

Areas in which research and development of the backfill component are required

were discussed during the informal discussion sessions on the second day of the

workshop. The content of the discussion can be divided into two categories,

the test methods used in the laboratory for evaluation of the backfill properties including hydraulic conductivity, swelling pressure, diffusivity, and

thermal conductivity; and the study of backfill properties in a hydrothermal

region in a repository environment. Measurement of sorption of radionuclides

and radiation effects were dropped from the discussion since much research has

been focused on the former and for the lack of expertise on the latter subject

among those participating in the discussion.

Table 1 summarizes the current status of laboratory testing to determine

swelling and hydraulic conductivity, as reported by the workshop participants.

In addition, the method of specimen preparation used in hydrothermal and

wetting-drying studies by one institution was included in the tabulation. As

it can be seen, the tests were conducted under a variety of conditions that

make comparisons of the test results difficult. Differences in testing conditions include the type of backfill, the amount of clay used in the specimen,

specimen size, initial density, moisture content at the time of compaction,

the type of water used, and the method of specimen compaction.

Various types of bentonites, silica or quartz sand (or fine aggregates), and

crushed host rocks were used as the candidate backfill. Depending on where the

backfill will be placed (either adjacent to the canister in the borehole called

buffer in many studies, or for backfilling the tunnels), the amount of clay

soils used was varied from 10 to 100 percent. The size of specimens used in

testing was closely related to the method of compaction and varied from 25 mm

to 102 mm in diameter and 5 mm to 115 mm in height. The height of the specimen

is also dictated by the time required to saturate the specimen prior to hydraulic conductivity measurements. For highly compacted samples, very long time

periods are required to reach the saturation. This is the reason why the

specimen height used in highly compacted materials is very small, 5 to 10 mm. Even with this short dimension, the researchers have reported that the specimen

190
would still need up to 3 weeks to reach complete saturation. Great variation

is also shown in the initial density of the specimens used in the testing,

which was reported to vary from 1.1 to 2.2 g/cm^.

In some cases, air-dried materials with moisture content as low as 6 to 10

percent were used in preparing the specimens and in other cases the moisture

content was as high as 30 percent. The researchers addressed the difficulties

encountered in compacting specimens having moisture content beyond 25 percent.

In addition, high initial moisture content can result in a nonhomogeneous

moisture distribution within the specimen. The type of water used to wet the

soil varies considerably and may be site specific. Tap water, distilled water,

deionized water; brine, and granitic, basaltic, tuff, or synthetic groundwater

were used.

Four different compaction techniques can be used for specimen preparation.

They are impact, vibratory, kneading, and static compaction. Other than the

vibratory method, the other three methods were listed in table 1. The ASTM

D1557 method, which is the modified Proctor, is an impact method in which a 4.54 kg rammer is dropped from a height of 457 mm for a specified number of

drops for each layer placed in a compaction mold. The Harvard miniature com- paction is a kneading method. At present, the uniaxial static compaction method

is probably the most popular technique used for backfill specimen preparation.

It can be conducted by using any press machine.

Another variable which was not included in the table is the temperature.

Conflicting results were obtained on the effect of temperature on the properties

of swelling and hydraulic conductivity. Lambe in 1960 [ 1 ] U found that the

samples under consolidation in consolidation tests swelled and compressed with

decreases and increases in temperature, respectively. However, a study by

Yong et al. , [2] of the swelling pressure of Na-montmorillonite at below room temperature resulted in the reverse effect. Further studies by Yong [3] on pure clays at and above room temperatures (25 to 45°C) indicated an increase

2/ References are listed at the end of this section.

191
in swelling pressure when the temperature was increased and thus was consistent

with his earlier findings [2], Recent studies on bentonite by Pusch [4] con- cluded that at 90°C the swelling pressure is lower than that at room temperature. Although he indicates that the number of tests was too small to give any- definite relationship between swelling pressure and temperature, Pusch thinks

that a 50 percent reduction in swelling pressure should be expected when the

temperature is raised from 20°C to 90°C. Pusch [5] also concluded from his test

results that at a 70°C temperature, the hydraulic conductivity is estimated to

be 5 to 10 times higher than that at 20°C for both the Na and Ca bentonites

used in this studies.

j Although it is impossible to unify the methods that should be used for the

determination of the swelling and hydraulic conductivity of the backfill, the

test variables should be limited in order to make the comparison of test results

meaningful. It was the consensus among the participants that for simple routine

purposes, standard test methods, such as those specified in the ASTM standards,

should be followed. It should be noted that no ASTM standard is available for

the evaluation of swelling potential of the material; however, methods such

as free swelling testing and swellconsolidation tests have been routinely used

in the engineering analysis. A good summary of methods used by the engineering

profession can be found in a paper presented by Heinz in 1981 [6].

Due to the intimate relationship between hydraulic conductivity and swelling,

it would be desirable to evaluate both properties in one test set up. Some

studies have been conducted by coupling these two measurements, such as the

work by Pusch [4,5] and Wheelwright et al. [7] presented in this workshop.

Testing of these properties at the temperature and pressure comparable to the

condition in a real repository should also be conducted. Obviously no standard

can be referred to for conducting these special types of testing, therefore,

a close working relationship among the researchers should be sought in order to

develop meaningful and reliable measurement data for the design of the long

term stability of the backfill in a repository environment.

192
One compromise which has to be made in laboratory determination of hydraulic

conductivity is the application of rather high hydraulic gradient across the

specimen during testing to reduce the time required for the measurement.

Nevertheless, the hydraulic conductivity determined using fictitiously high

gradient should result in a conservative measure of the estimated breakthrough

time of the multibarrier isolation system.

For backfill constituted of compacted clay materials such as bentonite, the

hydraulic conductivity of the barrier will be very low and will probably limit

the flow rate to a negligible amount. In this case, the minimum significant

flow rate can be reasonably determined by the estimated rate of molecular

diffusion in the aqueous phase. In other words, the molecular diffusion of

radionuclides rather than convective transport will be the predominant mechanism

for migration if the flow rate through the barrier is small enough. Therefore,

the mechanisms of diffusion should be thoroughly studied. A one-dimensional

linear mathematical model has been developed by Nowak [8] and Pusch [9], however,

only bulk diffusion was considered in the simulation. Surface diffusion, which

is another contributing mechanism, is identified as one area where further

research is required [10].

The backfill surrounding the canister will have to act as a heat transfer medium

since the decay of the radionuclides in the waste form generates heat. A backfill with sufficient thermal conductivity would help to distribute the heat

faster and more uniformly into the host rock mass, thus would prevent excessive

heat buildup in the canister. Designing a backfill with sufficient thermal con- ductivity to prevent excessive heat buildup under a given waste loading condition

is a very intriguing problem since the backfill is also expected to serve as a medium to slow down the flow of groundwater toward the canister.

Several parameters have been cited in the literature to affect the thermal

conductivity of a soil (backfill). They include soil composition and the

geometrical arrangement of soil particles, density, moisture content, and

temperature gradient between the backfill and the surface of the canister. Up

to this point, either air was assumed to fill the annulus between the canister

and the borehole wall [11], or a low but fixed value of thermal conductivity [12]

193
has been recommended for the heat transfer calculation. Although this approach

is conservative, a better simulation should be pursued considering the parameters listed above which have direct effect on the magnitude of thermal con- ductivity. Obviously, the backfill placed in a repository environment will go

through thermal properties changes with time since its composition may change

due to chemical alternation; the geometrical arrangement of soil particles and

density may change due to the application of lithostatic pressure and swelling

when it has the access to groundwater; its moisture content will change from

being dried out initially to eventually being saturated; and the decreasing

thermal gradient across the backfill throughout the decay of the radionuclides.

At present, two methods are used in the laboratory to determine the thermal

conductivity of a soil: steady-state heat flow measurement and transient heat

dissipation measurement. The steady-state method has several disadvantages.

Firstly, it introduces measurement due to migration of moisture in the direction

of thermal gradients in a moist soil. It also creates convection currents in a saturated medium. Finally, the method is time consuming. The transient method

is generally free of the disadvantages mentioned above. Furthermore, the measurement method is relatively simple and fast. For these reasons, the transient method is favored for the measurement of the thermal conductivity of the

backfill.

It was also the consensus among the participants that both mathematical modeling

and experimental work should be carried out to study the backfill properties in

a hydrothermal environments generated by the multi-barrier isolation system.

The hydrothermal region of the backfill is graphically depicted in figure 1.

The region is confined by the metal surface of the canister on one side and borehole wall on the other. The curve shown to the left of the borehole wall is

the average hydraulic gradient within the backfill as a function of time. During the operational period of the repository when the repository is still open,

the water table around the repository is lowered by pumping. It is only after

the closure of the repository that the hydraulic gradient will be gradually

restored to the boundary at the borehole wall because of the restoration of

the groundwater table. The curve on the right of the surface of the canister

depicts the decay of thermal loading with time. The two dashed curves across

194
the backfill indicate the cross-coupling of the thermal gradient and hydraulic

gradient. It is conceivable that the backfill will dry out immediately after

the emplacement of the wastes. This is considered to be the most critical

period for the backfill with regard to its long term capability to function as intended. Remember that no groundwater will be present during the operational

period of the repository. Thus the degree of drying out of the backfill would

be completely dependent upon the design of waste loading in the canister, and

the type of soil used as the backfill, its density, and the moisture content

used in placing the backfill.

After the closure of the repository, and dependent upon both site geology and

hydrology, groundwater will eventually reach the backfill. This is probably

the time when the hydraulic gradient across the backfill is the greatest. The

rate at which groundwater will be permeating and saturating the backfill is a very complicated problem because of the existence of the thermal gradient, and

is one of the major unsolved problems in the design of using backfill as one barrier. With the gradual reduction of thermal gradient with time, the backfill

will eventually be saturated. At that time, the backfill will exist in an aqueous environment influenced by reactions between groundwater and host rock,

between groundwater and the metal canister, and between groundwater and the

waste forms. What reactions that may affect the properties of the backfill

after saturation and the reaction rates remain to be determined.

The following is a summary of the discussions presented above

1. For routine determination of hydraulic conductivity, ASTM standards should

be followed.

2. For evaluation of hydraulic conductivity under repository condition, no standard method is available. Close working relationship among researchers

should be sought in order to provide a better basis for comparison of the

test results.

3. There is also no standard method on the evaluation of swelling potential

of fine-grained soils. Several techniques have been popularly used among

195
the researchers. Equipment has also been developed to incorporate the

measurements of both hydraulic conductivity and swelling. Again, close

working relationship among researchers is essential in developing reliable

material properties for the design of long term stability of the backfill

barrier

.

4. The mechanism of diffusion should be further studied. Surface diffusion,

which has been a neglected area because most of the studies up to the

present have focused on the bulk diffusion, should be evaluated.

5. Thermal properties of the backfill should be better defined and measured as

a function of its material type, the geometrical arrangement of its particles,

its moisture content at the placement, its density, and the thermal gradient

imposed across the backfill under a specific design waste loading.

6. Finally, both mathematical modeling and experimental work should be carried

out to study the backfill properties in a hydrothermal environment developed

in a multi-barrier waste isolation system.

References

1. Lambe , T.W., Compacted Clay: Engineering Behavior Transaction ASCE , 125,

pp. 718-741, (1960).

2. Yong , R.N., Taylor, L.O. and Warkenton, B.P., Swelling Pressures of Sodium

Montmorillonite at Depressed Temperatures, Proc. 11th National Conference

on Clays and Clay Minerals , (1962).

3. Yong, R.N., The Swelling of a Montmorillonitic Clay at Elevated Temperatures,

Proc. 3rd Asian Regional Conference on SMFE , (1967).

4. Pusch, R. , Swelling Pressure of Highly Compacted Bentonite, KBS Technical

Report 80-13, (1980).

196
5. Pusch, R. , Permeability of Highly Compacted Bentonite, KBS Technical Report

80-16, (1980).

6. Heinz, W.E., A Means of Predicting Swelling Pressures of Soils Found in

the Rocky Mountain Area, M.S. Thesis, University of Colorado , (1980).

7. Wheelwright, E.J., Hodges, F.N. , Westsik, J.H. Jr., and Bray, L.A. , Pacific

Northwest Laboratory's Backfill Development Project, This Publication , (1983).

8. Nowak, E.J., The Backfill Barrier as a Component in a Multiple Barrier

Nuclear Waste Isolation System, SAND 79-1109 , (1979).

9. Pusch, R. , Bentonite Shielding of Rock-Deposited Radioactive Wastes,

This Publication , (1983).

10. Nowak, E.J., Backfill Barrier Research and Development for Radioactive

Waste Isolation in Salt, Abstract Only, This Publication , (1983).

11. Draft Report, "Interime Reference Repository Conditions for Spent Fuel

and Commercial High-Level Nuclear Waste Repositories in Tuff," Prepared

by The Reference Repository Conditions Interface Workshop Group , (1980).

12. Radhakrishna , H.S., Evaluation of Thermal Properties of Buffer Materials

for a Deep Underground Nuclear Waste Disposal Vault, This Publication , (1983).

197
>

'-J -a 3

33 (3oU c

v. - e

a" to w

(0 l-l <D •V H

«J

,r j T,

r: >>

CO CO u (0 C0 c 3j o Q

pa

too o e0 u u aj V f=

a c Uh iH N

'J

a Cfl O. ou CO o.

0) J3 c *J o o

•H i-l 4-) co >> QJ

s~ 3!

(0 3

ft 4J QJ

QJ c (0

1-1 rl c

ft 0 o

'J c >x

<U

U S > u 4-t O

CA 3

•a c

U-4 c o o 0 0 H

mo 1 x 1 1 1 1 oX

CM 3 < 1

(N < cN -1 1 X> CN

(fl o

-X

T XJ a) CDu

C 1 c uX QJ o < -rH O o 0 cfl o u

rH

H CJ

u i- •H X 3 x s

a)

aQ 1 3 O O 0} cO o

7) u-lrm T3 CO cn

u

O o

-a

QJ

-r-t < CO in

X>

—t O t ul r-- <

-a -o

C c c o c CO CO H -H

4J -H >, X) O iH -H M 0) •H QJ

CO

J

H >—I

US i-f Cfl

a 3 (0 a) !> r0 (0 •/> 2 H O •H o CU 0 ^ O1 a> » It iJ a CO *J a. H CO a <0 o C iJ o c u O 3 CO CJ

4-1

UM

H

O t4 DC

C QJ

u *C

O J= -H U CO C QJ C 3 cfl C

OJ t-«

U -H

OOO

O ^ 8 C rH 3

QJ -i-t 1-4 33 VI CJ) 3 C bO

II 3 H -H

x: o —i u U J-4 -h u

~o

un c 3 qj u u O 3 U -O

198
199

APPENDIX A

A-1
NATIONAL BUREAU OF STANDARDS

Questionnaire

Nuclear Waste Management

Deep Geologic Repositories

Backfill Selection and Evaluation

To get the maximum benefit from your participation in this workshop,

we are soliciting your concise comments on the questions given herein

before adjournment on April 14, 1981.

Name

Affiliation

Address

Telephone Number

Area of Expertise Relating to Nuclear Waste Management

Percent of Time Devoted to Nuclear Waste Problem

A-2
1. What do we expect from the placed backfill in addition to that

given below:

- Hydrologic barrier to groundwater intrusion

- Its plastic nature will seal localized cracks in the surrounding

rock to minimize the potential of groundwater intrusion

- To buffer mechanical stresses and deformations

- To provide suitable bearing support and thus minimize subsidence

of the waste package

- To serve as an adequate thermal transporting medium

- To serve as a chemical conditioner of intruding groundwater

by modifying pH, EH and chemical composition

- To serve as a radionuclide sorptive barrier

- Others

A-3
2(a) List criteria for the material properties you consider important for

selection of backfill and then distribute 100 points among the criteria

based on their relative importance.

Criteria Points

2(b) For preliminary screening purposes, which of the above criteria would

you use.

A-4
3(a) Distribute 100 points based on the relative importance of the following

backfill material properties.

Material Properties Point

s

Permeability

Swelling

Thermal conductivity

Sorpt ion

Chemical Stability

Plast ic ity

Bearing capacity

Shear strength

3(b) For preliminary screening purposes, which of the above properties

would you use (no more than 4).

A-5
4(a) What are the advantages and disadvantages of manufactured backfill

with artificial additives versus various mixtures of natural materials?

4(b) Do you feel the location and type of parent material selected for the

repository affect your answer to Question 4(a) and to the backfill

evaluation in general?

A-6
5(a) Combining the materials given below and others, please construct the

ideal backfill you can envision:

- Clays - montmerillonite , bentonite

- Zeolites

- Sand

- Crushed host rock materials

- Additives for sorption of selected radionuclides (getters)

or Eh, pH control

- Others

5(b) Will the design concept change if you consider such factors as:

Availability

Quality Control:

Cost

A7
Describe in your vision the placement procedures and anticipated

difficulties in the installation of the backfill.

A-8
7. What is the ranges in the test variables listed within which laboratory

testing should be conducted.

Test Variables Range/Type

Actually used for tests

Desired already completed in your

proj ect .

° Temperature

° Groundwater /leachant

(site specific)

- hydraulic gradient

- liquid/solid ratio

° Solution Composition

(site specific)

- pH

- Eh

- ionic composition

species

concentrat ion

- corrosion products

- radiolysis products

0 Mechanical stresses

- compression (lithostat ic)

pressure

- swelling

- shear

- cyclic wetting/drying

° Material physical properties

- particle size distribution

- density

° Radiation dose rate and total dose

e Others

A-9
8(a) List test equipment, if known, used in evaluating the following backfill

Properties:

Propert ies Equipment

Swell ing

Permeability

Thermal conductivity

Strength

Sorpt ion

Chemical analysis

Others

A-10
8(b) Describe the test methods used in evaluating the backfill properties:

Availability

of

Property Test Methods Test Procedures

Swelling

Permeability

Thermal conductivity

Strength

Sorpt ion

Chemical analysis

Others

A-ll
9. Briefly discuss the change in backfill properties with time and your

vision of sequence of event happening to the backfill after its

placement. Identify major problem areas associated with the development of backfill.

A-12
10(a) Please list reference sources that would be helpful in the

collection of information regarding the backfill characteristics.

10(b) What references do you use most often?

A- 13

APPENDIX

B-1
Members of the Organizing Committee:

Mr. William P. Reed

Acting Program Manager

Nuclear Waste Management Program

National Measurement Laboratory

Dr . Riley M . Chung

Dr. Felix Y. Yokel

Mr. Lawrence A. Salomone

Geotechnical Engineering Group

Structures Division

Center for Building Technology

National Engineering Laboratory

B-2
List of Participants

Workshop on Research and Development Needs in Backfill for Long Term Storage

of Nuclear Waste in Deep Geologic Repositories, April 13 and 14 , 1981 at the

National Bureau of Standards.

Organizing Committee:

Riley M. Chung

William P. Reed

Lawrence A. Salomone

Felix Y. Yokel

Part icipant s

Name

Erik D. Anderson

Gary W. Beall

Anders Bergstrom

D. J. Cameron

Steven Cheung

National Bureau of Standards

Building 226, Room B162

Washington, D.C. 202 34

National Bureau of Standards

Building 221, Room B320

Washington, D.C. 20234

National Bureau of Standards

Building 226, Room B162

Washington, D.C. 20234

National Bureau of Standards

Building 226, Room B162

Washington, D.C. 20234

Organization and Address

National Bureau of Standards

Building 226, Room B162

Washington, D.C. 20234

Radian Corporation

8500 Shoal Creek Road

Austin, TX 78766

SKBF/KBS

Box 5864

S-102 48 Stockholm

Sweden

Atomic Energy of Canada Limited

Whiteshell Nuclear Research

Establishment

Pinawa, Manitoba

R0E1L0 , Canada

Atomic Energy of Canada Limited

Whiteshell Nuclear Research

Establishment

Pinawa, Manitoba

R0E1L0, Canada

FTS-921-2137

301-921-2137

FTS-921-3868

301-921-3868

FTS-921-3128

301-921-3128

FTS-921-2648

301-921-2648

Telephone No

FTS-921-3180

301-921-3180

512-454-4797

ext. 5631

08-67 95 40

204-753-2311

204-753-2311

B-3
Name Organization and Address Telephone No

Karen P. Coccia

Ramesh Dayal

National Bureau of Standards

Building 225, Room A140

Washington, D.C. 20234

Brookhaven National Laboratory

Department of Nuclear Energy

Nuclear Waste Management Div.

Upton, NY 11973

FTS-921-3395

301-921-3395

FTS-666-4993

516-345-4993

Bruce R. Erdal Los Alamos National Laboratory

CNC7 Mail Stop 514

Los Alamos, NM 87545

FTS-843-5338

505-667-5338

John T. Greeves U.S. Nuclear Regulatory Commission

Mail Stop SS-905

Washington, D.C. 20555

FTS -427-4177

301-427-4177

Floyd Hodges Battelle Pacific Northwest

Laboratory

P.O. Box 999

Richland, WA 99352

FTS -444-4467

509-444-4467

J. Keith Johnstone

William D. Kovacs

M. D. Merz

Donald P. Moak

Judith Moody

E. James Nowak

Edward 0. Pfrang

Sandia National Laboratories

Division 4537

Albuquerque, NM 87185

National Bureau of Standards

Building 226, Room B162

Washington, D.C. 20234

Nuclear Waste Material

Characterization Center

Battelle Northwest

P.O. Box 999

Richland, WA 99352

ONWI

Battelle Memorial Institute

505 King Avenue

Columbus, OH 43201

ONWI

Battelle Memorial Institute

505 King Avenue

Columbus, OH 43201

Sandia National Laboratories

P.O. Box 5800, Div. 5812

Albuquerque, NM 87185

National Bureau of Standards

Building 226, Room B370

Washington, D.C. 20234

FTS-844-7633

505-844-7633

FTS-921-2885

301-921-2885

FTS-440-2943

509-373-2943

FTS-976-7125

614-424-7125

FTS-976-5536

614-424-5536

FTS-844-1109

505-844-1109

FTS-921-2196

301-921-2196

B-4
Name Organization and Address Telephone No.

C. L. Pittiglio

Roland Pusch

H. S. Radhakrishna

U.S. Nuclear Regulatory Commission

Mail Stop SS-905

Washington, D.C. 20555

Division of Soil Mechanics

University of Lulea

Lulea, Sweden

Ontario Hydro Research

800 Kipling Avenue

Toronto, Canada

M8Z 5S4

FTS-427-4177

301-427-4177

46-920-91-357

416-231-4111

ext. 6695

Larry Ramspot Lawrence Livermore National

Laboratory

Mail Code 204

P.O. Box 808

Livermore, CA 94550

FTS-422-4176

415-422-4176

Delia M. Roy

Rustum Roy

Tom Schmitt

James Schornhorst

Ronald A. Schrack

Trueman Seamans

A. J. Sethi

Material Research Laboratory

Pennsylvania State University

University Park, PA 16802

Material Research Laboratory

Room 203

Pennsylvania State University

University Park, PA 16802

U.S. Nuclear Regulatory Commission

Mail Stop NL-5650

Washington, D.C. 20555

Westinghouse Electric Company

Advanced Energy Systems

Box 10864

Pittsburgh, PA 15236

National Bureau of Standards

Building 245, Room B110

Washington, D.C. 20234

U.S. Nuclear Regulatory Commission

Mail Stop SS-905

Washington, D.C. 20555

Geotechnical Research Center

McGill University

817 Sherbrooke Street, W

Montreal, PQ. Canada

814-865-1196

814-865-3421

FTS-443-5976

301-443-5976

412-892-5600

FTS-921-2677

301-921-2677

FTS-427-4177

301-427-4177

514-392-8295

B-5
Name Organization and Address Telephone No

.

Dale Simpson Lehigh University

Dept . of Geological Sciences

Williams Hall #31

Bethlehem, PA 18015

215-861-3664

Elizabeth N. Vine Los Alamos National Laboratory

Los Alamos, NM 87545

FTS-843-4845

505-667-4845

Earl J. Wheelwright Battelle Pacific Northwest

Laboratory

P.O. Box 999

Richland, WA 99352

FTS -444-3441

509-444-3441

M. I. Wood Rockwell Hanford Operation

Basalt Waste Isolation Project

Columbia Bank Building

Richland, WA 99352

FTS-444-7833

509-376-7833

Howard T. Yolken National Bureau of Standards

Building 220, Room A259

Washington, D.C. 20234

FTS-921-3775

301-921-3775

Raymond N. Yong Geotechnical Research Center

McGill University

817 Sherbrooke Street, W.

Montreal, PQ . Canada

H3A 2K6

514-392-4751

B-6
NBS-114A I REV. 2-8C) U.S. DEPT. OF COMM. BIBLIOGRAPHIC DATA

SHEET (See instructions)

1. PUBLICATION OR

REPORT NO.

NBS SP 668

2. Performing Organ. Report No. 3. Publ icatlon Date December 1983

4. TITLE AND SUBTITLE

Nuclear Waste Management Program Summary Report

(Workshop on Research and Development Needs in Backfill for Long Term Storage of

Nuclear Waste in Deep Geologic Repositories, held at the National Bureau of

Standards, Washington, DC, April 13-14, 1981)

Riley M. Chung, Editor

6. PERFORMING ORGANIZATION (If joint or other than NBS. see instructions) 7. Contract/Grant No.

NATIONAL BUREAU OF STANDARDS

DEPARTMENT OF COMMERCE 8. Type of Report & Period Covered

WASHINGTON, D.C. 20234 Final

9. SPONSORING ORGANIZATION NAME AND COMPLETE ADDRESS (Street. City. State, ZIP)

Office of Nuclear Waste Management

Department of Energy

Washington, DC 20585

10. SUPPLEMENTARY NOTES

Library of Congress Catalog Card Number: 83-600607

~J Document describes a computer program; SF-185, FlPS Software Summary, is attached.

11. ABSTRACT (A 200-word or less factual summary of most significant information. If document includes a significant

bibliography or literature survey, mention it here)

An international workshop on the backfill component of a multi-barrier underground isolation system for nuclear waste disposal/storage was conducted on April 13 and 14, 1981, at the National Bureau of Standards, Washington, DC.

This report includes the papers presented and/or submitted to the workshop for

publication. It also contains the surveyed results of a questionnaire distributed

to the participants at the workshop. The workshop held a one-day informal indepth discussion on the subject and a summary of these discussions is included

in this report, which states areas where further research and development needs

are required to better understand the fundamental mechanisms of the backfill in

the waste repository.

12. KEY WORDS (Six to twelve entries; alphabetical order; capitalize only proper names; and separate key words by semicolon s) backfill; compressibility; geology; nuclear waste; permeability (hydraulic

conductivity); research and development; shear strength; sorption; swelling;

thermal conductivity.

13. AVAILABILITY

Unlimited

|

| For Official Distribution. Do Not Release to NTIS

[X] Order From Superintendent of Documents, U.S. Government Printing Office, Washington, D.C. 20402.

~~

1 Order From National Technical Information Service (NTIS), Springfield, VA. 22161

14. NO. OF

PRINTED PAGES

228

15.

USCOMM-DC 6043-P80

NBS TECHNICAL PUBLICATIONS

PERIODICALS

JOURNAL OF RESEARCH—The Journal of Research of the National Bureau of Standards reports NBS research and develop- ment in those disciplines of the physical and engineering sciences in which the Bureau is active. These include physics, chemistry,

engineering, mathematics, and computer sciences. Papers cover a broad range of subjects, with major emphasis on measurement

methodology and the basic technology underlying standardization. Also included from time to lime are survey articles on topics

closely related to the Bureau's technical and scientific programs. As a special service to subscribers each issue contains complete

citations to all recent Bureau publications in both NBS and non- NBS media. Issued six times a year. Annual subscription: domestic

$18; foreign $22.50. Single copy. $5.50 domestic; $6.90 foreign.

NONPERIODICALS

Monographs—Major contributions to the technical literature on

various subjects related to the Bureau's scientific and technical activities. Handbooks—Recommended codes of engineering and industrial practice (including safety codes) developed in cooperation with in- terested industries, professional organizations, and regulatory

bodies. Special Publications—Include proceedings of conferences sponsored by NBS, NBS annual reports, and other special publications

appropriate to this grouping such as wall charts, pocket cards, and

bibliographies. Applied Mathematics Series—Mathematical tables, manuals, and

studies of special interest to physicists, engineers, chemists,

biologists, mathematicians, computer programmers, and others engaged in scientific and technical work.

National Standard Reference Data Series—Provides quantitative data on the physical and chemical properties of materials, compiled from the world's literature and critically evaluated. Developed under a worldwide program coordinated by NBS under

the authority of the National Standard Data Act (Public Law

90-396).

NOTE; The principal publication outlet for the foregoing data is the Journal of Physical and Chemical Reference Data (JPCRD)

published quarterly for NBS by the American Chemical Society

(ACS) and the American Institute of Physics (A1P). Subscriptions,

reprints, and supplements available from ACS, 1 155 Sixteenth St., NW, Washington. DC 20056.

Building Science Series—Disseminates technical r formation developed at the Bureau on building materials, components,

systems, and whole structures. The series presents research results,

test methods, and performance criteria related to the structural and

environmental functions and the durability and safety characteristics of building elements and systems.

Technical Notes—Studies or reports which are complete in themselves but restrictive in their treatment of a subject. Analogous to monographs but not so comprehensive in scope or definitive in treatment of the subject area. Often serve as a vehicle for final reports of work performed at NBS under the sponsorship of other government agencies.

Voluntary Product Standards—Developed under procedures

published by the Department of Commerce in Part 10, Title 15, of

the Code of Federal Regulations The standards establish nationally recognized requirements for products, and provide all concerned interests with a basis for common understanding of the

characteristics of the products. NBS administers this program as a supplement to the activities of the private sector standardizing

organizations. Consumer Information Series—Practical information, based on NBS research and experience, covering areas of interest to the con- sumer. Easily understandable language and illustrations provide

useful background knowledge for shopping in today's tech- nological marketplace. Order the above SBS publication!; from: Superintendent of Documents, Government Printing Office, Washington, DC 20402. Order the following SBS publications— Fl PS and SBSIR's—from

the Sational Technical Information Service . Springfield. VA 22 161.

Federal Information Processing Standards Publications (FIPS

PUB)—Publications in this series collectively constitute the Federal Information Processing Standards Register. The Register

serves as the official source of information in the Federal Govern- ment regarding standards issued b> NBS pursuant to the Federal Property and Administrative Services Act of 1949 as amended.

Public Law 89-306 (79 Stat. 1127), and as implemented by Executive Order 11717(38 FR 12315, dated Ma> II, 1973) and Part 6 of Title 15 CFR (Code of Federal Regulations).

NBS Interagency Reports (NBSIR)—A special series of interim or

final reports on work performed by NBS for outside sponsors

(both government and non-government). In general, initial dis- tribution is handled by the sponsor; public distribution is by the National Technical Information Service , Springfield, VA 22161.

in paper copy or microfiche form.
U.S. Department of Commerce

National Bureau of Standards

Washington, D.C. 20234

Official Business

Penalty for Private Use $300

POSTAGE AND FEES PAID U S DEPARTMENT OF COMMERCE

COM-215

SPECIAL FOURTH-CLASS RATE BOOK