Performance of High Level Waste Forms and Packages under Repository Conditions

289 Pages·2007·6.19 MB·English

Results of a previously unpublished CRP:
Performance of High Level Waste Forms
and Packages under Repository
Conditions

Results of a Co-ordinated Research Project
1991-1998
Foreword

Production of energy at nuclear power plants is unavoidably accompanied by the
generation of spent fuel, which has to be safely managed. There are two main strategies for
spent fuel management. In some countries, reprocessing of spent nuclear fuel is performed to
recover fissile materials for fuel fabrication and separate the fission products and small
amount of actinides in the form of high level waste (HLW). Other countries have chosen a
once-through fuel cycle strategy and pursue interim storage of “as generated” spent fuel, to
be followed by its final disposal.

The final step in both strategies is permanent disposal of spent fuel or conditioned
HLW (solidified in various kinds of glass or ceramic matrices) in deep geological formations.
Safe disposal of spent fuel and conditioned HLW requires an extensive international research
program to develop waste forms and waste packages suitable for disposal in various
geological environment. All components of the waste package (HLW form and the container)
and engineered barrier systems (host rock and backfills) should be designed and carefully
investigated to provide for long-term containment and isolation of the disposed waste from
the repository environment, in particular groundwater, and to retard the transport of
radionuclides from a failed container to the groundwater.

The IAEA has initiated a sequence of Coordinated Research Projects (CRP) to support
and co-ordinate international research activities in the above mentioned subjects. In the CRP
on “Performance of High Level Waste Forms and Packages under Repository Conditions”,
the problems of behaviour and interactions of the spent fuel and/or various HLW forms in the
conditions appropriate to specific geological repositories have been addressed. Fourteen
laboratories, representing thirteen IAEA Member States, participated in the Project. It should
be mentioned that despite the title of the CRP, little content on container behaviour was done.

The chief scientific investigators from Argentina, Australia, Belgium, Canada, China,
Czech Republic, Finland, France, Germany, India, Japan, Russia, and the United States of
America met in four Research Coordination Meetings in Karlsruhe (1991), Bombay (1993),
Tokai (1995) and in Avignon (1997) to present the scientific reports from participating
laboratories, discuss results and provide recommendations for future work. This TECDOC
summarizes principal achievements in participating laboratories and provides a consolidated
summary of the entire Project, as prepared by the group of consultants at the end of the CRP.
Final reports of participating laboratories, reviewed by J. Tait of AECL, Canada and R. Burcl
of the IAEA, are in the attachments.

V. Tsyplenkov of the IAEA Division of Nuclear Fuel Cycle and Waste Technology
initiated the CRP and guided the Project until February 1997. R. Burcl of the same Division
continued the Project and was responsible for compilation of this TECDOC while
J.L.Gonzalez and subsequently Paul J.C. Dinner, also from this Division, took the final
responsibility for its revision and publication.

Originally foreseen to be published as a “stand-alone” TECDOC, the publication of this
work was unfortunately delayed. Meanwhile, work continued on a related Technical Co-
operation (TC) project entitled “Chemical Durability and Performance Assessment under
Simulated Repository Conditions”. It was decided to consolidate publication of the work into
a single TECDOC, with the earlier work appended as a CD to the more recent investigations.

CONTENTS

1. INTRODUCTION..............................................................................................................1

1.1. Scientific background...............................................................................................1
1.2. Objective..................................................................................................................2
1.3. Scope and outline of the CRP...................................................................................2
2. HIGH-LEVEL WASTE MANAGEMENT STRATEGY...................................................4
3. PERFORMANCE OF HIGH-LEVEL WASTE FORMS UNDER REPOSITORY
CONDITIONS..............................................................................................................................5
4. SUMMARY OF LEACHING STUDIES............................................................................10

4.1. Glass.........................................................................................................................10
4.1.1. Glass leaching behaviour and effects of glass composition...........................10
4.1.2. Leachant composition...................................................................................11
4.1.3. Environmental conditions.............................................................................11
4.1.4. Radiation studies...........................................................................................12
4.1.5. Active, in-situ and natural analogue studies..................................................12
4.1.6. Modelling.....................................................................................................13
4.2. Spent fuel..................................................................................................................13
4.3. Ceramics...................................................................................................................15
5. PHILOSOPHY OF TESTS PLANNING............................................................................17
6. CONCLUSIONS.................................................................................................................19
7. RECOMMENDATIONS....................................................................................................21
REFERENCES.............................................................................................................................22

CONTRIBUTIONS BY PARTICIPANTS IN THE CO-ORDINATED RESEARCH
PROJECT ON LONG TERM BEHAVIOUR OF LOW AND INTERMEDIATE LEVEL
WASTE PACKAGES UNDER REPOSITORY CONDITIONS

Behavior of sintered glasses containing simulated high level wastes under repository conditions
D. Russo

Long-term behaviour of High Level Waste Forms in clay repository conditions
P. Van Iseghem

The performance of Synroc under repository conditions
K.P. Hart

Assessment of the performance of used CANDU fuel under disposal conditions
J.C. Tait
Study of properties of high level waste forms and packages under simulated disposal conditions
S. Luo

Experimental evaluation of interactions of solidified HLW forms under repository simulated
conditions
V. Balek

Dissolution of unirradiated UO fuel under simulated disposal conditions of spent fuel in
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crystalline granitic bedrock
K. Ollila

Long-term behaviour of HLW glasses in geological disposal conditions
T. Advocat

Corrosion behavior of the high level waste forms borosilicate glass and spent fuel in salt brines
A. Loida

Performance of high level waste forms and packages under repository conditions
K. Raj

Studies of waste form performance at Japan Atomic Energy Research Institute
T. Banba

Glass form performance for disposal of the Cesium and strontium concentrate resulting from the
partitioning of HLW
A.S. Aloy

Corrosion of radioactive phosphate glasses under repository conditions
P.P. Poluektov

The development of borosilicate glass for U.S. DOE radioactive wastes
P. Hrma

CONTRIBUTORS TO DRAFTING AND REVIEW

1. INTRODUCTION

1.1. SCIENTIFIC BACKGROUND

The disposal of radioactive waste containing fission products and actinides,
resulting from the use of nuclear power, will require its conditioning in proper matrix
and packaging, to assure isolation from the biosphere for periods of up to a million
years. For the conditioning of HLW, it has been noted by Lutze and Ewing [1] that:

“Given the compositional diversity of nuclear waste and the variations in
potential repository geologies in different countries, it is imperative that there be a
‘menu’ of more than one waste form. One must be able to choose the waste form that
best suits the type of waste and the geology of the repository.”

The waste form and waste container for a particular application should be
chosen to provide the optimum solution for each country’s program. The waste form
is the first barrier. As such, the knowledge of how the waste form will perform over
the geological time-scale will provide assurance for the safe disposal of the high level
waste.

Disposal of conditioned HLW waste is internationally envisaged to be within a
geological repository using a number of barriers to isolate the conditioned waste from
the repository environment. The design of the entire waste disposal system (waste
form, waste container, engineered barriers and host geology) will determine the
degree of isolation of the waste from the environment. Typically, the disposal system
consists of the waste package (conditioned high-level waste or spent fuel in a suitable
container), engineered barriers within the repository (e.g. clay, apatite, crushed salt,
backfill), the natural barrier of the host site (e.g. rock, clay, salt) and the surrounding
geological media.

A credible scenario for radionuclides to be released from a waste form is
through dissolution in groundwater that may be present or later enter the repository.
Performance testing of these waste forms will thus require aqueous leaching tests to
be conducted on waste form compositions. Short-term leach tests, using standardized
conditions, e.g. deionized water, MCC-1, PCT, ISO, Soxhlet [2], may be appropriate
for the inter-comparison of the durability of different waste form compositions.
However, the data obtained from such experiments are not sufficient for prediction of
long-term performance under repository conditions. In order to provide for long-term
extrapolation of waste form performance, and to collect data suitable for performance
assessment modelling, it is necessary to arrange for leaching studies over a long-
period of time (months to years), under repository relevant conditions (e.g. simulated
groundwater, redox conditions, appropriate flow-rates, temperature, host rock,
engineered barriers), and to develop an understanding of the leaching mechanism.

To develop a program to choose the HLW forms and assess their performance,
the following steps are foreseen:

• Develop an understanding of the existing knowledge of the waste form and
geology being considered for disposal;
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• Carry out detailed laboratory testing on inactive and simulated waste materials
using recognized and standardized procedures to understand the waste
performance under repository conditions and compare these results with the
knowledge and expectations gained from the previous step;

• Further extend laboratory testing to fully active samples to confirm results from
the previous steps;

• Conduct in-situ testing to investigate whether the performance in the geological
formation matches that obtained in the laboratory;

• Where appropriate natural analogues exist, compare the long-term predictions
and understanding of the waste form behaviour from laboratory and in-situ
studies with the known behaviour of these natural systems; and

• Develop scientific models based on experimentally determined mechanisms that
provide a sound basis for describing the long-term performance of the waste.

1.2. OBJECTIVE

The objective of the CRP on the “Performance of High Level Waste Forms and
Packages under Repository Conditions” was to contribute to the development and
implementation of proper and sound technologies for HLW and spent fuel management.
Special emphasis was given to the identification of various waste form properties and
the study of their long term durability in simulated repository conditions. Another
objective was to promote the co-operation and exchange of information between
Member States on experimental concerning behaviour of the waste form. The CRP was
composed of research contracts and agreements with Argentina, Australia, Belgium,
Canada, China, Czech Republic, Finland, France, Germany, India, Japan, Russia, and
the United States of America.

The main objectives of the Research co-ordination meetings held during the
CRP were:

• Provide an open forum for discussion on the subject matter of the meeting
between the CRP chief investigators;
• Give each participant the opportunity to formally present their program and to
receive the benefit of a review of their results and proposed work plan by
experts from other countries;
• Determine and co-ordinate the future activities within the CRP, including
program direction.

1.3. SCOPE AND OUTLINE OF THE CRP

The Project covered studies on glasses (borosilicate, boro-aluminosilicate, and
alumino-phosphate compositions), ceramics (SYNROC, perovskite, and zirconia- and
alumina-based ceramics) and spent UO fuel. More detailed information about the
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compositions and types of waste forms studied and studies performed in this CRP is
given in the attached final reports of the participants.

Terms of reference for the CRP under which contracts and agreements have been
executed between the Agency and laboratories/institutions of the Member States
include:

• Studies of the properties and durability of the waste forms;
• Interactions between the waste package components under repository
conditions;
• Near field effects on the waste package and components performance, including
the influence of radiation, thermal and geochemical environment as a function
of time; and
• Comparison of scientific and empirical models describing performance of the
waste form with experimental results.

It should be noted that even though the CRP was foreseen to address the role of waste
packages including the container, very little content on container behaviour was done.

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2. HIGH-LEVEL WASTE MANAGEMENT STRATEGY

From the point of view of activity content, spent fuel is the most significant
waste generated in nuclear reactor operation. Depending upon the selected fuel cycle
strategy, spent fuel can be considered either as a final waste form and directed for
disposal, or as a valuable source of fissionable material and reprocessed. In the last
case, fissionable material is separated in reprocessing. As a by product, a relatively
small volume of highly active waste, containing fission products and a small quantity
of transuranic elements, is generated.

HLW, generated in spent fuel reprocessing, is most commonly conditioned by
vitrification. As an alternative to HLW glass, special types of ceramic matrices can be
also used for HLW conditioning. Ceramics are also preferred for conditioning of
actinides.

Disposal in deep geological formations is presently considered to be the only
realistic final destination of both spent fuel as well as conditioned HLW. A crucial
problem of any radioactive waste disposal is the potential risk of activity release from
the disposed waste form to the environment, in particular to groundwater. In the case
of geological disposal of spent fuel and HLW, the properties and long term behaviour
of waste form (e.g. leaching properties, alteration of HLW matrices) represents only
one side of an extremely complex problem. Interactions of waste with the waste
container, durability of container material, near and far field interactions with
secondary container, backfilling material, host rock, impact of increased temperature,
radiation fields and many other factors have to be considered and carefully analysed.
Another significant problem is reliability of extrapolation of data, obtained in
relatively short term experiments, for an extended period, e.g. several millions years.

Any step in better understanding the above mentioned problems comes as the
result of complicated and expensive experiments and modelling, which are often
beyond the capabilities of individual countries. International co-operation and
exchange of technical information is, therefore, a consistent part of any spent fuel and
HLW management strategy, and obtained results are of special value for the whole
nuclear community. This has also been the main motivation for the IAEA support for
this CRP, where long time behaviour and durability of HLW glass and ceramic waste
forms as well as UO should eventually be studied in the complex environment of a
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real disposal facility.
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