For Official Use NEA/CSNI/R(2002)14

Organisation de Coopération et de Développement EconomiquesOrganisation for Economic Co-operation and Development 10-Jul-2002___________________________________________________________________________________________

English - Or. EnglishNUCLEAR ENERGY AGENCYCOMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS

REPORT OF THE TASK GROUP REVIEWING ACTIVITIESIN THE AREA OF AGEING OF CONCRETE STRUCTURES USEDTO CONSTRUCT NUCLEAR POWER PLANT FUEL-CYCLE ACTIVITIES

JT00129447

Document complet disponible sur OLIS dans son format d’origineComplete document available on OLIS in its original format

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ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT

Pursuant to Article 1 of the Convention signed in Paris on 14th December 1960, and which came into force on 30thSeptember 1961, the Organisation for Economic Co-operation and Development (OECD) shall promote policies designed:

− to achieve the highest sustainable economic growth and employment and a rising standard of living in Membercountries, while maintaining financial stability, and thus to contribute to the development of the world economy;

− to contribute to sound economic expansion in Member as well as non-member countries in the process of economicdevelopment; and

− to contribute to the expansion of world trade on a multilateral, non-discriminatory basis in accordance withinternational obligations.

The original Member countries of the OECD are Austria, Belgium, Canada, Denmark, France, Germany, Greece,Iceland, Ireland, Italy, Luxembourg, the Netherlands, Norway, Portugal, Spain, Sweden, Switzerland, Turkey, the United Kingdomand the United States. The following countries became Members subsequently through accession at the dates indicated hereafter:Japan (28th April 1964), Finland (28th January 1969), Australia (7th June 1971), New Zealand (29th May 1973), Mexico (18thMay 1994), the Czech Republic (21st December 1995), Hungary (7th May 1996), Poland (22nd November 1996), Korea (12thDecember 1996) and the Slovak Republic (14th December 2000). The Commission of the European Communities takes part in thework of the OECD (Article 13 of the OECD Convention).

NUCLEAR ENERGY AGENCY

The OECD Nuclear Energy Agency (NEA) was established on 1st February 1958 under the name of the OEECEuropean Nuclear Energy Agency. It received its present designation on 20th April 1972, when Japan became its firstnon-European full Member. NEA membership today consists of 27 OECD Member countries: Australia, Austria, Belgium,Canada, Czech Republic, Denmark, Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Luxembourg,Mexico, the Netherlands, Norway, Portugal, Republic of Korea, Spain, Sweden, Switzerland, Turkey, the United Kingdom and theUnited States. The Commission of the European Communities also takes part in the work of the Agency.

The mission of the NEA is:

− to assist its Member countries in maintaining and further developing, through international co-operation, thescientific, technological and legal bases required for a safe, environmentally friendly and economical use of nuclearenergy for peaceful purposes, as well as

− to provide authoritative assessments and to forge common understandings on key issues, as input to governmentdecisions on nuclear energy policy and to broader OECD policy analyses in areas such as energy and sustainabledevelopment.

Specific areas of competence of the NEA include safety and regulation of nuclear activities, radioactive wastemanagement, radiological protection, nuclear science, economic and technical analyses of the nuclear fuel cycle, nuclear law andliability, and public information. The NEA Data Bank provides nuclear data and computer program services for participatingcountries.

In these and related tasks, the NEA works in close collaboration with the International Atomic Energy Agency inVienna, with which it has a Co-operation Agreement, as well as with other international organisations in the nuclear field.

© OECD 2002Permission to reproduce a portion of this work for non-commercial purposes or classroom use should be obtained through theCentre français d’exploitation du droit de copie (CCF), 20, rue des Grands-Augustins, 75006 Paris, France, Tel. (33-1) 44 07 4770, Fax (33-1) 46 34 67 19, for every country except the United States. In the United States permission should be obtained throughthe Copyright Clearance Center, Customer Service, (508)750-8400, 222 Rosewood Drive, Danvers, MA 01923, USA, or CCCOnline: http://www.copyright.com/. All other applications for permission to reproduce or translate all or part of this book shouldbe made to OECD Publications, 2, rue André-Pascal, 75775 Paris Cedex 16, France.

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COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS

The NEA Committee on the Safety of Nuclear Installations (CSNI) is an international committee made upof scientists and engineers. It was set up in 1973 to develop and co-ordinate the activities of the Nuclear EnergyAgency concerning the technical aspects of the design, construction and operation of nuclear installations insofar asthey affect the safety of such installations. The Committee’s purpose is to foster international co-operation in nuclearsafety amongst the OECD Member countries.

CSNI constitutes a forum for the exchange of technical information and for collaboration betweenorganisations which can contribute, from their respective backgrounds in research, development, engineering orregulation, to these activities and to the definition of its programme of work. It also reviews the state of knowledgeon selected topics of nuclear safety technology and safety assessment, including operating experience. It initiates andconducts programmes identified by these reviews and assessments in order to overcome discrepancies, developimprovements and reach international consensus in different projects and International Standard Problems, and assistsin the feedback of the results to participating organisations. Full use is also made of traditional methods of co-operation, such as information exchanges, establishment of working groups and organisation of conferences andspecialist meeting.

The greater part of CSNI’s current programme of work is concerned with safety technology of waterreactors. The principal areas covered are operating experience and the human factor, reactor coolant systembehaviour, various aspects of reactor component integrity, the phenomenology of radioactive releases in reactoraccidents and their confinement, containment performance, risk assessment and severe accidents. The Committeealso studies the safety of the fuel cycle, conducts periodic surveys of reactor safety research programmes and operatesan international mechanism for exchanging reports on nuclear power plant incidents.

In implementing its programme, CSNI establishes co-operative mechanisms with NEA’s Committee onNuclear Regulatory Activities (CNRA), responsible for the activities of the Agency concerning the regulation,licensing and inspection of nuclear installations with regard to safety. It also co-operates with NEA’s Committee onRadiation Protection and Public Health and NEA’s Radioactive Waste Management Committee on matters ofcommon interest.

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Acknowledgement

Gratitude is expressed to the USNRC who made it possible for Dr Dan Naus with ORNL to be theassembler of the information. Thanks also to Dr Naus for his excellent work.

Gratitude is also expressed to the members of the Task Group whose contributions helped in building aninternational review on the topic.

The Task Group was composed of the following members:

- Dr. L. M. Smith, Chairman (UK)- Mr. E. Mathet, Secretary (OECD/NEA)- Mrs. B. Aghili (Sweden)- Mr. F. Bernier (Belgium)- Mr. C. J. Bolton (UK)- Mr. B. Hedberg (Sweden)- Mr. O. Jovall (Sweden)- Dr. I. Martinez (Spain)- Mr. I. McNair (UK)- Dr. D. J. Naus (US)- Mr. S. Nakamura (Japan)- Mr. A. Taglioni (Italy)- Dr. A. Vokal (Czech Republic)

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Foreword

The December 1999 meeting of the CSNI accepted a proposal by the Integrity and Ageing of Structures(IAGE) Working Group to broaden its scope to include the long term behaviour of concrete used toconstruct: fuel cycle facilities; fuel storage facilities; and other nuclear structures required to perform asafety function for a very long period of time, eg. decommissioned facilities.

A Task Group was then set up whose objectives were to produce a report of its discussion that wouldestablish the level of national interest in the topic; provide a forum for presentations of relevant nationalprograms and long-term strategies for fuel cycle, fuel storage, and other nuclear safety-related structures;and, where appropriate, identify and prioritize specific tasks that warrant additional attention by theConcrete Subgroup.

To help organize the discussion a questionnaire on fuel cycle, fuel storage and other nuclear safety relatedconcrete structures required to perform a safety function for a very long period of time was sent to the TaskGroup prior to this meeting. The concerns were with regard to safety and long term performance ofconcrete structures. The questionnaire addressed all aspects of concrete structure design, in-serviceinspection policies, operating experience and degradation, if any, observed for those types of structure inaddition to policy considerations.

This report was prepared using the results obtained from the questionnaire, input from task groupmembers, and information presented in the open literature.

The complete list of CSNI reports, and the text of reports from 1993 onwards, is available onhttp://www.nea.fr/html/nsd/docs/

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CONTENTS

Page

EXECUTIVE SUMMARY ................................................................................................................... 13

1. INTRODUCTION ................................................................................................................... 17

1.1. BACKGROUND..................................................................................................................... 171.2. MANDATE OF THE TASK GROUP.................................................................................... 181.3. ORGANIZATION OF THE TASK GROUP.......................................................................... 18

2. GENERATION, CHARACTERIZATION, AND MANAGEMENTOF RADIOACTIVE WASTE.......................................................................................................... 19

2.1 INTRODUCTION................................................................................................................... 192.2 SOURCES OF RADIOACTIVE WASTE.............................................................................. 19

2.2.1. Nuclear Reactors..................................................................................................... 192.2.2. Medicine, Industry, and Research .......................................................................... 202.2.3. Decommissioned Facilities..................................................................................... 21

2.3 CLASSIFICATION OF RADIOACTIVE WASTE .............................................................. 222.4 MANAGEMENT OF RADIOACTIVE WASTE.................................................................. 23

2.4.1 Exempt Waste......................................................................................................... 232.4.2 Low- and Intermediate-Level Waste ...................................................................... 232.4.3 High-Level Waste................................................................................................... 24

3. FUEL-CYCLE FACILITTIES AND LONG-TERMPERFORMANCE CONSIDERATIONS......................................................................................... 37

3.1. MATERIALS OF CONSTRUCTION.................................................................................... 373.2. POTENTIAL DEGRADATION CONSIDERATIONS ......................................................... 38

3.2.1. Concrete Degradation Mechanisms ........................................................................ 38

3.2.1.1. Chemical Attack ..................................................................................................... 383.2.1.2. Physical Attack ....................................................................................................... 40

3.2.2 Steel Reinforcement Degradation Mechanisms...................................................... 42

3.2.2.1. Corrosion ................................................................................................................ 423.2.2.2 Other Degradation Factors...................................................................................... 44

3.2.3. Liner Degradation Mechanisms.............................................................................. 44

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3.3. SERVICE CAPABILITIES AND DESIGN CONSIDERATIONS ...................................... 443.4. DESCRIPTIONS OF SEVERAL TYPICAL FUEL CYCLE FACILITIES ......................... 45

3.4.1 Mining and Milling Waste Sites ................................................................................ 453.4.2 Low- and Intermediate-Level Radioactive Waste Disposal Facilities ....................... 46

3.4.2.1 Use of Reinforced Concrete Structures ..................................................... 463.4.2.2 Applications............................................................................................... 48

3.4.3 Transuranic Waste Disposal Facilities ....................................................................... 533.4.4 High-Level Radioactive Waste Storage and Disposal Facilities................................ 53

3.4.4.1 Fuel Storage/Radwaste .............................................................................. 543.4.4.2 Tanks ......................................................................................................... 553.4.4.3 Geologic Repositories................................................................................ 56

3.5 PERFORMANCE ASSESSMENT AND INSTRUMENTATIONCONSIDERATIONS ............................................................................................................. 57

3.5.1 Performance Assessment ................................................................................................ 573.5.2 Instrumentation ............................................................................................................... 58

3.6 MODELING FOR SERVICE LIFE ESTIMATIONS ........................................................... 59

4. NATIONAL AND INTERNATIONAL RADIOACTIVE WASTEMANAGEMENT ACTIVITIES ...................................................................................................... 79

4.1. NATIONAL PROGRAMS .................................................................................................... 794.1.1. Belgium................................................................................................................... 794.1.2. Canada ................................................................................................................... 804.1.3. Czech Republic ....................................................................................................... 814.1.4. Finland ................................................................................................................... 814.1.5. France ................................................................................................................... 824.1.6. Germany ................................................................................................................. 834.1.7. Hungary .................................................................................................................. 854.1.8. Italy ................................................................................................................... 854.1.9. Japan ................................................................................................................... 864.1.10 Korea ................................................................................................................... 864.1.11 Norway ................................................................................................................... 864.1.12 Spain ................................................................................................................... 874.1.13 Sweden ................................................................................................................... 874.1.14 Switzerland ............................................................................................................. 884.1.15 United Kingdom ..................................................................................................... 894.1.16 United States ........................................................................................................... 90

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4.2. INTERNATIONAL PROGRAMS ......................................................................................... 904.2.1. OECD/NEA ............................................................................................................... 904.2.2. International Atomic Energy Agency ........................................................................ 904.2.3. European Commission ............................................................................................... 92

5. CONCLUSIONS AND RECOMMENDATIONS ........................................................................... 93

5.1. CONCLUSIONS ................................................................................................................... 935.2. RECOMMENDATIONS ........................................................................................................ 95

6. REFERENCES ................................................................................................................... 97

APPENDIX: SURVEY QUESTIONNAIRE .......................................................................................... 101

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EXECUTIVE SUMMARY

At the December 1999 meeting of the Nuclear Energy Agency (NEA) Committee on the Safety ofNuclear Installations (CSNI), a proposal by the Subgroup on Concrete Structures Aging was accepted tobroaden the Subgroup’s scope to include the long-term behavior of concrete used to construct fuel cyclefacilities, fuel storage facilities, and other nuclear structures required to perform a safety function for avery long period of time (e.g., decommissioned facilities). Consequently, at the fifth meeting of thesubgroup it was proposed that a Task Group be set up with experts in concrete structures and materials, andfuel cycle and fuel storage facilities. The principal objectives of the Task Group would be to establish thelevel of national interest in the topic; provide a forum for presentations of relevant national programs andlong-term strategies for fuel cycle, fuel storage, and other nuclear safety-related structures; and, whereappropriate, identify and prioritize specific tasks that warrant additional attention by the ConcreteSubgroup.

The first meeting of the Task Group was held April 2, 2001, at NEA in Issy-les-Moulineaux,France. Primary topics discussed at the Task Group meeting included: (1) status reports by each of thecountries in attendance; (2) summary of responses to a questionnaire that had been prepared and sent toMember Countries prior to the Task Group meeting; (3) development of background information listingpertinent fuel cycle structures, estimated lifetime requirements, and potential degradation mechanisms; (4)monitoring requirements; and (5) draft report makeup. Results obtained from the questionnaire, input fromtask group members, and information presented in the open literature were used to prepare this report.

Contained in the first chapter is background information on formation of the task group, themandate of the task group, and organization of the task group.

The second chapter addresses generation, characterization, and management of radioactive waste.Various forms of radioactive waste are identified. Primary sources of radioactive waste include nuclearreactors; medicine, industry, and research; and decommissioned facilities. Categories of radioactive wasteare identified (e.g., residues from processing uranium, materials and equipment contaminated duringoperation of nuclear facilities, nuclear fuel, and dismantling of nuclear reactors). Classification systemsdeveloped by organizations such as the International Atomic Energy Agency and the European Union arelisted. Basic techniques for management of exempt, low- and intermediate-level, and high-level waste aresummarized.

Fuel cycle facilities and long-term performance considerations are described in the third chapter.Primary materials used in the construction of reinforced concrete fuel cycle-related facilities are listed(e.g., concrete, steel reinforcement, and liners or coatings). Mechanisms that can produce prematuredeterioration of the reinforced concrete structures are described (e.g., chemical attack and corrosion of steelreinforcement). Service capabilities (operational and performance phases) and design considerations areprovided. Typical fuel cycle facilities for the four basic categories of radioactive waste associated with thenuclear fuel cycle are described and several examples of applications of reinforced concrete are presented.Performance assessment, monitoring, and instrumentation considerations for use in demonstratingsatisfactory performance are identified. The role of modeling and service life estimations in safetyassessments is noted.

The fourth chapter summarizes approaches of several OECD Member Countries to storage anddisposal of radioactive waste associated with the nuclear power plant fuel cycle. Pertinent activitiessponsored by the OECD/NEA, International Atomic Energy Agency, and the European Commission arelisted.

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General conclusions and recommendations are provided in the fifth chapter. Although concreteplays a role in the entire fuel cycle process, the reinforced concrete structures generally find their greatestapplication to facilities for storage and disposal of low- and intermediate-level waste that involvesfunctions of containment and shielding. Primary facilities include above-grade vaults, below-grade vaults,and modular canisters.∗

With respect to the use of reinforced concrete structures for management of high-level waste, such as wasteresulting from reprocessing operations, the primary applications have been to storage facilities thatmaintain the waste while radiation and heat decay, and until facilities are available for disposal. Thesefacilities may range from pads to spent fuel pools to dry storage casks to tanks.

With respect to durability, the concrete structures have increased requirements relative to conventionalcivil engineering applications in that the deteriorating influences are potentially more severe and therequired service lives may be up to several hundred years. Primary degradation factors include corrosion ofembedded steel, leaching, elevated temperature, and irradiation. Over the long term, leaching and crackinghave added importance as water will provide the transport medium for radionuclides should the otherengineered barriers fail. Although information on the performance of reinforced concrete structures in fuelcycle facilities is limited, it appears as if the overall performance of these structures has been fairly good.Some of the forms of degradation that have been observed include corrosion of reinforcing materials,cracking, and degradation of encast rubber moisture barriers. The operational period of many of thesestructures, however, has been relatively short, especially considering the time frame for which many ofthese structures are expected to function.

Due to the longevity requirements and importance of the fuel cycle facilities, condition assessment andperformance monitoring are considered to be of prime importance. Methods for use in conduct ofcondition assessments of general civil engineering reinforced concrete structures are fairly well establishedand generally start with a visual examination of exposed surfaces. Unfortunately, access to many of theradioactive waste storage or disposal facilities in all likelihood will be limited, and in some casesimpossible. Monitoring and instrumentation thus will play a vital role in developing the required data foruse in performance assessments as well as providing data for development and refinement of models forestimating performance.

Analysis of the role of the concrete barriers in low-level waste isolation requires that performanceassessment models be applied to concrete degradation. At the present state-of-the-art, models of thedegradation process tend to be somewhat empirical and the primary function of the models for estimatingservice life would be for comparative purposes (e.g., alternative design approaches).

The approach to management of radioactive wastes, and the spent fuel from nuclear power plants inparticular, varies from country to country. For some countries, engaged in once-through cycles using thedirect disposal option, spent fuel will be packaged and disposed in underground sites after a sufficientcooling period in surface-based facilities. For these countries, spent fuel is labeled as high-level waste. Onthe other hand, for countries that have embarked on the recycling strategy, valuable materials contained inthe spent fuel (e.g., uranium and plutonium) are separated and reused in conventional nuclear fuels, whilethe ultimate residues are properly treated and conditioned before being stored and eventually disposed.

∗ In waste management science, above-ground, on-ground, or below-ground repositories, also may be used

to represent near-surface facilities, rather than above-grade and below-grade.

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At the international level, activities have primarily been conducted under the auspices of organizationssuch as OECD/NEA, International Atomic Energy Agency, and European Commission. These activitieshave addressed topics such as safety aspects of waste disposal, field experiments in underground researchfacilities, and research on the basic phenomena. Some activities have addressed decommissioning ofnuclear-related facilities and disposal of the resulting waste forms.

Several areas have been identified where additional information is required: condition assessment, servicelife models, codes and standards, instrumentation and monitoring, decommissioning, degradationmechanisms, and repair techniques. It is recommended that the desired information be provided throughinternational conferences or workshops.

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1. INTRODUCTION

1.1 BACKGROUND

The Organization for Economic Cooperation and Development (OECD) is an intergovernmentalorganization of 30 countries primarily concerned with economics and trade. Membership in the NuclearEnergy Agency (NEA-28 member countries) of OECD is restricted to Western Europe, the United Statesof America, Canada, Mexico, Japan, Korea, and Australia. The NEA’s main objective is to promotecooperation between the governments of its participating countries in furthering the development ofnuclear power as a safe, environmentally acceptable, and economic energy source. The major policies andprograms of NEA are guided by the Steering Committee for Nuclear Energy, composed of representativesfrom each of the member countries, and under the authority of OECD Council. The Steering Committee isassisted in specialized areas by Standing Committees composed of experts provided by member countries.The present activity falls under the Committee on the Safety of Nuclear Installations (CSNI), which is oneof the six committees under the Steering Committee for Nuclear Energy.

Under CSNI there are four Working Groups, dealing with the following areas of interest:

• Risk assessment (RISK),• Analysis and Management of accidents (GAMA),• Integrity and aging of components and structures (IAGE), and• Operating experience (WGOE).

In addition, the CSNI has set up two Special Experts Groups:

• Human and Organizational Factors (SEGHOF), and• Fuel Safety Margins (SEGFSM).

Activities addressed in this document come under IAGE WG.

The Working Group on Integrity and Aging of Structures and Components has a general mandateto consider the logical basis for maintenance of the integrity of components, systems, and structures and topropose general principles for optimal ways of dealing with challenges to integrity, in particular fromaging. Specifically the mandate is as follows:

• to exchange views on generic technical aspects of integrity and aging of components and structures, and follow and take account of, as necessary, national and international programs concentrating on research, operational aspects, and regulation;

• in the relevant technical areas, stimulate the establishment of new required research and recommend possible international cooperative projects;

• to develop common technical positions on specific integrity issues and to identify areas where further work is needed; and

• to discuss the potential impact of aging and other challenges to integrity on the safety, regulation, and operability of nuclear power plants.

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IAGE WG is assisted in these tasks by a Subgroup on Integrity of Metal Components and Structures, aSubgroup on Concrete Structures Aging, and a Subgroup on Seismic Behavior of Structures.

At the December 1999 meeting of the CSNI a proposal by the Subgroup on Concrete StructuresAging was accepted to broaden the subgroup’s scope to include the long-term behavior of concrete used toconstruct fuel cycle facilities, fuel storage facilities, and other nuclear structures required to perform asafety function for a very long period of time (e.g., decommissioned facilities). Consequently, at the fifthmeeting of the Subgroup it was proposed that a Task Group be set up with experts in concrete structuresand materials, and fuel cycle and fuel storage facilities. The principal objective of the Task Group is toproduce a report on aging of concrete structures used to construct nuclear power plant fuel cycle facilities.

1.2 MANDATE OF THE TASK GROUP

The mandate of this activity is to establish the level of national interest in the topic, provide aforum for presentations of relevant national programs and long-term strategies for fuel cycle, fuel storageand other nuclear safety-related concrete structures; and, where appropriate, identify and prioritize specifictasks that warrant additional attention by the Subgroup on Concrete Structures Aging. Candidate activitiesthat could result include specialist's meetings, workshops, preparation of state-of-the-art reports, andconduct of international standards problems.

1.3 ORGANIZATION OF THE TASK GROUP

The Task Group held its first meeting on April 2, 2001, at NEA in Issy-les-Moulineaux, France.The Task Group was composed of the following members:

Dr. L. M. Smith, Chairman (UK)Mr. E. Mathet, Secretary (OECD/NEA)Mrs. B. Aghili (Sweden)Mr. F. Bernier (Belgium)Mr. C. J. Bolton (UK)Mr. B. Hedberg (Sweden)Mr. O. Jovall (Sweden)Dr. I. Martinez (Spain)Mr. I. McNair (UK)Dr. D. J. Naus (US)Mr. S. Nakamura (Japan)Mr. A. Taglioni (Italy)Dr. A. Vokal (Czech Republic)

Primary topics discussed at the Task Group meeting included: (1) status reports by each of thecountries in attendance; (2) summary of questionnaire responses; (3) development of backgroundinformation listing pertinent fuel cycle structures, estimated lifetime requirements, and potentialdegradation mechanisms; (4) monitoring requirements; and (5) draft report makeup.

Results obtained from a questionnaire (Appendix A) that had been sent to Member Countries,input from Task Group members, and information presented in the open literature were used to prepare thebalance of this report.

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2. GENERATION, CHARACTERIZATION, ANDMANAGEMENT OF RADIOATIVE WASTE

2.1 INTRODUCTION

Nuclear (or radioactive) waste is a byproduct from nuclear reactors, fuel reprocessing plants, andinstitutions such as hospitals and research facilities. It also results from reactors being decommissioned andother nuclear facilities that are permanently shut down. Radioactive waste, as a source of ionizingradiation, has long been recognized as a potential hazard to human health.

Radioactive waste occurs in a variety of forms with very different physical and chemicalcharacteristics, such as the concentrations and half-lives of the radionuclides. The waste may occur:

• in gaseous form, such as ventilation exhausts from facilities handling radioactive materials;• in liquid form, ranging from scintillation liquids from research facilities to high level waste

from the reprocessing of spent fuel; or• in solid form, ranging from contaminated trash and glassware from hospitals, medical

research facilities and radiopharmaceutical laboratories to vitrified reprocessing waste or spent fuel from nuclear power plants when it is considered waste.

Such waste may range from slightly radioactive, such as those generated in medical diagnostic procedures,to the highly radioactive, such as those in vitrified reprocessing waste or in spent radiation sources used inradiography, radiotherapy or other applications. Radioactive waste may be limited in volume, such as aspent sealed radiation source, or occupy large volumes such as tailings from mining and milling of uraniumores and waste from environmental restoration. Radioactive waste may also contain chemically orbiologically hazardous substances and it is important that hazards associated with these substances areadequately considered in radioactive waste management.

2.2 SOURCES OF RADIOACTIVE WASTE

2.2.1 Nuclear Reactors

Every part of the nuclear fuel cycle (Figure 2.1) produces some radioactive waste [e.g., operationof a 1000 MW(e) nuclear power plant at a 80% load factor for 40 years can produce 27 tonnes of spent fuelcontaining 240 kg of plutonium, 23 tonnes of uranium (0.8% U-235), 720 kg of fission products and

transuranics].1 The basic fuel of a nuclear power reactor contains U-235, which is in ceramic pellets inside

metal rods. Before the fuel rods are used, they are only slightly radioactive and may be handled withoutspecial shielding. During the nuclear reaction, the fuel “fissions” releasing two or three neutrons and asmall amount of heat. The released neutrons strike other atoms causing them to split and a chain reactionis formed releasing large amounts of heat. The splitting of relatively heavy uranium atoms during reactoroperation creates radioactive isotopes of several lighter elements, such as Cesium-137 and Strontium-90(fission products) that account for most of the heat and penetrating radiation in high-level waste.Plutonium is also formed due to uranium atoms capture of neutrons from nearby fissioning uranium atoms.

With time, the concentration of fission fragments and heavy elements formed in the same way asplutonium in a fuel bundle will increase to the point where it is no longer practical to continue to use thefuel (e.g., after about 12-24 months). Uranium fuel that has been used in a nuclear power reactor and is nolonger efficient in generating power to produce electricity (i.e., spent fuel) is considered to be high-level

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radioactive waste.∗ When removed from the reactor, a fuel bundle will be emitting both radiation,principally from fission fragments, and heat, thus requiring remote handling and shielding. The fuelbundles are stored in the reactor cooling pool to remove decay heat and provide shielding from radiation.Short to medium-lived low and medium level operational wastes also are produced. The most radioactiveoperational wastes are filter and ion exchange resins from the reactor water purification systems. Usedtools, exchanged parts, protective clothing, and scrap are also operational waste. The radioactivity of theoperational waste is low level mainly due to relatively short-lived radionuclides such as Cobalt-60 andCesium-137, which decay to safe levels after 200-300 years. Small amounts of long-lived waste may beincluded with the low-level waste.

The radioactive isotopes will eventually decay, or disintegrate, to harmless materials, however,while decaying they emit radiation. Some isotopes decay in hours or even minutes, but others decay veryslowly. Strontium-90 and Cesium-137 have half-lives of about 30 years, but less abundant Technecium-99and Iodine-129 have half-lives of 2.1 x 105 and 1.57 x 107 years, respectively. The heavier-than-uraniumtransuranic elements do not produce nearly the amount of heat or penetrating radiation that fission productsdo, but they take much longer to decay. A large portion of radioactive waste produced from the nuclearfuel cycle has radiation levels similar to, or not much higher than, the natural background level, and isrelatively easy to deal with. Only a small proportion is highly radioactive and requires isolation frompeople. Transuranic wastes account for most of the radioactive hazard remaining in high-level waste aftera thousand years (e.g., Plutonium-239 has a half-life of 24,000 years).

As of 1997, in the United States alone there was approximately 36,600 tonnes of spent nuclear fuelbeing stored at commercial nuclear power reactors, and by the year 2005 this quantity is expected to

increase to 52,000 tonnes.2 In the European Union, about 50,000 cubic meters of radioactive waste is

produced each year with 2,444 cubic meters of high-level radioactive waste estimated to be generated

during the period 2000 to 2005.3 Worldwide nuclear power generation facilities each year produce about

10,000 cubic meters of high-level waste and 200,000 cubic meters of low- and intermediate-level waste.4

2.2.2 Medicine, Industry, And Research

The use of radioisotopes in research centers, universities, industrial companies and in medicineresults in a comparatively high volume of compressible or combustible substances such as plastic, paperand protective clothing, small metallic objects or crushed glassware, together with worn out or damagedequipment, metal components, air filters, debris, and miscellaneous waste. Accordingly this waste ischaracterized by a very heterogeneous composition, including a broad spectrum of materials, contaminants,and specific activities. Additional heterogeneous radioactive wastes are produced from sources that do notdirectly involve isotopes (e.g., particle accelerators). Although there are a large number of establishmentsgenerating this waste, it is not a relatively large proportion of the total radioactive waste. Importantradionuclides used for various purposes ranging from medical applications to food irradiation includeTritium, Cobalt-60, Carbon-14, Americium-241, Radium-226, Cesium-137, and Iodine-125. Thesesources, after a substantial decrease in radioactivity, must be stored and disposed of as radioactive waste.This type of radioactive waste is still generating heat and emitting high-energy radiation, therefore it mustbe included in special shielding containers made of steel, lead, or concrete.

∗ The production of nuclear weapons has produced a legacy of high-level radioactive liquid and solid waste

that was created when spent nuclear fuel was treated chemically to separate uranium and plutonium. Thiswaste is expected to be included in any high-level waste disposal plans. Classification of waste isdescribed in the Section 2.3.

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2.2.3 Decommissioned Facilities

The operating lifetime of a nuclear facility is largely determined by economic and safetyconsiderations. Nuclear power plants are normally designed for an operating lifetime on the order of 30 to40 years. Through an appropriate aging management program, operations at these plants can probablycontinue for several decades. However, ultimately it will become technically or economicallyadvantageous to retire a facility from operation and, if necessary, replace it with a new plant. Althoughdecommissioning normally occurs at the end of the operating lifetime of the plant, it may also be requiredfor other reasons (e.g., accident or political reasons). The kind of reactor, the location of the facility, andthe total amount of radioactivity it contains are important elements in the selection of a decommissioningstrategy.

Generally when operations at a nuclear plant are terminated, the nuclear fuel, mobile radioactivematerials in the process systems, and the radioactive waste produced during normal operations are removedas soon as the plant ceases to operate. Certain equipment can also be removed and discarded. Theradioactivity produced in nuclear power plants is made up of both short- and long-lived radionuclides, butprincipally short-lived isotopes that would decay in 5 to 30 years. In order to take advantage of this andexpose workers to lower radiation levels, a significant reduction in radioactivity can be acchieved byplacing the facility in safe storage for that length of time. Thus, decommissioning is generallyaccomplished in stages. Each stage is defined by two characteristics: the physical state of the plant and itsequipment, and the surveillance needed to maintain the physical state. OECD considers decommissioning

in three stages:5

• Removing spent fuel from the reactor, draining the liquid systems, disconnectingoperating systems, blocking and sealing mechanical openings such as valves andplugs, and controlling the atmosphere inside the containment building. The facilityis kept under surveillance, access is limited, and routine inspections are carried outto assure that the plant remains in a safe condition.

• All equipment and buildings that can be easily dismantled are removed or decontaminatedand made available for other uses, leaving only the reactor core structure and its extensiveshielding. The containment building and the ventilation system may be modified orremoved if they are no longer needed for safety reasons, or they may be decontaminated toallow access for other purposes. Other buildings and equipment that are not radioactivemay be converted for new purposes as well. Surveillance during this stage is reduced, butit is desirable to continue spot checks of the building as well as surveillance of thesurrounding environment.

• Unless the site, buildings, or equipment are to be reused for other nuclear purposes, allmaterials with radioactivity levels exceeding those closely equivalent to the naturalradiation environment will be removed and the site released without restrictions or furthersurveillance.

The three stages may be carried out rapidly, progressing from one stage to the next, or carried out over aprolonged period of time lasting as long as 100 years or more. Although most countries intend to completeall stages, a facility could remain in either of the first two stages for a relatively long period of time, ordecommissioning could proceed directly from the first to third stage. Although it depends on the physicalstate of the power plant, as well as available resources and equipment, it is known that a deferral of thethird stage for 80-100 years would significantly reduce personnel management and protection costs, butraise maintenance and surveillance costs. Figure 2.2 presents an example of a plant layout before and after

decommissioning activities. Additional information on decommissioning of nuclear facilities is available.6

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2.3 CLASSIFICATION OF RADIOACTIVE WASTE

Most nations categorize wastes into classes in order to simplify waste-management actions, rules,and regulations while protecting public health. The main categories of waste from the nuclear fuel cycleare:

• Residues left from processing uranium ore that contain naturally occurring radioactiveelements mined with the uranium, and some chemicals used in the separation process. Theradioactivity is low-level but long-lived.

• Materials and equipment (e.g., protective clothing, cleaning materials, ion-exchangeresins) that become contaminated during operation of the nuclear facilities. Theradioactivity is mainly low-level and short-lived.

• Wastes arising from the nuclear fuel after it has been used in the reactor. This waste canbe used as fuel itself if it is not reprocessed, or the wastes resulting from reprocessing theused fuel so that it can be recycled for reuse. Used fuel that is not to be reprocessed can beregarded as high-level, long-lived waste. Reprocessing wastes are a mixture of high-,intermediate-, and low-level wastes, including long-lived intermediate-level wastes; themixture depends on the treatment technique used.

• Wastes resulting from dismantling nuclear reactors after the fuel has been removed andfrom fuel reprocessing plants at the end their operating lives. The radioactivity is low andintermediate level and mainly relatively short-lived. There will be some long-lived wastefrom the dismantling process.

The high-level wastes of the third category account for almost all (~99%) of the radioactivity produced bynuclear electricity generation, but a very small proportion of the total volume of waste, which by itself isvery small in comparison to wastes from other forms of thermal electricity generation. Since it is such asmall volume, it can be effectively and economically isolated. The long-lived, low-level wastes fromprocessing uranium ore account for most of the volume of all radioactive waste (i.e., 50 to 100 times asmuch as the rest), but very little of the radioactivity.

International bodies, national authorities and waste operators have established differentclassification systems, each meeting their respective criteria or area of responsibility. The systems groupwastes with similar characteristics and hazards within the same category in order to facilitate managementand improve safety. All the classification systems vary widely in approach and application. For example,to facilitate communication and information exchange among its Member States, the International AtomicEnergy Agency (IAEA) instituted a revised waste classification system in 1994 that takes into account bothqualitative and quantitative criteria, including activity levels and heat content. Table 2.1 presents theIAEA classification for its three principal waste classes (exempt, low and intermediate, and high level).The European Union also has prepared a classification system for solid radioactive waste to simplifycooperation among its member countries and to provide comparative information about waste holdings ineach country. Table 2.2 presents this classification system. Similarly, there are national classificationsystems. Table 2.3 presents definitions of radioactive waste classes according to U.S. statutes andregulations. Finally, classification of radioactive waste in terms of disposal route also can be done. Oneclassification of radioactive waste based on disposal route includes: (1) waste suitable for near-surfacedisposal (low- or intermediate-level short-lived waste only), and (2) waste not suitable for surface or near-surface disposal (deep geological disposal is best suited).

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2.4 MANAGEMENT OF RADIOACTIVE WASTES

Radioactive waste management incorporates all administrtive, operational, and safety-relatedactivities that are involved in the treatment, conditioning, storage, and disposal of all categories ofradioactive wastes, including transportation. While radioactive waste management methods vary fromcountry to country, the primary objective is to deal with radioactive waste in a manner that protects publichealth and the environment now and in the future without imposing undue burdens on future generations.Table 2.4 presents the basic principles that radioactive waste management needs to address to achieve itsobjective.

The primary characteristics of radioactive waste that affect what effects it could have and how itshould be managed are:

• how long the radioactivity lasts (content of short- or long-lived radionuclides),• concentration of radionuclides (high-, medium-, or low-level), and• whether it is heat generating or not (related to concentration of the radionuclides).

The length of life of the radioactivity determines how long it has to be managed. The concentration of theradionuclides and whether it is heat generating dictates how it should be handled (how much, if any,shielding is needed). These considerations determine what ways of disposal are suitable.

Effective management of radioactive waste considers the basic steps shown in Figure 2.3. Initiallythe waste should be characterized in order to determine its physical, chemical and radiological properties,and to facilitate record keeping and acceptance of radioactive waste from one step to another. Storage ofradioactive waste involves maintaining the radioactive waste such that: (1) isolation, environmentalprotection and monitoring are provided and (2) actions involving treatment, conditioning, and disposal arefacilitated. Pretreatment of waste is the initial step in waste management that occurs after waste generationand is used to separate waste streams. Treatment of radioactive waste includes those operations intendedto improve safety and economy by changing characteristics of the radioactive waste (e.g., volumereduction, radionuclide removal, and change of composition). Conditioning of waste involves thoseoperations that transform radioactive waste into a form suitable for handling, transportation, storage, anddisposal (e.g., immobilization through solidification of low- and intermediate-level liquid radioactive wastein cement and packaging into drums or engineered thick-walled reinforced concrete containers). Disposalis the final step in the radioactive waste management system and consists mainly of the emplacement ofradioactive waste in a disposal facility with reasonable assurance for safety without the intention ofretrieval and without the reliance on long-term surveillance and maintenance.

2.4.1 Exempt Waste

Exempt waste contains such low concentrations of radionuclides that that it can be excluded fromnuclear regulatory control because regulatory hazards are considered negligible.

2.4.2 Low- and Intermediate-Level Waste

Low- and intermediate-level waste contains enough radioactive material that it requires actions toensure the protection of workers and public for short or extended periods of time. These wastes include arange of materials from just above exempt levels to those with sufficiently high levels of radioactivity torequire shielding containers, and in some cases periods of cooling off. These wastes may be subdividedinto categories according to half-lives of the radionuclides it contains (e.g., “short-lived” could be less than30 years and “long-lived” greater than 30 years).

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Short-lived wastes are often treated to achieve volume reduction and/or conditioned (e.g.,immobilization in cement or bitumen) prior to disposal. Disposal methods for treated and conditionedwastes are typically in shallow concrete-lined trenches or engineered surface structures. The isolationperiod is usually up to 300 years, thus facilitating institutional and administrative control of the disposalsite. Main options that have been employed for the short-lived low- and intermediate-level wastes includenear-surface disposal facilities and geologic repositories. Figure 2.4 presents an example of a near-surfacedisposal facility.

Some low - and intermediate-level waste contain long-lived radionuclides in quantities that requirea high degree of isolation from the biosphere. This is typically provided by disposal in geologicformations at a depth of several hundred meters. Such wastes remain in storage pending ultimate disposal.Figure 2.5 presents a schematic of a repository for low- and intermediate-level waste. To ensure that nosignificant environmental releases occur over the period that such waste remains hazardous, a multiplebarrier concept such as also shown in Figure 2.5 is used. The waste is immobilized in an appropriatematrix (e.g., cement or bitumen) and sealed inside a corrosion-resistant canister. The canister itself isgenerally surrounded by an almost impermeable backfill (e.g., plastic clay) and placed into a repository inlow-permeability rock at depths of hundreds of meters.

2.4.3 High-Level Waste

In countries where spent nuclear fuel from reactor operations is chemically reprocessed, theradioactive wastes include highly concentrated liquid solutions of nuclear fission products that are latersolidified, primarily through vitrification. Steel containers holding the vitrified waste will typically bestored 30 to 50 years in air-cooled vaults for the decay heat to subside. Both the liquid solids and thevitrified solids are considered high-level waste. If the spent fuel is not reprocessed, it also is considered ashigh-level waste to be disposed of appropriately. Since the high-level waste generates such intense levelsof both radioactivity and heat, heavy shielding is required during its handling and temporary storage. Thewastes are generally stored in structurally engineered cooling pools or vaults for several decades prior todisposal. The fuel pools often are seismically designed reinforced concrete structures that are lined withsteel plate. The spent fuel is cooled while in the spent fuel storage pool by water that is force-circulatedusing pumps. While stored, both the temperature and radioactivity of the waste decreases, considerablysimplifying its future handling and disposal. The spent fuel remains in the fuel pool until it can betransferred on site to a dry cask storage location or transported off site to a high-level radioactive disposalsite. Canisters used for storage of spent fuel are fabricated from steel and hold about two dozen fuelassemblies. After loading the fuel assemblies into the canisters, water and air are removed, the canisterfilled with inert gas, welded shut, and rigorously tested for leak tightness. The canister is then placed intoa storage cask that can be fabricated from either metal or reinforced concrete. Figure 2.6 presents a spentfuel storage pool and Figure 2.7 presents dry storage of spent fuel.

Storage cannot be relied upon in the long-term to provide the necessary permanent isolation of thewastes from the environment, and future generations should not bear the burden of managing wastesproduced today. This has lead to the need for the nuclear industry to demonstrate the feasibility and safetyof high-level waste disposal, and some countries have enacted laws that require operational high-levelwaste disposal capability. Among the potential options for disposing high-level waste (e.g., disposal ingeological formations under the ocean floor and disposal in glaciated areas), deep geological disposal onland is the most appropriate means for isolating such wastes permanently from the environment. Reasons

for this include:7

• It is an entirely passive disposal system with no requirement for continuing human involvement to ensure safety;

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• Radioactive wastes present no hazard while they remain in a deep underground repository. Because of their depth of burial (several hundred meters or more), the possibility of intentional human intrusion is virtually eliminated, and with a suitable choice of location, the likelihood of inadvertent human intrusion can be made minimal;

• Flexibility and convenience are provided by a large variety of geological environments suitable for disposal;

• The disposal option is demonstrably practical and feasible with currently existing technology used in other mining and civil engineering practices; and

• Although waste disposal implies the lack of intention to retrieve the waste, the repository can be designed so that the waste can be recovered, while the repository is in operation or even after closure.

The concept of removing long-lived waste from the human environment by placing it into deepunderground repositories – geological disposal – that would keep it remote from humans and resistant to

malicious or accidental disturbance was proposed over 40 years ago.8 Since then there have been

significant developments, with the details varying from country to country and according to the type ofwaste. In general, the geologic disposal concept involves treating the waste in order to achieve a suitablephysical and chemical form, packaging it inside long-lived engineered barriers emplaced deepunderground, and sealing these facilities with appropriate materials. In these underground surroundings, asopposed to a surface environment, conditions remain stable over the long periods of time needed to allowthe radioactivity to decay to a sufficiently low level. In recent years, the concept is nearing implementationin several countries and has already taken place to a limited extent in Germany, Sweden, Finland, Norway,and the United States (long-lived radioactive components). All concepts are largely based on disposal inrepositories approximately 200 to 1000 meters below the surface. The host rock for the repositories varies(e.g., volcanic tuff, salt, and granite). Figure 2.8 presents one concept for disposal of high-level radioactivewaste.

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Table 2.1IAEA Classification System for Solid Waste

Exempt Waste (ew) Contains such low concentration of radionuclides that it can beexcluded from nuclear regulatory control because radiologicalhazards are considered negligible.

Low And Intermediate LevelWaste (LILW)

Contains enough radioactive material that it requires actions toensure the protection of workers and the public for short orextended periods of time. This class includes a range ofmaterials from just above exempt levels to those withsufficiently high levels of radioactivity to require use ofshielding containers and in some cases cooling off periods.LILW may be subdivided into categories according to the half-lives of the radionuclides it contains, with “short lived” beingless than 30 years and “long lived” being greater than 30 years.

High Level Waste (HLW) Contains sufficiently high levels of radioactive materials that ahigh degree of isolation from the biosphere, normally ingeologic repository, is required for long periods of time.Such wastes normally require both special shielding and coolingoff periods.

Source: International Atomic Energy Agency, Safety Series No. 111-G-1.1, Vienna, Austria, 1994.

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Table 2.2European Union Classification System for Solid Waste

1.Transition Radioactive The radioactive substances in this waste will decay within theperiod of temporary storage (up to 5 years) and the waste may thenbe suitable for management outside of the regulatory control systemfor radioactive wastes.

This waste mainly originates from medical applications

Waste that remains radioactive for longer than this period of 5 yearsis regarded as low and intermediate-level waste.

2. Low- and Intermediate-Level Waste (LILW):

The concentration of radionuclides in this waste is such that thegeneration of thermal power during its disposal is sufficiently low.Acceptable thermal power values are site-specific following safetyassessments.

2.1 Short-Lived Waste(LILW-SL)

This waste includes radioactive waste containing radionuclides thathave a half-life less than or equal to 30 years with a restricted alphalong-lived radionuclide concentration (up to 4000 Bq/g inindividual waste packages and up to an average of 400 Bq/g in totalwaste volume).

2.2 Long-Lived Waste(LILW-LL)

The long-lived radionuclide concentration in this waste exceeds thealpha limitations for short-lived waste.

3. High-Level Waste (HLW) The concentration of radionuclides in this waste is such that thegeneration of thermal power shall be considered during its storageand disposal. The thermal power generation level is site-specific.

This waste arises mainly from treatment or conditioning of spentnuclear fuel.

This classification system is based on the categories proposed by the IAEA and should be used inparallel with existing national system.

Source: European Union Internet Site; http://www.rwm-eu.org/en/n3/2_2.asp?index=2

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Table 2.3Definition of Waste Classes According to U.S. Statutes and Regulations

High-Level Waste(HLW):

(A) the highly radioactive material resulting from the reprocessing of spent nuclear fuel,including liquid waste produced directly in reprocessing and any solid material derivedfrom such liquid waste that contains fission products in sufficient concentrations; and(B) other highly radioactive material that the U.S. Nuclear Regulatory Commission(NRC), consistent with existing law, determines by rule requires permanent isolation.

The NRC has determined that irradiated reactor fuel shall, for the purposes of therepository, be considered HLW.

Spent Nuclear Fuel(SNF): Fuel that has been withdrawn from a nuclear reactor following irradiation, the SNF

constituent elements which have not been separated by reprocessing. Under Title 10U.S. Code of Federal Regulations Part 60, spent nuclear fuel is regulated as HLW.

Transuranic Waste(TRU):

This class is specfic to waste streams from U.S. Department of Energy (DOE) andcomprises material contaminated with elements that have an atomic number greater than92, including neptunium, plutonium, americium, and curium, and that are inconcentrations greater than 10 nanocuries per gram, or in such other concentrations asthe NRC may prescribe to protect the public health and safety. This definition wasrevised in 1984 by DOE Order 5820.2 to be "Without regard to source or form, wastethat is contaminated with alpha-emitting transuranium radionuclides with half-livesgreater than 20 years and concentrations greater than 100nCi/g at the time of assay.

By-Product Material: (A) any radioactive material (except special nuclear material) yielded in or maderadioactive by exposure to the radiation incident to the process of producing or utilizingspecial nuclear material, and (B) the tailings or wastes produced by the extraction orconcentration of uranium or thorium from any ore processed primarily for its sourcematerial content.

The tailings or wastes produced by the extraction or concentration of uranium or thoriumfrom any ore processed primarily for its source material content. Also called by productmaterials under Title 42 U.S.Code § 2014 (e)(2).

Low-Level Waste(LLW):

Radioactive material that - (A) is not high-level radioactive waste, spent nuclear fuel, orbyproduct material (as defined in U.S.Code § 2014) and (B) the NRC, consistent withexisting law and in accordance with paragraph (A), classifies as low-level radioactivewaste. This does not exclude commercial waste containing TRU materials. In thegovernment sector, TRU waste is excluded.

LLW is divided into two broad categories: waste that qualifies for near-surface burial,and waste that requires deeper disposal (Greater than Class C LLW - requires isolationfrom the biosphere for 500 years, or greater confinement waste).

LLW that is regulated by the NRC and qualifies for near surface burial is separated intothe three classes (A, B, and C). DOE LLW is subclassified according to facility-specificlimitations.

Naturally-Occurringand Accelerator-

Produced RadioactiveMaterials

(NORM/NARM)

Naturally-occurring radioactive material and accelerator-produced radioactive materiallie outside NRC's regulatory authority and are subject to health and safety regulation bythe States and other Federal agencies. The waste is generally subclassified as diffuse(<2nCi/g 226Ra or equivalent) or discrete (>2nCi/g 226 Ra or equivalent). They are underreview by Environmental Protection Agency and Resource Conservation and RecoveryAct.

Source: M. D. Lowenthal, “Radioactive-Waste Classification in the United States: History and CurrentPredicaments,” UCRL-CR-128127, Center for Nuclear and Toxic Waste Management, University ofCalifornia, Berkeley, 1997.

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Table 2.4Basic Principles of Radioactive Waste Management

Protection of human health Radioactive waste shall be managed in such a way as to securean acceptable level of protection of human health

Protection of the environment Radioactive waste shall be managed in such a way as toprovide an acceptable level of protection to the environment

Protection beyond nationalborders

Radioactive waste shall be managed in such a way as to assurethat possible effects on human health and environment beyondnational borders will be taken into account

Protection of future generations Radioactive waste shall be managed in such a way thatpredicted impacts on health of future generations will not begreater than the relevant levels of impact that are acceptabletoday

Burdens on future generations Radioactive waste shall be managed in such a way that it willnot impose undue burdens on future generations

National legal framework Radioactive waste shall be managed within an appropriatenational legal framework including clear allocation ofresponsibilities and provision for independent regulatoryfunctions

Control of radioactive wastegeneration

Generation of radioactive waste shall be kept to the minimumpracticable

Radioactive waste generation andmanagement interdependencies

Interdependence among all steps in radioactive wastegeneration and management shall be appropriately taken intoaccount

Safety of facilities The safety of facilities for radioactive waste management shallbe appropriately assured during their lifetime

Source: “Radioactive Waste Management Profiles- a Compilation of Data from the WasteManagement Database,” No. 3, International Atomic Energy Agency, Vienna, Austria,April 2000.

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Q u ic k T im e ™ a n d aP h o to - J P E G d e c o m p re s s o r

a re n e e d e d to s e e th is p ic tu re .

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QuickT ime™ a nd aPhoto - J PEG decompresso r

a re ne eded t o see this pictur e .QuickTime™ and aPhoto - JPEG decompressor

are needed to see this picture.

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Figure 2.3 Basic steps in radioactive waste program.

(Source: “Radioactive Waste Management Profiles – A Compilation of Data from the Waste ManagementDatabase,” No. 3, International Atomic Energy Agency, Vienna, Austria, April 2000.)

Waste and Materials

Pretreatment

Treatment

Conditioning

Disposal

Radioactive material (for reuse/recycle)

Exempt waste and material

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Disposal concept Safety barrier system

Figure 2.5 Schematic of repository for low- and intermediate-level waste.

(Source: NAGRA Internet Site; http://www.nagra.ch/english/lager.htm)

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Cross-section of canister and dry Above-ground storage bunker

storage cask for spent fuel

Figure 2.7 Dry storage of spent fuel.

(Source: “Information Digest,” NUREG 1350, Vol. 12, U. S. Nuclear Regulatory Commission,Washington, DC, June 2000.

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Figure 2.8 Concept for disposal of high-level waste.(Source: “The Disposal of High-Level Radioactive Waste,” No. 3, Nuclear Energy Agency, Issy-les-Moulineaux, France, January 1989.

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3. FUEL-CYCLE FACILITIES AND LONG-TERMPERFORMANCE CONSIDERATIONS

The design and construction of many of the facilities and structures for processing and long-termstorage of radioactive waste materials generated by the nuclear power plant fuel cycle employ reinforcedconcrete.∗ The reinforced concrete is used for many purposes including support, containment, andenvironmental protection. The types of facilities that rely on concrete range from surface structures, toshallow subsurface vaults, to deep underground repositories. These structures are required to functionsafely and reliably in challenging, demanding, and varying environments. In modern designs, thereinforced concrete in the walls, slabs, and other structural members is not intended to be in direct contactwith the waste forms because the concrete has limited capability to act as a barrier to the release of wastedue its potential for cracking, the presence of construction joints, permeability, and potential loss ofdurability due to environmental effects (e.g., chemical attack). The wastes therefore are generallycontained in a type of primary confinement (e.g., tanks, stainless or carbon drums, or special purposecanisters that could be made from special concrete mixtures) prior to being placed into storage. In someolder facilities, built before this policy was developed, waste is in direct contact with the concrete. Typesof repositories or disposal facilities, depending on the character of the material to be stored, typicallyinclude: low- and intermediate-level radioactive waste disposal facilities, spent fuel storage facilities,transuranic waste disposal facilities, and high-level waste disposal facilities. Table 3.1 presents an exampleof typical reinforced concrete structures associated with the nuclear fuel cycle and their functions.

3.1 MATERIALS OF CONSTRUCTION

Fuel cycle reinforced concrete structures are composed of several constituents that, in concert,perform multiple functions (i.e., load-carrying capacity, radiation shielding, and leak tightness). Primarily,these constituents can include concrete, conventional steel reinforcement, prestressing steel, and steel ornon-metallic liner materials. Concretes used in the fabrication of these structures typically consist of amoderate heat of hydration cement, fine aggregates (sand), water, various admixtures for improvingproperties or performance of the concrete, and either normal-weight or heavy-weight coarse aggregate.Both the water and aggregate materials are normally obtained from local sources and subjected to materialcharacterization prior to use. Steel reinforcement (mild steel in older facilities, high-tensile strength steelin more recent plants) is used primarily to provide tensile and shear load resistance/transfer and consists ofplain carbon steel bar stock with deformations (lugs or protrusions) on the surface. The post-tensioningsystem consists of prestressing tendons that are installed, tensioned, and then anchored to the hardenedconcrete forming the structure. The prestressing tendons provide primary resistance to tensile loadings andalso improve resistance to shear forces such as could develop during earthquake loadings. Corrosionprotection is provided by filling the ducts with wax or corrosion-inhibiting grease (unbonded), or portland-cement grout (bonded). Metallic or nonmetallic liners are provided on the inside surfaces of certainstructures (e.g., fuel pools) to provide a barrier against the leakage of fluids. Stainless steel or stainless-steel clad carbon steel are typically used for metallic liners. Nonmetallic liners utilize organic materials.Ruberized coatings have been utilized for reinforced concrete pond facilities. Although the liner orcoatings primary function is to isolate the concrete from its contents or provide a leaktight barrier, it alsocan act as part of the formwork during concrete placement.

∗ Cement-based solidification/stabilization is not considered.

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3.2 POTENTIAL DEGRADATION MECHANISMS

A reinforced concrete structure’s durability characterizes its long-term performance andthroughout its usable life a structure must resist the deteriorating effects of its environment in order toachieve satisfactory performance. With respect to fuel cycle facilities, the desired usable life may be up to500 years for low-level waste facilities

9 to 10,000 years for high-level waste facilities.

10 The structural

integrity and long-term durability of engineered concrete barriers used to contain radioactive waste can beadversely affected by exposure to different environments or degradation mechanisms. For example,deterioration can occur when concrete barriers are exposed to leaching or radioactive wastes that containacids, sulfates, chlorides, or other aggressive chemicals. The longevity, or long-term performance ofreinforced concrete fuel cycle facilities is primarily a function of the durability of these structures towithstand the potential effects of degradation.

Aside from earthquakes, tornadoes, and other short-term events, concrete degradation mechanismsare considered to be continuous processes that can potentially result in long-term deterioration. In nearlyall chemical and physical processes influencing the durability of concrete structures, dominant factorsinvolved include transport mechanisms within the pores and cracks, and the presence of a fluid. Concretepermeability therefore is of significant importance relative to the long-term durability of radioactive wastefacilities. Concrete permeability will vary according to such things as the proportions of constituents,degree of cement hydration, cement fineness, aggregate gradation, and moisture content. One of the most

important factors affecting ionic transport through concrete is the presence of cracks.11 The presence of

cracks not only controls the quantity of ions transported, but can also control whether there will be anyconvective transport. Factors that contribute to increases in concrete permeability or cracking therefore areof importance to the durability of the radioactive waste management facilities.

3.2.1 Concrete Degradation Mechanisms

The durability of concrete materials used to construct radioactive waste disposal facilities can belimited as a result of either adverse performance of the cement-paste matrix or aggregate constituents undereither chemical or physical attack. In practice, these processes may occur concurrently to reinforce eachother.

3.2.1.1 Chemical Attack

Chemical attack involves the alteration of concrete through chemical reaction with either thecement paste, coarse aggregate, or embedded reinforcing steel. Generally, chemical attack begins onexposed concrete surfaces by degrading the cover concrete that is located between the concrete surface andthe embedded reinforcing steel. As the degradation process continues, more and more of the coverconcrete is destroyed or degraded until the reinforcing steel begins to corrode. Cracks in the concreteresulting from structural loads, shrinkage, abrasion, erosion, alkali-aggregate reactions, or other causes caninfluence the rate of deterioration. Finally, after prolonged exposure, the entire concrete cross section canbe severely affected resulting in loss of structural and leaktight integrity. The rate of chemical attack is afunction of the pH of the aggressive medium and is influenced by the concrete permeability, alkalinity, andreactivity. Pertinent forms of chemical attack are summarized in the following sections.

Alkali-Aggregate Reactions. The chemical reaction between the active silica constituents of theaggregate and the alkalis in the portland cement is known as alkali-silica reaction. This reaction starts withattack of the siliceous minerals in the aggregate by the alkalis (Na2O and K2O) in the cement and producesan alkali-silicate gel that forms at the surfaces of the aggregate particles. As the gel imbibes water, itincreases in volume and causes internal pressures within the concrete that can eventually lead to expansion,

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cracking, and weakening of the bond between the aggregate particles and the cement paste. Theconsequences of alkali-aggregate reaction are pop-outs and spalling of the concrete surface and mapcracking. The rate of the alkali-silica reaction is influenced by the size of the siliceous mineral particles.Fine particles can react quickly causing expansion of the concrete within weeks, while larger particles tendto react slower, often taking more than five years to cause significant expansion of the concrete. Theporosity of the aggregates, the quantity of alkalis in the cement, and the availability of water to sustain thereaction also influence the reaction rate. However, the reaction occurs most rapidly when the concretesurface is continuously exposed to water. Low-alkali portland cement, aggregates with satisfactory servicehistory, and mineral admixtures such as silica fume, blast-furnace slag, and fly ash are often used tominimize the possibility of alkali-aggregate reactions. When this is not possible, potential mitigatingtechniques that have been suggested include use of a low water-cementitious material ratio to produceconcrete with low permeability, and addition of barriers or coatings to prevent contact between theconcrete and external sources of moisture.

Another type of chemical reaction that can cause concrete deterioration is known as alkali-carbonate reaction. When dolomitic limestone aggregates that contain clay react with the alkalis inportland cement, an alkali-carbonate gel forms. This reaction breaks down the aggregates and allowsingress of water resulting in swelling and subsequent cracking of the concrete. One factor that influencesthe rate of the alkali-carbonate reactions is the amount of water that is available to sustain the reaction.The best way to avoid problems resulting from alkali-carbonate reactions is to use aggregates that are notreactive. Low-alkali portland cement can also be used to minimize the possibility of alkali-carbonatereactions. However, unlike the alkali-silica reactions, pozzolans such as fly ash are not effective inmitigating alkali-carbonate reactions.

Although alkali-aggregate reaction typically occurs within the first 10 years of construction, somestructures have not experienced deterioration until they were 15-years-old or older. The delay in structuresexhibiting deterioration indicates that there may be less reactive forms of silica that may eventually cause

deterioration and the effects of alkali-silicate reactions may not be apparent for many years.12

Thepotentially less reactive form of alkali-aggregate reaction takes on added importance for structures in thefuel cycle facilities because they may be required to have service lives up to several thousand years. Also,alkali-aggregate reactions that initiate in fresh concrete can cease when the concrete dries out and canreinitiate at a later time when the concrete becomes moist again.

Leaching of Concrete Constituents. Water containing chlorides, sulfates, and bicarbonates of

calcium and magnesium does not attack the constituents of portland cement paste.13

However, pure waterwhich contains little or no calcium ions, or acidic groundwater present in the form of dissolved carbondioxide gas, carbonic acid, or bicarbonate ion, tend to hydrolyze or dissolve the alkali oxides and calcium-containing products in the cement paste. For process reasons, much of the water contained in water-retaining structures for fuel-cycle facilities is deliberately demineralized. Hydrolysis of cement paste willstop once the water and cement paste attain chemical equilibrium, but, if the water is flowing or seepingthrough the concrete due to pressure or capillary effects, hydrolysis of the cement paste continues untilmost of the calcium hydroxide has been leached away. The concrete that remains is more porous, morepermeable, less strong, and more vulnerable to chemical attack. The rate of leaching is dependent on theamount of dissolved salts contained in the percolating fluid, rate of permeation of the fluid through thecement paste matrix, and temperature. The rate of leaching can be controlled by minimizing thepermeation of water through the concrete and its diffusion into the adjacent environment. This can beaccomplished by using low permeability concretes and barriers or coatings.

Sulfate Attack. Most soils contain harmless quantities of sulfate in the form of gypsum that

typically ranges from 0.01 to 0.05 percent expressed as SO4.13

Where higher concentrations of sulfates arepresent in the groundwater or soil, generally due to the presence of magnesium and alkali sulfates,

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significant deterioration can occur. The sulfates react with the Ca(OH)2 and the hydrated tricalciumaluminate (C3A) to form gypsum and ettringite, respectively. Magnesium sulfate has the most damagingeffect because it also causes decomposition of the hydrated calcium silicates, eventually forming hydrated

magnesium silicate which has no binding properties.14

For concrete that may be exposed to seawater, thedeleterious expansion resulting from sulfate attack is reduced due to the presence of chlorides which makesthe concrete more susceptible to leaching and associated increase in porosity. Also, the attack by seawateris slowed by the blocking of pores due to the formation and deposition of magnesium hydroxide or

brucite.14

In addition to loss of concrete strength, sulfate attack can also manifest itself in the form ofexpansion and associated cracking of the concrete. The main factor in the production of concrete that is tobe resistant to a sulfate environment is for the concrete to be dense and of low permeability. Since it is theC3A that is attacked by sulfates, the concrete vulnerability can also be reduced by using a sulfate-resistingportland cement which is low in C3A. Also, improved sulfate resistance can be obtained through the use of

admixtures such as pozzolans and blast-furnace slag.12

Under extreme conditions supersulfated slag ce-ments can be used.

Acids and Bases. Acids present in groundwater (e.g., sulfuric or carbonic) and certain other acids(e.g., boric and sulfuric) can combine with the calcium compounds in the hydrated cement paste (i.e.,calcium hydroxide, calcium silicate hydrate, and calcium aluminate hydrate) to form soluble materials thatare readily leached from the concrete to increase its porosity and permeability. The main factordetermining the extent of attack is not so much the aggressiveness of the attacking acid, but more thesolubility of the resulting calcium salt. The rate of deterioration is also accelerated if the aggressivechemical solution is flowing. Since under acid attack there is a conversion of the hardened cement, theconcrete permeability is not as important as for other types of chemical attack (e.g., leaching and sulfateattack).

As hydrated cement paste is an alkaline material, high quality concretes made with chemicallystable aggregates normally are resistant to bases. However, sodium and potassium hydroxides in highconcentrations (> 20%) can cause concrete to deteriorate. Some flocs used to treat waste liquors fromreprocessing contain ammonium nitrate that can have a detrimental effect on both concrete and reinforcingsteel. Under mild chemical attack, a dense concrete with low water-cement ratio may provide suitableresistance. As corrosive chemicals can attack concrete only in the presence of water, designs to minimizeattack by acids and bases generally involve the use of protective barrier systems.

Micro-Organisms. Under certain conditions, concrete structures are known to be susceptible to

attack promoted by micro-organisms.15

In all the cases of biologically induced attack on concrete, the

bacteria involved have been of the aerobic type that produce sulfuric or acetic acid in their life cycle.16,17

The acid attacks the concrete by dissolving the calcium hydroxide and calcium-silicate-hydrate gel of thecement paste. The effect of acetic acid is more severe than that of sulfuric acid because the resultingreaction produces a more stable product than the sulphate. Both these processes may generate hydrogensulfide or methane (CH4) gas. Since oxygen is required, coatings and chemical treatments can be used tominimize the effects of micro-organisms.

3.2.1.2 Physical Attack

Physical attack is the second major category of causes of concrete deterioration. Although it isoften difficult to separate physical attack from chemical attack of concrete, for purposes of discussion,physical attack will include degradation factors that are related to environmental effects.

Salt Crystallization. Salts can cause damage to concrete through development of crystal growthpressures that arise through physical causes. Crystallization from a salt solution occurs when the

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concentration of solute exceeds the saturation concentration at a given temperature. In general, the higherthe ratio of solute concentration to saturation concentration, the higher the pressure. Deterioration of thistype occurs when concrete is in contact with water containing large quantities of dissolved solids such asCaSO4, NaCl, and Na2SO4. As the water permeates the concrete, the salts crystallize in open pores due toevaporation. Repeated evaporation can cause the salt deposits to build up to the point that the stressesgenerated can crack the concrete. Structures in contact with fluctuating water levels or in contract withgroundwater that contains salts are susceptible to this type of deterioration. The problem of saltcrystallization can be minimized by using sealers or barriers to either prevent water ingress or subsequentevaporation. However, if the sealer is not properly selected and applied it can cause the moisture to rise inthe concrete.

Freezing and Thawing. Concrete in a saturated or near-saturated condition can be susceptible todamage during freezing and thawing cycles. This damage is caused by hydraulic pressure that is generatedwithin the capillary cavities of the cement paste as the water freezes. Damage to concrete as a result offreeze-thaw action generally takes the form of cracking and spalling. Certain types of coarse aggregateswhen used in concrete slabs are known to cause cracks that are usually parallel to joints and edges. Thesecracks are commonly called D-cracks. Factors that can be used to enhance the resistance of concrete tofreeze-thaw damage include use of air entrainment, a low water-cement ratio, longer moist curing, higherstrength concrete, and drainage to divert the flow of water that can saturate concrete surfaces exposed to

freezing and thawing conditions.18

It is not the total air contained of the cement paste matrix that is the important factor in producingdurable concretes, it is the spacing of 0.1 to 0.2 mm air voids throughout the hardened cement paste that is

critical.13

The pore structure of hardened cement paste is influenced by the water-cement ratio and thedegree of cement hydration. At a given freezing temperature, the amount of freezable water will be largerfor concretes with higher water-cement ratios and for concretes that are in the early stages of curing.Although there is generally a direct relationship between concrete strength and durability, air-entrainedconcrete, that has a lower compressive strength than similar non air-entrained concrete, will exhibitimproved resistance to freeze-thaw damage. The cracking and spalling resistance of concretes exposed tovery low temperatures is dependent on the difference between the critical and actual water saturation of the

concrete. Dry or partially dry concretes do not suffer freeze-thaw damage.13

The critical degree ofsaturation prior to freezing is controlled by the concrete permeability and the availability of water.

Thermal Exposure. The mechanical properties of concrete can be adversely effected by exposureto elevated temperatures and thermal gradients. Loss of structural capacity and reduced stiffness can resultif the concrete in a structure is exposed to extreme fire or heat. These mechanical property variations resultlargely because of changes in the moisture content of the concrete constituents and progressivedeterioration of the cement paste and aggregates. The response of concrete to elevated-temperatureexposure depends on the type and porosity of aggregates, rate of heating, concrete permeability, moisturecontent at the time of heating, and various other factors. Significant deterioration of the concrete strengthdoes not generally occur until the exposure temperature reaches about 400oC. At this temperature,

dehydration of calcium hydroxide occurs.19

However, concrete exposed to temperatures of 90oC may lose

up to 10% of its room-temperature strength and modulus of elasticity.20

Also, the effects of moderateelevated temperature exposure over extended time periods such as required for the fuel cycle facilities areunknown. In addition to potential reductions in strength and modulus of elasticity, thermal exposure ofconcrete can result in cracking of the concrete, and when moist concrete with low permeability is heatedvery rapidly, surface spalling can occur.

Thermal cycling, even at temperatures as low as 65oC, can have some deleterious effects onconcrete. Under these conditions, concrete compressive strength, tensile strength, bond strength, and

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modulus of elasticity can be reduced below its room-temperature values.21

Most reinforced concretestructures are unaffected by thermal cycling due to daily temperature fluctuations. However, when cycledto higher temperatures between about 200 and 300•C, the most damage occurs during the first thermalcycle, with the extent of damage markedly dependent on aggregate type and is associated with loss of bondbetween the aggregate and the cement matrix.

Irradiation. Irradiation in the form of either fast and thermal neutrons or gamma can affect theconcrete. However, neutron radiation only occurs with nuclear criticality, so it can be discounted inrelation to fuel-cycle plants, where criticality would constitute a severe accident. Gamma rays produceradiolysis of water in cement paste and evolution of gas. Prolonged exposure of concrete to irradiation canresult in decreases in tensile and compressive strengths and modulus of elasticity. Although it is difficultto separate the effects of irradiation from the effects of temperature because the energy of absorbedirradiation is converted to heat, approximate threshold levels necessary to create measurable damage in

concrete have been reported as 1 x 1019 neutrons per square centimeter fluence22

and 2 x 1010 rad of dose

for gamma radiation.16,22

Damage to concrete due to excessive irradiation occurs as cracks and spalls of

exposed surfaces. Studies conducted at the Karlsruhe Research Center in Germany23

to investigate theeffects of irradiation on aggregates, cement pastes, and mortars indicate that for exposure levels to 1 x 1020

neutrons per square centimeter fluence, the aggregates investigated that showed no irradiation damagewere limestone and magnetite, whereas granite showed considerable damage. Granite aggregate is not apreferred material in fuel-cycle plants as granite releases radon gas that results in spurious alpha alarms.Studies evaluating the effects of irradiation on concrete properties and the influence on structural behaviorare very limited.

Settlement. All structures have a tendency to settle during construction and early life. Excessivesettlement or differential settlement can cause misalignment of equipment and lead to overstress conditionsin reinforced concrete structures (e.g., cracking). The amount of settlement is dependent on the physicalproperties of the foundation material at the site, which may range from bedrock (minimal settlementexpected) to compacted soil (some settlement expected), and on quality control of foundation construction.Settlement is considered in the design of the structures and is not expected to be significant. When areinforced concrete structure is sited on soils, the potential for settlement is acknowledged, and monitoringprograms are implemented to confirm that design criteria are met. In general, most of the settlement willoccur within a few months after construction and become negligible after this period.

3.2.2 Steel Reinforcement Degradation Mechanisms

Steel reinforcement is provided in concrete structures to resist tensile stresses and compressivestresses for elastic design, provide structural reinforcement where required by limit condition designprocedures, and to control the extent and width of cracks at operating temperatures. The most likely sourceof steel reinforcing system degradation is corrosion.

3.2.2.1 Corrosion

Corrosion of steel in concrete is an electrochemical process. The electrochemical potentials thatform the corrosion cells may be generated in two ways: (1) composition cells formed when two dissimilarmetals are embedded in concrete, such as reinforcing steel bars and aluminum conduit, or when significantvariations exist in surface characteristics of the steel; and (2) concentration cells formed due to differences

in concentration of dissolved ions in the vicinity of steel, such as alkalis, chlorides, and oxygen.13

As aresult, one of two metals or different parts of the same metal become anodic and others cathodic. Thetransformation of metallic iron to ferric oxide (rust) is accompanied by an increase in volume which,depending on the state of oxidation, may be as large as 600 percent of the original metal. This volume

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increase can cause cracking and spalling of the concrete. In addition, corrosion will result in a reduction ineffective steel cross-section and capacity. Depending on the source, local embrittlement may also beproduced.

Both water and oxygen must be present for corrosion to occur. In good-quality, well-compactedconcretes that are free of cracks having widths sufficient to permit corrosion, reinforcing steel withadequate cover should not be susceptible to corrosion because the highly alkaline conditions present withinthe concrete (pH>12) causes a passive iron oxide film to form on the surface of the reinforcing steel.However, when the pH of the concrete adjacent to the reinforcing steel falls below 11, or when chlorideions are present near the reinforcing steel, the passive iron oxide film can be destroyed causing corrosion toinitiate.

Reduction in the concrete pH can occur as a result of leaching of the alkaline substances from thecement paste or carbonation in which calcium hydroxide is converted to calcium carbonate or calcite.Carbonation is the reaction of carbon dioxide with the hydrated cement causing the pH of the affectedregion to be reduced. Carbonation begins at the concrete surface and progresses inward at an extremelyslow rate that depends on the permeability of the concrete, its moisture content, the carbon dioxide content,

and the relative humidity of the atmosphere.14

Carbonation is greater in concrete that is protected fromrain, but exposed to moist air, than in concrete exposed directly to rain. When the level of carbonatedconcrete reaches the first layer of reinforcing steel, corrosion initiates eventually causing the concretecover to crack and spall. The penetration of carbon dioxide from the environment into the concrete can beaccelerated if the concrete is porous or cracks are present.

The penetration of chloride ions can destroy the passive oxide film on the reinforcing steel, even athigh alkalinities (pH > 11.5). The chloride ions are common in nature and small amounts may beunintentionally contained in the concrete constituents. In addition to natural deposits of chloride salts,groundwater may be contaminated with chloride ions. For typical concrete mixtures, the thresholdchloride content to initiate steel corrosion is about 0.6 to 1.2 kilograms of chloride ions per cubic meter of

concrete.13,24

Once the passivity of the steel is destroyed, the electrical resistivity and oxygen control therate of corrosion. Since water, oxygen, and chloride ions are important factors in the corrosion ofembedded reinforcing steel, concrete permeability significantly influences the corrosion process.Concretes with low permeabilities usually have low water-cement ratios and contain adequate amounts ofcement, aggregates of the proper size and with the appropriate gradation, and mineral admixtures such asfly ash or silica fume. Other factors that can influence durability include the depth of concrete cover andthe maximum permissible crack widths.

Concrete cracking can accelerate the corrosion process. Concrete can crack for a variety ofdifferent reasons including plastic shrinkage, plastic settlement, drying shrinkage, thermal stresses,chemical reactions, weathering, corrosion of reinforcing steel, poor construction practices, construction

overloads, errors in design and detailing, and externally applied loads.25

Cracks widths in structures can becontrolled by appropriate use and placement of reinforcing steel and by limiting the tensile stresses carriedby the reinforcing steel. Concrete cracks have even been eliminated in some structures by using expansivecements that contain anhydrous calcium aluminosulfate, calcium sulfate, and uncombined calcium oxide

(e.g., ASTM C 845, Type E-1K).26

Under moist conditions and an absence of tensile stresses, dormant

cracks sometimes repair themselves by a natural process known as autogenous healing.25

This processoccurs through carbonation of the calcium hydroxide in the cement paste where calcium carbonate andcalcium hydroxide crystals precipitate, accumulate, and grow within the crack. Saturation of the crack andthe adjacent concrete with water is essential for autogenous healing to occur. Cracks that become healedcan transfer some tensile stresses and may even become sealed. However, autogenous healing will notoccur if there is a flow of water through the crack or if the crack is subjected to movement during thehealing process.

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Epoxy coated reinforcing steel, protective coatings, and cathodic protection of the reinforcing steelare approaches that have been used to prevent or delay the onset of corrosion. Additional information on

development of durable concrete structures in the presence of chloride ions is available.12,18,24,25

3.2.2.2 Other Degradation Factors

Degradation of steel reinforcing systems can also occur due to effects of elevated temperature,irradiation, and fatigue. Elevated temperature and irradiation levels experienced by reinforced concretestructures used in fuel cycle facilities are not likely to be significant enough to cause degradation. Fatigueis also not expected to lead to degradation.

3.2.3 Liner Degradation Mechanisms

Both non-metallic and metallic liners have been used in fuel cycle facilities to either isolate thewaste form from concrete or to provide a leaktight boundary.

The primary sources of degradation of non-metallic liners are cracks due to localized effects (e.g.,stress concentrations) or physical or chemical changes in the concrete. The integrity of non-metallic linerscan also be compromised due to localized impacts or possibly through accumulated radiation effects.

Metallic liners may be subject to the same general degradation mechanisms as steel reinforcingmaterials, of which corrosion is of most importance. For some components a corrosion allowance mayhave been added to the thickness during the design stage. However, little allowance may have beenprovided for the relatively thin plate lining pools for storage of spent fuel. Typically the liner plate isfabricated from stainless steel or carbon steel that has been clad with stainless steel. The influence of localcorrosion attack that can lead to loss of leaktightness is of most concern. Corrosion data for structural steel

in numerous environments are available.27

In general, depending on the environmental parameters, thisreference notes that surface corrosion rates generally range from 0.001 mm/year to 0.03 mm/year.

Liners may also be damaged by excessive thermal movement. The coefficients of thermalexpansion for concrete and stainless steel differ, so that temperature in excess of the design temperaturecould cause rupture of the liner material, resulting in liquor coming into contact with the concrete.

To resist the process chemistry in reprocessing plants, nitric acid grade stainless steel is usuallyspecified. The material is vulnerable to corrosion in the presence of chlorides.

3.3 SERVICE CAPABILITIES AND DESIGN CONSIDERATIONS

The required service capabilities of fuel cycle facilities must consider the operational phase whilethe waste is being placed into the facility as well as the performance requirements necessary topermanently isolate the waste after it has been placed in disposal. During the operational phase the neededservice capabilities are similar to storage facilities and can include radiation shielding, dissipation of decayheat, and the ability to both resist and confine certain aggressive chemicals. During the service phase,performance requirements necessary to permanently isolate radioactive wastes include providing bothphysical and chemical barriers, ability to resist aggressive attack from the waste, and long-term structuraldurability.

Reinforced concrete structures used in repository or disposal facilities are designed to satisfy theintended functions, with special attention given to constructability and durability. Although the design

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approach and construction of reinforced concrete fuel cycle facilities are in large measure similar to thoseused for conventional industrial structures, increased emphasis in certain areas may be required toaccommodate the environmental and extended service life requirements (e.g, accommodation of thermalgradients, increased cover, and reduced permeability). Key design considerations include loads (normaland accident conditions), steel reinforcement requirements and detailing, concrete mix design forworkability and durability, control of crack widths, thermal and shrinkage stresses, and expected service

life.28

Loads and load combinations for structural design are defined in terms of magnitude, direction,probability of occurrence, and accuracy in computation. With respect to ultimate strength design, loadfactors are applied to the nominal member strength to establish structural member size and geometry.Reinforced concrete structures important to safety must have sufficient capacity for normal and off-normalloads without permanent deformation and with no degradation of capability to resist future loads.

Guidance on loads and load combinations is available.29,30 As certain facilities require the use of steel

reinforcement to provide necessary structural strength and resistance to cracking, corrosion of thereinforcement leading to concrete deterioration is a concern. Measures taken to minimize the potential forcorrosion and its effects include minimizing the total steel volume, uniformly dispersing the steel,increasing the concrete cover, and employing other measures to mitigate steel corrosion (e.g., chemical ormineral admixtures). Concrete mixes are designed for durability, strength, impermeability, andworkability (e.g., material selection and water content). Cracks need to be controlled to limit the ingress ofsubstances that might attack the steel reinforcement, and to minimize the potential for long-term waterinfiltration. Migration of materials from the waste forms to the environment is also a concern. Thermalgradients produced by stored waste will result in applied forces and moments in the exposed structure anda potential loss in concrete mechanical properties. The potential for material degradation due totemperature (and radiation) effects needs to be addressed with respect to service life requirements.

3.4 DESCRIPTIONS OF SEVERAL TYPICAL FUEL CYCLE FACILITIES

In general, there are four basic types or categories of radioactive waste associated with thenuclear fuel cycle for which storage or disposal facilities, depending on the character of thematerial to be stored, are currently being used or considered:

• Mining and Milling Waste Sites,• Low- and Intermediate-Level Radioactive Waste Disposal Facilities,• Transuranic Waste Disposal Facilities, and• High-Level Radioactive Waste Disposal Facilities.

Examples of the facilities associated with the nuclear fuel cycle are presented in Table 3.2. Typical storageor disposal facilities associated with these waste forms are described below. In the discussion below, theemphasis is placed on low- and intermediate-level radioactive waste disposal facilities because thesefacilities currently represent the primary applications for reinforced concrete.

3.4.1 Mining and Milling Waste Sites

The mining and milling of uranium and thorium bearing ores has generated contaminated wastes.For example, in the U.S., the Uranium Mill Tailings Remedial Action Program (UMTRAP), administeredby the U.S. Department of Energy, was developed to address these wastes from 24 inactive processing sitesand about 4,500 adjacent properties. Eleven of these sites are to be stabilized in-situ using concrete andgeologic barriers. The remaining sites will be decontaminated. Regulations require that the abovementioned stabilized sites be effective for at least 200 years, but preferably up to 1,000 years, to the extentreasonably achievable. Topsoil will be added to support vegetation and a drainage ditch system that

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intercepts the gravel layer in the barrier will divert surface water away from the site. The groundwateraround the disposal facility will be monitored to verify the effectiveness of the waste confinement systems.

3.4.2 Low- and Intermediate-Level Radioactive Waste Disposal Facilities

Low- and intermediate-level radioactive wastes are normally shipped to disposal sites in wastecontainers such as steel drums, steel boxes, liners, and thick casks. Table 3.3 presents a description ofseveral waste containers that have been standardized in the UK for packaging low- and intermediate-level

waste.31

When a waste container is filled with conditioned waste, the complete assembly is denoted as awaste package. The waste packages can be considered as unshielded (i.e., remote handling andtransportation in a reusable shielded transport container is required) or shielded (i.e., built-in shielding andcontain low activity materials so packages can be handled using conventional techniques and act astransport packages themselves). Unshielded packages, typically drums or boxes, generally aremanufactured from stainless steel. Shielded packages are typically boxes fabricated from stainless steelthat are lined with concrete in sufficient thickness to provide the required shielding. Also, non-standardpackages for low- and intermediate-level radioactive wastes may be used that have been designed by thewaste producers for wastes that can not generally be packaged in standard containers. After some periodof interim storage, the waste packages are transported for disposal.

3.4.2.1 Use of Reinforced Concrete Structures

Numerous types of concrete structures are being considered for proposed low- and intermediate-

level waste disposal facilities.32

The primary types include:

• Above-grade vault disposal,

• Below-grade vault disposal, and• Modular concrete canister disposal.

Also these structure types may be used in combination to provide yet an additional barrier for containingthe waste that combines the benefits of both the concrete canister and vault technologies. However, it issignificantly more costly and involves additional operations on the waste.

Disposing low- and intermediate-level waste in above-ground vaults involves placing the wastecontainers in an engineered concrete structure located above the surface of the ground and at a later timecovering with earthen materials. The above-grade vault concept is envisioned as a large steel-reinforcedconcrete enclosure with either limited horizontal access (as through a doorway) or vertical access (asthrough an open top). The vaults are above the natural grade at the disposal site. The floor, walls, and roofof the structure are made from reinforced concrete. Fill material is used to fill the spaces among the wastecontainers inside the structure. In summary, the description of above-ground vaults implies that:

• The waste is disposed above ground,

• An engineered concrete cover exists over the waste, with an earthen cover applied later, and• The disposal units provide structural stability.

The inclusion of both a concrete vault and an earthen cover enhances the ability of this disposal concept tolimit water infiltration into the waste, to prevent plant or animal intrusion, and to reduce gamma radiationexposure rates at the ground surface. The concrete vault also provides structural stability that is not present

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in the shallow land burial design. An above-grade vault with modular concrete canisters is illustrated inFigure 3.1.

Disposing low- and intermediate-level waste in below-grade vaults involves placing the wastecontainers in a concrete engineered structure that is then covered with earth. The vault is below the naturalgrade. The structure consists of reinforced concrete floors, walls, and roof. Generally, the below-groundvault is a disposal design in which:

• The waste is disposed below ground level,

• Both an engineered concrete cover and an earthen cover exist over the waste, and• The disposal units provide structural stability.

Placement of the vaults below grade makes the facility less likely to attract intruders and less vulnerable todisruptive events and processes that can occur at the land surface. Covering the vault with soil maypreserve the integrity of the disposal unit for a longer period than if it were exposed to the atmosphere.Below-grade vaults could be appropriate for disposing all types of low- and intermediate-level wastebecause the structure itself may provide the required structural integrity and protection from inadvertentintrusion, however, in order to meet these goals, the structure would have to endure without substantialdegradation for at least 500 years. A general lack of long-term experience with concrete structures makesit difficult to project the performance of below-ground vaults for this time period with a high degree ofconfidence. Figure 3.2 presents a schematic of a below-grade facility as well as potential exposurepathways that could lead to transport of contaminants from the facility.

Disposing low- and intermediate-level waste in modular concrete canisters consists of placingindividual waste containers into modular reinforced concrete structures that in turn are placed into trenchesbelow natural grade. Except for the use of the concrete canisters, the physical details of below-groundmodular concrete canister disposal resemble those of below-grade vaults. Each canister can be viewed as amini-vault. Within the canisters, grout probably would be used as backfill between individual wastecontainers. The description implies that:

• The waste is disposed of below grade,

• An earthen cover exists over the waste, and• The disposal unit provides structural stability.

As in the case of below-grade vault disposal, the inclusion of both a concrete canister and an earthen coveras barriers enhances the ability of this disposal concept to restrict water infiltration into the waste; toprevent human, plant, or animal intrusion; and to reduce gamma exposure rates at the ground surface. Theconcrete canisters provide the structural stability that is not present in the shallow land burial concept. Theearthen cover could be made thick enough to prevent damage to concrete from freeze/thaw cycles, extremetemperature gradients, and tornadoes.

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3.4.2.2 Applications

Summarized below are several applications in which reinforced concrete has been utilized infacilities for the storage or disposal of low- or intermediate level radioactive waste. The applications arenot all related to disposal of low- or intermediate level radioactive waste resulting from the nuclear powerplant fuel cycle, but are typical of the structures that are or could be used.

El Cabril Near-Surface Disposal Facility, Cordoba City, Spain.33

In October 1992, theSpanish Ministry of Industry and Energy issued an operating license for the El Cabril Near-SurfaceDisposal Facility located at the site of an abandoned uranium mine approximately 100 km from CordobaCity and 400 km south of Madrid. The facility will be used for permanent disposal of low-levelradioactive waste generated at Spanish nuclear power plants, commercial establishments where medicalisotope tracers are used, and institutions that conduct nuclear-related research. El Cabril was selectedbecause environmental conditions at the site are not particularly severe, the surrounding area is rural, thewater table is sufficiently low, and the groundwater and soil do not contain significant quantities ofchloride ions, sulfate ions, or other chemicals that aggressively attack concrete.

The facility consists of a series of secondary waste containment cells consisting of reinforcedconcrete slabs and walls, Figure 3.3. Each cell is 24 m by 19 m by 10 m. In the interior, smaller (2 m by 2m by 2 m) concrete containers are stored that contain drums filled with waste. To ensure that thesurrounding environment is protected, a drainage control network is provided to collect liquids from thecells. Piping for the control network passes through an inspection gallery located below each row of cellsand terminates in a control tank. An additional deep drainage piping system has been installed below thecells and inspection galleries to collect infiltrating water. Before each container with waste is placed insidea cell, it is injected with grout to fill the remaining void spaces. Once a cell is filled, gravel backfill isadded, a plastic film placed above the cell, and a covering of concrete is cast to provide radiationprotection for workers. Final closure will be provided by a 0.5 m-thick reinforced concrete slab that is castabove the cell. Protection for this slab is to be provided by an impervious acrylic-fiberglass film.

The reinforced concrete structural components and the inspection galleries were designed towithstand earthquake forces. Other factors such as concrete durability and leaktightness also influencedmaterial selection and design. To minimize potential leakage paths through the concrete, the slabs wereconstructed without joints. Additional leakage protection is provided by a polyurethane film and ageotextile membrane that are installed on the slab surface which slopes at one percent to centrally locateddrains. Wear protection for these coverings is provided by a porous concrete layer that is located above theslab surfaces. Stainless steel water stops are used between the walls and the slabs and at every verticalconstruction joint to help assure that liquids are confined inside the cells. In addition, reinforcing bardetails were selected to keep crack widths from exceeding the specified limit of 0.05 mm, and thereinforcing bar cover was specified to be the maximum recommended for aggressive environments.Concretes used to construct the inspection galleries, vaults, and cells have specified compressive strengthsof either 29.6 MPa or 41.4 MPa. Because concrete durability has an influence on the long-termperformance and safety of the facility, emphasis was placed on concrete mixture selection. The keyobjectives were to prevent rapid carbonation of the concrete and provide additional protection for thereinforcing steel to delay the onset of corrosion. Other qualities that were considered desirable includeddense concrete with low shrinkage.

Nirex Phased Deep Geological Disposal Concept, United Kingdom. At Nirex, based in Harwellin Oxforshire, United Kingdom, a concept of multiple barrier containment designed to ensure the long-

term safety of a repository containing low-and intermediate-level waste has been developed.34

Themultiple barrier containment concept makes use of engineered barriers (both physical and chemical) and

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natural barriers, working in conjunction, to achieve the necessary degree of long-term isolation andcontainment of radionuclides by preventing or limiting their movement from the repository to the humanenvironment. To help ensure isolation of long-lived, solid radioactive waste, the repository will be locatedunderground.

Typically wastes are packaged by being immobilized in cement within a highly engineered wastecontainer made from stainless steel or reinforced concrete. Such wastes would be placed into disposalvaults excavated in rock at least 300 m below the ground (Figure 3.4). The siting of the disposal facilitywould be on the basis that there be a low rate of groundwater flow through the repository “host rock,” thatthere would be a sufficiently long time for any radioactive materials in the groundwater to travel to thedisposal vaults to the human environment at the surface, and that there would be appropriate levels ofreduction and dilution of dissolved concentrations of radionuclides along the travel path. In addition, thegroundwater chemistry would be relatively benign towards the engineering materials in the repository so asnot to be detrimental to their required safety functions. Also, the vault environment could be controlled tomaintain the sound condition of the waste packages by controlling the temperature, humidity, and chloridelevels.

To ensure that other options are left open, the underground repository approach is to be carried outin a staged, reversible way (e.g., future generations would have the option of final closure). At each stageof development, time would be available to build sufficient confidence before moving to the next stage,while retaining the ability to retrieve waste and pursue alternative options if they would become availableand preferred. At some appropriate time after a period of monitoring (e.g., 50 to 100 years), and if thesociety of the day so decided, a specially formulated cement-based grout could backfill the spaces aroundthe waste packages placed into the disposal vaults. This cement-based backfill represents an importantengineered barrier within the overall concept as it provides chemical conditioning of groundwater thatwould inevitably come into the disposal vaults and eventually come into contact with the radionuclides inthe wastes. The backfill is designed to ensure that the groundwater would be maintained in an alkalinecondition for very long periods of time. Thereafter, again at periods to be determined by the society of theday, the vaults could be sealed to specifications determined by safety considerations. Figure 3.5 presentsthe concept for waste emplacement and retrieval.

Centre de Stockage de la Manche (CSM) and the Centre de Stockage de I’Aube (CSA),33

Engineered concrete barriers are currently being used for near-surface low-level radioactive waste disposalin France. These barriers are part of a multi-component waste isolation system intended to protect thewaste packages against any external interference and, in the event of such an incident, limit theconsequences. System components include treated waste that has been immobilized in a matrix, primarywaste containers, concrete monoliths (concrete vaults of various sizes and designs), and a tumulusconsisting of gravel, impermeable clay, and soil. At CSM, the waste is placed into concrete vaults andburied in a tumulus. Waste at CSA, which is higher in activity, is also placed into concrete vaults, butthese vaults have double floors. To fill voids between the waste packages and to provide radiationshielding during the time the vaults are being filled, the CSA vaults are backfilled with grout after eachlayer of waste is added. After vaults at the two sites are filled, a concrete roof will be cast on the vaults tocreate an additional engineered barrier. As an extra precaution, a special collection system is locatedbeneath the waste repositories to collect water that happens to infiltrate the tumuli covers. Theserepository structures are designed to be earthquake resistant and impervious to rainfall and groundwater. Amonitoring network at the base of the repository allows the water tightness of the structures to be checked.Another network of drains at the base of the tumulus is used to collect and monitor runoff.

Central Interstate Low-Level Radioactive Waste Compact, United States.

To date in theUnited States, most low-level waste has been disposed in shallow earthen trenches, a technology calledshallow land burial. Some of the early shallow land burial sites have not contained radionuclides to theextent desired. Operating practices and institutional controls were also inadequate at some facilities. In

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part as a result of the experience with shallow land burial, several states and compacts of states haveadopted laws that forbid using shallow land burial as a means of low-level waste disposal. Even whereshallow land burial has not been specifically forbidden by law, there is a strong interest in usingalternatives to shallow land burial to dispose of low-level waste. Almost all of the new facilities beingplanned in the U.S. employ concrete barriers as an important part of the design. The one exception is aproposed low-level waste facility in the arid Southeast region of California. That facility allows shallowland burial with a 7-m earth cover for the lower-activity waste, however, the higher-level waste is proposedto be placed in concrete overpacks before being placed into a trench.

The U.S. Congress in 1988 gave its consent to the states of Arkansas, Kansas, Louisiana,Nebraska, North Dakota, and Oklahoma to enter into the Central Interstate Low-Level Radioactive Waste

Compact.9,33

This authorization allows the agreement states to jointly fund a regional compact formanaging their non-federal low-level radioactive wastes. It is the policy of the party states to enter into aregional low-level radioactive waste management compact for the purpose of:

• providing the instrument and framework for a cooperative effort;• providing sufficient facilities for the proper management of low-level radioactive waste

generated in the region;• protecting the health and safety of the citizens of the region;• limiting the number of facilities required to manage low-level radioactive waste generated in

the region effectively and efficiently;• promoting the volume and source reduction of low-level radioactive waste generated in the

region;• distributing the costs, benefits, and obligations of successful low-level radioactive waste

management equitably among the party states and among generators and other persons whouse regional facilities to manage their waste;

• ensuring the ecological and economical management of low-level radioactive waste, includingthe prohibition of shallow-land burial of waste; and

• promoting the use of above-ground facilities and other disposal technologies providing greaterand safer confinement of low-level radioactive waste than shallow-land burial facilities.

Following detailed site characterization studies, a 130 ha site near Butte, Nebraska, was selected as theproposed site for a near-surface low-level radioactive waste disposal facility, Figure 3.6. In 1990, plans forconstruction at the site were submitted to the state of Nebraska for consideration. As proposed, the facilitywill receive low-level radioactive waste for a 30-year operational period or until 142,000 m3 of wastes aredisposed. After which, the waste will be buried beneath an earthen mound and covered with a multi-layered cover. The engineered barriers that will be constructed at the site are intended to provideprotection for the surrounding environment for at least 500 years.

Above-grade reinforced concrete cells are the main engineering feature of the proposed facility. Up to 21cells will be constructed for disposal of the waste, with one of these cells constructed for disposal of themore hazardous of this waste. This cell must meet more rigorous requirements on waste form to ensurestability after disposal. A typical cell for the low-level waste is essentially a single story, 7.0–m-high,reinforced concrete frame with 0.9-m-thick walls, a 0.9-m-thick roof slab, and a 1.4-m- thick continuousbasemat. These cells will be 18.9-m wide and 85.6 m long. To minimize the potential for waterinfltration, only a minimum number of construction joints will be permitted. The cell for the morehazardous waste will be slightly bigger than the other cells, the roof constructed using 0.9–m-thickremovable precast concrete panels, and transverse concrete walls will be spaced at 7.6 m intervals downthe length of the cells. Loads considered in the design included dead loads, live loads, seismic loads basedon a ground level acceleration of 0.15 g, tornado-induced wind loads, and tornado-induced missile loads.Although these loads were one of the primary design considerations, wall and roof dimensions wereselected based on radiation shielding criteria. Consequently, the structural capacity of these components is

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substantially greater than necessary. An equally important design consideration involved the selection anddetailing of steel reinforcement to minimize crack spacing and width. Small-diameter steel reinforcing barsspaced at close intervals were specified. The concrete mix design is based on strength, durability,impermeability, and workability considerations. The concrete water-cement ratio was limited to amaximum of 0.40, moderate heat of hydration and sulfate resistant (Type II) portland cement will be used,fly ash will provide 20% of the cementitious materials, and chemical admixtures will be added to enhancedurability and workability. The 90-day design compressive strength is 27.6 MPa. An extended moist curetime of at least 14 days will be used to help improve overall concrete quality. An assessment of the long-term performance of the proposed disposal facility was undertaken to estimate the ability of the engineeredbarriers to successfully isolate the waste from infiltrating water. The analytical models utilized simulatedthe year-to-year progressive degradation of the concrete and reinforcing steel. Loss of function wasconservatively assumed to occur when cracks in the concrete penetrate to a depth equal to 75 percent of thesection thickness. Results of this analysis indicated that the cells for the lower-level waste will isolate thewaste from infltrating water for approximately 550 years, while the cell for the more hazardous waste willkeep the waste isolated for approximately 3,500 years.

Department of Energy, Oak Ridge, Tennessee, United States.35

Concrete structures for waste storage anddisposal applications are used at three U.S. Department of Energy facilities in Oak Ridge, Tennessee. Thewaste is generated from research activities, production operations, and clean up of contaminated areaslocated at the X-10, Y-12, and K-25 Plants.

Interim Waste Management Facility

An Interim Waste Management Facility is being constructed at the X-10 Plant near the Oak RidgeNational Laboratory. It is intended for permanent near-surface disposal of solid low-level radioactivewaste. The waste is primarily contaminated with short half-life (<30 years) fission product radionuclideswith surface dose rates of up to about 20 mSv/h. Eventually, the facility will include multiple reinforcedconcrete disposal pads where precast concrete waste vaults will be stacked to a height of 5.2 m and thenburied beneath a 2.4-m-thick layer of clay and stone. During the operational period, runoff from each pad,which serves as a secondary waste containment system, is collected and pumped to a 151 cubic meter steelstorage tank.

One pad and an accompanying drainage gallery and performance monitoring system have beenconstructed at the site. The pad is 27.4-m long, 18.3-m wide, and 0.4-m thick. A 0.3-m high by 0.3-mwide curb is located along three edges of the pad. Loads and load combinations resulting from dead, live,wind, and earthquake loads were considered during the design process. Epoxy-coated steel reinforcing barsand silica fume concrete with a design 28-day compressive strength of 34.5 MPa were used for all concreteconstruction. Silica fume concrete was selected to decrease the water permeability and porosity of theconcrete. In addition, the silica fume concrete has a slightly higher modulus of rupture and tensile strengthwhen compared to conventional portland cement concrete with the same specified compressive strength.An aerial view of a completed disposal pad is shown in Figure 3.7. When this photograph was taken, thepad was approximately half full of precast concrete waste vaults.

Pond Waste Management Project

The Pond Waste Management Project at the K-25 Plant involves temporary storage of mixedwaste. Drums of the solidified waste are placed into a storage facility in preparation for offsite disposal.The facility includes five large waste storage buildings that provide approximately 20,800 m2 of wastestorage area. Each building consists of a pre-engineered metal structure and an 200-mm-thick reinforcedconcrete slab on grade. Concrete curbs that are 152-mm high are provided around each slab to create asecondary confinement dike. Shrinkage-compensating concrete was used to construct the slabs and curbs.Use of this type of concrete was effective in minimizing cracking due to drying shrinkage and reducing the

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number of contraction joints that were needed. As an added precaution against contamination of theconcrete, an epoxy coating was applied to the floor and curb surfaces before the facility was placed intoservice.

Mixed Waste Storage Expansion Project

The Mixed Waste Storage Expansion Project involves the use of another facility at the K-25 Plantto store drums of mixed waste. In this case, existing 50-year-old reinforced concrete buildings wererefurbished and converted into storage areas. The original design drawings for the buildings werereviewed and calculations were prepared to demonstrate that the existing reinforced concrete floor slabswere structurally adequate for the anticipated fork-truck and storage loads. Acceptance criteria were basedon requirements provided in applicable building codes. Once approved, the concrete floor surfaces weregrit blasted in preparation for application of an epoxy coating, and the expansion joints were cleaned andsealed with an elastomeric material. Two coats of an epoxy coating were then applied over the entire floorsurfaces of the storage areas. This coating provides protection for the concrete and embedded steelreinforcing bars from possible chemical attack. It also makes spill cleanup easier and more effective.

Above-Grade Storage Pad Project

The Above-Grade Storage Pad Project involves a storage facility that was constructed at the Y-12Plant to receive low-level radioactive waste. The waste is composed primarily of construction debriscontaminated with radioactive material. A total of six pads were constructed. Each pad consisted of areinforced concrete slab that is approximately 51.8-m long, 16.2-m wide, and 254–mm thick with a 152-mm high curb around the perimeter. To facilitate removal of liquid that inadvertently accumulates on thepads, the top surface of the concrete is sloped to a closed trench drain. The pads were cast usingshrinkage-compensating concrete, and no joints were used. The design included a top and bottom layer ofepoxy-coated steel reinforcing bars spaced 152 mm on centers each way. In the event some crackseventually occur, the epoxy coating should extend the useful service life of the pads by providing addedprotection against corrosion. The slabs and curbs create a secondary waste containment system for the 1.2-m wide by 1.8-m long by 1.2-m high steel waste containment boxes that are stored on the pads. Weatherprotection is provided by a prefabricated enclosure consisting of a structural steel frame covered by apolyester fabric coated with polyvinyl chloride. The slab was designed to handle fork lift and truck loadsas well as distributed uniform loads from storage of steel boxes stacked up to five high.

Surge Tank Containment

A reinforced concrete containment structure for a 3,785 cubic meter surge tank was constructed in1995 at the X-10 Plant. The tank provides temporary storage for low-level radioactive liquid process wastegenerated at the Oak Ridge National Laboratory. Construction of the 56.1-m long by 25.3-m widecontainment involved placement of a 610-mm-thick reinforced concrete slab on grade followed byplacement of 305-mm-thick perimeter concrete walls. Wall heights varied from 2.4 to 3.6 m depending onthe adjacent terrain. The floor was reinforced with two layers of conventional 12.7-mm diameter deformedsteel reinforcing bars spaced 152 mm on centers each way. A shrinkage-compensating concrete with adesign 28-day compressive strength of 27.6 MPa was specified to reduce cracking and minimize thenumber of joints. Joints that were necessary contained a polyvinyl chloride bulb-type waterstop.

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3.4.3 Transuranic Waste Disposal Facilities

Transuranic (TRU) wastes are those containing isotopes above uranium in the periodic table.These are the by-products of fuel assembly and weapons fabrication and of reprocessing operations (e.g.,clothing, tools, rags, residues, debris, and other such items contaminated with small amounts of radioactiveelements – mostly plutonium). TRU waste activity is generally low but, like high-level waste, it containsseveral long-lived isotopes. Temporary above-ground storage of transuranic wastes has not posed a seriousimminent hazard to the public, but it is unacceptable for the long term. Many barrels and boxes containingwaste have exceeded their design lifetimes, and in some cases have corroded and leaked. Continuedstorage would require periodic repackaging, at considerable expense, and at some risk to workers, for theindefinite future.

Currently in the U.S., it is planned to dispose of TRU wastes at the Waste Isolation Pilot Project (WIPP)near Carlsbad, New Mexico. At this facility, the wastes will be stored within rooms excavated within arock salt formation. Salt is the material of choice because:

• Most deposits of salt are found in stable geological areas with very little earthquake activity,• Salt deposits demonstrate the absence of flowing fresh water that could move waste to the

surface,• Salt is relatively easy to mine, and• Rock salt heals its own fractures because of its plasticity.

Salt formations at the WIPP site were deposited in thick beds (sodium chloride rock) during theevaporation of an ancient ocean, the Permian Sea. The primary formation containing the WIPP mine isabout 610-meters thick, beginning 260 meters below the surface. The mission of the WIPP is to provide aresearch and development facility to demonstrate the safe disposal of TRU wastes from defense operations.The WIPP site received its first shipment of non-mixed transuranic waste on March 26, 1999.Approximately 38,000 shipments are expected over the next 35 years. Figure 3.8 presents a schematic ofthe WIPP facility.

3.4.4 High-Level Radioactive Waste Storage and Disposal Facilities

Primary sources of high-level wastes include spent nuclear fuel∗ and highly concentrated liquidsolutions of nuclear fission products resulting from chemical reprocessing of spent nuclear fuel. The spentnuclear fuel is generally stored in structurally engineered cooling pools or vaults until it can be transferredon site to a dry cask storage location or transported off site to a high-level radioactive disposal site.Reprocessing wastes are later solidified, primarily through vitrification. Among the potential options fordisposing of high-level waste, deep geological disposal on land is the predominate means planned forisolating such wastes permanently from the environment. Primary means for storage or disposal of thesewastes include fuel storage/radwaste facilities, tanks, and geologic repositories.

∗ Although spent nuclear fuel is classified as waste in the United States, this is not the case in other

countries.

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3.4.4.1 Fuel Storage/Radwaste Facilities

The fuel storage facility provides for receiving, storing, shielding, shipping, and handling newand spent fuel. The fuel storage facility is an integral part of the reactor building of most boiling-waterreactor Mark I and Mark II containment type plants. For Mark III containment type boiling-water reactorsand many pressurized-water reactors, the fuel storage facility is a separate structure. However, in someplants the fuel storage facility is part of an auxiliary building. Typically, the fuel storage facility is a multi-story reinforced concrete structure. The bearing walls are reinforced concrete and are also designed asshear walls. The fuel storage pool is supported by bearing walls and/or a basemat. The floors and roof arereinforced concrete supported by steel beams. The interior columns are structural steel or compositeconstruction. The concrete slabs are designed as diaphragms to transmit lateral loads to shear walls. Theinterior walls are concrete block masonry or reinforced concrete. In some plants the exterior walls of thefuel storage facility are structural framing with concrete panels, concrete block, or metal siding. The roofsof these buildings are composite design. The spent fuel pool has reinforced concrete walls for radiationshielding and a liner to provide leak tightness. The liners are normally subjected to preservice testing toconfirm their leak-tight integrity. Further, liner leakage is typically monitored during operation by meansof a system of leak chases located behind the liner plate seam welds (stainless or clad carbon steel liners).For shielding purposes, the pool and transfer canal are deep enough to assure water coverage of the fuelduring transfer and storage. In the UK, early irradiated fuel pool ponds were single skin, unlined concreteconstructed with extra care used to minimize shrinkage and early thermal cracking. More recent pondshave included double containment systems, with leak detection, and have been enclosed. Rubberwaterbars are typically used at expansion joints. Most recent pools are lined with resin-based sealingsystems. Ponds used for flask-handling operations may be lined with stainless steel to provide protectionagainst leaks and damage from dropped-load accidents. In addition, some are provided with a coated steelmembrane, placed within the thickness of the concrete wall, to ensure leak-tight performance even if theconcrete is cracked in a dropped-load accident.

Radwaste facilities are provided at plants for storage, processing, handling, packaging, andshipment of radioactive waste. These facilities are often housed in a dedicated building. These buildingsare designed to prevent the uncontrolled release of significant amounts of radioactivity even under extremeenvironmental conditions. The radwaste processing facility is constructed of reinforced concrete,structural steel, and/or masonry block walls. The exterior load-bearing walls are reinforced concrete andare also designed as shear walls to resist lateral loads. The floors are reinforced concrete supported bycolumns and steel framing. The intermediate walls, constructed of reinforced concrete or masonry block,provide the shielding required for occupancy and access. Heavy reinforced construction is normal forplants in the UK handling beta/gamma emitters. The plants are typically based on a cellular layout withthe mechanical drives outside the cell to minimize the need for cell entry. Prestressed concrete was usedfor some building roofs and for circular tanks.

The spent fuel is stored in the spent fuel storage pool until it can be transferred on site to a drycask storage location or transported off site to a high-level radioactive disposal site or for reprocessing. Asthere are no permanent repositories at present, fuel that is not reprocessed is presently maintained on site.This has resulted in the fuel pools at many of the facilities being close to or at capacity, even aftermodifications to increase the capacities. These utilities must now license and build interim dry storagefacilities. There are primarily two types of these facilities - Independent Spent Fuel Storage Installation(ISFSI) and Monitored Retrievable Storage (MRS). The ISFSIs are complexes designed and constructedfor the temporary storage of spent nuclear fuel and other radioactive material associated with spent fuel.They may be located at nuclear reactor sites or other sites. The MRS facilities are spent nuclear fuel drystorage complexes designed and constructed for the receipt, transfer, handling, and storage of spent fueland solidified high-level waste pending shipment to a permanent repository. Dry cask systems aregenerally used to provide confinement, radiological shielding, physical protection, and passive cooling.

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The dry cask systems may be fabricated from metal or concrete. Table 3.4 presents a listing of dry spentfuel storage designs that have been licensed by the U.S. Nuclear Regulatory Commission.

3.4.4.2 Tanks

Liquid high-level radioactive waste results when spent nuclear fuel from reactor operations ischemically reprocessed. Prior to disposal this waste is stored in tanks.∗ Examples of applications wherereinforced concrete has been utilized for storage of this waste are presented below.

Sellafield Site of British Nuclear Fuels plc.36 Reprocessing of irradiated nuclear fuel produces

liquid high-level waste that is a concentrated solution of fission products in nitric acid. A characteristic ofthis high-level waste is that it generates enough heat that if not adequately cooled it has the potential toreach a significantly increased temperature and, in the extreme, could boil. This material is stored in anumber of water-cooled tanks. The tanks are of two basic designs. The earliest steel tanks werecommissioned between 1955 and 1968 and are of 70 m3 capacity. Tanks commissioned between 1970 and1990 are all of 150 m3 capacity. The more recent tanks, each 6.2–m diameter by 6.2-m high and housed inseparate cells, have seven internal cooling coils, one or more external cooling jackets, and include systemsfor agitating the contents. The floors of the cells are clad with stainless steel that extends to a height of afew meters up the cell walls in order to contain up to the whole tank inventory should leakage occur. Thetank contents are maintained within a temperature range of 50-60°C and the tank bulk temperature is notallowed to fall below 45°C so as to avoid crystallization that would adversely affect heat transfer.Secondary containment is provided by a stainless steel external cooling and secondary containment jacket.The cells are contained within heavily shielded enclosures in which stainless steel lined reinforced concreteprovides the shielding and third level of containment. The degree of the radioactive fission products isrelated to the storage time. Thus, the composition of the waste changes and the heat generation ratedecreases as the storage time increases. Figure 3.9 presents principal components of the tanks for storageof highly active waste at BNFL Sellafield. Current policy is to incorporate the high-level waste intomolten glass, which is poured into stainless steel containers and stored in a purpose-built above-groundstore. This is designed to be cooled by natural convection, thus removing any reliance on mechanicalventilation systems. This store is not intended for long-term disposal. Waste arising from UK customerswill be retained until final disposal facilities are available; waste attributable to overseas customers will bereturned under contract to the country of origin.

US Department of Energy High-Level Waste Storage Tanks at Hanford, Washington.37

Stored radioactive waste tanks at Hanford originated from the chemical reprocessing of defense productionreactor irradiated fuels and various plutonium-uranium separations, chemical extraction, anddecontamination processes required to support the past weapons production mission. Both single-shell anddouble-shell reinforced concrete tanks are used to confine the liquid, salt cake, and sludge wastes so thatthey do not enter the environment. Many of the tanks have been in use since the 1940’s. Single shell tankswere constructed from 1943 to 1964 and range in capacity from 2x105 liters to 3.8x106 liters and consist ofreinforced concrete with a single steel liner plate. The single-shell tank essentially consists of a cylindricalreinforced concrete vault, with a concrete floor and a flat or domed roof. The inside wet surfaces of theconcrete are lined with carbon steel. The double-shell tanks were constructed from 1967 to 1986, consist ofa vault containing a steel tank, and are all of 3.8x106 liters capacity. The vault is a cylindrical reinforcedconcrete structure lined with carbon steel. Its roof is either domed or flat, and its floor is of thick,insulating concrete on which the inner tank rests. There is about a 76-cm annular space between the steeltank and the lined concrete vault. At the top of the cylindrical wall, above the liquid high-level waste, thesteel tank wall is joined to the carbon steel liner of the roof. Diameters of the tanks are on the order of 23

∗ Some countries incorporate the liquid high-level waste into glass that is cast into stainless steel containers

for storage.

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meters with a waste level of about 10.7 to 12.2 meters. Both single-shell and double-shell tanks arecompletely enclosed. In both cases, the concrete structure is the primary barrier to soil pressure, while itserves as the secondary confinement of the waste. Operating specifications for these tanks have alwaysrequired temperature limits for the waste contents in order to control the temperatures and thermalgradients in the reinforced concrete structures. Elevated temperatures can overstress and degrade theconcrete and accelerate corrosion/aging mechanisms in the waste steel tanks. Waste temperatures in thetanks presently range from ambient to 88°C. For most of the waste tanks, waste temperatures havehistorically been less than 180°C, but some of the tanks are suspected of having reached 315°C. Figure3.10 presents configurations for the high-level waste tanks at Hanford.

Reinforced concrete below-ground vault structures have been designed and constructed at theHanford Site for placement of grout made by mixing radioactive salt solution with cementitious

materials.33

The Hanford Grout Disposal Facility consists of a number of large underground reinforcedconcrete vaults. The grout as a slurry will be pumped into the underground vaults where it will cure andform solidified monoliths. The grout waste form and the vault structures serve as the primary confinementsystem or engineered barrier. The secondary confinement drainage and collection system under each vaultis a gravel filled reinforced concrete catch basin with a network of perforated steel pipe laterals. Also,below the catch basin drainage system with vertical extensions along the outside of the vault walls is acontinuous asphalt-aggregate barrier. This barrier provides additional assurance for confining the wastewithin the disposal facility, and protects against the intrusion of possible groundwater into the facility. TheHanford Grout Disposal Facility was planned to have over 30 vaults, each with a disposal capacity of6x106 liters. The inside wall and floor surfaces are coated with an asphalt sealant. The grout waste formwill occupy all but the top 1.2 m of a vault. The remaining space will be filled with a "cold cap" grout thatis not mixed with waste. Each vault has a roof made of precast, prestressed concrete panels that are liftedand grouted into place after the vaults are filled. Then the roof panels are topped with a layer of cast-in-place concrete. This layer has a 15-cm drainage crown from the center line out to the edges, and 200 mmmaximum thickness. The vault design requirements include the hydrostatic pressures from the grout slurry(before curing), the soil backfill pressure, and earthquake ground motions with peak accelerations of 0.25 ghorizontal, and 0.17 g vertical. To date, the Hanford Grout Disposal Facility has not been put into servicebecause of regulatory and political concerns and questions about the long-term performance of theconfinement systems or engineered barriers in place for the vaults.

3.4.4.3 Geologic Repositories

Deep geologic disposal is almost the unanimous choice for ultimate disposal of high-level waste.The basic concept of geologic disposal is to place carefully prepared and packaged waste in excavatedtunnels in geologic formations such as unsaturated volcanic tuff, rock salt, argillaceous formations (clays),basalt, and granite. The concept relies on a series of barriers, natural and engineered, to contain the wastefor thousands of years and to minimize the amount of radioactive material that may eventually be releasedfrom a repository and reach the human environment. These barriers generally include (1) the leach-resistant waste form itself, (2) corrosion-resistant containers into which the wastes are encapsulated, (3)special radionuclide- and groundwater-retaining material placed around the waste containers (backfill), and(4) the geological formation itself - the principal barrier - which should both retard the transport of

radionuclides in circulating groundwater, and isolate the waste from man’s environment.7 The four key

attributes that a geologic repository would need to exhibit to protect public health and environment for

thousands of years are:38

• Limited water contact with waste packages,• Long waste package lifetime,• Low rate of release of radionuclides from breached waste packages, and• Reduction in concentration of radionuclides as they are transported from breached waste

packages.

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Since water is the primary means by which radionuclides could reach the human environment, theprincipal functions of the barriers are to keep water away from the waste as long as possible, to limit theamount of water that finally does contact the waste, to slow the release of radionuclides from the waste,and to reduce the concentrations of radionuclides in groundwater.

All countries pursuing geologic disposal are taking the multibarrier approach, although the

barrier selection may differ.8 In Germany the disposal concept depends heavily on the geologic barrier, the

rock salt formation at the prospective disposal site. In Sweden, thick copper waste packages are relied uponto contain the waste. In the U.S., a defense-in-depth design strategy is used that includes the naturalcharacteristics of the Yucca Mountain, the chemical and physical forms of the waste, and the wastepackages and other engineered barriers. Although permanent disposal of high-level waste has not yetoccurred, demonstration of its feasibility is taking place through final disposal of low-and intermediate-level wastes in underground caverns. In Germany, low-level waste has been disposed underground in theAsse Salt mine, as a demonstration project, between 1967 and 1978, and a deep repository (>500 m) forlow- and medium-level waste has operated in a salt dome in Morsleben between 1981 and 1998. Thelicensing procedure to permit disposal of waste that does not emit heat in an abandoned iron ore mine atKonrad at a depth of 1000 meters is in its final stages. In Sweden, a repository at an intermediate depth fordisposal of low- and medium-level waste has been operating at the Forsmark nuclear site since 1988. Thedisposal caverns are excavated in granite bedrock, offshore, about 60 meters below the bed of the BalticSea, and are accessed by a tunnel from land. In Finland, a facility for the disposal of low-level andmedium-level waste was opened in 1992 at the Olkiluoto nuclear site and in 1998 at the Loviisa site.These consist of caverns excavated in granite at depths of around 100 m below ground. In Norway, theHimdalen facility for low- and intermediate-level waste started operation in 1999 and consists of fourcaverns below 50 meters of bedrock cover. In 1999 permits were granted in the U.S. to start disposal ofwaste from defense programs at the Waste Isolation Pilot Plant in Southeastern New Mexico (Figure 3.8).The waste to be disposed contains significant long-lived radioactive components, although high-level,heat-generating wastes are excluded. The waste is being placed into caverns excavated at a depth of 650meters below ground in a salt formation. This is the first purpose-built, deep geologic repository for long-lived wastes in the world. Figure 3.11 presents the general approach for geologic disposal of high-levelwaste. Table 3.5 provides an indication of the status of deep geological disposal programs in severalOECD Member Countries.

3.5 PERFORMANCE ASSESSMENT AND INSTRUMENTATION CONSIDERATIONS

Concrete barriers play an important role in the management of fuel-cycle related waste,particularly for low- and intermediate level wastes. Performance assessments must be conductedrecognizing both the short- and long-term functions of the vault and waste forms. In the short-term, thevault provides a physical barrier to flow into the facility while many of the short-lived radionuclides decayto insignificant levels. Eventually the physical barriers may fail while small quantities of long-livedradionuclides may still be present. During the longer times the vault can serve as a chemical barrier totransport. The effectiveness of chemical features of the vault can be expected to persist much longer thanphysical features.

3.5.1 Performance Assessment

While the radionuclides decay to insignificant levels, it is important that the concrete continue toprovide a physical barrier to their release. Detection and assessment of the magnitude and rate ofoccurrence of any environmental factor-related degradation are key factors in maintaining the capability ofreinforced concrete structures to meet their operational requirements. It is desirable to have an evaluationmethodology that, given the required data, provides the procedures for performing both a current conditionassessment as well as “certifying” future performance. Such a methodology would integrate service

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history, material and geometry characteristics, current damage, structural analyses, and a comprehensivedegradation model. Reliability methods can be utilized to incorporate uncertainties into the assessments.

Due to the longevity requirements and importance of the structures being considered, conditionassessment and performance monitoring are considered to be of prime importance. Some guidance forgeneral civil engineering structures is available through organizations such as the American ConcreteInstitute, the American Society of Civil Engineers, the Institution of Civil Engineers, and the InternationalUnion of Testing and Research Laboratories for Materials and Structures. Unfortunately, little informationis currently available that addresses condition assessments of concrete structures contained as part of aradioactive waste management facility.

3.5.2 Instrumentation

Normally, visual inspection provides the basis of a performance or condition assessment programfor civil engineering structures, however, access to the radioactive waste storage and disposal facilities inall likelihood will be limited, and in some cases may be impossible. Monitoring and instrumentationtherefore must be relied upon to develop the required data for use in performance assessments as well asproviding data for development and refinement of models for estimating future performance. In order toprovide reasonable assurance of the long-term performance of a waste storage or disposal facility, amonitoring and performance verification program should be developed and implemented that will measure,record, and develop a database of performance indicators.∗ The program should specify the types ofinstrumentation to be installed on or within the concrete structure, installation details and locations,calibration procedures, and the frequency and duration of measurements. In addition, a methodologyshould be defined for establishing the criteria to be used to define the pertinent parameters to be monitoredas well as establishing criteria (or limiting values) for use in evaluating the data provided. Theperformance indicators can then be used to assess facility performance as designed and constructed againstpre-established performance measures. The data can also be used to validate and update models that havebeen used to estimate the performance of the facilities.

Use of instrumentation for performance monitoring of the reinforced concrete structures inradioactive waste management facilities is therefore considered to be a high priority issue. Table 3.6identifies several types of instrumentation and the pertinent parameters that can be assessed. Althoughinstrumentation has been used with good success to monitor the performance of nuclear power plant-related prestressed concrete pressure vessels in the UK and nuclear power plant containments in France,

Belgium, and other countries,39

there are several concerns that need to be addressed: long-term stabilityand validation of the instrumentation, environmental effects, backfitting of instrumentation to existingfacilities, maintaining state-of-the-art systems (e.g., updatable software), and replaceability. Descriptionsof instrumentation that could potentially be used to monitor the performance of radioactive waste

management facilities are available.39,40

Only one study has been identified in which instrumentation hasbeen specifically applied to monitoring performance of radioactive waste management-related reinforcedconcrete facilities.

Researchers in Spain are addressing some of the above concerns through a study designed tomonitor the in situ aging of a concrete pilot container similar to the containers being used for low- and

∗ Monitoring related to sampling of the soil and groundwater in the vicinity of the waste storage or disposal

facility will not be addressed. It is acknowledged that this type of monitoring is important to an overallassessment since it can identify transport from the facility and the occurrence of ions that can potentiallylead to performance degradation, but the present discussion will be restricted to the role of instrumentationin assessing the performance of these facilities while they function as physical barriers. The concretestructures also act as chemical barriers to release of radionuclides.

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intermediate-level wastes at El Cabril.41

At the site a pilot container (Figure 3.12(a)) has beeninstrumented and buried in similar conditions to the actual containers at the site in order to determine howthe corrosion parameters evolve in the possible anerobic conditions and of relatively high humidityresulting from rain impregnating the soil. Sensors embedded during fabrication of the concrete containerincluded Cu/Constantan (temperature), vibrating wire strain gage (deformation), reference electrode ofMn/MnOMn and Ti wires (corrosion potential), small disc of Ti wire as reference (corrosion rate andconcrete resistivity), two identical small bars (concrete resistivity), and two identical small bars with Ti andMn/OMn as reference (oxygen availability), (Figure 3.12(b)). The sensors were introduced in three facesof the reinforced concrete container at three different levels and six groups were placed on the surface ofselected drums placed into the concrete container (Figure 3.12(c)). For measuring the corrosion potentialand the corrosion rate, either the main steel reinforcement of the container or the surface of the drum areused as working electrodes (Figure 3.12(d)). In order to obtain the real value of the corrosion rate, themeasurement used is the polarization resistance. Resistivity is measured by means of the currentinterruption method in a galvanostatic pulse. Oxygen flow at the steel reinforcement level is measured byapplying a cathodic constant potential of about –750mV (SCE) and measuring the current of reduction ofoxygen. Continuous monitoring of all parameters and recording of data is done by a GEOLOGGER.Temperature, deformation, and corrosion potential are presently measured four times per day, withresistivity, oxygen flow, and corrosion rate measured once per day. Of the 27 groups of sensors installed,less than 10% failed during the first year after installation. After five years, results indicate that: oxygenavailability at the level of the steel reinforcement is steadily decreasing, although it appears to be affectedby the seasonal cycles; temperature and deformation results have always shown good response and arevery indicative of the wet-dry events; the Mn/OMn sensors have been more stable for recording thecorrosion potential, however the Ti wires have exhibited superior behavior in potentiostatic measurements(oxygen availability); and the GEOLOGGER has worked satisfactorily, although there have been someproblems with noise that required shielding and filtering, and multiplexers became damaged requiringreplacement. The instrumentation under evaluation also offers the potential of adhering sensors to existingstructures in order to confirm operation, verify service life assumptions, or monitoring performance ofdecommissioned structures.

3.6 MODELING FOR SERVICE LIFE ESTIMATIONS

The long-term safety of radioactive waste storage and disposal facilities can be systematicallyassessed through predictive modeling of the gradual failure of the engineered barriers (i.e., waste form,waste package, and backfill) and subsequent transport to man’s environment of radionuclides bycirculating groundwater. Such safety assessments must be based on a good physical understanding of theprocesses involved in the release and transport of radionuclides, as well as those acting on, or likely to act

on, the repository and the geological formation.7

In addition, the potential interplay between theseprocesses must be understood. Validation and refinement of these models requires substantial siteinvestigation efforts that will involve the collection of data at the surface as well as in situ.

Concrete barriers will play an important role in all low-level radioactive waste disposal facilitiesas structural components and barriers to fluid flow and mass transport of radionuclides. Analysis of therole of the concrete barriers in low-level waste isolation requires that performance assessment models beapplied to concrete degradation. Mathematical models are key components of the safety assessments ofwaste disposal options. Due to the complexity of the problem, a series of chain models is required todescribe the long-term degradation of the engineered barriers and the eventual release of radionuclides intogroundwater and their transport to the biosphere. Any viable design method or assessment of service lifeinvolves a number of essential elements: a behavioral model, acceptance criteria defining satisfactoryperformance, loads under which these criteria should be satisfied, relevant characteristic materialproperties, and factors or margins of safety that take into account uncertainties in the overall system.Development of a model for predicting durability or service life is intimately linked to knowledge of thedegradation mechanisms affecting the concrete. The accuracy of predictions based on the model will

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depend on the degree to which the known mechanisms represent the actual damage production, as well ason the quality of the input data. Modeling must use information that results from testing, however, mostexisting tests do not incorporate all factors of importance to service life, those factors that are included maynot be related quantitatively to in-service exposure conditions, and data covering the time period of interestare nonexistant so that extrapolated results have a high level of uncertainity. At the present state-of-the-art,models of the degradation process tend to be somewhat empirical and the primary function of the codes forestimating service life would be for comparative purposes (e.g., alternative design approaches).

As a preliminary step, several OECD countries have developed Underground ResearchLaboratories in representative geological environments (e.g., salt, crystalline rock, and agrillaceous rock)to demonstrate the safety of the geological disposal option and develop data such as could be used in

modeling studies.7 These laboratories are being used to provide data in support of generic safety

assessments, to evaluate engineering feasibility, and to develop and refine techniques for site investigation.Laboratories in Belgium (Mol), Canada (Lac du Bonnet), the Federal Republic of Germany (Asse),Sweden (Stripa), and Switzerland (Grimsel) have been the focus of major international or bilateralcooperative research programs. Also, several large-scale performance assessments have been conducted(e.g., the Swedish KBS-3 analysis of spent fuel disposal in crystalline rock; the Swiss Project Gewähr 1985analysis of vitrified high-level waste disposal in a crystalline basement with overlaying sediments; and thePAGIS study of clay, crystalline, salt, and sub-seabed option by the Commission of EuropeanCommunities). Some input for help in validating models as well as the concept of deep geologicalrepositories is also available through natural analogues (e.g., modeling of movement of natural radioactiveelements over several million years and effective isolation of a uranium deposit such as in Canada at adepth of 430 meters that is surrounded by clay 5 to 30 meters thick).

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Table 3.1

Typical Facilities Associated With the United Kingdom Fuel Cycle

1. Enrichment process Not significant to present study

2. Fuel Manufacture Not significant in present study, except with respect toMixed Oxide Fuel (MOX) (see below)

3. Power Stations

3.1 Fuel Ponds/Dry FuelStore

For short-term storage of irradiated fuels prior to transfer toreprocessing/long-term storage.

3.2 Radwaste Storesincluding Vaults/Silos

Short-term stores pending disposal as low-level waste orintermediate storage pending treatment and transfer to long-term storage as intermediate-level waste.

4. Reprocessing facilities

4.1 Ponds For interim storage of fuel after receipt from nuclear powerstations, but prior to reprocessing

4.2 Cells/Vaults Cladding is stripped from fuel elements, and reprocessingis itself carried out

Encapsulation/vitrification carried out

4.3 Vaults/Silos Fuel element cladding and debris are stored prior toencapsulation and long-term storage

Resins and other bye-product liquors are stored prior totreatment/encapsulation and long-term storage

4.4 Miscellaneous Storage Short- and interim-storage pending disposal

4.5 MOX Fuel Production Use of recycled materials to manufacture new fuel

5. Interim/Long Term StorageFacilities

Encapsulated LLW/ILW including PCM/HLW

Nonreprocessed fuel

Reactors (“Safestor”) – residual bioshield, pressure vesseland core after removal of fuel

6. Repositories Near surface repository for encapsulated LLW

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Table 3.2Summary and Description of Typical Fuel Cycle Facilities

Facility Description and Commentary

On-Site Vaults Generally reinforced concrete; dry by intent; resin tanks and low-levelliquids; temperature < 100°C, low level irradiation; subject to water ingress,moisture, microbiological; lifetime < 200 years

Ponds Reinforced concrete with or without liner (metallic or nonmetallic) andsingle or double containment; water quality controlled; temperaturecontrolled to near ambient; irradiation of moisture barrier may be problem;subject to low level irradiation at floor level, water ingress, microbiologicalattack near free surface, chemical attack (sulfate and chlorides); lifetime 50to 150 years

Dry Fuel Storage Reinforced concrete; temperature < 40°C; subject to low level irradiation,chemical attack (CO and chlorides), water ingress, moisture, microbio-logical attack, external event loadings; lifetime < 200 years

Dismantling Cells Reinforced concrete with stainless steel liner; temperature < 40°C; subject toirradiation, water ingress (if ground level); lifetime < 70 years

Low-Level Waste Storage Reinforced concrete; temperature ambient; no concerns; lifetime < 50 years

Intermediate-Level Waste Storage

Reinforced concrete without liner; new designs address contaminatedwastes; temperature < 40°C; resins drummed or put into boxes; subject tolow level irradiation, moisture, water ingress, chemical attack from someliquids; lifetime < 100 years

High-Level Waste Storage

Reinforced concrete with or without liner; fuel waste or liquid stored asliquid or vitrified glass; temperature < 100°C; subject to water ingress withno specific external loading; lifetime < 100 years (policy dictated)

Reprocessed Materials Treat same as intermediate-level waste

Mixed Oxide (MOX) FManufacture

Reinforced concrete walls and roof to provide shielding and preventunauthorized access; ambient conditions; no contact with products; lifetime< 70 years

Fuel Storage Effectively same as dry fuel storage

Surface Disposal – LLW(ILW not considered aHLW not applicable)

Reinforced concrete; temperature ambient; subject to chemicals frgroundwater or site contents, water ingress, and biological from contents;lifetime < 500 years

Deep Disposal HLW (mass concrete); temperature < 100°C; subject to chemicaldeterioration, water ingress, hydrostatic (geological) loads, moistureingress or egress, and possibly biological attack; lifetime from 100 year(minimum) to infinity

Decommissioned Plants Reinforced concrete that may be lined or unlined; temperature ambient; lowirradiation; type of ILW store; subject to chemical attack (sulfates andchlorides), water ingress, moisture, microbiological, creep reversal; lifetime< 200 years

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Table 3.3Example of Type of Containers for Packaging Low- and Intermediate-Level Waste

Waste

Container

Typical Contents Overall Dimensions Gross Mass*

500 litre Drum The normal container formost operational ILW

800-mm diameter by 1200-mm high

2,000 kg

3 m3 Box A larger container for solidwastes

1720 mm by 1720 mm planby 1225-mm high

12,000 kg

3 m3 Drum A larger container for in-drum mixing andsolidification of liquid andsludge type wastes

1720-mm diameter by 1225-mm high

12,000 kg

4 meter Box For larger items of waste,especially fromdecommissioningoperations

4013 mm by 2438 mm planby 2200-mm high

65,000 kg

4 meter LLWBox

For LLW 4013 mm by 2438 mm planby 2200-mm high

30,000 kg

2 meter LLWBox

For LLW (dense materials) 1960 mm by 2438 mm planby 2200-mm high

30,000 kg

*Maximum mass when filled with conditioned waste.

Source: S. V. Barlow and J. D. Palmer, “The Packaging of Waste for Safe Long-Term Management”NIREX Report N/006, Nirex, Harwell, Oxfordshire, United Kingdom, May 2000.

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Table 3.4

Examples of Dry Spent Fuel Storage Designs

Vendor Model(Storage design)

Capacity(No. of Assemblies)

Transnuclear West, Inc. NUHOMS-7 (Horizontal ConcreteModule)

7 PWR

Foster Wheeler EnergyApplications, Inc.

Modular VaultDry Store (Concrete Vault)

83 PWR or 150 BWR

Transnuclear West, Inc. NUHOMS-24P (HorizontalConcrete Module)

24 PWR

NAC International NAC-128/ST (Vertical Metal Cask) 28 PWR

BNFL Fuel Solutions Corp. Transtor (Vertical Metal/ConcreteCask)

24 PWR

General Nuclear Systems, Inc. CASTOR V/21 (Vertical MetalCask)

21 PWR

Westinghouse Electric MC-10 (Vertical Metal Cask) 24 PWR

NAC International NAC S/T (Vertical Metal Cask) 26 PWRNAC International NAC-C28 S/T (Vertical Metal

Cask)28 Canisters (Fuel rods

from 56 PWRassemblies)

Transnuclear West, Inc. Standardized NUHOMS-24P,NUHOMS-52B (Horizontal

Concrete Module)

24 PWR52 BWR

Transnuclear Inc. TN-24 (Vertical Metal Cask) 24 PWRBNFL Fuel Solutions Corp. VSC-24 (Vertical Metal/Concrete

Cask)24 PWR

Holtec International HI-STAR 100 (Vertical MetalCask)

24 PWR68 BWR

Holtec International HI-STORM 100 (VerticalMetal/Concrete Cask)

24 PWR68 BWR

NAC International NAC-UMS (VerticalMetal/Concrete Cask)

24 PWR56 BWR

Transnuclear Inc. TN-32 (Vertical Metal Cask) 32 PWRNAC International NAC-MPC (Vertical

Metal/Concrete Cask)34-36 PWR

BNFL Fuel solutions Fuel solutions (VerticalMetal/Concrete Cask)

21 PWR74 BWR

Transnuclear Inc. TN-68 (Vertical Metal Cask) 68 BWR

Source: U.S. Nuclear Regulatory Commission Internet Site;http://www.nrc.gov/OPA/reports/cask.htm#caskinfo)

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Table 3.5Status of Deep Geologic Repositories for Several OECD Member Countries

Country Planned Year Status of Program

Belgium 2020 Underground laboratory at Mol

Canada 2020 Independent commission conducting studyof government plan to bury irradiated fuel ingranite at yet-to-be-identified site

Finland 2020 Field studies being conducted; final siteselection due in 2000

France 2010 Two laboratory sites to be selected forstudies; final disposal site to be selected in2010

Germany 2008 Gorleben salt dome site to be studied

Italy 2040 Irradiated fuel to be reprocessed and wastestored 50-60 years before burial in clay orgranite

Japan 2030-2045 Limited site studies. Cooperative programwith China to build underground researchfacility

Netherlands 2040 Interim storage of reprocessing wastes 50-100 years before eventual burial, possiblesub-seabed or in another country

Spain 2020 Burial in unidentified clay, granite, or saltformation

Sweden 2015 Site investigations at three sites in 2002;detailed investigation at one site in 2009;evaluation studies underway at Aspo nearOskarshamn nuclear complex

United States 2010 Yucca Mountain, Nevada site being studiedand if approved to receive 70,000 tons ofwaste

United Kingdom 2010 Irradiated fuel to be reprocessed. Deepunderground disposal for radioactive wastes(excluding high-level waste)

Source: Internet Site; http://www. Nuke-energy.com/data/waste.html

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Table 3.6Types of Instrumentation Available and Applicability

Parameter Monitored•Monitoring

MethodT I C EL WI M B P/M

Visual X X X X X X X X

Thermocouples X

Fiberoptics X X X X

Ultrasonics X

Coupons♣ X

Water Sampling♣ X X

Piezometer X X

Conductivity Testing X X

Load Cell X

Stress/Strain Gages X X

Acoustic Emission X X

Radiation Meters X

Physical Wipes X

Moisture Detectors X

Corrosion Potential X X X X

BRE Carbonation Meter X X

Conventional Surveying X

Laser Beam X

Physical Concrete Sampling♣ X X X

Radar X

Thermography X X

Displacement X

Tilt Meters X

Magnetic Resonance♣ X X

*T = temperature, I = irradiation, C = chemical, EL = external load, WI = water ingress,M = moisture, B = biological, P/M – physical/mechanical.

♣ Sample required.

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Figure 3.1 Cross section of a typical above-grade vault forstorage of low- and intermediate-level radioactive waste.

Source: N. Chau, R. D. Baird, and V. C. Rogers, “Performance of Reinforced Concrete Structures in Low-Level Radioactive Waste Disposal Units,” Concrete and Grout in Nuclear and Hazardous Waste Disposal,SP-158, American concrete Institute, Farmington Hills, Michigan, 1995.

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Figure 3.3 Aerial view of the concrete containers of low- and medium-levelradioactive wastes in El Cabril (Spain).

Source: C. Andrade, J. Rodríguez, F. Jiménez, J. Palacio, and P. Zuloaga,”Embedded Sensors for ConcreteStructures Instrumentation,” Workshop on the Instrumentation and Monitoring of Concrete Structures,NEA/CSNI/R(2000)15, OECD/NEA, Issy les Moulineaux, France, 2000.

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Figure 3.4 Nirex repository concept for low- and intermediate-level radioactive waste.

(Source: Nirex Internet Site, http://www.nirex.co.uk/publicn/reposit/htm)

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Figure 3.5 Nirex concept for low- and intermediate-level waste emplacement and retrieval.

(Source: Nirex Internet Site, http://www.nirex.co.uk/publicn/reposit/htm)

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Figure 3.6 Central Interstate Compact low-level waste disposal facility.

Source: J. W. Grindstaff, S. C. St. John, and N. J. Antonas, “Considerations for the Design andConstruction of a Reinforced Concrete Low-Level Radioactive Waste Disposal Facility, Concrete andGrout in Nuclear and Hazardous Waste Disposal, SP-158, American Concrete Institute, Farmington Hills,Michigan, 1995.

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Figure 3.7 Aerial view of a disposal pad at the Interim Waste Management Facility.

(Source: C. B. Oland and J. W. Baker, “Concrete Structures for Waste Storage and Disposal, ConcreteInternational, Submitted for Publication, American Concrete Institute, Farmington Hills, Michigan, 2001).

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Figure 3.8 Repository for Transuranic Waste - Waste Isolation Pilot Plant.

(Source: http://www.wipp.carlsbad.nm.us)

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Figure 3.11 General approach for geologic disposal of high-level waste(Schematic of Yucca Mountain High-Level Waste Repository, U.S.).

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4. NATIONAL AND INTERNATIONAL RADIOACTIVEWASTE MANAGEMENT ACTIVITIES

At the end of 1998, there were 345 nuclear power plants connected to the grid within OECDMember Countries, representing a total capacity of 292 GW(e) and generating some 24% of the electricity

produced by these countries.42

One third of these plants have been in operation for over 20 years. Up tothe year 2015, 85 nuclear power plants with total net capacity of 48 GW(e) representing 16% of the presentnet nuclear capacity are scheduled to be taken out of service in OECD Member countries. Most of theseare planned to retire before each lifetime reaches 40 years. Over the next decade, electricity capacity in

Member Countries is expected to grow at a rate of 1.4% per year in order to keep up with demand.43

Operation of the 427 nuclear power plants worldwide resulted in the generation of 10,500 tonnes of spent

fuel in 1997 alone.44

The approach to management of the spent fuel from nuclear power plants varies from country tocountry. For some countries, engaged in once-through cycles using the direct disposal option, spent fuelwill be packaged and disposed in underground sites after a sufficient cooling period in surface-basedfacilities. For these countries, spent fuel is labeled as high-level waste. On the other hand, for countriesthat have embarked on the recycling strategy, valuable materials contained in the spent fuel (e.g., uraniumand plutonium) are separated and reused in conventional nuclear fuels, while the ultimate residues areproperly treated and conditioned before being stored and eventually disposed. In such countries, thecategories for nuclear waste directly pertain to the non-recoverable materials to be disposed after beingproperly treated and conditioned. The wastes are subdivided into low-, intermediate-, and high-levelaccording to their level of reactivity and their life span. Reinforced concrete appears to have greatestapplication to the low- and intermediate-level storage and disposal facilities. Research related to disposalof high-level waste is primarily related to investigation of geologic repositories and development of dataand modeling methods for conduct of safety assessments of these facilities.

4.1 NATIONAL PROGRAMS

Information on several OECD Member Countries approach to storage and disposal of radioactivewaste associated with the nuclear power plant fuel cycle are summarized below.∗ Input was provided fromTask Group members, open-literature sources, and a survey questionnaire that was developed and sent toOECD Member Countries. A copy of the survey questionnaire is provided in Appendix A.

4.1.1 Belgium

In Belgium disposal of waste will be both above and below ground. Regulatory requirementsprimarily address the radiological hazards and there are no specific requirements concerning civilengineering structures for use in waste management. Reversibility of the high-level repository facilites isnot presently considered. Planned storage is for a 200 to 300 year time frame. Low-level wastes areconsidered to be those that decay to background level under institutional controls within a 200-year timeframe. Surface disposal will be used with the site opened for unrestricted use after 200 years. Sites areselected for low seismicity with draining of upper layer, impermeable lower layer, and above the highestlevel reached by groundwater. High-level waste will utilize vitrified reprocessing with natural barriers

∗ Several activities being conducted as part of national programs addressing aging management of nuclear

power plant concrete structures reported in a prior Task Group Report45

also have application to the presentstudy (e.g., performance of instrumentation, effect of elevated temperature on concrete properties, andperiodic condition assessments). These activities will not be repeated.

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emphasized for disposal. The Mol-Dessel nuclear site has been selected as a potential host formation forthe disposal of high-level and long-lived radioactive waste. An underground laboratory (HADES) hasbeen constructed in Boom Clay at Mol at a depth of 223 m and has been in operation for more than 15years. The facility consists of one shaft and one horizontal gallery. Construction of the cast iron linedaccess shaft through water-bearing sands was performed using a ground freezing technique. Due toswelling of the clay upon freezing, the ground freezing technique was not feasible for digging severalkilometers to form the storage galleries. A small shaft and a 7-m-long exploratory drift with an internaldiameter of 1.4 m were constructed extending from the end of the underground research laboratory to studythe feasibility of digging galleries in unfrozen clay at great depth. The gallery was excavated manually at arate of 25 cm/day. The gallery is lined with 60-cm-thick concrete. In order to implement full-scaledemonstration testing, an extension of the existing facility was necessary that required sinking a secondshaft (3-m diameter) and construction of an 84-m-long gallery to connect the second shaft to the existingfacility. The connecting gallery was excavated by semi-mechanized techniques with an excavation rate ofabout 2 m/day.

At the HADES Laboratory there has been fissurization in some of the blocks in the test drift,primarily due to high stress at punctual or line contact between blocks. The fissurization was detected byvisual inspections that are conducted on a weekly basis. Convergence data is provided by stress and loadcells. No corrective actions have been required to date. In general, research has concentrated on chemicalprocesses rather than physical and material characteristics. Specific research activities have included:assessment, primarily through literature review, of the lifetime of concrete structures in terms of safetyassessment studies for waste disposal; leaching of different types of cement paste (cemented resins andconcentrates from the nuclear plants at Doel and Tihange); long-term soil-structure interaction consideringthe elasto-viscoplastic behavior of the clay and the dissipation of the pore water pressure; and developmentof data from monitored facilities (e.g., visual inspection, total pressure, and convergence measurements).

4.1.2 Canada

Radioactive waste produced in Canada at all stages of the fuel cycle is managed according to acomprehensive set of policies and programs. Current low- and intermediate-level waste resulting fromnuclear power plants is stored at the Radioactive Waste Operations Site at the Bruce reactor site. Sitecharacterization is in progress to locate a near-surface disposal facility at this location by 2002. AtomicEnergy of Canada Limited (AECL) has been responsible for generic research for permanent disposal ofnuclear fuel waste. Ontario Power Generation (formerly Ontario Hydro) has been responsible for researchon transportation and storage of such waste. Research on storage and disposal of low- and intermediate-level waste is carried out by both organizations. AECL is developing plans for a deep geologic repositoryin which used fuel would be buried in a vault 500 to 1000 meters deep in plutonic rock of the CanadianPrecambrian Shield. It is expected that there will be one central disposal facility that can accommodate allused fuel produced over almost a century of nuclear power generation in Canada. A multiple barrierapproach that includes both engineered and natural barriers will be used to isolate the fuel. The wasteform, containers, buffer, backfill, and rock mass between the disposal vault and the biosphere are thebarriers. The containers housing the sealed waste resulting from either used Canadian Deuterium Uraniumfuel or solidified high-level wastes from reprocessing this fuel are designed to last at least 500 years. Thecontainers will be placed into disposal vaults or boreholes drilled from the rooms. Each container issurrounded by a buffer layer of clay-based material and after filling each room will be sealed with backfilland other vault seals. All tunnels, shafts and exploration boreholes will be sealed in such a way that apassively safe facility requiring no insitutional controls will result. Computer modeling, utiliznginformation from field and laboratory studies, has been used to assess performance of the entire disposalsystem. Spent fuel is currently stored on-site at the plants, either in pools or in dry concrete casks.

Radioactive waste from AECL’s laboratories and from small users are to be disposed of in anear-surface facility at Chalk River where it is are currently in storage in concrete trenches, containers, and

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buildings. The planned Chalk River facility is based on reinforced concrete in-ground modules, each being30-m long by 20-meter wide by 9-meter deep. Each module will have a capacity of 2,000 m3 of packagewaste in the form of drums, bales, and standardized boxes. Spaces between the waste packages will befilled with sand and the base of each unit will be of compacted buffer material. During the operatingphase, the facility will have a light building over it with a crane for handling the waste packages. After themodules are filled, this structure will be removed for use on other modules and the facility covered with aconcrete cap overlain by an engineered cover containing barrier and drainage features.

4.1.3 Czech Republic

In the Czech Republic the only regulatory body for all nuclear matters is the State Office forNuclear Safety and Radiation Protection (SJUB), and the state organization Radioactive Waste RepositoryAuthority (RAWRA) is responsible for waste management and disposal. Between 1953 and 1963 some200 m3 of so-called institutional wastes were placed into an abandoned limestone mine (Hostim) and in1964 another mine (Richard II) was developed about 50 miles north of Prague as a repository forartificially-produced radionuclides. A facility for disposal of naturally-occurring radioactive materialwastes and some sealed sources has also been in operation since 1972 in an abandoned uranium mine atBratrstvi. An above-ground concrete vault facility for disposal of low-level waste and short-livedintermediate level waste has been built next to the Dukovany nuclear power plant. Design life is 400 yearsconsisting of 100 years operation and 300 years institutional control. Low- and intermediate-level wastesare bitumenized and stored in four rows of reinforced concrete storage pits having a sloped concrete baseto facilitate drainage into collection reservoirs and four underlying layers for drainage and support. Sealedconcrete walls are 70-cm thick and 30 cm of sealed concrete is used for closure. Each of the 112 pits hasstorage capacity for 1,600 barrels with shielding provided by transport casks containing several barrels.Storage in the concrete vaults is permanent. Monitoring is to be used as part of the overall strategy foraging management of the concrete vault reposistory at Dukovany. Studies have been performed relative tomeasurement of concrete properties (e.g., carbonation depth, porosity, and hardness). Nonregularinspections are performed in conjunction with performance assessment studies. Potential degradationmechanisms considered include concrete carbonation, steel reinforcement corrosion, and microbiologicalattack. Degradation experienced to date has been limited to concrete carbonation, but no corrective actionshave been required. Detection methods utilized include Schmidt hammer measurements, concrete coverthickness, capillary pore volume, and depth of carbonation. Areas where additional research is requiredinclude systematic studies of coupled degradation processes such as carbonation and microbiologicalattack, and introduction of standardized methods focused on specific safety functions of a given concretestructure such as measurement of radionuclide diffusion in repository concrete materials. At the Dukovanysite, spent fuel is to be stored above ground on site for 40 to 200 years. Studies establishing a site for adeep geologic repository started in 1993. Included was the national policy calling for deep disposal ofintermediate- and high-level waste. An underground laboratory will be built in the Czech Republic on thesite chosen for a deep geological repository (http://www.surao.cz/english/index-en.html.).

4.1.4 Finland

There are two commercial reactor sites, each with two reactors, one at Loviisa, operated byImatra Power Company (IVO), with Russian VVER’s and one at Olkiluoto, operated by Industrial PowerCompany Limited (TVO), with BWR’s. The utilities are responsible for the safe management of wastesand for the necessary research and development, as well as covering the costs of the complete operation.In 1978 the two utilities established a joint body to coordinate research and development activities. IVOand TVO each have developed their own facilities for disposal of low- and intermediate-level waste at theirreactor sites. IVO completed construction of a rock cavern repository for the low- and intermediate-levelwastes from the Loviisa plant in 1996. The repository was excavated to a depth of 110 meters where it isbelow a gently dipping fracture zone that constitutes a boundary between an overlying zone of fresh waterand a lower zone of very saline groundwater. The present repository consists of two finalized rock

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caverns, one for dry maintenance waste and one for immobilized wet waste. Construction and installationwork of engineered barriers for the intermediate-level cavern, consisting mainly of reinforced concretestructures, will be completed at a later date. At Olkiluoto, a repository for low- and intermediate-levelwaste has been in operation since 1992. The repository is similar to that at Loviisa except the rock massfavors vertical silo-type caverns rather than horizontal tunnels such as used at Loviisa. The repositoryconsists of two silos at a depth of 60-100 meters. One of the silos will be used for bitumenized waste andconsists of a thick-walled silo inside the rock cavern. Space between the concrete silo and cavern is filledwith crushed rock. All waste will be placed into concrete boxes that can each contain up to 16 drums.

Spent nuclear fuel at both plants is presently maintained in fuel pools on site. At Olkiluto thefuel pools are 13.5 meters deep and have stainless steel clad walls of thick reinforced concrete. The poolsrest on bedrock and are located below ground level. Spent fuel will be encapsulated in a copper-ironcanister consisting of an inner cast iron insert as a load-bearing element and an outer container of oxygen-free copper to provide a shield against corrosion. The canisters will be placed into boreholes drilled in thefloors of tunnels that are constructed at a depth of about 500 meters in crystalline rock of good quality.The annulus between the canister and the rock walls of the boreholes will be filled with compactedbetonite. During closure the tunnels will be backfilled with a mixture of sand and betonite. Disposal of thespent fuel will be handled through a joint company, Possiva, that is carrying out siting and developmentwork for the deep repository. Experiments consisting of full-scale demonstrations of boring of disposalholes for spent fuel canisters are being done in a research tunnel located in conjunction with the low- andintermediate-level repository at Olkiluto. Eventual wastes from decommissioning of the reactors will bedisposed of in underground repositories for operational reactor wastes at the power plant sites.

4.1.5 France

France in 1979 set up a special agency, National Agency for Radioactive Waste Management(ANDRA), within the Atomic Energy Commission (CEA) to be responsible for designing, constructingand operating long-term disposal facilities as well as undertaking all necessary studies to this end.Legislation in 1991 established ANDRA as a public service company separate from the CEA undersupervision of Ministries of Industry, of Environment, and of Research. The 1991 legislation requiredimplementation of a 15-year research program in three different areas: (1) research of solutions to separateand transmute long-life radionuclides in the waste; (2) studies of retrievable and non-retrievable disposal indeep geologic layers with the help of underground laboratories; and (3) studies of the processes forconditioning and long-term surface storage of waste. ANDRA establishes specifications for waste formand disposal design, siting, construction, and operation of repository. Activities have addressed siteselection with three sites selected in 1996 for consideration: granite below sedimentary cover at laChapelle-Baton, in the Vienne; in marl at Bure, in the Haute-Marne; and in marl at Chusclan, in the Gard,near Marcoule.

The Manche disposal facility, described previously, is now entering the institutional controlperiod after 25 years of operation and receiving over 500,000 m3 of wastes. The first packages of wastewere disposed in ground, and the rain water retrieved in ditches and then controlled in a reservoir. Thisdisposal concept was abandoned in favor of a disposal system on platforms embedded in bitumen andequipped with a water drainage system. Packages with the most radiation were stabilized with cement andstored in concrete containers. Installation of cover representing the last step in closure was completed inJune 1977. The Aube disposal facility, also described previously, with a capacity for 1,000,000 m3 ofwaste has receiving waste packages treated in conditioning facilities (compaction or grouting) since 1992.A 60-year operating period is envisaged for this facility. The wastes are protected by a series of barriersmade by the package and its conditioning (wastes are embedded in their packages), by the disposalstructures (disposal cells, underground surveillance galleries and final cover), and the geologicalenvironment of the site.

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ANDRA for several years has been in charge of developing a surface storage concept specific tovery low-level waste forms. A project to locate the storage center near the Aube center is under study,both because of its local geological qualities and the already existing competence and infrastructure. Thevery low-level waste will result from dismantling of nuclear power plants, research laboratories, andresearch and nuclear production centers; other industries such as the food, chemcial, or metallurgicalindustry, whose fabrication processes concentrate the natural radioactivity present in some materials; andthe cleaning and rehabilitation of sites formerly polluted. After inerting or stabilizing, the waste will bestored in cells dug into the impermeable clay (15 meter thick), the bottom of which will be prepared toretrieve water that might have filtered during the storage time. The cells are isolated from the environmentby a synthetic membrane placed on a thick layer of impermeable clay which is then protected by a coveralso consisting of clay. A control system will allow the storage to be monitored and to retrieve the waste ifnecessary. The center will be able to handle about 25,000 tons of very low-level waste per year. It isanticipated that the facility could be operated as early as 2003.

ANDRA was authorized in 1999 to build a research laboratory at a depth of 500 meters in a 150-million year old clay formation in eastern France (Meuse/Houte-Marne). The underground researchlaboratory is a research tool and a multidisciplinary laboratory. Two separate galleries, oriented upwardsand downwards respectively, of a few hundred meters length will enable the clay formation to be surveyedin the northwestern and northeastern directions, chosen on the basis of regional faults. At a depth of about445 meters, a part of the rock will also be excavated to observe and measure the upper part of theformation. At 490 meters depth a network of excavated galleries will constitute the core of theunderground laboratory. The profile of the galleries will be that of a horseshoe 4-meters high. Steel archesspaced at about 1-meter intervals will ensure proper ground support. The floor will be covered withconcrete and, if necessary, arches will be supplemented by anchor bolts and concrete. A specificexperiment involving excavation of an experimental gallery will be conducted in the laboratory. It willalso allow for a full-scale study of the disturbances on the host rock due to the excavation work. Theprinciple of the experiment is to make boreholes in order to measure the hydro-mechanical characteristicsbefore opening the experimental gallery. These parallel or perpendicular boreholes will be about 20 to 40meters long. Parameters to be measured include permeability, rock deformation, and sound speed asindicators of rock fractures. By comparing measurements before and after the excavation, it will bepossible to distinguish the effects directly induced by the excavation and to assess evolution of thepotential damage to the rock wall. Throughout the operational phase of the underground laboratory thebehavior of all the engineered structures will be followed by various measuring devices (e.g., deformationsand stresses) installed at regular intervals (i.e, 20-meter intervals in wells and every 40 meters in galleries).Results from the underground research laboratory will be used to help make a decision, planned by 2006,on the creation of a deep geological underground disposal center. The disposal concept will be based onthe principle of multiple barriers that include the waste package (waste and the material used to stabilize itin an adequate package), the structural barrier which is placed between the waste package and the rock, andthe geological barrier rock itself. In addition to deep geological disposal, separation-transmutation is alsobeing pursued to reduce the quantity of long-lived radioactive wastes at the source.

4.1.6 Germany

The energy policy of the German Federal Government includes, firstly, putting an end to thegeneration of electricity in nuclear plants – if possible in concurrence with the power utility industry – and,secondly, drafting of a new national plan for the treatment and disposal of radioactive waste material. In1998, 19 nuclear power plants were connected to the German power grid and producing 145.2 TWh ofelectrical energy (~33% of total domestic power generation). Depending on its initial enrichment and finalburn-up, the annual discharge of spent fuel amounts to about 360 – 450 tonnes. To date, about 7,300tonnes, including some 70 tonnes of plutonium, have been discharged. Spent fuel from light-water reactorsis managed in two ways: (1) reprocessing with subsequent disposal of fission products and actinides, and

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(2) direct disposal. The government is planning on establishing direct repository as the only disposaloption.

Low-level waste has been disposed underground in the Asse Salt mine, as a demonstration project,between 1965 and 1978. Asse has been used for disposal of about 141,000 drums of low- and short-livedintermediate-level waste. It is now used as a research and development facility for testing boreholeemplacement of high-level waste and the retrievability of intermediate-level waste. Short-lived low- andintermediate-level waste has been stored at the Morsleben facility contained in an abandoned salt mine.Both heat-generating (1,200 – 1,500 m deep) and non-heat-generating (870 m) waste can be stored atGorleben. The Gorleben district has a salt dome 250 to 3,000 meters under the earth’s surface.Goreleben is planned as a repository for high-level waste produced from reprocessing. The repository willhave a square gallery approximately 7-m wide by 8-m high extending 400 to 600 meters into the salt domeat a depth of 880 meters. Two shafts, galleries, and drillings are to be used to investigate the salt dome.The exploratory level is 840-m deep. Gorleben also hosts a dry store for spent fuel with the fuel placed inabove-ground concrete buildings having natural air convective cooling. The Federal Government,however, questions the suitability of Gorleben as a repository and intends to have further sites in differenthost rocks investigated.

Morsleben, a deep repository (>500 m) for low- and medium-level waste, has operated in a saltdome since 1987. Nonheat-generating waste will be stored at a depth of 1000 meters at the Konradfacility, an abandoned iron ore mine. Interim storage of spent fuel elements is taking place at Gorlebenand Ahaus, which are of similar design. The total area of the Ahaus facility is approximately 18 ha and theone-story storage building is made of reinforced concrete approximately 200-m wide by 38-m deep by 20-m high. Dry storage is used with ventilation depending on natural circulation. A total of 420 light-waterreactor fuel element casks can be stored. Due to decommissioning of the high-temperature gas-cooledreactor and the thorium high-temperature reactor, 50 casks of the 420 are allocated to these reactors. Asmall amount of heat is released by these casks that in the winter has resulted in condensation developingof the surface of the casks to produce some surface rust. These casks are inspected and recoated asrequired. Condensation has not been a problem for the light-water reactor casks.

In view of policy laid down in the coalition agreement of the Federal Government as ofOctober 20, 1998, namely, to put an end to the utilization of nuclear energy in Germany as soon as possibleand, subsequently to strive for the introduction of new energy structures and, substantially reducingfunding for nuclear reactor safety and repository research in the next few years, the Federal Minister ofEconomics and Technology (BMWi) had the entire field subjected to a review. Repository research in linewith recommendations resulting from the BWMi study is conducted at Forschungszentrum Jülich (FZJ),Forschungszentrum Karlsruhe (FZK), Forschungszentrum Rossendorf (FZR), Gesellschaft für Anlagen-und Reaktorsicherheit (GRS), and Bundesanstalt für Geowissenschaftenund Rohstoffe (BGR). At FZJwork is addressing reduction of generation of long-lived nuclides, destruction of plutonium in safe nuclearfacilities, and modeling of different mobilization, retention, and transport mechanisms. As Germany hasnot decided on the siting for a high-level repository, the development of accident scenarios at FZK aims atdemonstrating site-independent long-term safety (i.e., transmutation of long-lived actinides and fissionproducts, and immobilization of plutonium). The behavior of actinides in different barriers is beinginvestigated. Research related to immobilization of high-level waste aims at the development of adaptedtechnologies for the vitrification of high-level waste from reprocessing. At FZR assessments of long-termrisks of old waste deposits from uranium mining is addressed. GRS primarily concentrates on thedevelopment and testing of methods related to safety and system analysis including necessarycomputational models for release, mobilization, and transport mechanisms of radionuclides from therepository throughout the entire geological system. BGR is addressing geoscientific research and has beeninvolved in site investigations at Konrad, Gorleben, and Morsleben. Areas include preparatory research foralternative rock sites, repository siting for high-level waste in crystalline formations, optimized and appliedcalculation methods to demonstrate stability, site-independent modeling of groundwater movements as a

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function of salinity-dependent water density, and investigations of backfill materials. The comprehensivereview commissioned by BWMi also addressed facilities receiving funds from BWMi and FederalMinistries for Education and Research (BMBF) for nuclear reactor safety research.

4.1.7 Hungary

Low- and intermediate-level radioactive waste treatment and disposal takes place at thePüspökszilágy facility. The waste is disposed in near-surface concrete trenches. Since there is nopermanent disposal site for solid radioactive, intermittently, solid low- and intermediate-level waste fromthe nuclear power plant also has been placed at this site. This practice has stopped and sites are underinvestigation for permanent disposal of low- and intermediate level as well as high-level waste. Disposalof high-level waste is planned for a deep geologic repository. A location in Southwestern Hungary, BodaClaystone Formation in the Mecsek Mountain area, is being considered. The formation dips under auranium ore-bearing sandstone formation where a mine existed for 42 years. As a result of the miningoperation, considerable information about lithology and the structure of the underlying sandstone has beencollected, with four of the mine boreholes having penetrated the formation itself to a few hundred meters.Site characterization studies are being conducted using existing as well as exploratory boreholes.

Spent fuel from the Paks plants (VVER-440) is currently being stored on site. As the spent fuelassemblies are discharged from the reactors they are located one by one into vertically arranged carbonsteel tubes in a storage facility. The storage facility is functionally divided into three major structural units.The first is a vault module where the spent fuel assemblies are stored. The vault module is a structureenclosed by thick reinforced concrete walls and shell structures filled with concrete, the basic function ofwhich is to provide radiation shielding. Each vault is capable of accommodating 450 spent fuelassemblies. Natural draft-driven air flow is used to dissipate heat generated by the fuel assemblies. Thesecond component is a reception building in which the reception, preparation, and unloading of the spentfuel transfer cask takes place. This building is made of a reinforced concrete substructure with thebasement and structural steel superstructure forming a hall. The fuel handling system and the variousauxiliary systems are installed in this building. The third component is the charge hall where the fuel-handling machine travels during fuel transfer operations. The hall is bordered by the reinforced concretewall of the ventilation stack on one side and by a steel structure with steel plate sheeting on the other side.The basic function of the sheeting is to protect the fuel-handling machine against the environment.

4.1.8 Italy

Italy has a considerable amount of radioactive wastes produced over about 30 years of nuclearactivity. In 1996 activities began for site selection and design of a low-level waste repository. Thedisposal concept selected is similar to that of Spain and is based on a modular concept with the concretemodule being the basic disposal unit housing the waste package and the backfill material. This concept isnot considered for disposal of very low-level waste that can be disposed of using a less severe containment.Two main aspects of the module design are under evaluation: (1) behavior of the structure under expectedthermal and mechanical conditions during the operational and institutional period of control, and (2)selection of the appropriate mix design to permit durability over the institutional period of control and itspredictability over the time period considered (thousands of years). Numerical modeling is being used toexamine behavior. An extensive experimental program for selection and characterization of cementitiousmaterials is being conducted by universities and industrial laboratories. Two concrete mixes are underconsideration. One with silica fume and one without. The concretes are being characterized with respect tostrength, leaching, carbon dioxide and chloride penetration, shrinkage, and creep. Preliminary resultsindicate that concrete having water-to-cement ratios less than 0.40, in conjunction with the use of blastfurnace slag-portland cement and fly ash permit manufacture of cementitious barriers having high strengthand long-term durability. Addition of silica fume improves resistance to penetration by carbon dioxide andchlorides. No site is currently available in Italy for disposal.

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4.1.9 Japan

In Japan, low level radioactive waste goes to the Rokkasho Center in Oishitai that is operated byJapan Nuclear Fuel, Ltd. Institutional control is deemed necessary for 300 years. The site is 30-60 mabove sea level and burial is 14 -19 m below surface. There are eight burial groups of five concrete vaults(50-60 cm thick, 24 by 24 by 6 m) where the solidified waste in drums is placed, backfilled with mortar, 2m of betonite-sand placed outside the vault with an additional 4 m soil cover. The waste must be solidified(cement, bitumen, plastic), structurally sound, surface contamination level less than regulations specify,and contain no harmful materials for either the package or facility. Drain pipes run from the vaults to aninspection tunnel.

Spent fuel is reprocessed and vitrified. A repository philosophy has been developed with thefocus on chemical engineering of the near-field environment and design of an engineered-barrier system tocomplement natural retardation by migration. By focusing on the waste package and engineered near-fieldenvironment, Japan hopes to develop a generic approach. The Power Reactor and Nuclear FuelDevelopment Cooperation is performing research and development to study geological environments andto establish the geological disposal technology (e.g., evaluation of deep geologic disposal system, disposaltechnology, survey and study of environmental conditions, and deep geological environment).Geoscientific research in support of a deep geological disposal is being conducted at the Tono GeoscienceCenter where underground water flow and water chemistry are investigated. An underground researchlaboratory is planned at Mizunami in gifu Prefecture to investigate deep crystalline rock structures at thislocation. Reprocessing, vitrification, and storage of waste is the responsibility of the Japan Nuclear FuelService Company. Licensing is to be handled by the Ministry of International Trade and Industry (MITI)and Science and Technology Agency (STA). The tentative target for commissioning of a disposal site inJapan is sometime in the 2030’s or by around 2045 at the latest.

4.1.10 Korea

Overall project management for low- and intermediate-level waste is administered by KoreaElectric Power Company. Activities performed include siting for low- and intermediate-level waste,construction and operation of the disposal facility, and collection and treatment of the wastes. A designconcept for final disposal using either vault or rock cavern storage is under consideration, with theselection based on site conditions. A site is anticipated to be in operation in about 10 years.

High-level waste disposal is planned for a deep geological repository. Activities in support ofthis include repository system development and performance assessment, geoenvironmental scienceresearch, engineered barrier development, underground raeionuclide migration study, and geotechnicalengineering research. The Korea Atomic Energy Research Institute is active in several areas in support ofrepository development – drafting a disposal concept and total system performance assessmentmethodology, survey of uranium mine area for natural analog study and possible host rock,characterization of domestic betonite as a buffer material, sorption study on encapsulation (or backfill)minerals, and migration experiment through geologic medium.

4.1.11 Norway

All radioactive waste in Norway is treated and stored by Institute for Energy Technology. Aradioactive waste facility at Himdalen, about 40 km east of Oslo, is constructed in a rock formation with itsentrance in a hillside. The four waste caverns are accessible through a tunnel that declines slightly towardsthe entrance and has about 50 m rock cover. The waste will be placed into concrete structures withwaterproof cover, and a self-drainage system has been constructed in order to keep the waste and thecaverns in a dry state. With a total capacity of 10,000 drums, the facility is planned for all low- and

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intermediate-level waste generated in Norway until 2030. About 1,000 drums buried in a shallow groundrepository at Kjeller are to be retrieved, treated, and transported to the Himdalen facility.

4.1.12 Spain

Empresa Nacional de Residuos Radioactivos (ENRESA) is responsible for all waste managementactivities in Spain. Spent fuel qualifies as high level waste and will be disposed of directly. The spent fuelinitially will be placed either into onsite cooling pools or storage casks where pool storage space is limited.A dual-purpose metal cask has been developed to increase on-site storage capacity. Spain intends todispose of spent fuel directly in a deep repository. Long-lived intermediate-level waste will also bedisposed in this same facility. Disposal in granite and salt and clay is being considered. For granite a steelcanister embedded in a 0.75 m buffer at a 500 m depth is proposed. Packages are intended to last forgreater than 1,000 years with betonite backfill to provide retardation and mechanical buffer for about10,000 years. Salt storage would be at 850–m depth and would use self-shielded casks. An undergroundlaboratory is planned prior to final repository development to investigate such things as instrumentationand numerical methodologies, feasibility and performance of the engineered barriers at full scale and underrepresentative conditions of temperature and depth, and acquisition of basic data of the most relevantprocesses of the different repository subsystems. On-going activities include full-scale testing ofengineered barriers in granite, large-scale in situ demonstration test for repository sealing in argillaceoushost rock, short-term performance assessment of spent fuel as a waste form, and corrosion evaluation ofmetallic materials for disposal canisters. Potential sites for the repository are being identified. Acentralized interim storage facility is planned to store high-level vitrified waste from Vandellos-1 nuclearpower plant. This facility will also be used to accommodate the spent fuel from nuclear power plantswhere storage capacity has become saturated.

Conditioned low- and intermediate-level wastes are sent to the El Cabril (Cordoba) interimstorage facility. Drums are placed into concrete containers 2 by 2 by 2 m that are placed into 24 by 19 by10 m concrete cells and encapsulated with mortar. The disposal structure has a floor that drains intogallery tunnels running underneath. When decommissioned, alternating permeable and impermeablematerial will cover the structures where they will remain buried for 300 to 500 years. In order to know theintegrity of the concrete in these structures, several studies are being conducted. At the site a pilotcontainer has been instrumented and buried under similar conditions to the real containers in order toascertain how the corrosion parameters evolve in the possible anerobic conditions and relatively highhumidity due to rain water penetrating the ground. Parameters monitored include temperature,deformation, corrosion potential, corrosion rate, concrete resistivity, and oxygen availability. After oneyear, less than 10 % of the 27 groups of sensors installed have failed. Data has been obtained from thepilot container for over five years.

4.1.13 Sweden

Sweden disposes all of its waste domestically without reprocessing. After storage in fuel poolsonsite, the spent nuclear fuel is placed into heavy transport casks, encapsulated, and sent by sea in aspecially designed ship to the Central Interim Storage Facility for Spent Nuclear Fuel (CLAB). CLABprovides interim storage for the high-level waste. Upon arrival at the receiving section the fuel is unloadedunder water with the actual storage chamber located in a rock cavern 25 – 30 m below ground. The 120-mlong rock cavern contains four fuel storage pools and one reserve pool. Extension works are ongoing toexpand the storage capacity by an additional 3,000 tonnes. After about 30 years in CLAB, the spent fuelwill be encapsulated in durable canisters and deposited in a repository. Since the canister should remainintact in the repository environment, it must not be penetrated by corrosion and must be able to toleratemechanical stresses from being located at a depth of about 500 m in the crystalline bedrock. A coppercanister is proposed in order to provide resistance to corrosion with an insert of cast spheroidal graphiteiron for mechanical strength. Estimated life of the canister is millions of years. The final repository for

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short-lived low-level and intermediate-level waste (operational) is Final Repository for RadioactiveOperational Waste (SFR) which is situated near the Forsmark Nuclear Power Station. The waste that isdisposed of in SFR can be divided into waste from nuclear plants (substances in contact with fuel) andwaste from industry, medical care, and research. SFR is situated in crystalline bedrock at a depth of morethan 50 m below the seabed that in turn is under 5 m of water. The facility is connected to the groundsurface by two parallel one-kilometer-long access tunnels. The disposal chambers consist of four 160-m-long rock vaults of varying design, plus a 70-m-high rock cavern in which a concrete silo has been built.Low-level waste enclosed in unopened freight containers is emplaced in one of four rock vaults. Three ofthe rock vaults receive intermediate-level waste. Dewatered filter resins in concrete tanks are deposited intwo of the intermediate-level waste vaults. The third vault contains more hard-to-handle waste that isplaced into pits which are sealed with concrete lids after filling. The 50-m-high by 26-m-diameter byabout 1-m-thick silo is to contain most of the radioactive material and is mainly used for filter resins frompurification of reactor water. The space between the silo wall and rock is filled with bentonite. Internallythe silo is divided into square vertical pits measuring 2.5 m on a side with the pits separated by concretewalls. When the waste arrives at SFR it is enclosed in protective containers, with intermediate-level wastefurther embedded in concrete or asphalt. Other low-level waste is pre-incinerated and comes to SFR inmetal drums. The radioactivity in the waste declines such that it is equivalent to the natural radioactivity inthe rock after 500 years. Waste from decommissioning of nuclear plants is mainly low- and intermediate-level waste and is planned to be deposited in SFR. Once SFR has been sealed and closed, the radioactivematerials are isolated in waste packages that are surrounded by different natural and engineered barriers.Since the silo is located below sea level and groundwater is stagnant, there is no need for institutionalcontrol.

4.1.14 Switzerland

In Switzerland the producers of radioactive waste are responsible for its safe management anddisposal. In 1972 the producers joined together to form the National Cooperative for the Storage ofNuclear Waste (NAGRA) which is responsible for the planning and disposal of all kinds of radioactivewaste. Dedicated companies located at the sites are responsible for the construction and operation ofdisposal facilities. Because of the population density, no shallow land burial is envisaged.

In order to provide interim storage for low- and medium-range radioactive waste until repositoriesare available, the Swiss nuclear power plant operators planned and developed an interim storage facilityZWILAG located in Würenlingen. All categories of radioactive residues of Swiss origin come to thisinstallation for interim storage. This consists mostly of operating waste and fuel elements from nuclearpower plants; waste resulting from reprocessing of spent fuel elements; and radioactive waste frommedical, industrial, and research applications. A pretreatment and incineration plant have been constructedat the site for waste treatment and processing, as well as a reception area for interim storage, and storagebuildings for high-, medium-, and low-level radioactive waste.

Two underground repositories are planned, based on different characteristics of the waste types:one for low- and intermediate-level waste and one for high-level and long-lived intermediate-level waste.Both repositories are to be constructed in a rock formation that is practically impermeable in order toprevent the transport of any radioactivity released from the repository to the surface through groundwater.In addition, geological conditions at the sites must be stable for the required period of isolation, namelyuntil the radioactivity has decayed to safe levels. As understanding of rock properties is of greatimportance for the safe disposal of radioactive waste, two underground rock laboratories have beenconstructed – one in granite (Grimsel Test Site) and one in Opalinus Clay (Mont Terri Rock Laboratory).Phenomena such as water movement in rock, transport of radioactive materials in rock, dispersion ofalkaline water resulting from concrete degradation, and effects such as gas migration and deformation dueto tunnel construction are under evaluation. Furthermore, a bentonite barrier is being heated by electricallyheated canisters simulating waste packages.

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4.1.15 United Kingdom

In the United Kingdom the industrial fuel cycle policy is to reprocess and recycle uranium toAdvanced Gas-Cooled Reactors (AGRs) and Light-Water Reactor (LWR) systems, develop and maintaincomplete fuel cycle capability (UF6 conversion, enrichment, UO2, and MOX fuel fabrication, spent fuelprocessing), and sell fuel cycle services abroad. The waste management strategy is to reprocess spentMagnox/AGR fuels as rapidly as plant capacity permits; reprocess other thermal reactor fuel after severalyears cooling; vitrify high-level waste (French process); long-term interim storage of high-level glass for atleast 50 years before disposal; shallow-land burial for low-level waste; and options for disposal of long-lived low- and intermediate-level waste are being reconsidered. Uranium conversion and fuel fabrication(Magnox) take place at Springfields. Uranium enrichment is done at Capenhurst. Fuel fabrication,reprocessing (Magnox fuels), and vitrification takes place at Sellafield. Fast breeder reactor fuels arereprocessed at Dounreay. Fuel cycle and waste management responsibilities are as follows: BritishEnergy Generation (nuclear energy production, reactor waste management), British Nuclear Fuels plc(BNFL) (Risley – engineering, Sellafield – reprocessing, waste conditioning, MOX fuel production, low-level waste disposal), United Kingdom Atomic Energy Authority (UKAEA) (decommisioning andradwaste, environment and energy, fuel services, fusion, industrial technology, safety and reliability, andreactor services), NIREX (disposal of long-lived low- and intermediate-level waste), and UraniumEnrichment Companies (URENCO) (uranium enrichment). Currently most radioactive high- andintermediate-level waste is stored by the waste producers in a variety of tanks, vaults, and silos in specialbuildings. Liquid high-level waste is immobilized by mixing with glass-making ingredients, melting it in afurnace, and pouring the glass into high-integrity stainless steel containers. Raw intermediate-level wasteis typically immobilized in cement inside steel drums. The immobilized high- and intermediate-levelwastes are kept in stores next to the treatment plant. Most of the intermediate-level waste is generally sentto BNFL’s shallow disposal site at Drigg in Cumbria. No other sites are currently being investigated fordisposal of radioactive wastes and there are no plans to identify any potential sites. The NIREX conceptfor phased geological disposal of low- and intermediate waste and the tanks used to store liquid high-levelwastes at Sellafield were described in Section 2.

The Health and Safety Executive (HSE) coordinates UK nuclear research work related to nuclearsafety via the Industry Management Committee (IMC). HSE retains a Levy Program that maintainsinternational collaboration and sources of independent expertise, and gives the capacity to carry outresearch in the eventuality that the IMC declined to commission work that the Nuclear Safety Directorate(NSD) believed to be necessary. For fuel cycle work, the program is defined by the Nuclear ChemicalPlant Research Index (NCPRI). This in practice serves as a focus for discussion with BNFL, who makeavailable information on their internal research program to the NSD in response to the NCPRI. There is noresearch program commissioned in response to the NCPRI. Research coordinated by the HSE relating toreactor plants is in some cases also relevant to fuel cycle plants(http://www.hse.gov.uk/nsd/resindex/1b/index.htm). In particular, issues of interest would be:nondestructive testing of concrete (Issue 1.5.2), concrete properties (Issue 1.5.5, Sub-Issue No. 1.5.5.2),concrete repair (Issue 1.5.9), field studies of the effectiveness of concrete repairs (Issue 1.5.9, Sub-IssueNo. 1.5.9.2), integrity of safety-related civil engineering structures (Issue 1.5.16), structural capacity ofreinforced concrete structures with corroded reinforcement (Issue 1.5.16, Sub-Issue No. 1.5.16.1), andstructural performance enhancement using strengthening applied with adhesives (Issue 1.5.18, Sub-Issue 1.5.18.1).

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4.1.16 United States

Current commercial nuclear activities in the United States include all aspects of the nuclear fuelcycle – from uranium mining to disposal of radioactive waste. Spent fuel reprocessing is determined bythe nuclear industry that at present has elected not to reprocess because of economic considerations.Mining, milling, fabrication of UO2 fuel, and low-level waste disposal are done predominately by privatefirms while high-level waste/spent fuel disposal is the responsibility of the Federal Government. Twogovernment agencies are directly involved in the commercial fuel cycle. The U.S. Department of Energy(USDOE) provides science and technology towards promoting secure, competitive, and environmentallyresponsible nuclear technologies. The U.S. Nuclear Regulatory Commission is directed toward theregulation of all civilian uses of byproduct, source, and special nuclear materials to ensure adequateprotection of public health and safety, and to protect the overall environment. The USDOE is restoringcontaminated lands and managing wastes produced during the cold war. Privatization initiatives thatbegan in 1977 are being used to address this in order to help assure private sector efficiencies and that newtechnologies are applied directly to the cleanup problems. Pursuit of disposal of commercial spent fuel in ageologic depository is continuing. The Civilian Radioactive Waste Management Program is focusing ondetermination of the suitability of the Yucca Mountain site located in Nevada on federally-owned land.The repository would be managed by USDOE and built approximately 305 m below the land surface and305 m above the water table. Under current plans, the waste would be repackaged and placed into disposalcanisters that would be transported by rail car down a ramp into a 160 km network of tunnels. The waste isto be isolated from the environment by positioning the waste above the water table, containing the waste inextremely thick and corrosion-resistant canisters, and burying the waste deep. When an estimated 64,000tonnes of waste has been disposed, the repository would be closed. Plans are to monitor the repository forat least 50 years, after which the tunnels will be sealed and a guard posted for as long as necessary. If itproceeds as scheduled, waste disposal would begin in 2010. The wastes are currently stored at commercialnuclear power plants and USDOE facilities in specially-designed water-filled pools and above-ground drystorage facilities. Liquid high-level waste is stored in large underground tanks made of stainless or carbonsteel. Small amounts of existing commercial high-level waste and all defense-related high-level waste willbe vitrified and disposed of in the Spent Nuclear Fuel (SNF) repository. The repository for defense-relatedtransuranic wastes is the Waste Isolation Pilot Plant (WIPP) located in New Mexico. The plant received itsfirst shipment of non-mixed wastes in 1999. Both the Yucca Mountain and WIPP facilities were describedpreviously.

4.2 INTERNATIONAL PROGRAMS

Collaborative approaches have also taken place at the international level so that commonapproaches can be developed, and expertise and experience can be shared. Considerable expertise hasbeen accumulated in the field of radioactive waste management over the years, particularly in the areas of:46

• The handling, treatment, storage, and disposal of short-lived low- and intermediate-levelwaste,

• The coordinating (vitrification) of high-level waste and the storage of high-level waste,• The minimization of the waste production during plant operation, and• The management of “historical” waste and the management of older waste facilities under

changed legislative and regulatory frameworks.

Information obtained in support of this report indicates that deep geological disposal represents theconsensus approach to the long-term management of long-lived radioactive waste. Much of theinternational research required to underpin development of design concepts for the specific wastesidentified for disposal in facilities such as described previously in this report has therefore been related to

assessment of the long-term post-closure radiological safety because of the need to: 34

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• Understand the physical and chemical behavior of the combination of materials containedwithin the repository (noting that after closure there is the potential for the materials to evolve and mix, modifying the physical and chemical environment of the repository);

• Identify how the materials and conditions within the repository could affect the surrounding natural environment;

• Identify the potential for radionuclides entrained in groundwater or gas to leave the engineered repository and reach the accessible environment, and the associated pathways and kinetics for this transport; and

• Consider the behavior of the repository contents and surroundings over very long time scales.

The international activities have generally been conducted under the auspices of organizations such as theOECD, International Atomic Energy Agency, and the European Commission. In addition, work beingconducted under a number of technical committees and in support of aging management of nuclear powerplant civil structures also has application to the present study. Since a prior report prepared to address

aging of nuclear power plant concrete structures addressed many of these activities,45

the technicalcommittee and aging management-related activities will not be repeated in this report. Only briefsummarizes of the international activities will be provided as much of this work does not directly addressconcrete structures in fuel cycle facilities.

4.2.1 OECD/NEA

General objectives of OECD in the area of radioactive waste management include:

• Developing common approaches for radioactive waste management strategies,• Building confidence in long-term safety assessment of waste disposal systems and

site evaluation methods,• Arranging international “peer reviews” of national program, and• Promoting internationally accepted technical solutions for the decommissioning of

obsolete nuclear facilities through cooperative program for exchange of scientific andtechnical information in this field.

Relatively recent activities in support of these objectives include meetings on measurement and physicalunderstanding of groundwater flow through argillaceous media, site evaluation and design of experimentsfor radioactive waste disposal, chemical modeling of sorption, and use of hydrogeochemical informationfor testing flow models. NEA for over 12 years has had a cooperative program addressingdecommissioning that provided a forum for information exchange. A performance assessment advisorygroup addresses long-term safety and performance assessment of waste disposal systems.

4.2.1.1 International Atomic Energy Agency (IAEA)

The IAEA nuclear fuel cycle and waste technology program is aimed at: facilitating the transferand exchange of information and technology among Member States; and providing assistance andguidance, when requested, on the formulation and implementation of strategies in nuclear fuel cycle-related activities and radioactive waste management programs with due regard to efficiency, safety,environmental soundness and sustainability, and consistency with internationally accepted norms whereapplicable, and good practices. Under the subprogram on implementation and application of radioactivewaste management technologies, the IAEA assists Member States in their efforts to implement technicallysafe, environmentally sound, and cost effective handling, processing, storage, and disposal of radioactivewastes with reliable, efficient, and proven technologies, by facilitating the transfer and exchange of

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information and technology. Recent documents have addressed safety requirements for near surfacedisposal of radioactive wastes, planning and operation of low-level waste disposal facilities, experience inselection and characterization of sites for geologic disposal of radioactive waste, techniques for in situmobilization and isolation of radioactive waste at disposal and contaminated sites, liquid radioactive wastetreatment using inorganic sorbents, high-level waste form and package behavior in simulated repositoryconditions, characterization of radioactive waste forms and packages, and a state-of-the-art report ontechnologies for decommissioning. Recent or current activities include chemical durability andperformance of spent fuel and high-level waste forms under simulated repository conditions, long-termbehavior of low- and intermediate-level waste packages under repository conditions, extrapolation of short-term observations to time periods for isolation of long-lived radioactive waste, new methods foroptimization of decontamination for maintenance or decommissioning, decommissioning activities forresearch reactors, safety assessment methodologies for near surface repositories, and biosphere modelingand assessment

4.2.3 European Commission

The European Commission is conducting research and development related to radioactive wastemanagement and decommissioning (e.g., safety aspects of waste disposal, field experiments inunderground research facilities, research on basic phenomena, partitioning and transmutation, anddecommissioning of nuclear installations). Recently completed projects have addressed building safetycases for hypothetical repositories in crystalline rock and clay, and development of a reliable overallmethodology for assessing performance of engineered barriers in radioaactive waste repositories. Severalcurrent projects include: use of a deep Russian borehole for injection of liquid radioactive waste forfurther understanding of the chemical behavior and migration of radionuclides in the geologicalenvironment; qualification of corrosion-resistant materials for long-lived high-level waste and spent fuelcontainers; development of technical/economical optimization concepts (e.g, clustering) for futureEuropean radioactive waste repositories; a prototype repository for full-scale testing of the Swedishreference concept (KBS-3) for deep disposal of spent nuclear fuel; and a review of the on-going work inthe field of preparation of an European Commission communication on decommissioning.

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5. CONCLUSIONS AND RECOMMENDATIONS

5.1 CONCLUSIONS

In reviewing national and international activities addressing aging management of concretestructures used for fuel cycle, fuel storage, and other long-term applications, several general conclusionscan be derived.

• Applications – Although concrete plays a role in the entire fuel cycle process, the reinforcedconcrete structures have generally found their greatest application to facilities for storage oflow- and intermediate-level waste that involves functions of containment and shielding.Primary facilities include above-grade vaults, below-grade vaults, and modular canisters.These facilities tend to be thick-section reinforced concrete structures, often divided into cells,that generally are constructed from conventional concrete constituent materials. Enhanceddurability is provided through: increased section and steel reinforcement cover thicknesses;use of additives to improve placeability, provide increased strength, and reduce permeability;and use of coatings or containers to prevent direct contact with the waste form. Thesefacilities are generally one component in an engineered barrier system that is designed toisolate the waste from the environment.

With respect to the use of reinforced concrete structures for management of high-level waste,such as waste resulting from reprocessing operations, the primary applications have been tostorage facilities that maintain the waste while radiation and heat decay and until facilities areavailable for disposal. These facilities may range from pads, to spent fuel pools, to dry storagecasks, to tanks.

• Durability Factors – A reinforced concrete structure’s durability characterizes its long-termperformance, and throughout its usable life the deteriorating influences of its environmentmust be resisted in order to achieve satisfactory performance. The concrete structures in fuelcycle facilities have increased requirements relative to conventional civil engineeringapplications in that the deteriorating influences, in addition to potentially being more severe,can also result from the materials they contain (e.g., elevated temperature, irradiation, andradionuclides from waste forms) as well as external influences (e.g., chemical and physicalattack). Also, the desired service life of these structures, depending on the application, mayrange from tens to hundreds or even thousands of years. During this time the reinforcedconcrete is to provide both physical and chemical barriers to isolate the waste from man’senvironment. Although a number of deteriorating influences can affect these structures,corrosion of embedded steel, leaching, elevated temperature, and irradiation (depending of theapplication) probably represent the greatest threat. Over the long term, leaching and crackinghave added importance as water will provide the transport medium for radionuclides shouldthe other engineered barriers fail.

• Performance History – Although information on the performance of reinforced concretestructures in fuel cycle facilities is limited, it appears as if the overall performance of thesestructures has been fairly good. Some of the forms of degradation that have been observedinclude corrosion of reinforcing materials, cracking, and degradation of encast rubber moisturebarriers. Causes of corrosion have been related to ingress of chlorides and carbonation, withcracking resulting from settlement, thermal, and excessive dead loads. Corrosion of liners ofsingle-wall reinforced concrete tanks has also occurred. The operational period of many ofthese structures, however, has been relatively short, especially considering the time frame forwhich many of these structures are required to function.

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• Condition Assessments, Monitoring, and Instrumentation – Due to the longevity requirementsand importance of the fuel cycle facilities, condition assessment and performance monitoringare considered to be of prime importance. These assessments need to be conductedrecognizing both the short- and long-term functions of the facilities (i.e., physical andchemical). Methods for use in conduct of condition assessments of general civil engineeringreinforced concrete structures are fairly well established and generally start with a visualexamination of exposed surfaces. To be of most use, the condition assessments should beconducted at regular intervals. Unfortunately, access to many of the radioactive waste storageor disposal facilities in all likelihood will be limited, and in some cases impossible.Monitoring and instrumentation thus will play a vital role in developing the required data foruse in performance assessments as well as providing data for development and refinement ofmodels for estimating performance. Due to accessibility constraints, monitoring to date hasgenerally been limited to capture of fluids below the facilities and use of wells to monitorgroundwater for aggressive ions or presence of radionuclides. With the exception of use ofembedded instrumentation in a pilot container at El Cabril in Spain, the application ofinstrumentation has primarily been restricted to assessing deep geological disposalcharacteristics.

• Service Life Modeling – Analysis of the role of the concrete barriers in low-level wasteisolation requires that performance assessment models be applied to concrete degradation.Mathematical models are key components of the safety assessments of waste disposal options.Such safety assessments must be based on a good physical understanding of the processesinvolved in the release and transport of radionuclides, as well as those acting on, or likely toact on, the disposal facility and the geological formation. Interactions (e.g., synergistic effects)between the various processes must also be understood. Validation and refinement of thesemodels requires substantial site investigation efforts that will involve the collection of data atthe surface as well as in situ.

Any viable design method or assessment of service life involves a number of essentialelements: a behavioral model, acceptance criteria defining satisfactory performance, loadsunder which these criteria should be satisfied, relevant characteristic material properties, andfactors or margins of safety that take into account uncertainties in the overall system.Development of a model for predicting durability or service life is intimately linked toknowledge of the degradation mechanisms affecting the concrete. The accuracy of predictionsbased on the model will depend on the degree to which the known mechanisms represent theactual damage production, as well as on the quality of the input data. Modeling must useinformation that results from testing, however, most existing tests do not incorporate all factorsof importance to service life, those factors that are included may not be related quantitativelyto in-service exposure conditions, and data covering the time period of interest are nonexistentso that extrapolated results have a high level of uncertainty. At the present state-of-the-art,models of the degradation process tend to be somewhat empirical and the primary function ofthe models for estimating service life would be for comparative purposes (e.g., alternativedesign approaches).

• National and International Programs – The approach to management of radioactive wastes, andthe spent fuel from nuclear power plants in particular, varies from country to country. Forsome countries, engaged in once-through cycles using the direct disposal option, spent fuelwill be packaged and disposed in underground sites after a sufficient cooling period in surface-based facilities. For these countries, spent fuel is labeled as high-level waste. On the otherhand, for countries that have embarked on the recycling strategy, valuable materials containedin the spent fuel (e.g., uranium and plutonium) are separated and reused in conventional

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nuclear fuels, while the ultimate residues are properly treated and conditioned before beingstored and eventually disposed. In such countries, the categories for nuclear waste directlypertain to the non-recoverable materials to be disposed after being properly treated andconditioned. The wastes are subdivided into low-, intermediate-, and high-level according totheir level of reactivity and their life span. Reinforced concrete appears to have greatestapplication to the low- and intermediate-level storage and disposal facilities. Research relatedto disposal of high-level waste is primarily related to investigation of geologic repositories anddevelopment of data and modeling methods for conduct of safety assessments of thesefacilities. At the national level, activities in large measure have addressed facilities for storageof low- and intermediate-level waste and evaluation of potential sites for deep geologicdisposal of high-level waste. In some countries, however, high-level waste is and for theforeseeable future will continue to be stored aboveground, either as liquid or vitrified, inreinforced concrete vaults. At the international level, activities have primarily been conductedunder the auspices of organizations such as OECD/NEA, International Atomic EnergyAgency, and European Commission. These activities have addressed topics such as safetyaspects of waste disposal, field experiments in underground research facilities, and research onthe basic phenomena. Some activities have addressed decommissioning of nuclear-relatedfacilities and disposal of the resulting waste forms.

5.2 RECOMMENDATIONS

The design and construction of many of the facilities and structures for processing and long-termstorage or disposal of radioactive waste materials generated by the nuclear power plant fuel cycle employreinforced concrete. The reinforced concrete is used for many purposes including support, containment,and environmental protection. The types of facilities that rely on concrete range from surface structures, toshallow subsurface vaults, to deep underground repositories. These structures are required to functionsafely and reliably in challenging and varying environments for periods of time that can potentially rangeup to thousands of years. During their operation period, these structures in all likelihood will be subjectedto a number of environmental stressors or aging factors that may adversely affect their performance toresult in shortened service lives. Detection and assessment of the magnitude and rate of occurrence of anyenvironmental factor-related degradation are key factors in maintaining the capability of these structures tomeet their operational requirements. As the knowledge base for modern concretes, such as would be usedin fabrication of fuel cycle-related facilities, is relatively short (e.g., 100 versus 500 or more years),additional input is required in several areas to help provide the continuing assurance that these structureswill continue to meet their requirements throughout their design life. Several areas where additionalinformation is required include:

• Condition assessment approaches and criteria for acceptance;• Service life models, development and validation, that take into account reliability methods

and updating as additional data become available (Bayesian);• Codes and standards specific to radioactive and hazardous waste facilities;• Instrumentation and monitoring methods for use in performance assessments;• Decommissioning, procedures and assessment criteria for “safestor” and entombment

approaches;• Degradation mechanisms, particularly where the mechanisms can operate over extended

periods of time or synergistic effects are present, and improved characterization of service environments; and

• Repair techniques to extend the performance period should premature degradation occur.

It is recommended that a forum such as international workshops or conferences be provided forcoordination and exchange of information related to life-cycle management of reinforced concretestructures utilized in nuclear power plant fuel cycle-related facilities. Attendees could describe activities

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such as aging management practices and/or plans, in-service inspection and monitoring programs utilized,status of codes and standards development, experiences relative to performance, development andeffectiveness of repair practices, and procedures for assessing future performance. The internationalworkshops or conferences could be held at two to three year intervals.

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11. K. A. Snyder, “Effect of Drying Shrinkage Cracks and Flexural Cracks on Concrete BulkPermeability,” NISTIR 6519, National Institute of Standards and Technology, U.S. Department ofCommerce, Gaithersburg, Maryland, May 2000.

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17. R. D. Rogers, M. A. Hamilton, and J. W. McConnell, Microbial-Influenced Cement Degradation -Literature Review, NUREG/CR-5987, EGG-2695, March 1993.

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APPENDIX A

SURVEY QUESTIONNAIRE REQUESTING INFORMATION ON AGING OFNUCLEAR POWER PLANT FUEL CYCLE CONCRETE STRUCTURES

CountryLevel of National Interest

Particular ConcernsNational Programmes

Ageing Management Research into Ageing Phenomena/Long Term Behaviour

Long Term StrategiesPlant Identification

Location/CountryPlant/Facility Type/FunctionOperatorLicence/statutory control

Structure Identification

Structure NameStructure FunctionMaterial(s) of ConstructionLoading Conditions (Structural & Environmental)In-service Inspection Requirements/Standards

Design Codes & StandardsCurrent ConditionDesign LifeTotal Life (including Decommissioning)Date of Construction

Degradation/Ageing Mechanisms

MechanismsIdentified prior to DesignIdentified after DesignDetection MethodsAcceptance Criteria

Other Relevant Information/Comments