Waste Management Strategies in five European Countries

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DIPARTIMENTO

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INGEGNERIA

CIVILE

E

INDUSTRIALE

T

ESI DI

L

AUREA

M

AGISTRALE IN

I

NGEGNERIA

N

UCLEARE

E DELLA

S

ICUREZZA

“Waste Management Strategies in

five European States”

TUTORS: CANDIDATE: Prof. Ing. Giuseppe Forasassi Andrea Mossi Department of Civil and Industrial Engineering

Dr. Ing. Rosa Lo Frano

Department of Civil and Industrial Engineering

Véronique Blanc

AREVA NC D&S DT

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iii

List of Figures

Figure 1: Power reactors in operation, under construction or in long-term or permanent

shutdown ... 2

Figure 2: Operational reactors by age ... 3

Figure 3: Staff reduction profile during decommissioning (OECD-NEA) ... 15

Figure 4: Waste management hierarchy ... 19

Figure 5: Conceptual illustration of the waste classification scheme ... 21

Figure 6: Options for Radioactive Material Control ... 24

Figure 7: Example of liquid water treatment by evaporation (natural circulation). ... 27

Figure 8: Example of liquid waste treatment by filtration and ion exchange. ... 28

Figure 9: Example of liquid waste solidification by bitumization ... 28

Figure 10: Example of liquid waste solidification by cementation. ... 29

Figure 11: Example of Vitrification. ... 30

Figure 12: Schematic of the super compactor ... 31

Figure 13: Schematic of incineration ... 31

Figure 14: Melting treatment concept for large components. ... 45

Figure 15: Characteristics of AWIJ and AWSJ ... 55

Figure 16: Fuel Reprocessing Cycle ... 58

Figure 17: Example of AR wet storage facilities ... 61

Figure 18: Example of AFR wet storage facilities ... 62

Figure 19: Outdoor and indoor MPC storage ... 64

Figure 20: Vault facility ... 65

Figure 21: Concrete vault of El Cabril LILW disposal facility ... 68

Figure 22: Surface Disposal conceptual design ... 69

Figure 23: Schematic diagram illustrating the relationships between natural analogue studies and the various components in the safety assessment of a disposal concept for nuclear fuel wastes. ... 72

Figure 24: Schematic of WIPP facility ... 73

Figure 25: Organisational Structure of the Relationships between FANC and ONDRAF ... 82

Figure 26: Buildings 150 and 151 ... 93

Figure 27: Internal View of Building 151 ... 93

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Figure 30: Buildings 129 ... 94

Figure 32: Buildings 136 ... 94

Figure 33: Monoliths ... 96

Figure 34: Module illustration during operation ... 97

Figure 35: Final category A disposal. ... 97

Figure 36: Options for Category B&C waste storage ... 98

Figure 37: National system for spent fuel and radioactive waste management ... 108

Figure 38: Classification of radioactive wastes with their respective actual or foreseen management routes. ... 109

Figure 39: Methodology for the development of clearance common projects in NPP. ... 110

Figure 40: Inventory of Low and Intermediate level waste open for disposal at El Cabril. ... 112

Figure 41: Inventory of High and Intermediate level waste not susceptible to disposal at El Cabril. ... 112

Figure 42: Aerial view of the El Cabril disposal facility. ... 114

Figure 43: El Cabril disposal conceptual design. ... 115

Figure 44: VLLW disposal design ... 116

Figure 45: Barriers of VLLW Disposal ... 116

Figure 46: Design concept of CTS. ... 117

Figure 47: Swiss waste inventory stored as of December 2010. ... 121

Figure 48: Total volume of radioactive waste to be managed in Switzerland ... 122

Figure 49: Transport and storage casks in the hall for dry storage Central Storage Facility ZZL . ... 124

Figure 50: Inventories of spent fuel in storage as of 31 December 2010. ... 124

Figure 51: Position and interrelationships for the Sectoral Plan for deep geological repositories. ... 126

Figure 52: Levels of Radioactive Waste in the UK. ... 129

Figure 53: The definition of radioactive material and radioactive waste and the associated decision making process ... 131

Figure 54: UK LLW strategy in summary. ... 137

Figure 55: The LLWR site layout. ... 137

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vii Figure 57: Schematic diagram for the final cap design. ... 139 Figure 58: Process for implementing geological disposal. ... 141

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List of Tables

Table 1: Reactors closed having fulfilled their purpose or being no longer economic to

run (* = ran approx full-term) ... 7

Table 2: Reactors closed following an accident or serious incident ... 8

Table 3: Reactors closed prematurely by political decision ... 9

Table 4: Advantages and disadvantages related to chemical decontamination ... 39

Table 5: Use of chemical techniques in decontamination of different materials and surfaces ... 41

Table 6: Advantages and disadvantages related to chemical decontamination ... 43

Table 7: Use of mechanical techniques in decontamination of different materials and surfaces ... 44

Table 8: Performance parameters of oxy-fuel cutting ... 48

Table 9: Performance parameters of oxygen lance ... 49

Table 10: Performance parameters of plasma cutting ... 50

Table 11: Performance parameters of oxy-arc cutting ... 50

Table 12: Performance parameters of laser cutting ... 51

Table 13: Advantages of the water jet technology ... 55

Table 14: World commercial reprocessing capacity ... 58

Table 15: Advantages and Disadvantages for MPC method ... 63

Table 16: Advantages and Disadvantages for Vault method ... 65

Table 17: VLLW, LLW and LILW repository sites and projects in selected OECD member countries ... 70

Table 18: Average costs for decommissioning nuclear power plants ... 76

Table 19: Cost structure for decommissioning nuclear power plants ... 76

Table 20: Combination of criteria to select the category of nuclear waste ... 83

Table 21: Sum-up of category depending on the activity level and radiological half-life ... 84

Table 22: Waste physically presents in Belgian nuclear sites divided by site class. ... 86

Table 23: Waste physically presents in Belgian nuclear sites divided by facility ... 86

Table 24: Inventory of waste present at UMTRAP facility. ... 87

Table 25: Waste originate from decommissioning of facilities in Belgium divided by site class ... 88

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ix Table 26: Waste originate from decommissioning of facilities in Belgium divided by

facility ... 88

Table 27: Estimate of the quantities and activities for category A, B and C waste to be managed by 2070. ... 90

Table 28: Volume and activity per storage building as of December 31, 2010 ... 95

Table 29: Publication ICRP, EURATOM directive and Italian legislation. ... 99

Table 30: Limits under which waste can be disposed of without a conditioning process. ... 101

Table 31: Radionuclide concentrations limits for the IInd Category of conditioned waste. .. 101

Table 32: Criteria in Italian categorisation ... 102

Table 33: Volume of radioactive waste stored in Italian sites until 31/12/2011 (ISPRA) ... 103

Table 34: Nuclear facilities and temporary repositories in Italy until 31/12/2011 ... 104

Table 35: Inventory of spent fuel at end of 2012 ... 113

Table 36: UK inventory at 1st April 2010 ... 132

Table 37: UK estimated future waste arising ... 133

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List of Abbreviations

ADS Accelerator Driven System

AFR Away-From-Reactor

AGR Advanced Gas Reactor

ALARA As Low As Reasonably Achievable

APAT Agenzia per la Protezione dell'Ambiente e per i servizi Tecnici

AR At-Reactor

ATA Alphatoxische Abfälle

AWIJ Abrasive Water Injection Jets AWSJ Abrasive Water Suspension Jets BNFL British Nuclear Fuel Limited

BR1 Belgian Reactor 1

BWR Boiling Water Reactor

BZL Bundeszwischenlager

CANDU CANadian Deuterium Uranium

CEN Centre d'Étude de l'Énergie Nucléaire

CoRWM Committee on Radioactive Waste Management

CS Carbon Steel

CSN Consejo de Seguridad Nuclear CTS Centralised Temporary Storage

DECC Department of Energy & Climate Change

DETEC Department of the Environment, Transport, Energy and Communications

DF Decontamination Factor

DFR Dounreay Fast Reactor

EA Environmental Agency

ENEA Ente per le Nuove tecnologie l'Energia e l'Ambiente ENEL Ente Nazionale per l'energia Elettrica

ENRESA Empresa Nacional de Residuos Radiactivos, S.A. ENSI Swiss Federal Nuclear Safety Inspectorate

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xi EPR10 Environmental Permitting Regulations 2010

EU European Union

EURATOM European Atomic Energy Community

EW Exempt Waste

FANC Federal Agency for Nuclear Control GDF Geological Disposal Facility

GLEEP Graphite Low Energy Experimental Pile GRR-2001 Royal Decree of 20 July 2001

HAA Hochaktive Abfälle

HAW Higher-Activity Waste

HDPE High Density Polyethylene

HLW High level waste

HSE Health and Safety Executive HSWA Health and Safety at Work Act

HV-VLLW High Volume - Very Low Level Waste IAEA International Atomic Energy Agency

ICRP International Commission on Radiological Protection ILW Intermediate level waste

ISFSI Independent Spent Fuel Storage Installation

ISPRA Istituto Superiore per la Protezione e la Ricerca Ambientale ITREC Impianto di Trattamento e Rifabbricazione Elementi di

Combustibile

ITS Individual Temporary Storage

LILW-LL Low and Intermediate Level Waste – Long Lived LILW-SL Low and Intermediate Level Waste – Short Lived LL-VLLW Low Volume - Very Low Level Waste

LLW Low level waste

LLWR LLW Repository

LTF Long Term Fund

MOX Mixed Oxide Fuel

MPC Multi-Purpose Canister

MRWS Managing Radioactive Waste Safely

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xii MYTIC Ministry of Industry, Tourism and Commerce

Nagra National Cooperative for the Disposal of Radioactive Waste NDA Nuclear Decommissioning Authority

NEA Nuclear Energy Act

NEA Nuclear Energy Agency

NEO Nuclear Energy Ordinance NIA65 Nuclear Installations Act 1965

NIEA Northern Ireland Environmental Agency NORM Naturally Occurring Radioactive Materials

NPP Nuclear Power Plant

OECD Organization for Economic Co-operation and Development ONDRAF Organisme National des Déchets Radioactifs et des Matières

Fissiles

ONR Office for Nuclear Regulation PFR Prototype Fast Reactor

PGRR Plan General de Residuos Radiactivos PHWR Pressurized Heavy Water Reactor PSI Paul Scherrer Institute

PUREX Plutonium Uranium Redox EXtraction PWR Pressurized Water Reactor

R&D Research and Development RAA Residuos de Alta Actividad RBBA Residuos de Muy Baja Actividad RBMA Residuos de Baja y Media Actividad RD&D Research, Development and Demonstration RNRF Regulation on Nuclear and Radioactive Facilities RP 122 Radiation Protection 122

RPA Radiological Protection Act

RPO Radiological Protection Ordinance RSA93 Radioactive Substances Act 1993

SEPA Scottish Environmental Protection Agency SFOE Swiss Federal Office of Energy

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xiii SMA Schwach und Mittelaktive Abfälle

SOGIN SOcietà Gestione Impianti Nucleari

SS Stainless Steel

THORP Thermal Oxide Reprocessing Plant

UK United Kingdom

UMTRAP Uranium Mill Tailings Remedial Action Program

VLLW Very Low Level Waste

VSLW Very Short Lived Waste

WIPP Waste Isolation Pilot Plant

WWER Voda Voda Energo Reactor

ZEBRA Zero Energy Breeder Reactor Assembly ZWILAG Zwischenlager Würenlingen AG

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SUMMARY

Radioactive waste arising from the operation and decommissioning of nuclear facilities as well as from the use of radioisotopes for medical, industrial or research purposes has to be managed responsibly, as the health risk as one aspect of the management of such materials could have an impact not only on today’s population but also on future generations.

Therefore, decision-making and management of the various tasks involved need to be based on reliable data regarding amount, characteristics and history of the collected wastes. In addition, the disposal has to be documented thoroughly to guarantee provision of sufficient knowledge to future generations. Each Country has a strategy for the handling and the storage of its radioactive waste and they are undertaken according to the respective requirements.

The work carried out in this study aims to provide an overview of the status of the implementation of national radioactive waste and of spent fuel management strategies in Belgium, Italy, Spain, Switzerland and United Kingdom as well as making some recommendations for future waste management systems. The study covers collections, reportings and record keepings of radioactive waste and spent fuel data concerning these nations.

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1 Introduction to Decommissioning and Waste

Management

1.1 Decommissioning and current development

Decommissioning can be defined as all the administrative and technical operations that allow taking off a facility from the list of licensed facilities. The administrative operations concern the elaboration of the decommissioning plan and authorizations drawing, the free release certificates for the facilities and the site. The technical operations, on the other hand, involve the decontamination, the dismantling and waste management. The aim of the decommissioning is not the simply building’s demolition but their liberations from all the obligations and controls corresponding to the class they belong to.

Decommissioning is a long, slow and complex process. Moreover, decommissioning is not an homogeneous procedure, since it entails with the end cycle of the nuclear plant. This was confirmed by the experimental evidences demonstrating the decommissioning operations are different in the number of facilities involved in.

National and International authority controls have identified decommissioning timescales and strategies that have been carried out in the Countries where decommissioning is already a reality.

First of all it is necessary to understand what does decommissioning mean and which role takes in the phases of the cycle life of nuclear facilities. Nuclear decommissioning concerns all the nuclear plants independently of plant size and purposes. In this context, it is important to classify the nuclear installations, as follows:

 Nuclear power plants for electricity production;

 Research, experimental or isotope production reactors;

 Fuel fabrication plants;

 Spent-fuel reprocessing plant;

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 Hot cells for activities on activated materials, contaminated materials or radioisotopes.

Although these facilities are very different in design, purpose and radioactive power, the main issues related to radiation, environment and technology are similar.

1.1.1 Decommissioning status and perspective in the world

As of August 2013, about 100 mines, 102 commercial power reactors, 46 experimental or prototype reactors, over 250 research reactors and a number of fuel cycle facilities, have been retired from operations. Some of these have been fully dismantled [1].

The typical life plant (by design) of a nuclear power reactor is 30 to 40 years with some granted a 20-years extension to 60 years. There are currently 438 reactors in operation worldwide, with a total installed electrical capacity of 374 332 MWe and 71 reactors under construction (Fig. 1). 330 of these 438 nuclear power reactors are more than 25 years old whilst 64 are more than 40 years old. The average age of the nuclear power plant reactors currently in operation is almost 30 years (Fig. 2) [2].

Figure 1: Power reactors in operation, under construction or in long-term or permanent shutdown [2].

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Figure 2: Operational reactors by age [2].

Many nuclear power plants around the world have been designed and constructed before the problem of how to eventually dismantle them had been solved or seriously considered. The closure of the operating life of a nuclear installation should be planned at the design stage, after a refurbishing or, in any case, some years in advance of the event. Although decommissioning normally happens at the end of the operating life of the facility, it can be needed also for other reasons:

 The initial technology, using radioactive material, has become obsolete or uneconomical. There are 102 reactors shut down for this reason (Tab.1);

 An accident, serious incident or unplanned event. There are 11 reactors shut down for this reason (Tab.2);

 The closure prematurely by political decision or due to regulatory impediment. There are 25 reactors shut down for this reason (Tab.3) [1].

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Country Reactor Type MWe net each Start-up Years operating each Shut down Belgium BR-3 Prot PWR 10 1962 24 1987

Canada Douglas Point Prot PHWR 206 1967 17 1984

Gentilly 1 Exp SGHWR 250 1971 6 1977 Gentilly 2 PHWR 638 1982 30 2012 Rolphton NPD Prot PHWR 22 1962 25 1987 France Bugey 1 GCR 540 1972 22 1994 Chinon A1 Prot GCR 70 1963 10 1973 Chinon A2 GCR 180 1965 20 1985 Chinon A3 * GCR 360 1965 25 1990 Chooz A Prot PWR 305 1967 24 1991

Brennilis EL-4 exp GCHWR 70 1967 18 1985

Marcoule G-1 Prot GCR 2 1956 12 1968 Marcoule G-2 Prot GCR 39 1959 20 1980 Marcoule G-3 Prot GCR 40 1960 24 1984 Phenix * FNR 233 1973 37 2010 St Laurent A1 GCR 390 1969 21 1990 St Laurent A2 GCR 465 1971 21 1992

Germany Juelich AVR Exp HTR 13 1968 21 1989

Uentrop THTR Prot HTR 296 1985 3 1988

Kalkar KNK 2 Prot FNR 17 1978 13 1991

Kahl VAK Exp BWR 15 1961 24 1985

MZFR Exp PHWR 52 1966 18 1984 Groswelzheim Prot BWR 25 1969 2 1971 Lingen Prot BWR 183 1968 10 1979 Niederaichbach Exp GCHWR 100 1973 1 1974 Obrigheim * PWR 340 1968 36 2005 Stade * PWR 640 1972 31 2003 Wuergassen BWR 640 1972 22 1994

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Italy Garigliano BWR 150 1964 18 1982

Japan Fugen Prot ATR 148 1978 24 2003

Hamaoka 1 BWR 515 1974 26 2001

Hamaoka 2 BWR 806 1978 25 2004

JPDR Prot BWR 12 1963 13 1976

Tokai 1 * GCR 137 1965 33 1998

Kazakhstan Aktau BN-350 Prot FNR 52 1973 27 1999

Netherlands Dodewaard * BWR 55 1968 28 1997

Russia Obninsk AM-1 Exp LWGR 6 1954 48 2002

Beloyarsk 1 Prot LWGR 108 1964 19 1983

Beloyarsk 2 Prot LWGR 160 1968 22 1990

Melekess VK50 Prot BWR 50 1964 24 1988

Novovoronezh 1 Prot

VVER-440/V210

210 1964 23 1988

Novovoronezh 2 Prot

VVER-440/V365

336 1970 20 1990

Spain Garona BWR 446 1971 42 2012

Jose Cabrera * PWR 141 1968 38 2006

Sweden Agesta Prot HWR 10 1964 10 1974

UK Berkeley 1-2 * GCR 138 1962 26 1988-89 Bradwell 1-2 * GCR 123 1962 39 2002 Calder Hall 1-4 * GCR 50 1956-59 44-46 2003 Chapelcross 1-4 * GCR 49 1959-60 44-45 2004 Dungeness A 1-2 * GCR 225 1965 41 2006 Hinkley Pt 1-2 * GCR 235 1965 35 2000 Hunterston A 1-2* GCR 160 1964 25 1989-90 Oldbury 1-2* GCR 217 1967 44 2011-12

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6 Sizewell A 1-2 * GCR 210 1966 41 2006 Trawsfynydd 1-2 * GCR 196 1965 26 1993 Wylfa 2* GCR 490 1971 41 2012

Windscale Prot AGR 28 1963 18 1981

Dounreay DFR Exp FNR 11 1962 18 1977

Dounreay PFR Prot FNR 234 1975 19 1994

Winfrith Prot SGHWR 92 1968 23 1990

USA Big Rock Point* BWR 67 1962 35 1997

BONUS Exp BWR 17 1964 4 1968

CVTR Exp PHWR 17 1963 4 1967

Crystal River PWR 860 1977 35 2013

Dresden 1 BWR 197 1960 18 1978

Elk River BWR 22 1963 5 1968

Enrico Fermi 1 Prot FNR 61 1966 6 1972

Fort St. Vrain Prot HTR 330 1976 13 1989

Haddam Neck* PWR 560 1967 29 1996

Hallam Exp sodium

cooled GR 75 1963 1 1964 Humboldt Bay BWR 63 1963 13 1976 Indian Point 1 PWR 257 1962 12 1974 Kewaunee* PWR 566 1974 39 2013 Lacrosse BWR 48 1968 19 1987 Maine Yankee* PWR 860 1972 25 1997 Millstone 1 BWR 641 1970 28 1998 Pathfinder Prot BWR 59 1966 1 1967

Peach Bottom 1 Exp HTR 40 1967 7 1974

Piqua Exp Organic

MR

12 1963 3 1966

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7 San Onofre 1* PWR 436 1967 25 1992 Saxton Exp PWR 3 1967 5 1972 Shippingport Prot PWR 60 1957 25 1982 Trojan PWR 1095 1975 17 1992 Vallecitos Prot BWR 24 1957 6 1963 Yankee NPS* PWR 167 1960 31 1991 Zion 1-2 * PWR 1040 1973 25 1998 Sturgis FNPP PWR 10 1967 9 1976

Table 1: Reactors closed having fulfilled their purpose or being no longer economic to run (* = ran approx full-term) [1].

Country Reactor Type MWe net Years operating Shut down Reason

Germany Greifswald 5 VVER-440/V213

408 0.5 11/1989 Partial core melt

Gundremmingen A BWR 237 10 1/1977 Botched shutdown Japan Fukushima Daiichi 1 BWR 439 40 3/2011

Core melt from cooling loss

Fukushima

Daiichi 2 BWR 760 37 3/2011

Core melt from cooling loss

Fukushima

Daiichi 3 BWR 760 35 3/2011

Core melt from cooling loss Fukushima Daiichi 4 BWR 760 32 3/2011 Damage from hydrogen explosion

Slovakia Bohunice A1 Prot GCHWR

93 4 1977 Core damage

from fuelling error

Spain Vandellos 1 GCR 480 18 mid 1990 Turbine fire

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8 GCHWR Ukraine Chernobyl 4 RBMK LWGR 925 2 4/1986 Fire and meltdown

USA Three Mile Island

2

PWR 880 1 3/1979 Partial core melt

Table 2: Reactors closed following an accident or serious incident [1].

Country Reactors Type MWe net each

Years operating each

Shut down

Armenia Metsamor 1 VVER-440/V270

376 13 1989

Bulgaria Kozloduy 1-2 VVER-440/V230

408 27, 28 12/2002

Kozloduy 3-4

VVER-440/V230

408 24, 26 12/2006

France Super Phenix FNR 1200 12 1999

Germany Greifswald 1-4 VVER-440/V230 408 10, 12, 15, 16 1990 Muelheim Kaerlich PWR 1219 2 1988 Rheinsberg VVER-70/V210 62 24 1990 Italy Caorso BWR 860 12 1986 Latina GCR 153 24 1987 Trino PWR 260 25 1987 Lithuania Ignalina 1 RBMK LWGR 1185 21 2005 Ignalina 2 RBMK LWGR 1185 22 2009

Slovakia Bohunice 1 VVER-440/V230

408 28 12/2006

Bohunice 2

VVER-440/V230

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9 Sweden Barseback 1 BWR 600 24 11/1999 Barseback 2 BWR 600 28 5/2005 Ukraine Chernobyl 1 RBMK LWGR 740 19 12/1997 Chernobyl 2 RBMK LWGR 925 12 1991 Chernobyl 3 RBMK LWGR 925 19 12/2000 USA Shoreham BWR 820 3 1989

Table 3: Reactors closed prematurely by political decision [1].

1.1.2 Decommissioning strategies

The international decommissioning panoramas of nuclear facilities are characterized not only by the available technology, age and status of the plants and their weigh in the various national contexts, but it has often conditioned from national policy. In particular, they relate to requirements concerning acceptable conditions for unrestricted re-use. In terms of decommissioning options, the Nuclear Energy Agency (NEA) of the Organization for Economic Co-operation and Development (OECD) specifies that “there is no unique or preferred approach to Decommissioning and Dismantling of nuclear facilities”[3]. One of the main differences between Countries concerns the clearance levels for facilities and waste that, as an example, do not exist in France. These differences have important implications for waste management and reuse of nuclear components.

Nevertheless, there are different phases and stages of decommissioning approved at international level and three principal strategies have been identified by the International Atomic Energy Agency (IAEA).

Late in the 1980s and early in the 1990s, the IAEA identified three decommissioning stages (identified as Stages 1, 2 and 3) [4].

Each of these three decommissioning stages can be defined by following two parameters:

 The physical and radiological state of the plant and its equipment;

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Stage 1: Storage with surveillance

The first stage begins at the shutdown. The spent fuel (and thus 99% of the radioactivity) is removed from the reactor, the liquid systems are drained, the operating systems are disconnected and the mechanical openings are securely sealed. The facility is kept under surveillance and inspected to ensure that it remains in a safe condition.

Stage 2: Restricted site release

All equipment and buildings that can be easily dismantled are removed or decontaminated and made available for other uses. Any remaining fluid is drained from the systems. At nuclear reactors, the biological shield is extended and sealed to completely enclose the reactor structure. Surveillance around the first contamination barrier may be reduced, but it is deliverable that periodic spot checks be continued.

Stage 3: Unrestricted site release

All buildings, equipment and materials that cannot be decontaminated below established clearance levels are removed and the resulting waste is handled and stored or deposited. The remaining parts of the plant and site are released for unrestricted use. No further inspection or monitoring is required for effectively returning the area to “green field” status.

A lot of factors must be considered when selecting the strategy for the decommissioning of a nuclear facility. These include the national nuclear policy, radioactive waste

management, health and safety, characteristics of the facility, environmental protection, availability of staff, future use of the site, improvements of decommissioning technology that may be achieved in the future, cost and availability of funds for the project and various social considerations. The relative importance of these factors must be judged case by case.

In the mid-1990s, the IAEA adopted three decommissioning strategies:

 Immediate dismantling;

 Safe enclosure;

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11 These strategies have been well defined and are currently used in all the IAEA safety standards. Each of these options has advantages and drawbacks [5]:

Immediate dismantling

Normally, within five years from the shutdown of the facility, the immediate dismantling strategies start.

All radioactive materials above a specified level are removed and the end point of the project is that the site or facility can be cleared or used without any regulatory restrictions. This option does not allow for any significant decay of radionuclides. It also implies that waste and spent fuel management, as applicable, must be available. This does not mean that a disposal site must be in place, but some type of waste management system (i.e. interim storage) must be available.

This internationally agreed approach has the advantage that experienced operational staff from the facility is still available, people who knows the history of the site, including any past incidents, if occurred, that could sometimes complicate the decommissioning process. Immediate dismantling also avoids the unpredictable degradation of the reactor parts over an extended period, eliminates the risk of future exposure to radiation and restores the landscape. A drawback of this approach is that levels of radioactivity in the reactor parts are higher than in the case of safe enclosure. This means that greater precautions must be taken during dismantling and that larger volumes of waste must be classified as radioactive.

Deferred dismantling (also called safe storage, safe store or safe enclosure)

The facility is placed and maintained in a safe stable condition, usually of the order of 40-60 years, until it is dismantled and decontaminated to level that allows the removal according to regulatory controls. All spent fuel is removed from the site before the long-term storage period begins. This reduces the safeguards and security concerns as well as and allows for a large reduction in the overall risk of the facility.

Deferred dismantling has the advantage of allowing radioactive materials to decay to lower levels of radioactivity than in the case of immediate dismantling. This approach allows to reduce both the disposal problems and the risks of exposure of workers. In the meantime, robotic and other types of techniques, which make dismantling safer and cheaper, may undergo further development. A drawback is that some materials, such as

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12 concrete and steel, may deteriorate, making the eventual decommissioning more difficult. Moreover, personnel knowledge of site’s history may be lost as time passes.

Entombment

In this situation, the overall controlled area is reduced and the remaining radioactive material is encased on-site, normally in concrete: while the remaining structure must be monitored and safely maintained for a period of time. This site essentially becomes a near surface waste repository. All the requirements for such a waste repository will have to be met, to include the siting and design requirements. Most regulators do not prefer this type of approach.

1.1.3 Factors influencing the selection of a

decommissioning strategy

For establishing and preparing a decommissioning strategy of a nuclear facility, a number of factors must be considered and duly evaluated. The strategy will change in relation to the facility type and, of course, these factors must be evaluated on a case-by-case basis. In addition to the technical elements, it is growing an increasing awareness about the importance of the non-technical factors on the selection of a decommissioning strategy. The following list identifies the main issues that seem to be particularly relevant for the selection of a strategy:

 The scope of the decommissioning activities;

 The basic decommissioning options;

 The facility type, size, number of units on a site;

 The operational history;

 Project planning;

 Analysis of material flow;

 Regulatory and policy requirements (timing, clearance criteria);

 Socio-economic issues;

 Waste management provisions;

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 Staff availability and personnel issues;

 Knowledge retention;

 Site reuse;

 Stakeholders, decision makers, regulators and the public.

1.1.3.1 Decommissioning policy and regulatory requirements

Decommissioning policy includes all governmental (national or regional) choices, as described in laws, regulations, standards, guidelines and mandatory requirements that will influence the framework in which decommissioning takes place, as for example, the requirements relates to:

 The disposal of shutdown nuclear facilities;

 The use of decommissioned sites;

 The responsibility of the decommissioning: industry or government;

 The national radioactive waste management policy;

 The national policies for re-use and recycling of materials;

 The national policies for public and worker health protection;

 The environmental, safety, and regional development aspects.

The kind of legislation that it is developed to enforce the decommissioning requirements derives a lot from the juridical system in each County. In some systems, the legislations are considered as the explanation of the goals. In other cases, the legislations are very much detailed and prescriptive. Furthermore, there could be regional needs. For example, in the European Union, there are different directives relating to the radiation protection and environmental impact assessment that must be added into the legislation of the Member States. The way this is done will depend on the individual legal system of a State.

Therefore, the regulations systems for the different Member States can be different, but the safe decommissioning objectives are the same.

A regulatory factor that could influence the strategy chosen by the operator of nuclear facilities is the uncertainty on the stability in the long term of the legislation. An operator

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14 can never be sure of the stability of the existing legislation, although he can be confident that if there is a change, it would lead to more stringent constraints.

1.1.3.2 Environment, public and worker protection

The requirements established by the legislation on Radiation Protection and Industrial Safety on the decommissioning process condition the choice of decommissioning strategy. The first objective, which must be satisfied in any decommissioning programmes, is to guarantee health and safety of the workers and to protect the general public and the environment. Public exposure and environmental impacts are expected to be lower as reasonably achievable (ALARA criteria) and well within the regulatory limits for operating facilities.

A cost-benefit analysis should be realized to determine what extent delayed dismantlement will have a positive effect. Although this depends on the physical state of the nuclear facility, as well as on the available resources and equipment, the property of radioactive substances to decay has led to the suggestion that there is some advantage in leaving the plant or buildings in care and maintenance for periods of time on the grounds that this will make eventual decommissioning safer and easier. This argument may be valid for short-life radio-nuclides in situations where the material can be contained and physical deterioration will not make the decommissioning task more hazardous.

1.1.3.3 Public acceptance

The process of deciding between the different decommissioning strategies may take into consideration the possible effects on factors such as [6]:

 Environmental factors (e.g. the value of the neighbouring land);

 Employment problems;

 The public's perception of the hazards, whether the installation is maintained in a safe shutdown condition or is dismantled.

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15 It also analyses the public opinion about the proposed choice in the final strategy to be adopted, the influence of stakeholders such as community members, the media, activists, political and business leaders, and the employees themselves.

Other references raise the issue of the social and economic situation around sites of nuclear facilities. Concerning the on-site employment, immediate dismantling requires a larger staff and work force than other strategies, resulting in a slower and smoother reduction of the operations staff (Fig 3).

Figure 3: Staff reduction profile during decommissioning (OECD-NEA) [7].

1.1.3.4 Technology availability

The availability of technology is not, as generally though, one of the major factors for the selection of a decommissioning strategy. In reverse, the choice of a strategy can influence the development of new technologies necessary for the dismantling of facilities, their characterization or decontamination. For example, the immediate dismantling will allow developing remote control equipment or robotic systems that can access areas of the plants the objective of which is to reduce the radiation exposure to the personnel. In the case of deferred dismantling strategy, it would be necessary to develop programmes for a better understanding of the degradation of buildings and structures with time.

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16 It is widely accepted in the decommissioning that the current technology is adequate for today's decommissioning and dismantling needs. Many decommissioning projects have been carried out in the entire world demonstrating this fact as clear evidence in the industrialized countries.

In general, the choice of the best decommissioning strategy depends on a large variety of factors but it appears that the type of reactors is not a major factor influencing the decision, despite the technological differences between them.

The list of retired reactors and their decommissioning status show that water cooled reactors (PWR, BWR, WWER, CANDU) can be either decommissioned immediately or after a period of safe enclosure, depending on other factors than the "type of facility". Gas cooled reactors are generally dismantled after a period of safe storage due to the higher complexity of their design, the presence of quantities of graphite and the limited possibilities to decontaminate the systems.

The techniques for dismantling fuel cycle facilities are essentially similar to those for dismantling nuclear power plants except that a safe enclosure period would not be helpful in reducing the radioactivity of those facilities contaminated with long-life radio-nuclides. For these facilities, early dismantling is therefore the preferred strategic choice.

1.1.3.4.1 Effect of the design of a facility on the decommissioning

Experiences from the decommissioning projects suggest that the decommissioning issues should take into account at the design and construction phases of a nuclear facility. The plant records may sometimes be incomplete or inaccurate, e.g. they may not reflect the final plant configuration. For example, the plant operators need to give special attention to the collection and preservation of information on contamination events, as such contamination may otherwise only be identified during demolition of the concrete structure [8]. It is highly desirable that designers of new plants are aware of the issues, the strategies, the techniques and the needs involved in decommissioning.

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1.1.3.5 Waste management

The availability of a final repository is and important factors to enable a fast, economic and final clearance of a site. This aspect is more clear when there is the need to start the dismantling of a facility which will generate large volumes of waste materials.

If no disposal facility is available, it may not be considered to be an obstacle to early dismantling, if other factors than waste disposal and costs influence in the decision making process. In this case, if it does not judge appropriate to defer the decommissioning until a disposal route has been established, temporary storage facilities must be built. Naturally this means to increase the cost of the decommissioning operations, because at the end of the process, it will have to be dismantled.

Another important factor is the availability or not of a repository for the spent fuel. The absence of a disposal for this kind of waste can result in the temporary storage of spent fuel on site and in the decision to postpone the dismantling of the facility until a final solution has been found.

1.1.3.6 Site reuse

The site reuse after the dismantling of a facility is a factor influencing the decommissioning strategy.

Different actions will be taken depending on the target of the end state of the decommissioning.

This may be "green field" if the complete removal of the site from regulatory control is decided (more expensive) or "brown field" if some of the previous installations can be reused.

The following alternatives are possible for ending the nuclear supervision of a facility:

 Clearance of the site for unrestricted reuse without radiological supervision after total removal of the facility;

 Clearance of the site and the remaining structures (buildings, systems) for another commercial use without radiological supervision;

 Conversion of the site and the remaining structures into another facility which is licensed under the nuclear legislation without radiological clearance

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1.1.3.7 Founds availability

The availability of founds for the decommissioning of nuclear facilities is a key factors for the selection of a strategy.

If no funding has been collected at the time of closure, the required amount of money must be put from the State or other Body when the immediate dismantling is decided. If money is not available, deferred dismantling must be considered through a new instituted fund collecting the necessary means to implement the dismantling activities.

In the same way if an operating plant which has well anticipated its future liabilities, uncertainties will affect the fund that has been established. These uncertainties can modified the cost estimates by unpredictable factors like changes in:

 Staff or equipment costs;

 Regulatory requirements;

 Governmental priorities or policies;

 Disposal costs for radioactive wastes;

 Criteria for release of materials, buildings and sites from regulatory control.

One way of minimizing these uncertainties is to complete decommissioning as early as possible, after final plant shutdown.

1.2 Waste Management

Radioactive waste is any material in solid, liquid or gaseous state containing significant amount of radionuclides with a concentration of radioactivity greater than certain limit established by the Authority in order to protect human health and the environment and for which no further other use is foreseen. Radioactive wastes are usually producted from nuclear facilities and other applications in nuclear fields such as research, industry and medicine.

Radioactivity naturally decays over time, so radioactive waste has to be isolated and confined in appropriate disposal facilities for a sufficient period of time until it no longer poses a hazard. The period of time waste must be stored depends on the type of waste and

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19 radioactive isotopes. It can range from few days, for very short-lived isotopes, to millions of years, for spent nuclear fuel.

Current major approaches to manage radioactive waste are the segregation and the storage for short-lived waste, the near-surface disposal for low and some intermediate level waste, and the deep geological disposal or transmutation for the high-level waste.

To achieve this aim, the waste must be processed, conditioned and packaged to allow a secure handling; the radioactive waste volume must be reduced as low as possible to optimize the space in the repositories.

As a general guideline for the handling of radioactive waste has been set in the European Directive 2008/09 in which hierarchy of waste management has been fixed (Fig.4) with the priority order of avoiding wastes, minimizing the amount, re-use, recycle and with disposal as the least favoured option.

Figure 4: Waste management hierarchy [9].

1.2.1 Waste Classification

The radioactive waste management process normally starts with the waste classification to decide on the future steps that the waste should withstand up to its final disposal facility.

Radioactive waste, as defined by the IAEA, is waste that contains or is contaminated with radionuclides at concentrations or activities greater than clearance levels as established by the regulatory body [10]. This definition already includes a reference to national

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20 authorities and indicates that any legal regulations are defined by those. The current classification of radioactive waste suggested by IAEA holds six classes [11]:

 Exempt waste (EW): Waste that meets the criteria for clearance, exemption or exclusion from regulatory control for radiation protection purposes;

 Very short lived waste (VSLW): Waste that can be stored for decay over a limited period of up to a few years and subsequently cleared from regulatory control according to arrangements approved by the regulatory body, for uncontrolled disposal, use or discharge;

 Very low level waste (VLLW): Waste that does not necessarily meet the criteria of EW, but that does not need a high level of containment and isolation and, therefore, is suitable for disposal in near surface landfill type facilities with limited regulatory control;

 Low level waste (LLW): Waste that is above clearance levels, but with limited amounts of long lived radionuclides. Such waste requires robust isolation and containment for periods of up to a few hundred years and is suitable for disposal in engineered near surface facilities. This class covers a very broad range of waste. LLW may include short lived radionuclides at higher levels of activity concentration, and also long lived radionuclides, but only at relatively low levels of activity concentration;

 Intermediate level waste (ILW): Waste that, because of its content, particularly of long lived radionuclides, requires a greater degree of containment and isolation than that provided by near surface disposal. However, ILW needs no provision, or only limited provision, for heat dissipation during its storage and disposal. ILW may contain long lived radionuclides, in particular, alpha emitting radionuclides that will not decay to a level of activity concentration acceptable for near surface disposal during the time for which institutional controls can be relied upon. Therefore, waste in this class requires disposal at greater depths, of the order of tens of metres to a few hundred metres;

 High level waste (HLW): Waste with levels of activity concentration high enough to generate significant quantities of heat by the radioactive decay process or waste with large amounts of long lived radionuclides that need to be considered in the design of a disposal facility for such waste. Disposal in deep, stable

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21 geological formations usually several hundred metres or more below the surface is the generally recognized option for disposal of HLW;

A conceptual illustration of the waste classification scheme is presented in Fig. 5.

Figure 5: Conceptual illustration of the waste classification scheme [11].

The degree of containment and isolation provided in the long term varies according to the disposal option selected. The classification scheme is based on the consideration of long term safety provided by the different disposal options currently adopted or envisaged for radioactive waste.

In the classification scheme, the following options for management of radioactive waste are considered, with an increasing degree of containment and isolation in the long term:

 Exemption or clearance;

 Storage for decay;

 Disposal in engineered surface landfill type facilities;

 Disposal in engineered facilities such as trenches, vaults or shallow boreholes, at the surface or at depths down to a few tens of metres;

 Disposal in engineered facilities at intermediate depths between a few tens of metres and several hundred metres (including existing caverns) and disposal in boreholes of small diameter;

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 Disposal in engineered facilities located in deep stable geological formations at depths of a few hundred metres or more.

1.2.2 Exclusion, Exemption and Clearance

The decommissioning of NPPs produces a big quantity of radioactive waste, mainly solid waste such as steel and concrete. Frequently, these materials own a radioactivity level due to contamination and activation but the major parte is characterized by a feeble level of activity.

The possibility to let off certain radioactive waste from the control and manage of normally decommissioning activities is an important factor in view of the reduction waste management costs and volumes. These are the mains goals to achieve because allow minimizing the quantity of radioactive materials that should be stored for a long time before the disposal in conventional field.

The choice of the “clearance levels” has a meaning both radiological and economical:

 Radiological, because they derive from the adoption of a limit dose and from the assumption and estimates on the ways in which weakly contaminated material can lead to exposure in the population and environmental after the re-use in conventional field;

 Economical, because by the choice of the release value derives the amount of contaminated material that can be reused rather than disposed of as radioactive waste and, consequently, the cost of decommissioning.

The IAEA has established guidelines that give reference value for the clearance but in each country, the authority control agencies could decide different limit for the clearance of radioactive waste. In order to understand the concept of clearance, two similar concepts, like exclusion and exemption, equally implying no submittal to regulatory control will be explained.

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1.2.2.1 Exclusion

According with the IAEA Safety Guide No. RS-G-1.7, the term “exclusion” refers to the deliberate exclusion of a particular category of exposure from the range of an authority performing regulatory control based on the fact that is not considered amenable to control thought the regulatory authority in question.

The IAEA Safety Guide No. RS-G-1.7 provides examples of excluded exposure such as: exposure from 40K in the body, from cosmic radiation at the surface of the earth and from unmodified concentrations of radionuclides in most raw materials. All of these examples are of exposure to natural sources of radiation although there is no explicit requirement to limit the concept to such exposure. [12]

1.2.2.2 Exemption

According with the IAEA Safety Guide No. RS-G-1.7, the term “exemption” can be defined as the determination by a regulatory body that a source or practice need not be subject to some or all aspects of regulatory control on the basis that the exposure due to the source or practice is too small to warrant the application of those aspects.

Basically, exemption may be regarded as a generic authorization issued by the regulatory body, which automatically exonerates the practice or source from requirements such as those relating to notification or authorization. Two main exemption criteria are stated within the IAEA Safety Guide No. RS-G-1.7:

 The effective dose expected to be incurred on any member of the public due to an exempted practice or source is of the order of 10 μSv or less in a year;

 Either the collective effective dose committed by one year of performance of the practice is no more than bout 1 man Sv or an assessment for the optimization of protection shows that exemption is the optimum option.

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1.2.2.3.1 Clearance

According to the IAEA Safety Guide No. RS-G-1.7, the term “clearance” can be defined as the removal of radioactive materials or objects within the authorised practices from further regulatory control by the regulatory authority. Furthermore, the IAEA states that clearance levels shall take account of the exemption criteria and shall not be higher than the exemption levels defined by the regulatory body.

Figure 6: Options for Radioactive Material Control

1.2.3 Waste management

The term “radioactive waste management” concerns the treatment, conditioning, handling, storage, transport and disposal of the radioactive waste. There are essentially three principles adopt in the radioactive waste management to reduce the hazard of the handling and storage of radioactive waste:

 Dilution is designed to carry out an extensive resolution and distribution of the waste or the radioactive contents into the environment (i.e. water, air) in a quantity lower than the law limit;

 Storage and decay of waste or radioactive materials containing only radionuclide with very short half-life;

 Concentration means that the waste or the radioactive materials are concentrated and isolated towards the environment.

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25 The most important consideration in all waste management processes is that the applicable safety and radiological protection regulations must be complied with. Minimize the volume of material requiring disposal to reduce the mobility of the radionuclides contained in the waste, and to segregate the waste by the type of radioactivity contained (alpha-bearing, non alpha-bearing, low-, intermediate-, high- activity waste). The purpose of minimising waste volume, mobility and segregating wastes is to achieve an optimum combination of safety and economic disposal.

The waste management strategy is based on national waste management regulations and takes into account government policy. It normally includes items such as:

 An evaluation of waste types, their physical and chemical characteristics and the volumes of each waste category including the foreseen rate at which waste will be generated;

 Criteria for the restricted or unrestricted reuse and recycling of the materials and components from the decommissioning;

 The project and process for handling, treating, conditioning, storing and disposing of each category of waste;

 Methods for monitoring of cleared waste before unrestricted release;

 Requirements of packaging and package design for transport and disposal.

1.2.3.1 Waste treatment

Radioactive waste treatment involves all the operations intended to concentrate the radioactivity inside the waste; they depend on the nature and activity of wastes. These operations allow obtaining almost the radioactivity contents on one side and effluents with low activity on the other side. These technologies are currently used technologies and may be combined for treating the whole spectrum of radioactive waste.

1.2.3.1.1 Treatment and Conditioning of liquid waste

Liquid waste consists of large volumes of dilute solution containing dissolved and/or finely divided contaminated materials that should be treated to extract the radioactive material from the carrier solution. These allow also reducing the volume of materials that requiring further conditioning and disposal.

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26 The usual techniques for decontamination of liquid waste are ion exchange, filtration, chemical treatment and evaporation. These techniques are used singly or in series combination, depending upon the characteristics of solution being treated. The purified liquids resulting from these processes are usually decontaminated sufficiently so that discharge to the environment is possible after monitoring to ensure compliance with regulations.

Certain liquids used in some decontamination processes, such as phosphoric acid in electropolish and organic solvents in surface cleaning, may require special treatment because of their specific chemistry.

The concentrated radioactive residues, contained in the ion exchange media, filers, or concentrator bottom liquids, require conditioning to immobilize them. This immobilisation is accomplished by mixing radioactive residues with matrix material, such as cement or bitumen or polymers and subsequently solidifying the mixture within a suitable container.

Liquid waste treatment by evaporation

The evaporation is one of the most important and effective process commonly used in the nuclear industry; it is an effective method for chemical and radiological purification of liquid waste from a nuclear facilities.

The process may include two stages: the evaporation and the vapour purification, with this latter increasing the decontamination factor. Fuel cycle offers technologies for both stages that have been demonstrated in many types of facilities, particularly in reprocessing plants, where evaporation is a key stage in several complex processes. The solutions offered are well-suited to the treatment of liquid waste generated by nuclear power plants and other nuclear facilities.

If the quantity of radioactive water released must be minimized, or, if zero-release is required, the evaporation process is the best method to achieve this goal. The system has a throughput of 1 to 110 l/min and the decontamination factor (bottoms/distillate) is up to 107 (adapted to special plant requirements) [13].

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27

Figure 7: Example of liquid water treatment by evaporation (natural circulation)

[13].

Liquid waste treatment by filtration and ion exchange

The filtration train serves to remove ions from the waste water by means of mechanical filtration and ion exchange and is used mainly to treat waste water to be recycled, which is generated in boiling water reactors.

The waste water filter removes any not dissolved particle or colloid by mechanical filtration and serves as ionic pre-treatment. It is a disc-type pre-coat filter, which contains disked filter plates horizontally stacked on top of each other and mounted to the centre to a hollow shaft which can be rotated. The filter is a self-contained unit and operated with a continuous flow rate. The upper side of the disk is covered by a fine pre-coat layer. The surface of the screen is pre-coated with ion-exchange resins. The decontamination factor of the water waste filter is up to 10.

Downstream of the water waste filter is a mixed-bed filter. Its purpose is to remove the remaining ions from the waste water by ion exchange. The cleaned water is neutral and demineralised. The decontamination factor of the mixed-bed filter is up to 100. The system throughput is up to 80 m3/h.

Mixed bed filters were used in former times also in the waste water treatment systems in PWR, but due to the good quality of treated water by evaporation the filter in the plants, they are no longer used [13].

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Figure 8: Example of liquid waste treatment by filtration and ion exchange [13].

Liquid waste solidification by bitumization

This process is used when radioactive liquid waste needs conversion into a product suitable for final storage. Bitumen is readily available and has properties that make it a preferred binding material for use in solidifying liquid waste from evaporator bottoms, decontamination solutions, filter sludge/slurries, spent resin suspensions. It is also used for the encapsulation of solid radwaste like filter elements, worn-out parts, etc. The waste/bitumen ratio is up to 50:50 (adapted to the special plant requirements) and the volume reduction factor is up to 2.5 (related to waste with solid content of 20% by weight) [13].

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Liquid waste solidification by cementation

This process is used when radioactive liquid waste needs conversion into a product suitable for final storage. Cementation was the first solidification method used and it is now widespread. Cementation is a process used to immobilize low and intermediate level waste and a wide range of cements and additives are available to best fit with the type of waste to immobilize.

Figure 10: Example of liquid waste solidification by cementation [13].

Liquid waste solidification by Vetrification

The vitrification process is continuous and comprises two distinct steps. The very high activity liquid waste (FP and minor actinides) is first burned in a calciner, then mixed with a glass frit into a metal melter, where the glass is heated to the melting point by conduction; the highly stable glass matrix is then poured directly into a metal container.

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Figure 11: Example of Vitrification [13].

1.2.3.1.2 Treatment and Conditioning of solid waste

Solid wastes consist principally of contaminated and activated structural materials and system components, including concrete, reinforcement and structural steels, graphite, various metals, rubber, plastics, paper and other fibrous material. The principal purpose in the treatment and conditioning of solid waste are:

 Minimize the waste volume to make easier and cheaper waste processing and handling;

 Segregate the waste into groups according to the types of contaminant present to facilitate emplacement in the appropriate disposal facility;

 Process and package the waste in containers suitable.

Treatments of combustible and compactable materials are usually accomplished by using mechanical compaction and/or incineration. For materials that are not combustible and compactable, treatment and conditioning processes are limited to segmentation, to facilitate packaging, and to grounding within the containers, when appropriate.

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Super Compaction

To reduce the volume of solid waste, the super compaction could be the better solution to adopt. Dry compactable solid waste is packed into compactable drums and pressed with a force of up to 2000 t. The result is compacted pellet that is piled and storaged in larger drums and then fill of mortar.

Figure 12: Schematic of the super compactor [14].

Incineration

Incineration is dedicated to the treatment of combustible solid waste. It is a well proven technology in use in many States. The incineration of solid waste produces ashes that should be solidified, for example by grouting or cementation and gases are produced in off-gas stream to be filtered by a specific filtration system.

Plasma treatment is a method comparable with indigenization but with higher treatment temperature. The effect is that in addition to the burning of the combustible material, the incombustible material can melt.

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1.3 Decontamination

Decontamination is defined as the removal of contamination from surfaces of facilities or equipments by washing, heating, chemical or electrochemical action, mechanical cleaning or other techniques. In decommissioning programmes, the objectives of decontamination are [15]:

 To reduce radiation exposure;

 To salvage equipments and materials;

 To reduce the volume of equipments and materials requiring storage and disposal in licensed disposal facilities;

 To restore the sites and facilities, or parts of them, to an unconditional-use condition;

 To remove loose radioactive contaminants and to fix the remaining contamination in place in preparation for protective storage or permanent disposal work activities;

 To reduce the magnitude of the residual radioactive source for public health and safety reasons;

 To reduce the protective storage period or to minimise long-term monitoring and surveillance requirements.

A decontamination programme may also require facilities capable to treat secondary wastes from decontamination (e.g., processing chemical solutions, aerosols, filters, debris, etc). The concentrated wastes, representing a more significant radiation source, must be solidified and shipped for disposal in licensed disposal facilities unless properly treated in the waste reduction/recycling/reclamation processing alternative.

There are three main reasons for considering the use of decontamination techniques:

 The importance of removing contamination from components or systems to reduce dose levels in the installations. The access will be easier and it will become possible to use hands-on techniques for dismantling instead of the more expensive use of robots or automations;

Waste Management Strategies in five European Countries

DIPARTIMENTO

DI

INGEGNERIA

CIVILE

E

INDUSTRIALE

T

L

M

I

N

S

“Waste Management Strategies in

five European States”

Véronique Blanc

AREVA NC D&S DT

Contents

List of Figures

List of Tables

List of Abbreviations

SUMMARY

1

Introduction to Decommissioning and Waste

Management

1.1

Decommissioning and current development

1.1.1

Decommissioning status and perspective in the world

1.1.2

Decommissioning strategies

1.1.3

Factors influencing the selection of a

decommissioning strategy

1.2

Waste Management

1.2.1

Waste Classification

1.2.2

Exclusion, Exemption and Clearance

1.2.3

Waste management

1.3

Decontamination