UK EPR HINKLEY POINT C SITE
SUBMISSION OF GENERAL DATA AS APPLICABLE UNDER ARTICLE 37 OF THE EURATOM TREATY
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ARTICLE 37 SUBMISSION FOR HINKLEY POINT C Contents
CONTENTS
0. INTRODUCTION 13 0.1. BACKGROUND 13 0.2. CURRENT PROJECT STATUS 14 0.3. DEVELOPMENT PROGRAMME 15 0.4. PREPARATION OF GENERAL DATA FOR HINKLEY POINT C 15 1. THE SITE AND ITS SURROUNDINGS 17 1.1. GEOGRAPHICAL, TOPOGRAPHICAL AND GEOLOGICAL FEATURES OF THE SITE AND THE REGION 17 1.1.1. Geographical location 17 1.1.2. Relevant features of the region 18 1.1.3. Other installations relevant for their discharges 21 1.1.4. Facility position in relation with other Member States 23 1.1.5. Population in the nearest non-UK conurbations 24 1.2. SEISMOLOGY 25 1.3. HYDROLOGY 26 1.3.1. Description of surface water bodies 26 1.3.2. Description of the littoral area, tides, currents 32 1.4. METEOROLOGY 34 1.4.1. Winds 34 1.4.2. Precipitation 36 1.4.3. Extreme weather phenomena 37 1.5. NATURAL RESOURCES AND FOODSTUFFS 40 1.5.1. Water utilisation 40 1.5.2. Principal food resources in the region 40 1.6. OTHER ACTIVITIES IN THE VICINITY OF THE SITE 46 1.6.1. Air traffic 46 1.6.2. Rail traffic 46 1.6.3. Road traffic 46 1.6.4. Maritime activity 46 1.6.5. Industrial activities 47 2. THE INSTALLATION 49 2.1. MAIN FEATURES OF THE INSTALLATION 49 2.1.1. Nature and purpose of the installation 49 2.1.2. General design and installation plan for the EPR 50 2.1.3. Main operational and safety procedures 59 2.1.4. Reactor operating principles and safety provisions 66
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2.1.5. Summary of improvements for the EPR reactor 67 2.2. VENTILATION SYSTEMS AND THE TREATMENT OF GASEOUS AND AIRBORNE WASTES 68 2.2.1. Nature of gaseous radioactive effluent 68 2.2.2. Sources of gaseous radioactive effluent 69 2.2.3. Treatment of gaseous radioactive effluent 70 2.3. LIQUID WASTE TREATMENT 75 2.3.1. Nature of liquid radioactive effluent 75 2.3.2. Sources of liquid radioactive effluent 75 2.3.3. Treatment of liquid radioactive effluent 77 2.3.4. Storage and release of radioactive liquid effluent 81 2.3.5. Summary of optimisation measures for limiting the impact of liquid radioactive waste 81 2.4. SOLID WASTE TREATMENT 82 2.4.1. Waste treatment in the NAB 83 2.4.2. Waste treatment in the ETB 83 2.5. CONTAINMENT 87 2.5.1. Description of radioactive product containment barriers 87 2.5.2. Monitoring of containment barrier seals 89 2.6. DECOMMISSIONING AND DISMANTLING 91 2.6.1. Design for decommissioning 91 2.6.2. Outline of Regulatory provisions 91 3. RELEASE FROM THE INSTALLATION OF AIRBORNE RADIOACTIVE EFFLUENTS IN NORMAL CONDITIONS 93 3.1. AUTHORISATION PROCEDURE IN FORCE 93 3.1.1. Description of the current procedure 93 3.1.2. Waste discharge limits and associated requirements 94 3.2. TECHNICAL ASPECTS 95 3.2.1. Annual discharges foreseen 95 3.2.2. Origins of the radioactive effluents, their composition and physico-chemical forms 95 3.2.3. Management of these effluents, methods and paths of release 97 3.3. MONITORING OF DISCHARGES 99 3.3.1. Sampling, measurement and analysis of discharges 99 3.3.2. Principal features of monitoring equipment 100 3.3.3. Alarm levels and intervention actions 101 3.4. EVALUATION OF TRANSFER TO MAN 102 3.4.1. Models and parameter values used to calculate the consequences of the releases 102 3.4.2. Evaluation of activity concentration and exposure levels associated with discharge limits 112
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3.5. RADIOACTIVE DISCHARGES TO ATMOSPHERE FROM OTHER INSTALLATIONS 116 3.5.1. Source of aerial discharges 116 4. RELEASE FROM THE INSTALLATION OF LIQUID RADIOACTIVE EFFLUENTS IN NORMAL CONDITIONS 119 4.1. AUTHORISATION PROCEDURE IN FORCE 119 4.1.1. Description of the current procedure 119 4.1.2. Waste discharge limits and associated requirements 119 4.2. TECHNICAL ASPECTS 120 4.2.1. Annual discharges foreseen 120 4.2.2. Origins of the radioactive liquid effluents, their composition and physico-chemical forms 120 4.2.3. Management of the effluents, methods and paths of release 121 4.3. MONITORING OF DISCHARGES 123 4.3.1. Monitoring of liquid radioactive effluent from the LRMDS tanks and ExLWDS storage tanks 123 4.3.2. Monitoring of turbine hall drainage water 126 4.3.3. Alarm levels and intervention actions 127 4.4. EVALUATION OF TRANSFER TO MAN 128 4.4.1. Models and parameter values used to calculate the consequences of the releases 128 4.4.2. Evaluation of activity concentrations and external dose rates at the proposed discharge limits 132 4.5. RADIOACTIVE DISCHARGES INTO THE SAME RECEIVING WATERS FROM OTHER INSTALLATIONS 134 5. DISPOSAL OF SOLID RADIOACTIVE WASTE FROM THE INSTALLATION 139 5.1. SOLID RADIOACTIVE WASTE 139 5.1.1. Categories of solid radioactive waste including spent fuel and estimated amounts 139 5.1.2. Processing and packaging 147 5.1.3. Storage arrangements 151 5.2. RADIOLOGICAL RISK TO THE ENVIRONMENT 153 5.2.1. Assessment of risks to the environment 153 5.2.2. Precautions taken 154 5.3. ARRANGEMENTS FOR THE MOVEMENT AND DESTINATIONS OF THE DIFFERENT WASTE CATEGORIES TRANSFERRED OFF- SITE 155 5.3.1. Destinations of waste packages transferred off-site 155 5.3.2. Movement of LLW and VLLW packages 157 5.4. CRITERIA FOR CONTAMINATED MATERIALS TO BE RELEASED FROM THE REQUIREMENTS OF THE BASIC SAFETY STANDARDS 157
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5.4.1. National strategy, criteria and procedures for the release of contaminated and activated materials 157 5.4.2. Clearance levels established by competent authorities for disposal, recycling and reuse 159 5.4.3. Envisaged types and amounts of released materials 159 6. UNPLANNED RELEASES OF RADIOACTIVE EFFLUENTS 161 6.1. REVIEW OF ACCIDENTS OF INTERNAL AND EXTERNAL ORIGIN WHICH COULD RESULT IN UNPLANNED RELEASES OF RADIOACTIVE SUBSTANCES 162 6.1.1. Plant safety principles 162 6.1.2. Development of the safety report for Hinkley Point C 164 6.1.3. Design scope 165 6.1.4. Internal faults – design basis analysis 165 6.1.5. Multiple failure accidents under risk reduction category A (RRC-A) 168 6.1.6. Core melt accidents under category RRC-B 168 6.1.7. Consideration of additional safety related scenarios 169 6.2. REFERENCE ACCIDENTS TAKEN INTO CONSIDERATION 170 6.2.1. Category 4 (PCC-4) accidents 171 6.2.2. RRC-B operating conditions - core melt accident 173 6.3. EVALUATION OF THE RADIOLOGICAL CONSEQUENCES OF THE REFERENCE ACCIDENT(S) 175 6.3.1. Release to atmosphere 175 6.3.2. Release into an aquatic environment 182 7. EMERGENCY PLANS, AGREEMENTS WITH OTHER MEMBER STATES 183 7.1. INTERVENTION LEVELS ESTABLISHED FOR DIFFERENT TYPES OF COUNTERMEASURES 183 7.1.1. Control of contaminated or potentially contaminated food supplies 184 7.1.2. Permanent relocation 184 7.2. UK EMERGENCY PLANNING 184 7.2.1. Regional level: authorities 185 7.2.2. Emergency planning zones 185 7.2.3. National level: site operator 186 7.2.4. National level: authorities 186 7.3. ARRANGEMENTS FOR THE EXCHANGE OF INFORMATION WITH OTHER MEMBER STATES 187 7.3.1. Cross-border communications and agreements with other member states 187 7.4. EMERGENCY PLAN TESTING ARRANGEMENTS 187 8. ENVIRONMENTAL MONITORING 189 8.1. INTRODUCTION 189 8.2. TERRESTRIAL ENVIRONMENT 190
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8.2.1. Sample analysis 190 8.2.2. Dose rate measurements 192 8.3. MARINE ENVIRONMENT 192 8.3.1. Sample analysis 192 8.3.2. Dose rate measurements 192 8.4. DETECTION LIMITS 193 8.5. EXISTING AND FUTURE MONITORING PROGRAMME 193 8.6. MONITORING PROCEDURES 197 ANNEX A – ASTRAL FOOD CHAIN MODEL 199
Table 1.1 Demographic data for non-UK reference groups 25 Table 1.2 Tidal elevations at Hinkley Point C 33 Table 1.3 Current speeds at tidal diamond positions B and H 33 Table 1.4 Wind speed and direction – 1999 – 2002 data from Hinkley Point 36 Table 1.5 SAAR for all the catchments of interest in the Hinkley Point area 37 Table 1.6 Estimates of extreme low water level for 2002 38 Table 1.7 Extreme high water levels 38 Table 1.8 Varying design event rainfall depts (mm) for varying duration events 39 Table 1.9 Farm types 40 Table 1.10 Crops 41 Table 1.11 Livestock 41 Table 1.12 Exports of fish from the UK by importing country 2008 43 Table 1.13 Exports of fish and shellfish from the UK by importing country 2008 44 Table 1.14 Total landings for areas VIIf and VIIg 45 Table 3.1 Proposed limits for gaseous radioactive discharges from Hinkley Point C 94 Table 3.2 Distribution by activity of noble gases discharges in gaseous form 96 Table 3.3 Distribution by activity of fission and activation products discharged in gaseous form 96 Table 3.4 Detection limits for measurements of gaseous radioactive discharges from Sizewell B (for information purposes) 101 Table 3.5 Reference group location data 103 Table 3.6 Assumptions and parameters for atmospheric discharges 103 Table 3.7 Hinkley Point meteorological data PC-CREAM format for years 1999 to 2008 104 Table 3.8 Wind speed and frequency parameters used in long range model 105 Table 3.9 Other meteorological parameters used in long range model 106 Table 3.10 Assumptions/parameters ground deposition 106 Table 3.11 Food intake data for the local reference group 108 Table 3.12 Food intake data for Member State reference groups (kg y-1) 108 Table 3.13 Food transfer factors used in long range assessment (in Bq kg-1 per Bq m-2 s-1) 109 Table 3.14 Farming family habit data 109 Table 3.15 Inhalation and ingestion dose coefficients (Sv Bq-1) 111 Table 3.16 External exposure factors (dose rate per unit activity concentration in air) 112 Table 3.17 Local environmental concentrations from Hinkley Point C discharges 112 Table 3.18 Environmental concentrations in other Member States from Hinkley Point C using the long range model 113 Table 3.19 Annual dose to the local reference group by exposure pathway (µSv y-1) 113 Table 3.20 Annual effective dose to reference group in the Channel Islands by exposure pathway (µSv y-1) 114 Table 3.21 Annual effective dose to reference group in France by exposure pathway (µSv y-1) 114 Table 3.22 Annual effective dose to reference group in Republic of Ireland by exposure pathway (µSv y-1) 114 Table 3.23 Comparison of modelled environmental air concentrations in France from Hinkley Point C discharges with monitored readings 115
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Table 3.24 Discharges of gaseous radioactive wastes 117 Table 4.1 Proposed limits for liquid radioactive discharges from Hinkley Point C 119 Table 4.2 Distribution by activity of other radionuclides discharged in liquid form 121 Table 4.3 Determination of activity discharged and activity concentrations for each radionuclide category – liquid radioactive waste 125 Table 4.4 Detection limits for measurements of liquid radioactive waste 126 Table 4.5 DORIS – discharge and output related parameters 129 Table 4.6 DORIS – Sediment distribution coefficient (Kd) and concentration factors 130 Table 4.7 Habit data for the Republic of Ireland marine reference group 131 Table 4.8 Habit data for local reference group 132 Table 4.9 Activity concentrations calculated by DORIS for Bristol Channel (local compartment) (Bq kg-1) 132 Table 4.10 Activity concentrations calculated by DORIS for Irish Sea South (Bq kg-1) 133 Table 4.11 Activity concentrations calculated by DORIS for Celtic Sea (Bq kg-1) 133 Table 4.12 Effective doses to the local reference group from liquid discharges at the proposed limits 133 Table 4.13 Effective doses to the nearest Member State reference group from liquid discharges at the proposed limits 134 Table 4.14 Discharges of liquid radioactive wastes from other nuclear sites 136 Table 5.1 Categories of solid radioactive waste 140 Table 5.2 Estimated annual volumes of solid low level radioactive wastes produced during operation of two EPR units at Hinkley Point C 143 Table 5.3 Estimated annual volumes of solid intermediate level radioactive wastes produced during operation of two EPR units at Hinkley Point C 145 Table 5.4 Specified elements 158 Table 5.5 Annual quantities of released materials 159 Table 6.1 Category 2 events (PCC-2): design basis transients 166 Table 6.2 Category 3 events (PCC-3): benchmark incidents 167 Table 6.3 Category 4 events (PCC-4): benchmark accidents 167 Table 6.4 Category RRC-A internal accidental transients 168 Table 6.5 Activity release rate in the event of cladding failure 172 Table 6.6 Fraction of radionuclides discharged 174 Table 6.7 Key radionuclides and other radionuclides considered 175 Table 6.8 Lung class used in assessment for key radionuclides 176 Table 6.9 Amounts of significant radionuclides assessed in each scenario 176 Table 6.10 Deposition parameters used in the local assessment model 177 Table 6.11 Dose per unit deposition via food ingestion (from ASTRAL code) (mSv Bq-1 m-2) 178 Table 6.12 Long range atmospheric dispersion modelling parameters 179 Table 6.13 Maximum time integrated air concentrations 180 Table 6.14 Total surface concentrations 180 Table 6.15 Maximum committed effective doses to an adult member of reference groups by pathway for a fuel handling accident 180 Table 6.16 Maximum committed effective doses to an adult member of reference groups by pathway for the steam generator tubes rupture accident 181 Table 6.17 Maximum committed effective doses to an adult member of reference groups by pathway for the LOCA 181 Table 6.18 Maximum committed effective doses to an adult member of reference groups by pathway for a core melt accident 182 Table 8.1 Environmental monitoring in existence for Hinkley Point B 194
Figure 1.1 Hinkley Point site location 17 Figure 1.2 Main towns in the vicinity of Hinkley Point site 18 Figure 1.3 Site topography 20 Figure 1.4 Somerset and its districts 21 Figure 1.5 Other relevant installations 23 Figure 1.6 Nearest non-UK reference groups 24
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Figure 1.7 Watershed draining to the Hinkley Point foreshore (green) and to Great Arch Sluice (orange) 27 Figure 1.8 Tidal flood zone 30 Figure 1.9 Fluvial flood zone 31 Figure 1.10 Combined tidal and fluvial flood zone 32 Figure 1.11 Windrose for Hinkley Point C 35 Figure 1.12 International Council for the Exploration of the Sea areas 45 Figure 2.1 Schematic layout of the EPR power generation process 50 Figure 2.2 Hinkley Point C site layout plan 52 Figure 2.3 View of the lower section of the reactor building 54 Figure 2.4 Nuclear steam supply system 59 Figure 2.5 Safety injection system and residual heat removal system 60 Figure 2.6 EPR safeguard systems 61 Figure 2.7 Main design choices for improved safety 68 Figure 2.8 Effluents management process 68 Figure 2.9 Nature of gaseous radioactive effluent 69 Figure 2.10 Processes for discharging primary gaseous effluent 71 Figure 2.11 Treatment of gaseous effluent from ventilation 71 Figure 2.12 Treatment of gaseous effluent from the secondary system 73 Figure 2.13 Unit 1 stack discharges 73 Figure 2.14 Unit 2 stack discharges 74 Figure 2.15 Sources of liquid radioactive effluent 76 Figure 2.16 Treatment of primary liquid effluent 78 Figure 2.17 Treatment of spent liquid effluent 79 Figure 2.18 Treatment of steam generator blowdown water 80 Figure 2.19 Treatment of turbine hall drainage water 81 Figure 2.20 Flow diagram showing the active resin packaging process 85 Figure 2.21 Flow diagram showing the concentrate packaging process 86 Figure 2.22 Flow diagram showing the water filter packaging process 87 Figure 3.1 Windrose for Hinkley Point Site 105 Figure 3.2 Other nuclear establishments contributing to air discharges 118 Figure 4.1 Other nuclear sites contributing to marine discharges 137
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ARTICLE 37 SUBMISSION FOR HINKLEY POINT C Acronym List
ACRONYM LIST
AC Alternating current AEP Annual Exceedance Probability AGR Advanced Gas-Cooled Reactor ALARA As Low As Reasonably Achievable ALARP As Low As Reasonably Practicable AOD Above Ordnance Datum BAT Best Available Techniques BSS Basic Safety Standards CD Chart Datum CEFAS Centre for Environment, Fisheries & Aquaculture Science CILWDS Conventional Island Liquid Waste Discharge System COMAH Control of Major Accident Hazards CSTS Coolant Storage and Treatment System CVCS Chemical and Volume Control System DAW Dry Active Waste DDF Depth Duration Frequency DECC Department of Energy and Climate Change DEPZ Detailed Emergency Planning Zone DN Nominal Diameter EBS Extra Boration System ECURIE European Community Urgent Radiological Information Exchange EDF Electricité de France EFWS Emergency Feedwater System EIADR 99 Environmental Impact Assessment for Decommissioning Regulations 1999 EPR European Pressurised Water Reactor ERL Emergency Reference Level ETB Effluent Treatment Building ExLWDS Additional Liquid Waste Discharge System FAOSTAT Food and Agriculture Organisation of the United Nations statistical database FDP Funded Decommissioning Programme FEH Flood Estimation Handbook FPC(P)S Fuel Pool Cooling (and Purification) System FSR Flood Studies Report GDA Generic Design Assessment GDF Geological Disposal Facility GR Grid Reference GWPS Gaseous Waste Processing System HEPA High Efficiency Particulate Air HLW High Level Waste HSE Health and Safety Executive HVAC Heating, Ventilation and Air Conditioning IAEA International Atomic Energy Agency
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ARTICLE 37 SUBMISSION FOR HINKLEY POINT C Acronym List
ICRP International Commission on Radiological Protection ILW Intermediate Level Waste INSEE National Institute of Statistics and Economic Studies – France IRSN Institut de Radioprotection et de Sûrete Nucléairé IRWST In-containment Refuelling Water Storage Tank LLW Low Level Waste LLWR Low Level Waste Repository LOCA Loss of Coolant Accident LRMDS Liquid Radwaste Monitoring & Discharge System LWPS Liquid Waste Processing System LWR Light Water Reactor MCERTS Monitoring Certification Scheme (Environment Agency) ML Local Magnitude MOX Mixed-oxide MSRT Main Steam Relief Train MSSS Main Steam Supply System NAB Nuclear Auxiliary Building NRPB National Radiological Protection Board NSSS Nuclear Steam Supply System NVDS Nuclear Vent and Drain System OD Ordnance Datum ONR Office for Nuclear Regulation PCC Plant Condition Category PCmSR Pre-Commissioning Safety Report PCSR Pre-Construction Safety Report POSR Pre-Operation Safety Report PSA Probabilistic Safety Analysis PWR Pressurised Water Reactor REPPIR Radiation (Emergency Preparedness and Public Information) Regulations RHRS Residual Heat Removal System RIFE Radioactivity in Food and the Environment RNAS Royal Navy Air Station rpm Revolutions per minute RRC Risk Reduction Category SAAR Standard-period Average Annual Rainfall SG Steam Generator SGBS Steam Generator Blowdown System SIS Safety Injection System SiteLWDS Site Liquid Waste Discharge System TLD Thermo-Luminescent Dosimeter UK United Kingdom of Great Britain and Northern Ireland US EPA United States Environmental Protection Agency VLLW Very Low Level Waste
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ARTICLE 37 SUBMISSION FOR HINKLEY POINT C Chapter 0 – Introduction
0. INTRODUCTION
0.1. BACKGROUND 1. The generation of electricity from the existing fleet of nuclear power stations in the United Kingdom will reduce over the next twenty years. This is because many of the power stations will reach the end of their design life during this period. The United Kingdom of Great Britain and Northern Ireland (UK) Government has recognised the need for a new fleet of nuclear power stations to be built, to replace the generating capacity that will be lost from the existing fleet. This will ensure that the UK is able to fulfil its energy needs and meet its obligations to reduce carbon emissions. Subject to development consent from the UK Infrastructure Planning Commission, one of these new power stations will be built next to the existing power stations at Hinkley Point A and Hinkley Point B in Somerset.
2. This document contains the 'general data' as required by Article 37 of the Euratom Treaty for the operation of the new nuclear power station that is under development at Hinkley Point. This information is provided by Her Majesty's Government to the European Commission under Article 37 of the Euratom Treaty, as specified by the Commission Recommendation of 11 October 2010 (2010/635/Euratom). An opinion is sought from the Commission as to the acceptability of the radiological impact of the plan to dispose of radioactive waste from the proposed operation of the new nuclear power station at Hinkley Point, as described, is liable to result in the contamination of the water, soil or airspace of another Member State. The proposal is for the construction of Hinkley Point C; two European Pressurised Water Reactors (EPR) and associated facilities including interim storage facilities for spent fuel and Intermediate Level Waste (ILW).
3. This submission has been drafted prior to the event which occurred recently in Fukushima, in Japan, but has been reviewed in light of the feedback received at the time of writing.
4. Despite differences in the seismicity of the UK and the design of the plant, we recognise that there will be lessons to learn to further enhance the safety of the Hinkley Point C project. As information becomes available any lessons learned that are relevant to the design, construction and operations at the proposed new nuclear power plant at Hinkley Point C will be applied. This is in line with the UK regulatory approach of taking reasonably practicable steps to reduce the risk with a view to continuous improvement of safety standards.
5. In the UK, the Secretary of State for Energy and Climate Change has commissioned a report from the Office of Nuclear Regulation’s (ONR) Chief Nuclear Inspector, Dr Mike Weightman, investigating the implications of the Fukushima event for the UK’s nuclear fleet and highlighting any lessons learned. Dr Weightman published an Interim Report on 18th May 2011. The conclusions and recommendations of the Interim Report and their implications for Hinkley Point C are addressed in Chapter 6 of this submission.
6. The UK Government has undertaken a rigorous and systematic assessment of sites for the development of new nuclear power stations in England and Wales. A total of eight sites, including Hinkley Point, have been selected as being suitable for such development. The basis for selection is contained in the revised draft Nuclear National Policy Statement (1).
7. The Hinkley Point C nuclear power station, will be constructed by a consortium of the EDF (Electricité de France) Group of companies and Centrica plc (known as NNB Generation Company Ltd), on land that is owned by EDF. Hinkley Point is situated on the Severn Estuary approximately 8km from Bridgwater and the M5 to the south, 24km from Minehead in the west and approximately 10km from Burnham-on-Sea.
(1) Department of Energy and Climate Change, Revised Draft National Policy Statement for Nuclear Generation [EN-6], October 2010.
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8. An Article 37 submission has been made for the Hinkley Point B site and a decommissioning submission was made for Hinkley Point A site. Both submissions received favourable opinions.
9. The proposed Hinkley Point C power station will consist of a number of buildings which will house:
the nuclear reactors which produce energy; the auxiliary and safeguard systems which are connected to the reactors to ensure that the plant is controlled correctly during normal and accidental operating conditions; waste management facilities for the management and treatment of wastes that will arise from operations on the site; turbines and generators which convert the energy into electricity; sea water pumping station which is used for cooling; a laundry for the washing of work wear and other items from contaminated areas; and facilities for the decontamination of tools and equipment.
10. In addition, the site will also have:
an ILW interim storage facility for the safe storage of solid intermediate level radioactive waste until it can be disposed of to a dedicated waste management facility; and a spent fuel interim storage facility for the safe storage and monitoring of the spent fuel.
11. The EPR is a reactor of approximately 1650MWe and is of Franco-German design. It is a third generation reactor and has benefited from many design improvements progressively introduced as pressurised water reactor technology has developed. These improvements particularly relate to the safety of the installation, the radiological protection of workers and environmental performance in that the design incorporates measures to reduce the amount of waste produced and discharged for each unit of electricity that is produced.
0.2. CURRENT PROJECT STATUS 12. Construction of the Hinkley Point C power station will require a Development Consent Order from the Infrastructure Planning Commission. The process of obtaining a Development Consent Order from the Infrastructure Planning Commission requires parallel application for other key regulatory permissions. The Infrastructure Planning Commission must satisfy itself that, amongst other things, the relevant regulatory bodies will be able to grant the permissions for the proposed activities at the site. This view will be without prejudice to final regulatory decisions. Key regulatory permissions that will influence the decision of the Infrastructure Planning Commission are:
Nuclear Site Licence. Hinkley Point C will be classified as a ‘Nuclear Installation’ as defined by the Nuclear Installations Act 1965 (as amended). Operations on a Nuclear Installation are subject to the provisions of a Nuclear Site Licence that is granted by the Health and Safety Executive’s (HSE) Office for Nuclear Regulation (ONR)(2). Environmental Permit. Discharges and disposals of radioactive waste from Hinkley Point C will be subject to the provisions of an Environmental Permit, that is granted by the Environment Agency under the Environmental Permitting (England and Wales) Regulations 2010. The decision to grant a permit, and the extent to which it applies, will take account of the opinion of the European Commission on the data presented in this submission. Separate environmental permits will also be needed for the emissions from standby diesel generators and the discharge of cooling water to Bridgwater Bay.
(2) The ONR is the regulatory body with responsibility for regulating the nuclear power industry in the United Kingdom. On 1st April 2011, the ONR was established as an agency of the HSE. At the time of writing, legislation is planned to establish the ONR as a statutory body. The ONR will bring together the relevant nuclear regulatory functions of HSE and the Department for Transport.
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13. The HSE and the Environment Agency have formal arrangements for collaborative working on matters related to the regulation of nuclear sites in England and Wales. Both parties have been engaged in the preparation of the data contained in this submission.
14. In addition to the site specific applications described above to be submitted by the owner- operator of the Hinkley Point C site (NNB Generation Company Ltd), the developers of the EPR (EDF SA and Areva), have submitted the EPR design for Generic Design Assessment (GDA). The GDA process has been developed by the Environment Agency and the HSE and involves a rigorous and structured examination of design information over a period of several years. The GDA allows the safety, security and environmental implications of new power station designs to be assessed before an application is made for the permissions required to build that design at a particular site. At the end of the assessment period, the Regulators will issue reports on their findings which will state whether they judge the generic design of the plant to be satisfactory from safety, security and environmental aspects.
0.3. DEVELOPMENT PROGRAMME 15. It is intended that the power station will be operational for 60 years. Commissioning is preceded by a six year construction phase which includes site preparation work, construction of the buildings and installation of plant and equipment. The construction phase will consist of the following key steps:
Preparatory work, including excavations for the EPR platform, underground construction work (service tunnels) and soil consolidation (installation of foundation and bedding concretes). Construction of the buildings, including civil engineering work, and the installation of electrical and mechanical plant and equipment. The main sea outfall structure will consist of an underwater tunnel connecting the EPR platform to a diffuser that will be positioned approximately 2km from the shore. Commissioning tests will take place over a period of 20 months. They will be carried out on the various parts of the installation, to ensure that the expected performance levels are achieved. Nuclear fuel loading will be followed by a programme of commissioning tests prior to full operation of each reactor.
16. The first discharges of radioactive waste from the operation of the EPR reactors at Hinkley Point C are expected in 2018.
0.4. PREPARATION OF GENERAL DATA FOR HINKLEY POINT C 17. The general data presented in this submission have been prepared in accordance with Commission Recommendation of the 11 October 2010 on the Application of Article 37 of the Euratom Treaty. The general data relates to the following operations:
the operation of nuclear reactors (except research reactors whose maximum power load does not exceed 1MW continuous thermal load); and the predisposal management of radioactive waste arising from those operations.
18. The general data have been prepared from a range of sources. Key sources of information include:
The application for the Environmental Permit (Radioactive Substances Regulation) to allow the disposal of radioactive waste to the environment and by transfer to other premises. This document includes comprehensive information on the sources, management techniques, disposal routes and impacts for all radioactive wastes. The application for a Nuclear Site Licence to permit the use of nuclear matter. This document provides comprehensive information on safety related hazards, risks and mitigation measures. The application for a Development Consent Order that will allow the construction of the site and its associated infrastructure. This document provides detailed information on the description of the site and its surroundings, building design and environmental impacts.
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ARTICLE 37 SUBMISSION FOR HINKLEY POINT C Chapter 1 – The site and its surroundings
1. THE SITE AND ITS SURROUNDINGS
1.1. GEOGRAPHICAL, TOPOGRAPHICAL AND GEOLOGICAL FEATURES OF THE SITE AND THE REGION
1.1.1. Geographical location 19. The Hinkley Point power station site is located near Bridgwater, Somerset on the south-west coast of England. The new site, Hinkley Point C and the majority of associated permanent terrestrial infrastructure, will be situated on land currently used for agricultural purposes adjacent to the coast. Temporary infrastructure will be located to the south of this.
20. The centres of the two reactor buildings proposed for the Hinkley Point C units are located at the following co-ordinates, grid reference (GR) and latitude-longitude:
Unit 1 GR 320369, 145812 51.2059 deg - 3.1413 deg Unit 2 GR 320139, 145812 51.2059 deg - 3.1446 deg
Figure 1.1 Hinkley Point site location
Hinkley Point Somerset
Dorset Devon
Cornwall
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ARTICLE 37 SUBMISSION FOR HINKLEY POINT C Chapter 1 – The site and its surroundings
Figure 1.2 Main towns in the vicinity of Hinkley Point site
Hinkley Point C site
21. This submission refers to Hinkley Point C, a pair of third generation (GEN III) reactors. Neighbouring the Hinkley Point C site, is Hinkley Point A consisting of twin Magnox reactors (GEN I) which are being decommissioned and Hinkley Point B site, a twin Advanced Gas- cooled Reactor (AGR) (GEN II) which is still in operation.
1.1.2. Relevant features of the region
1.1.2.1. Topographical features 22. The topography (see Figure 1.3) is characterised by undulating countryside, terminating at Bridgwater Bay to the north, at a natural cliff line. The Hinkley Point C development site is divided into two by an east-west ridge which peaks at a maximum elevation of 35.3m Above Ordnance Datum (AOD).
23. To the north of the ridge, ground levels generally range between 14m AOD and 31m AOD. An east-west depression associated with the valley of the Hinkley Point C drainage ditch, passes through the centre of this area. The Hinkley Point C drainage ditch flows west to east, before turning to the north (at an elevation of 8.6m AOD) and discharging to the Hinkley Point foreshore at an elevation of 8.5m AOD. Elevations along the top of the cliff range from 10.7m AOD to 16.6m AOD and the foreshore at the foot of the cliff ranges between -4.2m AOD to and 5m AOD.
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24. To the south of the ridge, ground levels range from approximately 14m AOD to 22m AOD. The gently undulating relief continues from the depression referred to above with the land gently rising to 5.8m AOD and then increasing in gradient to a maximum of between 21.1 and 24.8m AOD. The land then gently falls towards the southern boundary of the application site where elevations typically range between 15m AOD and 16m AOD adjacent to Bum Brook. A small hillock is located towards the south-west corner of the application site where the land crests at an elevation of 28.7m AOD.
25. The majority of manmade topographical features present within the study area, are located within the land occupied by Hinkley Point B. These features include a double humped mound, the Nuclear New Build Induction Centre and an overflow car park.
1.1.2.2. Geological features 26. The Hinkley Point C site is located within the Bristol Channel Basin, a west-northwest to east-southeast trending basin approximately 30km wide and 150km long that contains an approximate 3km thickness of both Tertiary and Mesozoic sediments which were deposited in a synclinal trough, with an increase in sediment thickness towards the centre of the basin.
27. The geology of the area is dominated by the Lower Lias of the Liassic Group (Lower Jurassic). The coastal section at Hinkley Point is largely unfaulted. However, the wider site and surrounding area has several geological faults especially the Hinkley Point fault that is oriented in a northeast to southwest direction. The other faults oriented east to west, are located to the south and west areas of the site.
28. Mesozoic rocks of Jurassic and Triassic age are exposed along the cliff at the west of the Hinkley Point C Site. To the east of the Hinkley Point C Site, the Lower Lias cliff line presents the flat low-lying ground of the River Parrett estuary which forms an extensive area of Quaternary sedimentation known as the Somerset Levels. Similarly, it was noted that the outer Parrett estuary consists of estuarine and marine Holocene deposits characterised by reclaimed coastal marshes and mudflats. This extensive alluvial plain to the immediate east of Hinkley Point is interrupted by a ridge formed by head deposits overlying ‘Liassic limestones'. The ridge projects into the near shore. Hinkley Point is bounded on both the east and south by low lying land. This alluvial land runs along the line of a valley that forms the southern boundary between the existing power stations and Stolford.
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Figure 1.3 Site topography
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1.1.2.3. Local demographic features 29. The three surrounding districts of Sedgemoor, West Somerset and Taunton Deane (see Figure 1.4) have approximate population sizes of 112,000, 35,400 and 108,200 respectively, and a combined population of approximately 256,000 according to the mid-year 2007 estimates reported by the Office for National Statistics, 2008.
30. Sedgemoor and Taunton Deane are of a similar population size, with a population density in keeping with the South West trend. In contrast, West Somerset is more rural in nature, exhibiting a lower population number and a significantly lower population density than local, regional or national averages.
Figure 1.4 Somerset and its districts
HPC
Mendip
West Somerset Sedgemoor
SOMERSET
Taunton Deane South Somerset
1.1.3. Other installations relevant for their discharges 31. When considering the radioactive discharges (both gaseous and liquid) from other facilities, the following approach was taken in order to identify which facilities to take into consideration.
32. The most up to date list of firms holding Nuclear Site Licences (according to the Nuclear Installations Act 1965 - as amended) was obtained from the HSE website. The locations of these sites and the locations of the nearest non-UK receptors(3) were plotted using Geographical Information System software.
33. Two circles were drawn on the map centred on each receptor and with radii equal to the distance from each receptor to Hinkley Point C (see Figure 1.5). All the licensed sites within the circle were considered relevant in terms of their contribution to discharges to liquid or gaseous media, based on their distance from the receptor.
(3) A receptor is defined as the nearest point of landfall in the nearest neighbouring country.
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34. For aerial discharges, the other installations considered are:
Devonport Royal Dockyard Ltd; Hinkley Point A; Hinkley Point B; Inutec Ltd (Winfrith); and Research Sites Restoration Ltd (Winfrith).
35. For marine discharges, the other installations are:
Hinkley Point A; Hinkley Point B; Quotient Biosciences Ltd - Cardiff; and Wylfa.
36. Additional sites at Berkeley and Oldbury are considered for marine discharges as these both discharge into the same body of water as Hinkley Point C. While Trawsfynydd is within the radii considered for marine doscharges, it is an inland site and does not make direct discharges in to the sea and is therefore not considered.
37. The location of Hinkley Point C in relation to other member states is presented in Section 1.1.4. Given the prevailing wind direction in the British Isles from the southwest, the majority of gaseous discharges will be carried away from continental Europe. Therefore only a very small percentage of activity discharged from any facility in the UK, could affect reference groups in France. Air concentration falls rapidly with distance, this combined with the prevailing wind direction means that only sites that are closer to the reference group than Hinkley Point C have been considered.
38. The operations of such licensed premises are subject to the requirements of Article 37 of the Euratom Treaty. Solid radioactive waste from other installations is not considered. This is because it is either stored on the site or disposed of by transfer to other, appropriately licensed premises. During this process the waste is stored and transported in packages that isolate it from the environment. There will be no disposals of solid radioactive waste on the Hinkley Point C site.
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Figure 1.5 Other relevant installations
1.1.4. Facility position in relation with other Member States 39. The distances from the Hinkley Point site to the nearest land belonging to another Member State of the European Union are as follows:
Hinkley Point to County Wexford in the Republic of Ireland is 250km; and Hinkley Point to Arrondissement of Cherbourg in France is 185km.
40. It should be noted that the Channel Islands are not a Member State of the European Union. Nevertheless for completeness of information, this submission considers the impacts on the population of Alderney in the Channel Islands (distance: 175km).
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41. Distances are calculated as the shortest direct line for France and the Channel Islands (air dispersion) and as the shortest route through the sea for the Republic of Ireland (water dispersion pathway) (see Figure 1.6).
42. Within these regions, the nearest towns to the Hinkley Point C power station site are Alderney in the Channel Islands, Auderville in Cherbourg and Rosslare (Rosslare Harbour) in County Wexford.
Figure 1.6 Nearest non-UK reference groups
6°0'0"W 5°0'0"W 4°0'0"W 3°0'0"W 2°0'0"W
# Rosslare 52°0'0"N
52°0'0"N
250 km # E Hinkley Point C 51°0'0"N
51°0'0"N
185 km 175 km 50°0'0"N
50°0'0"N
Auderville
Alderney
# # 6°0'0"W 5°0'0"W 4°0'0"W 3°0'0"W 2°0'0"W
1.1.5. Population in the nearest non-UK conurbations 43. The nearest conurbations, towns and villages belonging to other Member States of the European Union are listed in Table 1.1 along with their population size, population density and area.
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Table 1.1 Demographic data for non-UK reference groups
Density Country County Towns Population Area (km2) (inhabitant km-2) Republic of County Wexford - 131,749 2,365.27 56 Ireland(4) Rosslare 1,791 18.99 94 Cahore 374 18.29 20 Wexford 8,854 2.26 3,918 Ladys Island 569 13.6 42 Churchtown 285 21.81 13 France(5) Arrondissement - 197,357 1,762 112 of Cherbourg Auderville 304 4.33 70 Cherbourg-Octeville 41,563 14.26 2,915 Beaumont- Hague 1,340 7.9 170 Herqueville 159 2.91 55 Querqueville 5,677 5.56 1,021 Equerdreville- Hainneville 18,084 12.83 1,410 Digulleville 300 7.89 38 Greville-Hague 808 10.03 81 Jobourg 458 10.15 45 Omonville-la- Petite 126 6.16 20 Saint-Germain- des-Vaux 428 6.36 67 Urville-Nacqueville 2,306 11.58 199 Vauville 389 16.35 24 Omonville-la- Rouge 539 4.29 126 Eculleville 47 2.33 20 Channel Jersey Whole Island 87,186 116.2 750 Islands Jersey Saint-Helier (Civil Parish) 28,310 8.6 3,292 Alderney Whole Island 2,294 8 287 Guernsey Whole Island 59,807 62.9 951 Sark Whole Island 600 5.45 110
1.2. SEISMOLOGY 44. The region around Hinkley Point has a low level of seismicity. It is geologically formed of old basement formations belonging to the variscan domain and is a typical intra-plate domain, far from active plate closest boundaries.
45. The seismic hazard of the UK, has been studied in detail by the Seismic Hazard Working Party, a group of experts formed in 1982. The Seismic Hazard Working Party has produced two reports specifically for Hinkley Point site, in 1987 and 1991. The 1991 report gives a Uniform Hazard Spectrum at 10-4 annual probability.
46. The UK is located in the north-western sector of the Eurasia plate. The two closest boundaries are the Mid Atlantic Ridge to the west and the collision zone between Africa and Eurasia to the south. It is an intraplate area and as such, levels of seismicity are
(4) Central Statistics Office Ireland at http://www.cso.ie/census, 2007 (5) 1999] census on France's National Institute of Statistics and Economic Studies (Institut National de la Statistique et des Études Économiques: INSEE)
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characteristically low. Generally, the crust of Britain and Ireland consists of three ‘blocks’, namely:
Laurentian crust north of the Highland Boundary Fault; Avalonian crust south of the Iapetus Suture Zone; and an intervening zone of accreted terranes.
47. In Cornwall and Devon, the maximum historical magnitude is estimated to be around 4.2 ML (local magnitude). On south coast of Wales, the historical seismicity is more important, with several earthquakes above 4.0 ML and a maximum observed magnitude around 5.2 ML. Instrumental recorded seismicity gives the same picture: to the south of the Bristol Channel, earthquakes are rare and the maximum recorded magnitude is 3.6 ML near Bideford (west coast of Devon). In Wales, earthquakes are more numerous, and the maximum magnitude reached 4.1 ML near Cardiff.
1.3. HYDROLOGY
1.3.1. Description of surface water bodies 48. The general hydrological overview is subdivided into the following sub-sections:
surface water courses; surface water outfalls; groundwater recharge; and flood baseline.
1.3.1.1. Surface water courses 49. This section addresses the baseline characteristics relating to surface water courses that may be affected by the development.
50. Each watercourse of interest to the site is described in the sections that follow. Each of the catchment areas that correspond to the watercourses discussed below, are shown in Figure 1.7.
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Figure 1.7 Watershed draining to the Hinkley Point foreshore (green) and to Great Arch Sluice (orange)
Hinkley Point C Drainage Ditch 51. Hinkley Point C drainage ditch flows through the main Hinkley Point C development area. Hinkley Point C drainage ditch’s catchment is undulating with generally shallow slopes. The stream runs from west to east with its source at 14.5m AOD. At GR 320310, 145840, Hinkley Point C drainage ditch’s alignment is rotated through 90° such that it flows from south to north where it discharges onto the foreshore at GR 320290, 146150. The maximum elevation of the catchment is 37.4m AOD and the lowest point where it discharges to the foreshore is at an elevation of 8.3m AOD. The catchment will be bisected by the new Hinkley Point C Power Station platform at a location upstream of the source, leaving a natural catchment of 0.23km2.
Holford Stream 52. The stream’s flow path is from west to east and is culverted under Wick Moor Drove from where it remerges downstream of Wick Moor Drove and having flowed not more than 10m, it once again enters a culvert under a track.
53. To the west of Wick Moor Drove the catchment slopes are relatively steep, while to the east of Wick Moor Drove the catchment is characterised by low lying flat pasture land. At present, land use for the catchment is 100% rural with no urbanisation. In general, the riparian vegetation is dense, comprising long grasses, thistles and reeds.
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Bayley’s Brook and Bum Brook 54. Bayley’s Brook flows from southwest to northeast through the village of Shurton and converges with Bum Brook at GR 320320, 144530.
55. Bum Brook flows from west to east and has two tributaries - Bailey’s Brook and Stogursey Brook. The stream then divides into two channels at Wick (East and West Brook). The stream has several hydraulic structures along its course including bridges, culverts, a weir and a ford.
56. Along this reach of Bum Brook, the river flows through farmland and gardens. The watercourse is culverted under Wick Moor Drove and under the lane to and from Wick.
57. The vegetation along the course of Bum Brook upstream of Wick Moor Drove is characterised by farmland verges which have become overgrown with long grasses, wheat and reeds. Downstream of Wick Moor Drove, the bank vegetation comprises overgrown woodland vegetation such as long grasses, nettles, hawthorn and brambles.
Stogursey Brook 58. Stogursey Brook flows southwest to northeast through the town of Stogursey, then through farmland, through the hamlet of Newnham and once again through farmland. The stream then turns to the east and passes under Wick Moor Drove and then turns to the northeast where it converges with Bum Brook 130m downstream of Wick Moor Drove.
59. The vegetation along Stogursey Brook is very dense. The dense vegetation continues downstream of the road until it turns northeast where the banks become vegetated with a few trees and brambles.
East Brook and West Brook 60. To the east of Stogursey, Bum Brook (having converged with Stogursey Brook to the west) divides into East Brook and West Brook at GR 321687, 144515. East and West Brook then flow parallel with each other for 1.52km before converging immediately upstream of Great Arch sluice. The landscape along East and West Brook is flat, low lying farmland, used primarily as pasture.
1.3.1.2. Surface water outfalls 61. With the exclusion of Hinkley Point C Drainage Ditch which discharges to the Hinkley Point foreshore, the fluvial network in the area has two outfalls. The largest outfall point is Great Arch Sluice which is located immediately downstream of East and West Brook, which have a collective catchment area in excess of 20km2. The Great Arch Sluice is located at GR 322717, 145947. The water flows through a culvert under the flood embankment into a large inspection chamber. This chamber contains the flap that prevents tidal flow upstream.
62. The other outfall point is Cole Lane Sluice located to the immediate east of the Hinkley Point B power station. Here the stream flows under the flood defence embankment and into the Bristol Channel, draining the series of ditches located to the south of the existing Hinkley Point power stations.
1.3.1.3. Groundwater recharge 63. The rainfall recharge provides the driving mechanism for groundwater flow. Groundwater springs out at outcrops of lower permeability strata and also provides baseflow to surface watercourses. It is likely that the watercourses are structurally controlled, with those flowing west to east following the trends of the strike faults and fold axes and those flowing southwest-northeast, following structures parallel to the Hinkley Point Fault (as well as merely following the topography to the Bristol Channel).
64. Given the topography and likely groundwater flow regime it is likely that the surface watercourses are at least in partial hydraulic continuity with the groundwater, probably with
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significant groundwater contributions to baseflow and possible groundwater recharge in places. Streams from the south of Stogursey running off the Mercia Mudstones some 2km to the south of the Hinkley Point C Site, could contribute to groundwater recharge under some circumstances (e.g. runoff during low water table conditions).
65. Section 1.4 shows rainfall data obtained from the Meteorological Office.
1.3.1.4. Flood baseline 66. There are areas of the site that are located within Environment Agency Flood Risk Map Zones 2 and 3. Flood Zone 3 is characterised as having greater than a 1% annual probability (1 in 100 years) of river flooding or greater than a 0.5% annual probability (1 in 200 years) of tidally influenced river flooding or flooding directly from the sea. Flood Zone 2 is characterised as having greater than a 0.1% annual probability (1 in 1,000 years) of river flooding or tidal flooding.
67. The Environment Agency tidal, fluvial and combined tidal and fluvial Flood Zones are illustrated in Figure 1.8, Figure 1.9 and Figure 1.10 respectively. A section of Holford Valley in the south of the site is at risk from tidal flooding based on the 0.5% AEP (annual exceedance probability) tide under the assumption that no flood defences are present. However, a flood defence embankment is located between Hinkley Point and Stolford Point which affords a standard of protection against the 1 in 100 year event water level.
68. The site is protected against flooding from the sea by its elevation. This elevation of the platform on which the plant is to be constructed, will be created at 14.0m AOD. The elevation has been determined in part on the basis of requirements to set the platform at a height which is sufficient to eliminate the potential for tidal and wave flooding taking account of storm surges, tsunamis, sea swell and predicted sea level rise over the development's life from climate change. A sea wall is provided to prevent erosion of this platform and wave overtopping of this wall is limited to acceptable amounts.
69. The site drainage system will be designed to withstand rainfall events up to the 1% AEP standard without the system being surcharged. Additionally, designs will ensure that surface water flooding that might arise during events that exceed the design standard will be effectively routed such that the consequences will be minimised up to the 0.01% AEP event. In addition, other measures are built into the design of the site to ensure that in the event of failure of part of the drainage network, water is not able to enter key safety related buildings.
70. The site drainage will include a groundwater drainage system, designed to maintain groundwater levels no higher than about six metres below the station platform level, in order to limit the flotation forces on deep buildings. This drainage system will discharge into the station cooling water system.
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Figure 1.8 Tidal flood zone
Flood Zone 2 Flood Zone 3
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Figure 1.9 Fluvial flood zone
Flood Zone 2 Flood Zone 3
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Figure 1.10 Combined tidal and fluvial flood zone
Flood Zone 2 Flood Zone 3
1.3.2. Description of the littoral area, tides, currents
1.3.2.1. Intertidal area 71. The intertidal area of the coastline between Brean Down to the north of Hinkley Point C and Lilstock to the west, is predominantly mud and sand. Between Lilstock and Blue Anchor, further to the west, the intertidal area consists of stone and rock interspersed with mud and sand.
72. The area has a large tidal range, resulting in fast running water and large areas of exposed mud and sand. Beach and water activities are restricted at some locations due to the dangerous nature of the soft mud and sand and the strong tides.
1.3.2.2. Tidal ranges and wave climate 73. The tidal range in the Bristol Channel and Severn Estuary increases further up the estuary. At Hinkley Point itself, Mean Low Water Springs is 0.8m Chart Datum (CD) and Mean High Water Springs is 11.5m CD (Table 1.2). Therefore, there is a mean spring tidal range of 10.7m, thus making this system classed as ‘megatidal’. Note that Chart Datum relates to Watchet (14km to the west of Hinkley Point) and is given as -5.8m Ordnance Datum Newlyn (Admiralty Chart 1123).
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Table 1.2 Tidal elevations at Hinkley Point C
m OD m CD Highest Astronomical Tide 7.1 13.0 Mean High Water Springs 5.6 11.5 Mean High Water Neaps 2.5 8.4 Mean Sea Level 0.1 6.0 Mean Low Water Neaps -2.3 3.6 Mean Low Water Springs -5.1 0.8 Lowest Astronomical Tide -6.1 0.2
74. Wind data for Hinkley Point are presented in Section 1.4. As far as wind effect on wave climate is concerned, winds at Hinkley Point are dominated by winds from the west- northwest. When these occur there is an effective fetch of approximately 400km. Winds from the northwest-northeast occur around 10% of the time and fetch is around 23km. Winds are least frequent and are weakest from the northeast-south sectors where fetch is also minimal. In combination with the regional bathymetry and coastal configuration, the result is that Hinkley Point is mostly subjected to waves from the west-northwest.
75. The wave climate of the Bristol Channel contains a significant swell component and is dominated to a major extent by the open fetch to the North Atlantic. Wave energy is focussed at headlands, although the offshore banks dissipate energy. The annual 10% exceedance significant wave height was reported as 1m to 2.5m.
76. Waves play a significant role in sediment transport at the coast itself. Tidal circulation within the Bristol Channel is complex but is considered to be ebb-dominant with large volumes of sand and mud being transported westward in the channel’s centre.
77. Conversely, the dominant waves from the west and southwest cause littoral drift, predominantly eastward and flood-dominant sediment transport occurs for limited periods throughout the year along the coastal fringe.
1.3.2.3. Currents 78. Tidal currents flow along the west-east axis of the Bristol Channel and are locally very strong. Maximum tidal currents increase up the channel from about 0.72m s-1 off Lundy, to 2.42m s-1 in the Bristol Deep off Avonmouth.
79. Various studies indicate that of the flood tide, the strongest flow is generally on the north side of the Bristol Channel off Barry, with much reduced velocities on the southern side of the channel, i.e. near Hinkley Point.
80. Typical tidal speeds from the Admiralty tidal diamonds B (6km north of Watchet) and H (between Steep Holm and Brean Down) are presented in Table 1.3.
Table 1.3 Current speeds at tidal diamond positions B and H
Neaps B Springs B Neaps H Springs H Flood (m s-1) 0.75 1.45 0.85 1.6 Ebb (m s-1) 0.80 1.5 0.80 1.5
81. More specifically at Hinkley Point, there have been a number of surveys associated with thermal studies for the existing Hinkley Point A and B stations and the proposed new units on the C site. A significant feature is that the currents off Hinkley Point are much reduced in comparison to those offshore to the north.
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82. On Neap tides the data show that ‘upstream’, the ebb continues after the flood has commenced ‘downstream’ and the delay in turning is longer at Spring tides due to frictional effect and relatively strong gradient. This contributes to the main flood starting earlier off Hinkley Point on Neaps at High Water +6 hours and the existence of a ‘young flood’ hugging close inshore at High Water -6 hours on Springs. There is evidence of a net flow seawards in the Bristol Channel and off Hinkley Point, with the latter somewhat accentuated by the freshwater discharge from the River Parrett. The buoyancy associated with the freshwater affects residual flows in the region.
1.3.2.4. Mixing properties 83. The extent of vertical and horizontal mixing are governed by a number of factors, including the speed of the tide, water depth and the effective roughness of the bed. While the estuary is vertically well mixed everywhere with respect to temperature, significant water density gradients can exist due to the unique character of the load of suspended particulate.
84. Freshwater flow into the estuary is significant and can often exceed 1000m3 s-1 for the Severn and Wye combined. Within Bridgwater Bay itself, the River Parrett can have significant flow, peaking at approximately 200m3 s-1, contributing to a mean salinity near Hinkley Point of 30.3 (+/-3.18) with a minimum value of 23.3 and a maximum value of 33.3. These values are based on all the profiling data through the water column and across the full range of tidal conditions (low and high water, spring and neap tides).
85. The area around Hinkley Point and Bridgwater Bay has high tidal ranges and fast tidal speeds, with flows that are strongly constrained by bathymetry. Due to the different orientation of the major estuary axis between the Bristol Channel and the Severn Estuary, it could be expected that in the Bridgwater Bay region, particles adverted landwards on the flood tide, may not be returned to the same location by the ebb. At the smaller scale, model results undertaken to investigate mixing of plumes from the ‘A’ power station discharge, indicated very little mixing with offshore waters, i.e. those to the northwest in the centre of the estuary axis, but that there would be considerable dilution resulting from along streamline mixing.
1.4. METEOROLOGY 86. The nearest meteorological data for Hinkley Point are recorded at the Hinkley Point Weather Station. The area has a temperate climate which is typically wetter than the UK average.
1.4.1. Winds 87. The Hinkley Point Weather Station has recorded an average temperature of 11.2°C. The station shows that the dominant wind direction is from the northwest with 42.4% of all recorded wind directions coming from between 240° and 315°. Hinkley Point also has minor southerly and north easterly components. The wind rose showing wind speed and frequency of direction is presented as Figure 1.11.
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Figure 1.11 Windrose for Hinkley Point C
350° 0° 10° 340° 8000 20° 330° 30° 320° 40° 6000 310° 50°
300° 60° 4000
290° 70°
2000 280° 80°
270° 90°
260° 100°
250° 110°
240° 120°
230° 130°
220° 140° 210° 150° 200° 160° 190° 180° 170° 0 3 6 10 16 (knots) Wind speed 0 1.5 3.1 5.1 8.2 (m/s)
88. Table 1.4 summarises the wind speed frequencies for each wind sector at 15° intervals. The data have shown that stronger winds tend to associate with north-westerly winds with 95.5% of wind speeds greater than 13m s-1 occurring when wind directions are between 240° and 315°.
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Table 1.4 Wind speed and direction – 1999 – 2002 data from Hinkley Point
Start of Frequency of wind speeds in sector (%) sector (°) Calm 0.5-1 m s-1 1-2 m s-1 2-3 m s-1 3-7 m s-1 7-13 m s-1 >13 m s-1 Total* 0 0.03 0.07 0.28 0.26 0.85 0.17 0.00 1.66 15 0.02 0.07 0.28 0.36 1.20 0.29 0.00 2.22 30 0.02 0.06 0.29 0.37 1.57 0.45 0.00 2.75 45 0.01 0.06 0.26 0.43 2.06 0.67 0.01 3.50 60 0.03 0.05 0.29 0.43 2.31 0.79 0.00 3.90 75 0.02 0.07 0.27 0.41 1.86 0.51 0.00 3.14 90 0.03 0.05 0.27 0.38 1.35 0.28 0.00 2.36 105 0.02 0.09 0.30 0.37 1.35 0.20 0.00 2.33 120 0.02 0.06 0.29 0.50 1.49 0.29 0.00 2.65 135 0.01 0.06 0.26 0.53 1.83 0.28 0.00 2.98 150 0.02 0.05 0.27 0.50 1.81 0.41 0.00 3.06 165 0.01 0.07 0.27 0.61 2.28 0.61 0.00 3.85 180 0.02 0.05 0.28 0.63 3.08 0.81 0.01 4.87 195 0.02 0.05 0.28 0.61 3.34 1.01 0.00 5.30 210 0.02 0.05 0.25 0.52 3.05 0.86 0.00 4.74 225 0.02 0.04 0.23 0.45 2.83 0.94 0.00 4.51 240 0.02 0.05 0.20 0.35 2.58 1.27 0.03 4.49 255 0.02 0.05 0.20 0.35 2.61 1.85 0.08 5.15 270 0.02 0.06 0.24 0.41 2.64 2.56 0.35 6.27 285 0.03 0.05 0.29 0.60 6.14 4.47 0.32 11.89 300 0.02 0.08 0.37 0.96 5.62 2.91 0.07 10.03 315 0.01 0.06 0.33 0.69 2.56 0.91 0.02 4.58 330 0.02 0.05 0.33 0.42 1.00 0.23 0.00 2.04 345 0.02 0.07 0.31 0.36 0.70 0.12 0.00 1.58 All 0.46 1.41 6.62 11.49 56.06 22.91 0.91 99.85 Missing 0.15 data * total dataset – 87,670 hours
89. An assessment of Pasquill stability classes has also been made to indicate the presence of temperature inversions and to input into the assessments of the dispersion of radioactive releases to atmosphere. Inversions are created by a very stable boundary layer where temperature increases with altitude. The Pasquill stability class data from 1999-2007 have shown that only 1.1% of the nine year period is classed as stable (classes F+G) with 91.1% of the dataset being classed as neutral or stable. This indicates a low frequency of potential inversion conditions.
1.4.2. Precipitation 90. This section illustrates details on rainfall in the different catchments described in Section 1.3, with special emphasis on the effects on surface water flows.
91. Annual average rainfall data in the Hinkley Point area, have been derived from a number of sources. It is generally accepted throughout the UK that, in the absence of long duration rainfall records, the catchment information provided by the Flood Estimation Handbook
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(FEH) CD-ROM(6) provides the most accurate and current information with respect to mean annual rainfall.
92. The FEH CD-ROM gives values for the annual average rainfall (also known as ‘Standard- Period Average Annual Rainfall - SAAR) for the period 1960–1990. The FEH CD-ROM has been used to generate a catchment specific value for this study for all the catchments. The catchments are illustrated in Figure 1.7 and their corresponding SAAR values are given in Table 1.5 below.
93. It is noted that average annual rainfall was specified in the Meteorological Office HM33(7) report for the site as 763mm y-1 which is in good agreement with the range of data presented in Table 1.5.
Table 1.5 SAAR for all the catchments of interest in the Hinkley Point area
Catchment ID Catchment Name SAAR (mm) 1 Hinkley Point C Drainage Ditch 753* 2 Holford Stream upstream of source 753 3 Holford Stream upstream of Wick Moor Drove 752 4 Holford Stream upstream of Sharpham Sluice 748 5 Bum Brook upstream of Wick Moor Drove (including Bayley’s Brook) 812 6 Stogursey Brook upstream of Wick Moor Drove 854 Fluvial network upstream of Sharpham Sluice (with exception of Holford Stream 7 825 and drainage ditches to south of Hinkley Point a and B) 8 Fluvial network upstream of Great Arch Sluice 817 * This catchment is not identifiable on the FEH CD-ROM and SAAR is assumed to be the same as that for the Holford Stream catchment.
94. It can be seen from Table 1.5 that a value for SAAR in the range 747–812mm y-1 is the most applicable for the Hinkley Point C site. The higher values given for the other catchments (catchment ID 4-7) are affected by the orographic influence of the East Quantocks, from where Stogursey Brook is sourced.
1.4.3. Extreme weather phenomena
1.4.3.1. Extreme low water level 95. Extreme low water level is determined by considering the combination of the astronomical tide and low water setdown (due to wind, pressure etc.), with a probability of occurring once every 10,000 years. The extreme low water level directly gives the design parameter to be taken into account for design studies.
96. In accordance with safety principles, the upper limit of the 70% confidence interval for the low water level with a 10,000 year return period, is studied for the site. As can be seen from Table 1.6 below, the extreme low water level for Hinkley Point is -7.62m OD (Ordnance Datum).
(6) Flood Estimation Handbook CD-ROM (2010) Version 3.0
(7) Meteorological Office HM33 rainfall reports, 2009
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Table 1.6 Estimates of extreme low water level for 2002
Return Period (years) Mean Value (m OD) 70% upper (m OD) 10 -6.228 -6.342 50 -6.339 -6.602 100 -6.379 -6.720 500 -6.460 -7.012 1,000 -6.490 -7.145 10,000 -6.571 -7.621
1.4.3.2. Extreme high water level 97. For the Hinkley Point site, the most critical marine data are those which indicate elevations at states of high water and particularly so-called ‘extreme’ water levels.
98. In the open ocean, tidal ranges are up to 1m. Much larger ranges occur on some continental shelves and in some, shelf sea tidal ranges may exceed 10m, e.g. Bay of Fundy, Baie du Mont St Michel and the Argentine Shelf. The Severn Estuary is one such system.
99. Surges are common in the Severn Estuary, where the astronomical tide is affected by meteorological conditions, particularly the movement of atmospheric depressions north- eastward across the British Isles. Small surges are frequent, with positive surges of about 1m occurring every one or two years and those of 2m or above occurring on a decadal time scale.
100. There is also a potential response to meteorological forcing at frequencies similar to the tide, resulting in intense surges which are generated and then decay during a single semi-diurnal tidal cycle.
101. The present calculated extreme high water levels at Hinkley Point are given in Table 1.7 below. It can be seen that these water levels are significantly lower than the proposed 14.0m AOD level of the operational power station platform. Further considerations related to the combined tidal and fluvial flood zones are contained in Section 1.3.1.4.
102. Feedback from the Fukushima event known at the time of writing has been considered in the assessment of external flooding hazards. Relevant conclusions for the Hinkley Point C site are presented in Chapter 6 of this submission.
Table 1.7 Extreme high water levels
Return period (years) AEP Water levels (m AOD) 1 in 1 100% 7.7 1 in 10 10% 8.0 1 in 50 5% 8.2 1 in 100 1% 8.4 1 in 500 0.2% 8.6 1 in 1,000 0.1% 8.7 1 in 10,000 0.01% 9.0
1.4.3.3. Tornadoes 103. On average, 33 tornadoes are reported each year in the UK. This average is based on a 30 year period, though in reality yearly figures vary dramatically.
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104. Most tornado reports are from the western Midlands, eastern Midlands, central-southern England, south-eastern England and East Anglia. Some occur in south-western England, north-western England, north-eastern England and Wales. Tornadoes are rare in Northern Ireland and Scotland.
105. For the generic UK site, it is assumed that an average of 33 tornadoes per year occur over a land area of 150,000 square kilometres which corresponds approximately to the area of England and Wales. On average it is supposed that a tornado is of 50m width and that it travels for 3km. This gives an approximate frequency of a structure being struck by a tornado in England and Wales as 3.3 x 10-5 y-1.
106. There is an absence of local knowledge of any history of tornado events in the region of the Hinkley Point Power Stations. Two tornadoes have, however been reported in the vicinity of Hinkley Point since 1980. The first, a T3 tornado/waterspout passed through the Severn Road Bridge on 1 March 1981 putting at risk vehicles on the bridge and overturning a caravan parked below the bridge on Beachley peninsula. The second was a waterspout in the Bristol Channel on 11 January 2004, between Barry and Hinkley Point.
1.4.3.4. Severe storms 107. It is also important to consider rainfall characteristics on a larger scale, such as those storm durations that influence flood events. A range of rainfall depths for storms of varying magnitude and variation are provided in Table 1.8.
108. These data were determined using the FEH CD-ROM Depth-Duration-Frequency (DDF) rainfall model. This is the recommended method for determining rainfall depth in the UK and has largely superseded the Flood Studies Report (FSR) method although the FSR is still widely used, not least because it is able to provide rainfall depths for event magnitudes with a return period of up to 1 in 10,000 years whereas the DDF model is not recommended for return periods of greater than 1 in 1,000 years.
Table 1.8 Varying design event rainfall depts (mm) for varying duration events
Hinkley Point C Drainage Ditch/Holford Stream upstream of source Storm duration Return period (years) (hours) 1 5 10 30 50 100 200 1000 0.5 5.0 14.6 18.2 25.7 30 37 45.7 74.3 1 6.6 18.1 22.4 30.8 35.7 43.5 52.9 83.5 2 8.8 22.5 27.4 36.9 42.3 50.9 61.2 93.6 6 13.7 31.5 37.6 49 55.3 65.2 76.7 111.9 12 18.1 39 45.8 58.5 65.4 76 88.3 125 24 25.3 51 59.1 73.9 81.8 93.9 107.7 147.9 48 35.4 66.7 76.2 93.3 102.3 115.9 131.2 174.8 120 47.7 82.3 92.4 109.9 119 132.4 147.4 188.7
1.4.3.5. Extreme high water temperature 109. The following high water temperatures have been observed in the waters around Hinkley Point since 1977:
23°C in August 1983; 23.8°C in August 1995; 23.5°C in July 2003; and 23.1°C in July 2005.
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110. Studies have been completed to identify extreme high water temperature in the vicinity of the Hinkley Point site in the UK. Warm extremes have been obtained by combining onsite observations with re-analysis and climate models more typical of open sea conditions. The extreme high water temperature taking into account a 10,000 year return level, a medium scenario climate change and an 84% confidence in models can be considered as around 28.5°C.
1.5. NATURAL RESOURCES AND FOODSTUFFS
1.5.1. Water utilisation 111. Seventeen Groundwater Abstraction Licenses were identified within 2km of the Hinkley Point C site all of which are identified as being for general farming and domestic purposes. No “Surface Abstraction Licenses” were identified in the area.
112. Within 5km from the Hinkley Point site, five households are using borehole water as their domestic supply, but no one isdrinking spring or well water. Livestock are identified drinking borehole water at seven farms and surface water at three farms; livestock were not found to be drinking well or spring water.
113. There is no abstraction of water from the region for utilisation by other Member States.
1.5.2. Principal food resources in the region
1.5.2.1. Livestock and crops 114. The number of farms of different types, the area of crop production (hectares) and numbers of livestock in Somerset are presented in Table 1.9, Table 1.10 and Table 1.11 below.
Table 1.9 Farm types
Farm type Number of holdings (2008) Cereals 473 General cropping 109 Horticulture 432 Specialist pigs 100 Specialist poultry 282 Dairy 685 Grazing livestock (less favoured areas) 296 Grazing livestock (lowland) 1,982 Mixed 347 Other 4,753
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Table 1.10 Crops
Arable crops Area (ha) Wheat 27,300 Winter barley 4,667 Spring barley 3,364 Oats 1,675 Other cereals 891 Total cereals 37,897 Potatoes 1,210 Sugar beet 75 Field beans 2,034 Peas for harvesting dry 186 Oilseed rape 4,369 Linseed 360 Root crops, brassicas, fodder beet 755 Maize 12,650 Other arable crops 1,079 Bare fallow 2,219 Horticultural crops Area (ha) Peas and beans 589 All other vegetables and salad 521 Total vegetables grown in the open 1,110 Crops grown under plastic 44 Total fruit 1,320
Table 1.11 Livestock
Livestock Number of livestock Total cattle* 312,544 Total pigs 97,034 Total sheep 489,416 Total goats 3,910 Poultry – total layers 548,894 Poultry – broilers (table chicken) 3,153,409 Turkeys 6,980 * Includes female dairy and beef cattle and male beef cattle
115. A terrestrial survey(8) was undertaken covering all land and watercourses within 5km of the Hinkley Point site.
116. Twenty five farms were identified in the survey area. Of these:
five produced beef cattle; seven produced beef cattle and dairy cattle; one produced beef cattle, dairy cattle and pigs; five produced beef cattle and sheep; two produced dairy cattle;
(8) CEFAS Radiological Habits Survey: Hinkley Point, 2006 - RL03/07 (www.cefas.co.uk)
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one produced sheep; one produced pigs; two produced chickens; one produced pheasants and sheep; and fourteen also produced arable crops.
117. The survey also showed that beef cattle and lambs were sold to Highbridge market. Beef cattle were sold to a wholesaler and butcher approximately 1km outside the survey area, to Taunton market and to an abattoir at Langport. Milk was sold to a creamery at Nether Stowey, most of which was used to make cheese and a small amount of cream. The cheese was sold nationally and direct from the creamery.
118. Two farms sold milk to the national chain Milk Link. Pigs were sold to an abattoir in Bristol. Chickens were sold to a processing plant in Devon. Some pheasants were sold to shoots within the survey area, but most of the pheasants were sold to shoots outside the survey area.
119. Fourteen farms produced arable crops, which included wheat, barley, beans, oil seed rape, maize, silage and linseed. The crops were kept for animal feed or sold to a grain merchant just outside the survey area. One farm also produced rhubarb, which was sold outside the survey area to farm shops and restaurants.
1.5.2.2. Fishing 120. An aquatic survey was undertaken over an area extending from Brean Down to Blue Anchor on the southern shore of the Bristol Channel, encompassing all inter-tidal areas and fishing areas up to 9km from the shore.
121. There were scarcely any commercial fishing activities in the survey area. Five full time commercial fishermen were identified as working in the survey area. Of these three were operating stake nets and set nets approximately 1km offshore on the Steart Flats at Stolford. These were used to catch cod in winter and bass, grey mullet and eel in summer. The other two commercial fishermen had stake nets on the beach at Blue Anchor for cod in winter and thornback ray and bass in summer.
122. Fishing for Elver is popular on the River Parrett. Approximately 140 Elver fishing licences are granted by the Environment Agency. Elvers were sold to two companies who exported them live to Holland and Scandinavia to restock rivers and fish farms.
123. Commercial fishing for crustaceans was only identified in Stolford where two fishermen set netting over mud for brown shrimps. The commercial collection of molluscs was not identified in the survey area.
124. Shore angling was a popular activity on many of the beaches in the survey area (observed at Brean, Burnham-on-Sea, Shurton Bars, Lilstock, Kilve, Doniford, Watchet Harbour and Blue Anchor). Two angling charter boats were operating from Watchet Harbour. The outcrops near Hinkley Point were reported to be popular with anglers mainly in the winter for cod, as was Donisford.
125. Sources were identified reporting quantities of fish exported from the UK to other European Union countries but only at a national level as there is no regional detail. The tables below are from the Marine and Fisheries Agency statistic collections and they provide a basic indication of the quantities of different landings exported to various countries.
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Table 1.12 Exports of fish from the UK by importing country 2008
Quantity of landings (tonnes) Country Other Demersal Cod Herring Mackerel Salmon Sardine and Pelagic Belgium 24 - 391 2,649 10 994 Denmark 7,767 156 4,874 402 65 1,725 France 1,053 1,009 3,679 23,045 122 23,092 Germany 3,409 3,995 3,185 3,198 1,053 9,371 Greece 1 90 178 208 - 275 Ireland 3,226 648 787 6,138 265 12,812 Italy 9 1 8 1,567 65 1,673 Luxembourg - - 67 161 - 26 Netherlands 69 18,232 12,268 688 7,231 22,719 Portugal 2,772 66 14 30 209 801 Spain 3,752 201 346 1,092 143 11,210 Sweden 1,335 - 1,279 34 391 754 Other EU 15 5 - 62 152 29 694 Other EU 109 4,895 9,237 996 275 3,278 EFTA(9) 7 6 76 36 160 410 Other countries 1,117 7,765 42,238 17,389 1,146 18,624 Total 9,166 12,666 78,689 57,785 9,584 108,458
(9) European Free Trade Association
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Table 1.13 Exports of fish and shellfish from the UK by importing country 2008
Quantity (tonnes) Country Cockles and Shrimps and Saithe Crabs Other shellfish Mussels Prawns Belgium 2 11 9 1,148 561 Denmark 1,500 - - 1,263 462 France 4,614 3,470 1,368 1,898 11,388 Germany 249 6 5 3,002 222 Greece - 9 - 509 23 Ireland 17 160 795 3,219 1,666 Italy - 516 750 2,192 11,050 Luxembourg - - - 18 20 Netherlands 1 43 10,362 1,161 1,972 Portugal - 1,586 2 59 102 Spain 27 7,089 311 930 13,648 Sweden - 7 - 35 17 Other EU 15 - 1 1 19 8 Other EU - 2 73 419 239 EFTA - 14 1 203 43 Other countries 15 310 16 315 7,341 Total 15 9,052 91 16,390 48,762
126. The Marine and Fisheries Agency also provides a summary of catches, but not exports, for the International Council for the Exploration of the Sea area of capture. In particular, the combined landings for areas VIIf Bristol Channel and VIIg South East of Ireland (as shown in Figure 1.12) are summarised below in Table 1.14.
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Figure 1.12 International Council for the Exploration of the Sea areas
9°0'0"W 8°0'0"W 7°0'0"W 6°0'0"W 5°0'0"W 4°0'0"W 3°0'0"W - 54°0'0"N
53°0'0"N
53°0'0"N VIIa
52°0'0"N
52°0'0"N
VIIg
H 51°0'0"N E
Hinkley 51°0'0"N Point C VIIf
50°0'0"N
50°0'0"N VIIe VIIh 8°0'0"W 7°0'0"W 6°0'0"W 5°0'0"W 4°0'0"W 3°0'0"W
Table 1.14 Total landings for areas VIIf and VIIg
Demersal Pelagic Shellfish Total Landings (tonnes) 3,100 2,300 7,200 12,600
1.5.2.3. Hunting 127. Two wildfowling clubs were identified that were shooting in the area between Brean Down and the River Parrett. These clubs are mainly shooting under permit from English Nature on the salt marsh along the west side of the River Parrett, east of Steart village.
128. The main species being shot were Mallard, Teal and Wigeon as well as Pintail, Pochard and Gadwall.
129. Game from within the survey area is locally consumed. This includes pheasant, pigeon, rabbit and venison. One private pheasant shoot was identified on farmland in the survey area.
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1.5.2.4. Other activities 130. The survey also identified two people who are consuming Porphyra Umbillicalis, one of which is also consuming sea lettuce both species were collected from Stolford. Two other people are using seaweed from Stolford as fertiliser on their vegetables.
131. Two farmers in the survey area are grazing livestock on salt marsh. One of these grazes beef cattle and sheep on salt marsh between Stolford and Steart village. This farmer’s livestock is sold locally to Taunton market. The other farmer is grazing sheep on salt marsh at Steart village between May and September. This farmer’s lamb is sold nationally through a meat company and nine people from the survey area were identified as consuming salt marsh grazed lamb from this farm.
132. Three beekeepers were identified in the survey area with a combined total of 42 hives near the villages of Knighton, Coultings and Stringston. The average production of honey per hive per year is 27kg. The beekeepers and their families consume honey from their hives. Excess is given to friends or sold from their door and some is also sold to farm shops outside the survey area.
133. The consumption of wild foods includes blackberries, sloes, mulberries, damsons, bullus plums, almonds, hazelnuts and mushrooms.
1.6. OTHER ACTIVITIES IN THE VICINITY OF THE SITE
1.6.1. Air traffic 134. A few airfields are present in the vicinity of the site, including Watchford Farm, Franklyns Field, Dunkeswell (light aircraft, microlight and parachuting), Halesland (gliding activities), Culmhead and Upottery. The latter two are disused but may have some activity.
135. Royal Navy Air Station (RNAS) Yeovilton, known as Her Majesty’s Ship Heron, is located near Yeovil, Somerset. The station consists of around 1,400 acres of airfield sites at Yeovilton (approximately 40km from Hinkley Point C) and the satellite at Ilton (Merryfield, approximately 30km from Hinkley Point C). RNAS Yeovilton is a large multi-role air station and one of the busiest military airfields in the UK. It is home to Royal Navy Lynx Helicopters and the Royal Navy Commando Helicopter Force. RNAS Yeovilton operates over 100 aircraft in four different categories and is manned by around 4,300 personnel, service and civilian, including Ministry of Defence employees and permanent contractors.
136. The Air Navigation (Restriction of Flying) (Nuclear Installations) Regulations 2007 states that ‘no aircraft is to fly over a nuclear installation to which the regulator applies below the height above mean sea level specified’. The specification for the Hinkley Point no fly zone, is a 2 nautical mile radius from the centre of site and below 2,000 feet.
1.6.2. Rail traffic 137. The nearest rail line to site is around 10km distance and the closest rail siding is Bridgwater at approximately 18km.
1.6.3. Road traffic 138. The closest main road is the A39, running approximately 5.5km south-southwest of the site. The A39 connects Bridgwater to Minehead.
1.6.4. Maritime activity 139. A study of local maritime activity and the risk of collision with the undersea east and west cooling water intake structures, has been undertaken for the waters near Hinkley Point. Vessels considered in the study included recreational craft, merchant shipping, fishing and
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military vessels. No military vessels were identified in the area. The level of shipping in the area is considered to be low.
140. In the two month study period:
a moderate level of recreational craft were identified, however the risks associated with these craft were considered low due to the shallow draft of the vessels and the low impact energies associated with them; three merchant vessels were tracked within the study area a total of six times; no fishing vessels were sighted within ten nautical miles of the location of the intake structures; and no military vessels were identified in the area.
141. The calculated frequency of impact with a single intake, was from 3.3x10-7 for the west intake tunnel and 2.7x10-6 for the east intake tunnel with corresponding collision return periods of 3,900,000 and 340,000 years respectively. The risk of multiple structures being impacted simultaneously was found to be between 10-12 and 10-6.
1.6.5. Industrial activities 142. Hinkley Point C is located in a rural area with very limited industrial developments.
143. A search of dangerous or hazardous sites was carried out around Hinkley Point C and the only two Control of Major Accident Hazards (COMAH) & Notification of Installations Handling Hazardous Substances sites, are the existing sites of Hinkley Point A and B.
144. Outside of the remit of the above regulations, the following industrial sites using hazardous chemical were identified within 10 miles from the Hinkley Point C site:
A small industrial biodigester facility at Cannington, approximately 4 miles from the site. A small limestone quarry which mainly supplies lime powder and granules for the animal feed and agriculture sector. This is located in Cannington, approximately 4 miles from the site. A paper mill at Watchet, which is over 6 miles from the site. The Ambersil polymer factory at Bridgwater, which manufacture polymer processing products. The British Cellophane factory at Bridgwater, used hazardous chemicals but it is closed and is being demolished. The Royal Ordnance Factory at Bridgwater, was built in the early 1940s between the villages of Puriton and Woolavington to manufacture explosives. The factory, located over 6 miles from Hinkley Point C, was closed in early 2008.
145. A gas main runs from Bridgwater through to Nether Stowey. This is located approximately 4 miles from the closest point on the Hinkley Point C site.
146. No petrol or fuel is sited and underground high pressure oil pipelines were identified within 10 miles from the site.
147. No mining or major industrial developments other than Hinkley Point A and B were identified within 10 miles from the site.
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2. THE INSTALLATION
2.1. MAIN FEATURES OF THE INSTALLATION
2.1.1. Nature and purpose of the installation 148. The Areva/EDF design EPR, is a Pressurised Water Reactor (PWR), the design of which incorporates technology from French N4 and German KONVOI reactors. Two EPR units are to be constructed on the Hinkley Point C site of approximately 1650MWe each, cooled through an open system which works broadly in the same way as other PWR’s already in operation. As the chosen model, the third generation EPR offers improvements to the first and second generation reactors currently in operation in Europe. Examples of these improvements are presented in Section 2.1.5.
149. The principle on which a nuclear power plant operates is very similar to that of a conventional power station: the boiler that burns fossil fuel is replaced by the nuclear reactor, where the heat produced comes from the fission of fuel nuclei.
150. The heat produced converts water into steam which is then released in the turbine, driving the AC generator and producing electricity. Between the heat source (the nuclear fuel) and the heat sink (the sea), a PWR has three physically separate systems, as shown in Figure 2.1:
The primary system extracts the heat produced by the fuel in the reactor. The secondary system uses this heat to convert water into steam for the turbine. The cooling system condenses the steam released into the turbine.
2.1.1.1. The primary system 151. The primary system is located within the sealed containment in the reactor building, and extracts the heat produced by the fuel in the core.
152. The reactor core is responsible for the production of heat. The reactor core typically consists of fuel assemblies, with each fuel assembly formed by an array of Zircaloy M5 tubes made up of fuel rods. The fuel rods consist of enriched uranium dioxide (UO2) pellets stacked into the Zircaloy M5 cladding. The reactor also has equipment to control the nuclear reaction and to shut it down automatically.
153. The containment of the reactor consists of a cylindrical internal pre-stressed concrete wall and an exterior wall constructed from reinforced concrete, separated by an inner space called the “inter-containment annulus”. The internal surface of the interior containment is covered by a metallic leak tight skin.
154. The heat produced is transferred by means of a coolant of pressurised water that passes in a closed circuit around the fuel. The heat is then transferred to the secondary system via a heat exchanger; the steam generator.
2.1.1.2. The secondary system 155. The secondary system is the water-steam system. The steam produced by the steam generator is sent into the turbine which drives the AC generator that produces the electric current. The electrical energy is then transferred via electrical power lines.
156. When it leaves the turbine, the spent steam is condensed in a second type of heat exchanger; the condenser. It is converted into water and then returned to the steam generator to start a new cycle. The electromechanical equipment is installed in the turbine hall.
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2.1.1.3. The cooling system 157. The condenser is itself continuously cooled by water circulating in a third system; the plant cooling system. This comprises:
the seawater intake tunnel; the pumping station that filters and pumps the sea water and sends it to the condenser; and an outfall pond which releases cooling water back into the sea by means of an overflow and an underwater tunnel, ending in a diffuser which is anchored to the seabed.
Figure 2.1 Schematic layout of the EPR power generation process
2.1.2. General design and installation plan for the EPR 158. The EPR unit is a PWR with a thermal power rating of 4,500MW and with a design life of 60 years. The plant can operate at base load, i.e. at 100% capacity and can also perform load follow operations, monitoring variations in the grid's demand for electrical energy by working at between 20% and 100% of nominal power.
159. Hinkley Point C will consist of two EPR reactor units, each provided with ancillary systems and common buildings and facilities.
160. The description below is related to an individual reactor unit, the two units being essentially identical. Shared facilities are described separately.
161. One EPR unit comprises the following structures, as shown in the site layout plan provided in Figure 2.2. A more detailed description of these structures is provided in Section 0.
Reactor building, used principally to house the nuclear steam supply system (NSSS). Safeguard buildings (auxiliary safeguard electrical and mechanical). These are divided into four divisions, each containing a series of emergency systems along with their associated electrical systems. Fuel building. Nuclear auxiliary building (NAB). Diesel generator buildings. Turbine hall containing the turbine generator, the condenser and the feedwater plant. Conventional island electrical building houses the electrical systems for the conventional
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(non-nuclear) parts of the plant. Transformer platforms containing the main, unit and auxiliary transformers and switchgear. Pumping station including the forebay. Cooling water discharge plant including the filtering debris recovery pit and the outfall pond. Fire fighting water storage building. Access building.
162. The following facilities are shared by the two reactor units (see Figure 2.2). A description of the main shared structures is provided in Section 2.1.2.2.
Effluent treatment building (ETB). Hot laundry. Hot workshop. Hot warehouse. Decontamination facility. Containment tank building. Discharge tanks building. Outfall tunnel. Spent fuel interim storage facility. ILW interim storage facility. Tunnels (for personnel use). Oil and grease store. Chemicals store. Gas stores. Contaminated tools store. Non-radioactive waste store. Transit area for Very Low Level Waste (VLLW). Operational service centre. Demineralisation station. Permanent sewage wastewater treatment plant. Services buildings (garage, emergency services, security).
163. The following facility is non-shared and specific to Unit 2 (see Figure 2.2). A description of this structure is provided in Section 2.1.2.3.
Radioactive waste treatment building of Unit 2.
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Figure 2.2 Hinkley Point C site layout plan
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2.1.2.1. Description of structures for each unit 164. The structures described below are provided for each of the two units.
Reactor building 165. The reactor building houses the NSSS and some of its associated operational and safety equipment. The building is the third barrier protecting the public against the effects of an accident (the first is the cladding around the fuel and the second is the primary circuit structures).
166. In normal operation, this building protects the reactor cooling system against external events and screens radiation from the reactor core. In abnormal or accident conditions, the reactor building provides containment and limits the radiological consequences at the site boundary.
167. The reactor building, the four safeguard buildings and the fuel building share the same foundation raft. This minimises the relative movement of the buildings in the event of an earthquake. They are also designed to withstand the effects of shock waves (from an external explosion).
168. The reactor building consists of an external framework, known as the containment, and internal structures. The containment is designed to withstand internal accidents involving the steam supply system, as well as external damage from natural and manmade causes.
169. The containment consists of:
an inner chamber of pre-stressed concrete, which is pressure-resistant in the case of an accident, has a metal liner and is adequately leak proof when under pressure; an outer reinforced concrete shell; and a space between the containments, maintained below atmospheric pressure, so that leakage from the internal chamber can be contained. After treatment using high efficiency particulate air (HEPA) filtration and iodine traps, leaks are discharged into the atmosphere via the NAB stacks.
170. The reactor building houses the primary system, which serves as an envelope to maintain the reactor coolant at its operating pressure and temperature, and reduce leaks and radioactive discharge into the enclosures atmosphere. This system fulfils three main functions:
transfers heat from the reactor core to the steam generators; controls reactivity by adjusting the boron concentration in line with the control rods; and controls the pressure via the pressuriser.
171. The elements that make up the primary system are the reactor vessel, the pressuriser, four heat transfer loops, each with a steam generator and a primary coolant pump and the connecting pipework. The concrete shield wall surrounding the primary system protects the containment from accidentally generated projectiles and reduces radiation levels in the areas surrounding the primary loops.
172. The primary coolant is borated, demineralised water which serves at the same time as a moderator, a neutron reflector and a diluter for the boric acid used to control the reactivity.
173. The liquid waste produced during normal operation of the primary circuit, is directed to the coolant storage and treatment system (CSTS), and the gaseous discharge is sent to the gaseous waste processing system (GWPS), through either the chemical and volume control system (CVCS) or the nuclear vent and drain system (NVDS).
174. Leaks during operation and discharge from venting and draining during maintenance or repair, are collected by the NVDS.
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175. The reactor building also includes the primary coolant reserve, which is shown in blue in Figure 2.3. This is installed in the reactor building to:
avoid having to switch from an injection phase to a recycling phase after draining a tank, when primary coolant has been lost, or after a breach (the in-containment refuelling water storage tank collects the water that may be discharged inside the containment during an incident or accident and remains full); and have a water reserve available to cool the corium (a lava like molten mixture of portions of nuclear reactor core, formed during a nuclear meltdown) in the unlikely case of a core meltdown.
176. The installation’s structural arrangements to mitigate a core meltdown accident, consist of the area provided for the molten core material to spread and the channel leading from the reactor pit to this area. The spreading area extends over approximately 170m2 and is situated to the side of the reactor pit. It is surrounded on two sides by the in-containment refuelling water storage tank.
Figure 2.3 View of the lower section of the reactor building
Safeguard buildings 177. These buildings house the safety equipment including the pumps and valves for the safety injection system, the pumps and heat exchangers for the cooling system and the corresponding ventilation system.
178. All systems classed as safeguards are designed with quadruple redundancy and are sited in physically separate divisions.
179. Each division has:
a low-pressure safety injection system; a medium pressure safety injection system, which is also used for cooling the reactor on shutdown; a component cooling water system; and an emergency feedwater system.
180. In addition, divisions one and four, house a residual heat removal system.
181. The buildings also contain the electrical safety systems, the instrumentation and control, the control room and ventilation systems.
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Fuel building 182. Each fuel building houses:
equipment relating to fuel, including the fuel building fuel pool, the fuel transfer area, the fresh fuel area and the loading area for the irradiated fuel containers; equipment associated with the reactor cavity and fuel pool cooling (and purification) system (FPC(P)S); equipment for the reactor's CVCS and storage facilities for the reactor boron and water make-up system and the extra boration system; and some ventilation equipment for the fuel building and for the space between the containments.
183. The fuel building is designed to enable new or irradiated fuel elements to be handled and stored in a controlled atmosphere. It houses the equipment necessary for these operations and enables their use.
Nuclear auxiliary building 184. The NAB is built on an independent foundation raft next to the fuel building. The NAB for each reactor, houses some of the operational systems and dedicated maintenance areas.
185. The main systems installed in the NAB are:
CSTS; FPC(P)S; the GWPS; part of the steam generator blowdown system (SGBS); the NAB ventilation system; the operational chilled water system; the reactor boron water make-up system; and the sampling equipments and laboratories.
186. All air extracted from the ventilation systems of radiologically controlled areas in the nuclear island, is channelled, collected and checked in the NAB before it is discharged via the stack.
Diesel generator buildings 187. Each of the two diesel generator buildings (four across the two units) houses two main diesel generator sets and one station black out diesel generator set for plant cooling. The diesel generator buildings are constructed from reinforced concrete and are built on an independent foundation raft.
188. The two diesel generator buildings are geographically separated. Their positions are determined by maintenance imperatives and to enable the easy movement of diesel motors in and out of the building.
189. Each diesel generator building houses two main diesel generator sets, each of which supplies a safety train within a division of the safeguard building, as well as a station black out unit. The two generators and the station blackout generator with their auxiliaries are protected against internal hazards by a separating wall.
190. Both diesel generator buildings are designed to withstand external hazards and contain internal hazards.
Turbine hall 191. The turbine hall contains:
the turbine generator set and associated auxiliary systems; the condenser and heating station that supplies water to the steam generators; and a device for bypassing steam from the turbine to the condenser, to enable house load
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operation in the case of an electrical system fault.
Conventional island electrical building 192. The conventional island electrical building adjoins the turbine hall near the transformer platforms. The building houses the permanent and secured electrical distribution panels, which supply the conventional island's auxiliary equipment and the automated controllers that manage and monitor it.
193. The electrical building also supplies electricity continuously at 10kV to each of the four electrical buildings in the nuclear island.
Transformer platforms 194. The platforms for each unit adjoin the turbine hall and are located close to the conventional island electrical building. Each platform houses the main transformer, the two unit transformers and the line and coupling breakers. A dedicated platform houses the auxiliary transformer, which is always live.
Pumping station 195. The pumping station supplies raw water to the power plant primarily for cooling the condensers via the forebay which is a semi-circular basin located adjacent to the pumping station. Seawater is transferred to the forebay via the intake tunnel.
196. The pumping station has four separate drainage channels:
Two central channels which mainly supply the essential service water system and the circulating water condenser cooling system. Two side channels, each fitted with a chain-filter which mainly supply the essential service water systems, the auxiliary (raw water) cooling system and the ultimate cooling water system.
197. The supply of seawater to the essential service water systems and ultimate cooling water system pumps (used respectively for the component cooling water system and the residual heat removal system) is designed so that if a drum or chain-filter stops working, the seawater is still supplied.
Cooling water discharge plant 198. The cooling water discharge plant for each unit consists of a filtering debris recovery pit (pre- discharge section) for collecting marine debris from the pumping station and an outfall pond (discharge pond), which is connected to the outfall galleries. The outfall galleries of both outfall ponds (Unit 1 and Unit 2) meet at a connecting structure that connects to the common outfall tunnel shared by both units.
Fire fighting water storage building 199. The main function of the fire fighting water storage building is to store and supply large volumes of water to be used in the event of a fire or other safety issues and in case of loss of the heat sink for the emergency feedwater system pumps.
Access building 200. The access buildings main function for each unit is to control access to the nuclear island. The building contains a room for maintaining and decontaminating small pieces of equipment, operations rooms and technical rooms.
2.1.2.2. Description of shared structures 201. The following buildings and facilities are shared by both EPR units.
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Effluent treatment building 202. The ETB processes and stores liquid and solid wastes prior to disposal or onward transfer. It is designed to treat waste from the two EPR units.
203. The building is divided into two parts, one section is used for storing solid waste and the other treats liquid and solid waste.
204. The waste storage section comprises:
a room for storing drums and containers of solid waste that is accessible to road convoys; areas for checking drums before disposal; an area for storing the resins used to treat the steam generator drains; a reception room for the mobile resin treatment stations; and a handling crane and baler used to package the low level waste (LLW).
205. The part of the building used for the treatment of liquid and solid waste, consists of a heavy- duty area that mainly houses the liquid waste processing system (LWPS) (head storage, processing) and the solid waste treatment system (storage of resins and concentrates, filter encapsulation cell). It also houses a plant for producing concrete and storing aggregates. In addition, it contains the ETB control room and the electrical rooms.
206. The ETB at Hinkley Point C is shared between the two units and is adjoined to Unit 1. Therefore, the solid radioactive wastes (LLW and ILW) generated in Unit 2 will be pre- conditioned in the radioactive waste treatment building of Unit 2 before transfer by road to the shared ETB.
Hot laundry 207. The main function of the hot laundry building is to launder radioactive contaminated work clothing such as overalls and overshoes for reuse. Non-contaminated work clothing is not laundered in the hot laundry. Common practice in the UK is for radioactive contaminated work clothing to be laundered off-site. This facility provides Hinkley Point C with the ability to be self sufficient should it choose. The hot laundry building is adjacent to the ETB. The hot laundry will discharge effluents to the liquid radwaste monitoring and discharge system (LRMDS).
Hot workshop, hot warehouse and decontamination facility 208. The hot workshop, hot warehouse and decontamination facility are encompassed in a single structure adjacent to the hot laundry.
209. The hot workshop is designed to perform machining of radioactive contaminated components or tools. The hot workshop may generate liquid effluents, which will be stored underneath the hot laundry and sent to the LWPS for treatment prior to discharge off-site.
210. The hot warehouse is designed to store radioactive contaminated tools.
211. The decontamination facility is designed to reduce or suppress radioactive contamination of tools, components or wastes. Decontamination of equipment enables reuse of tools and minimisation of the volume of material requiring disposal.
Discharge tanks building 212. The discharge tanks of the LRMDS, additional liquid waste discharge system (ExLWDS) and site liquid waste discharge system (SiteLWDS) are located in the discharge tanks building which adjoins the ETB. Their pumps are located under and adjacent to the hot laundry.
213. The site discharge tanks store effluent prior to discharge and comprise the following:
LRMDS (system for collection, monitoring and discharge of effluent from the nuclear island).
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ExLWDS (system for additional storage capacity, if required, and for return of effluent from the LRMDS and SiteLWDS back to the LWPS if retreatment is required) SiteLWDS (system for monitoring and discharge of effluent from the secondary circuit. The conventional island liquid waste discharge system (CILWDS) collects and treats effluent from the turbine hall before sending to the SiteLWDS).
214. The contents of the LRMDS, ExLWDS and SiteLWDS tanks may be transferred to the LWPS for retreatment via the ExLWDS if required.
215. Liquid effluents from Unit 2 are routed via underground galleries to the ETB or directly to the relevant storage tank.
Containment tank building 216. The containment tank building which collects and processes the site’s waste water and rain water from the roads and roofs, contains a system for the collection of oils and hydrocarbon effluents, sedimentation (settling) tank and an oil filter (separator). The containment tank building also recovers polluted water after a fire or an accidental spillage on a road. The water processed in the containment tank building is transferred to the forebay.
Outfall tunnel 217. A common outfall tunnel is shared between the two EPR units at Hinkley Point C and is the main discharge outlet for radioactive liquid waste. The location of the two outfall heads are approximately 2km off-shore. The internal diameter of the outfall tunnel is approximately 7m.
Spent fuel interim storage facility 218. The spent fuel interim storage facility provides safe and secure underwater storage of spent fuel once it leaves the spent fuel pool of the fuel building from both EPR units, after a period of cooling. The spent fuel will remain here until disposal at the national Geological Disposal Facility (GDF). The spent fuel interim storage facility consists of a pool, or pools, housed in a seismically qualified building. Spent fuel assemblies are placed in storage racks, located at the bottom of the pool(s), which are designed to be resistant to movement. Whilst in storage, water cooling and clean-up systems remove the heat generated by the spent fuel assemblies and maintain water quality. Throughout the operational life of this facility, an inspection and monitoring regime will be implemented to ensure that fuel is safely stored. The pools are lined with welded stainless steel plates and have leak detection and collection systems to ensure that there is no unplanned release of radioactivity to the environment.
Intermediate level waste interim storage facility 219. ILW generated during the operational phase will be placed in the ILW interim storage facility which will be designed to be in operation for up to 100 years. ILW generated during the 60 years of EPR operation, primarily from the solid waste treatment system in the ETB, will be conditioned and packaged in the ETB before transfer to the ILW interim storage facility. The ILW interim storage facility will provide interim storage for all ILW pending removal to a final national GDF.
Tunnels 220. The buildings on the site are linked by various tunnels. The tunnels, designed for use by personnel, are the tunnels between the operation centre and the access tower and the service tunnels used for ducts and/or cableways.
2.1.2.3. Description of other structures
Radioactive waste treatment building of Unit 2 221. The radioactive waste treatment building of Unit 2 is a dedicated building for pre-conditioning and preparing solid radioactive wastes (LLW and ILW) generated in Unit 2, for transport to the shared ETB for treatment and conditioning.
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222. The two main functions of the radioactive waste building of Unit 2 are:
pre-conditioning of filters (used on CVCS, FPC(P)S, CSTS of Unit 2) in concrete drums (type C1 or C4), closed with a temporary biological plug; and preparation and transfer to the shared ETB adjoined to Unit 1, of the waste packaged in concrete drums (filters) or metallic boxes (low activity ion exchange resins).
2.1.3. Main operational and safety procedures
2.1.3.1. Description of the nuclear steam supply system 223. The NSSS includes the following elements (see Figure 2.4):
The core which contains 241 fuel assemblies with each fuel assembly formed by a 17x17 array of Zircaloy M5 tubes made up of 265 fuel rods and 24 guide thimbles. The fuel rods consist of Uranium dioxide (UO2) pellets stacked into a Zircaloy M5 cladding tube which is then plugged and seal welded. Four cooling circuits containing water at a pressure of 155 bar absolute. Each circuit comprises a primary pump, a steam generator and the connecting pipework. The steam from the secondary side of the steam generators is saturated and at a pressure of around 80 bar absolute. A pressuriser which maintains the water in the primary system at a constant pressure. A protection system comprising the instrumentation that controls the various operating parameters in the NSSS. This automatically activates the safety devices which prevent normal operating limits from being exceeded. 89 control clusters, each comprising 24 control rods. A control rod is made up of two parts: the lower part is made of a silver/cadmium/indium alloy and the upper part is made of carbon and boron (B4C). The control clusters, together with diluted boron in the water of the main cooling system, are used to control the core reactivity. Various auxiliary systems required either for correct operation (normal operating systems), or for the safety of the reactor (safeguard systems).
Figure 2.4 Nuclear steam supply system
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2.1.3.2. Safety injection system and residual heat removal system 224. The mechanical safety injection system and residual heat removal system, combine safety injection functions with reactor cooling on shutdown (see Figure 2.5). Configured to operate as a safety injection system, it provides water to cool the reactor core and maintain it below sub-critical conditions in the event of the accidental loss of primary coolant, for example:
if a pipe is breached, leading to a discharge of coolant that cannot be replaced by the normal make-up system; if the mechanism driving a control rod assembly fails, causing the accidental ejection of a rod; or if a steam generator tube fractures.
225. When configured to cool the reactor core on shutdown, it removes heat from the primary system in a controlled way, most of which is the residual heat from the core.
226. The safety injection system and residual heat removal system comprises four separate and independent trains. Each train injects into the primary system using an accumulator, a medium-pressure safety injection pump and a low-pressure safety injection pump, with a heat exchanger at the pump outlet.
227. The accumulators inject into the cold leg of the primary system. All the pumps discharge into the in-containment refuelling water storage tank in the reactor building and also inject coolant into the cold legs of the circuits in the primary system. The low-pressure trains may be operated manually to inject into the hot leg and the cold leg simultaneously.
228. In normal operation, low-pressure safety injection also acts as the reactors residual heat removal system, via the low-pressure exchangers, by pumping out the primary coolant in the hot leg. Electricity is supplied to each train by an independent train, which is backed up by a dedicated main diesel generator.
Figure 2.5 Safety injection system and residual heat removal system
IRWST
SG: Steam Generator IRWST: In-containment refuelling water storage tank
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2.1.3.3. In-containment refuelling water storage tank 229. The in-containment refuelling water storage tank contains a large quantity of borated water (see Figure 2.6). It collects water which can be discharged inside the containment in the case of an accident. It also acts as a water reserve for the safety injection system, the containment heat removal system (removal of power from the containment) and if necessary the CVCS and floods the corium spreading area in the event of a severe accident.
230. Filters and anti-clogging devices protect the safety injection system and containment heat removal system pumps against the migration of debris in accident conditions.
Figure 2.6 EPR safeguard systems
MSSS MSRT
EBS
SIS
EFWS
RHRS
MSSS: Main steam supply system EFWS: Emergency feedwater system MSRT: Main steam relief train RHRS: Residual heat removal system EBS: Extra boration system SIS: Safety injection system IRWST: In-containment refuelling water storage tank SG: Steam Generator
2.1.3.4. Extra boration system 231. The extra boration system comprises two distinct and independent trains, each designed to inject boron at high-pressure using a boron tank, a positive displacement pump and two injection lines, into two primary loops for each train (see Figure 2.6).
232. The extra boration system pumps and tanks are located in the fuel building. Electricity is supplied to each train by an independent train, which is backed up by a dedicated main diesel generator.
2.1.3.5. Emergency feedwater system 233. This is an emergency system that supplies water to the steam generators should the normal water supply system fail.
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234. The emergency feedwater system (see Figure 2.6) comprises four separate and independent trains, each supplying the secondary side of a steam generator with demineralised water from the emergency feedwater system tank. Header systems for pump suction and discharge, enable the use of the entire tank volume and allow the injected flow to be redirected if necessary, during pump maintenance, or if the secondary pipework is fractured.
235. Electricity is supplied to each train by an independent train, which is backed up by a dedicated main diesel generator. Furthermore, in order to withstand a generalised voltage loss (loss of the external power supplies and non-availability of the four main diesel generators), the power supply to trains in divisions one and four, is backed up by two station black out generators which start-up manually and provide an alternative to the four main diesel generators.
2.1.3.6. Component cooling water system 236. The component cooling water system is designed to remove heat from the components and systems installed in the nuclear island, particularly those that are important for safety.
237. The component cooling water system comprises four closed circuit trains, each separate and independent and acting as a barrier. Each train has a pump and a heat exchanger, cooled by the essential service water system.
238. Each train independently cools a safety injection system train, particularly when it is operating in residual heat removal system mode. The headers between trains one and two, and between trains three and four, cool the other systems and since either of the associated component cooling water system trains can cool these systems, the arrangement is reliable and flexible.
239. Isolating valves are fitted between the headers to keep the component cooling water system trains mutually independent. The valves are designed so that when the headers are normally open to one of the two connected trains, the other is shut off.
2.1.3.7. Essential service water system 240. The purpose of the essential service water system is to cool the component cooling water system. It comprises four separate and independent trains, each providing a cooling function between the component cooling water system heat exchangers and the natural environment. The essential service water system pumps are installed in the pumping station.
2.1.3.8. Chemical and volume control system 241. The CVCS is responsible for the following:
Controlling reactivity by regulating the boron concentration in the primary coolant during controlled boration, or dilution procedures performed by the reactor boron and water make-up system. The CVCS pumps can also discharge directly into the in-containment refuelling water storage tank and inject borated water into the primary system in some circumstances. Regulating the volume of primary coolant by loading and discharging it. Fluid discharged from the primary system is cooled first by an exchanger-regenerator that reheats the loaded fluid, then by non-regenerating exchangers (cooled by the component cooling water system). Purifying the primary coolant, by removing the fission and activation products in the form of ions or particles, using demineralisers or filters. Adding the chemicals lithium hydroxide to control the pH and hydrazine to remove oxygen when the reactor is started from cold. Injecting water into the no.1 seals on the primary pumps to ensure they are leak resistant. Transferring the primary coolant to the CSTS.
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Auxiliary pressuriser spraying in addition to the normal spraying.
242. The CVCS tank and charging pumps are located in the fuel building. Electricity is supplied by independent trains. Most of the CVCS actuators are backed up by the main diesel generators.
2.1.3.9. System for collecting, processing and discharging radioactive waste 243. Effluent that is radioactive or likely to be contaminated is collected separately, depending on its state (gas, liquid or solid) and quality (reusable or spent) by different systems and is sent to the liquid, gas or solid waste processing plants. Liquids are stored if necessary, before being reused within the power plant or discharged. Gases are discharged after passing through retardant filters to allow radioactive decay. Solid waste is treated before disposal off- site.
244. Reusable effluent from the primary water discharged from the CSTS and CVCS during plant operation, is collected without being polluted by oxygen, sent to the CSTS tanks and recycled in make-up water and boric acid for the primary system. This category also includes most of the controlled leaks and bleeds from system equipment carrying the primary coolant. These are carefully collected by NVDS and sent to the CSTS.
245. Spent effluent from the nuclear installations that cannot be recycled is collected by the NVDS and sent to the LWPS, situated in the ETB. To facilitate processing and prevent contamination from spreading, the effluent is sorted according to the degree of chemical and radioactive pollution. In order to do this, spent effluent is divided into three categories:
process drains (primary quality water that cannot be recycled); Floor drainage (system leaks, water from the floors of nuclear buildings and from showers); and chemical effluent (chemically polluted water containing primary coolant).
246. After processing, the spent liquid effluent is sent to the appropriate tank of the LRMDS, ExLWDS or the SiteLWDS.
2.1.3.10. Fuel pool cooling (and purification) system 247. The FPC(P)S is divided into two sub-systems; the fuel pool cooling system and the pool water purification system.
Fuel pool cooling system 248. This system has two main trains that are separate and independent. Each cools the fuel pool by means of two pumps and a heat exchanger. The system also has a third emergency train, for use in the event of a total loss of the two primary trains, which has a pump and a heat exchanger.
249. The trains empty the liquid from the fuel pool and then re-inject it. The main pumps and exchangers are located in the fuel building. Each main train is linked to one of the headers in the component cooling water system which cools it and each train may therefore be cooled by either of the two component cooling water system trains associated with the header.
250. The third train is cooled by the containment cooling ventilation system, then by the dedicated emergency cooling system backed up by the essential service water system.
251. Electrical power is supplied by independent trains. Each pump can be supplied by two electrical power trains. The electrical power supply is backed up by the main diesel generators. In the event of a generalised voltage loss, the third train is also backed up by the station black out generators.
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Pool water purification system 252. The pool water purification system comprises a purification circuit for the spent fuel pool, a purification circuit for the reactor pool and in-containment refuelling water storage tank and skimming circuits for the spent fuel pit and reactor pool.
253. The system also includes two cartridge filters, a demineraliser and a fine resin filter. An additional cartridge filter is installed in the spent fuel pool skimming circuit.
254. The pool water purification system transfers water between the in-containment refuelling water storage tank and the reactor pool, which is necessary during refuelling shutdown.
2.1.3.11. Steam and energy conversion systems
Turbine generator unit 255. The purpose of the turbine generator unit is to receive steam from the steam generators and convert thermal energy into electrical energy. It comprises a turbine and a three-phase AC generator.
256. The turbine converts the energy from the steam into mechanical energy. It consists of one high-pressure cylinder, one medium-pressure cylinder, three dual-flow low-pressure cylinders and two vertical moisture separators/reheaters which are used for drying when the steam expands.
257. Residual energy is transferred to the heat sink via the condenser. The turbines nominal rotation speed is 1,500rpm.
258. The AC generator is cooled by hydrogen and pressurised water. The turbine generator unit is not a safety system, but it is designed to trip automatically if certain reactor protection systems are stressed.
Main steam supply system 259. The main steam supply system:
provides the main steam to the turbine and the other devices that use the main steam supply in the turbine hall during normal operation; removes the residual heat by transferring steam to the condenser or into the atmosphere in abnormal operating conditions; protects the steam generators against excess pressure, using an atmospheric steam dump system and safety pressure release valves; cools the primary system until it reaches the injection pressure for the medium-pressure safety injection system in the event of a small breach in the primary system or a break in the steam generator pipes; isolates the steam circuit if there is an excessive increase in the flow of steam; and contains the activity in the event of an small breach in the primary system or a break in the steam generator pipes by isolating the steam side.
260. The main steam supply system has four identical trains (one for each steam generator). Each train comprises:
a main steam isolation valve; an atmospheric steam dump train, comprising a pressure relief valve and an isolating valve (main steam relief train system); two safety valves; pipework from the gas system flow limiter to the main steam valve cubicle outlets; valves and pipework for the treatment lines; and valves for the operating condensate drainage system.
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Main condenser 261. The main condenser receives the steam discharged from the three low-pressure turbine cylinders and the turbine bypass, and cools it using the plant's circulation water system. It comprises six modules (two for each low-pressure cylinder).
Turbine bypass system 262. The role of the turbine bypass, with regard to the condenser, is to compensate for the difference in power between the turbine and the NSSS. During normal operation, the power provided by the NSSS is equal to the power consumed by the turbine. During rapid transients, or when the power fluctuates at low loads, the turbine bypass system opens to compensate for the power imbalance between the turbine and the reactor.
2.1.3.12. Feedwater plant 263. The feedwater unit comprises:
a set of pumps to extract water from the condenser (three pumps 50%/condenser extraction system); a set of low and high-pressure reheaters; a feedwater tank, which is also used for degassing and reheating the water from the low- pressure reheaters; four motorised feed pumps, each designed to provide 33% of the required nominal flow (motor driven feedwater pump system) and a start-up and shutdown pump; and a main feedwater system valve unit for very-low-flow, low-flow and high-flow, which feeds water to the steam generators.
2.1.3.13. Pumping station 264. The pumping station has four channels for supplying water from the forebay:
two central channels, each equipped with four pre-filtration waterways (narrow passages) and a filter drum, for filtering both the essential service water system and the circulation water system; and two side channels, each equipped with a pre-filtration waterway and a chain-filter, for filtering both the essential service water system and the auxiliary cooling system.
265. Each waterway has a fixed grid with a trash rake. Sluice gates can be used to isolate the extracted water. A pontoon with anti-hydrocarbon booms is installed in front of the pumping station.
266. The pumped and filtered water is sent to:
the circulation water system; the essential service water system for the nuclear auxiliaries; the circulation water filtration system; the feedwater system for the pumps that wash the filter drums (circulation water filtration system); the demineralised water production system; the auxiliary cooling water system for the conventional auxiliaries; and the emergency service water drainage system (emergency cooling water system), for cooling the two containment heat removal system trains and the third FPC(P)S train. The emergency cooling water pumps can draw from two different filtration channels and as a last resort, from the unit's outfall pond.
2.1.3.14. Description of the electrical supply 267. The electrical power supply installation needs to provide the auxiliary systems with the electricity required for the various plant operating phases. It is divided into two parts:
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The external electricity supply, comprising: o the main electricity supply system (400kV), used to supply the auxiliary devices (step- down transformer) and to feed energy into the grid (main transformer); and o the auxiliary electricity supply system (auxiliary transformer), used to shut down the power plant if the main supply system and the AC generator are lost at the same time. The emergency electricity supply, which comprises four main diesel generators providing 10kV and station black out generators supplying 690V.
268. The power plant is connected to the main electricity supply system via a line breaker. When the power plant is connected to the electrical system, the 400kV supply is connected via the main transformer (20kV/400kV), the coupling breaker and the line breaker.
269. The electrical supply for the production unit's auxiliary devices is connected via two step- down unit transformers with three sets of windings (400kV/3x10kV). Each secondary winding on the step-down transformers is connected to a 10kV distribution panel, which supplies each of the four trains comprising the supply to the auxiliary devices.
270. The preferred method for supplying electricity to the unit is via the step-down transformers. If this supply is lost, an automatic process switches the source so that the auxiliary transformer supplies the unit's distribution panels.
2.1.4. Reactor operating principles and safety provisions
2.1.4.1. Plant operation 271. The reactor will be permanently managed by an operational management team, which includes at least one Operations Shift Manager and operators for each shift and each unit. Each unit will be managed from a control room that contains all information and control devices required to operate the unit during normal power operations, shutdown and incidental and accidental situations. The control room is connected to the outside of the power plant by various means, including telephones, radio links, etc.
272. If the control room is unavailable (e.g. fire) and needs to be evacuated, the unit can be brought to a safe shutdown state from the fallback station, situated in the unit's electrical rooms. The fallback station and the actuators and sensors are linked, so that the safety and reliability associated with the normal connections between this equipment and the control room are not reduced.
273. It is also possible to operate the plant manually. In particular, operation at below 20% of nominal power is only performed manually (with the exception of steam generator regulation).
274. The reactor can operate either in base mode (the operating power is determined and established manually), or by remote control over a limited range (the operating power is then continually reassessed, based on demand from the grid).
275. When operating automatically, the production unit allows the following transients, without adjusting the steam generator safety valves or opening the condenser bypass valves:
A rapid change in load, amounting to 10% of the nominal power, but not exceeding 100% of the nominal load. An increase or decrease in load at a rate of 5% of the full load per minute.
276. Refuelling takes place when the reactor is shut down and the reactor pool is full of water, after removing the head from the reactor vessel. When spent fuel is removed from the reactor, it is stored underwater in the fuel building storage pool for up to ten years, but usually around three years, before transfer to the on-site spent fuel interim storage facility.
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2.1.4.2. Safety principles for plant operation 277. During the unit's lifetime, the safety mechanisms must be guaranteed at all times when the plant is in operation. This is achieved by:
defining the minimum conditions required for the various normal operating conditions, in particular when equipment or functions are not available; defining the conditions under which systems can be used; periodic testing to check the performance of systems and equipment; validating and inspecting the systems and equipment; defining operating procedures in the event of an incident or accident; and applying quality assurance procedures.
278. These requirements are set out in the General Operating Rules and the operator's Management System.
279. The operation of a unit is subject to the following main constraints:
Operating instructions must be followed. The safety and protection systems and automated controls must be available and the safety injection tanks topped up. The reactor building must be closed, the building is however accessible via hatches when the reactor in operation. Activity in the primary system must be restricted if flaws are found in the fuel cladding. Leakage from the primary system into the sealed containment must be restricted.
2.1.5. Summary of improvements for the EPR reactor 280. The reactor is described as third generation and its design derives directly from the reactors using light water and slightly enriched uranium that are in operation today. The decision to use it makes it possible to take full advantage of the wealth of previous experience in design and operation. The main developments in the EPR design are improvements to the safety of installations, particularly as a result of:
the in-depth reinforcement of the principle of defence in depth; the consideration of severe accidents in the design; the increased consideration of internal and external attacks at the design stage; the redundancy and diversification of safety systems; the enhancement of reactor behaviour during transient operating conditions; and an improvement in the quality of components.
281. Reduced environmental impact and increased radiological protection for all persons concerned by:
more efficient use of fuel; a reduction in the production of waste and effluent from plant operation; reduced exposure for plant personnel; and consideration of decommissioning in the design.
282. Figure 2.7 presents a high-level schematic view of the main design safety improvements in the EPR.
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Figure 2.7 Main design choices for improved safety
Double containment with ventilation and filtration Corium spreading area
Containment heat evacuation system
In-containment refuelling water storage tank Four redundancy trains of the main safeguard systems
2.2. VENTILATION SYSTEMS AND THE TREATMENT OF GASEOUS AND AIRBORNE WASTES 283. The gaseous effluent management process can be broken down schematically as shown in Figure 2.8.
Figure 2.8 Effluents management process
Collection Processing Control Discharge
2.2.1. Nature of gaseous radioactive effluent 284. Gaseous radioactive waste includes:
Noble gases, which are formed through fission mainly comprising xenon-133 and xenon-135 and to a lesser degree, krypton-85. These are largely eliminated via radioactive decay in the GWPS retarding beds. Opening the primary system to load fuel may increase the activity concentration of noble gases discharged into the stack. Argon-41 (with a half-life of under two hours), formed by the activation of atmospheric argon, may also be detected when the air in the reactor building is changed. Tritium which is formed by ternary fission in the fuel, by activation of the beryllium in the secondary neutron sources, and by the activation of the boron and lithium added to the primary coolant. This can be found in the form of tritiated water in the various reactor tanks and pools. It is transported by the ventilation system and continuously discharged as steam via the main stack and other outlets. Carbon-14 is formed mainly by the activation of oxygen and nitrogen dissolved in water. Carbon-14 is discharged to atmosphere mainly in the form of methane and also to a lesser extent, as carbon dioxide.
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Radioiodines, consisting mainly of iodine-131 and iodine-133, are formed through fission. The iodines are retained in the iodine traps installed in the ventilation systems (these traps are activated as required). Aerosols are formed mainly through activation (cobalt-58 and cobalt-60) and through fission (caesium-134 and caesium-137). The air carried by the ventilation systems is constantly filtered. Radioactive discharge into the environment in the form of aerosols, represents a mass of less than one microgram per year (largely cobalt-60). The devices for treating gaseous radioactive waste (filters, iodine traps, recombination, retarding beds) help to limit the activity discharged into the environment.
2.2.2. Sources of gaseous radioactive effluent 285. Gaseous radioactive waste is divided into three categories, as shown in Figure 2.9.
Figure 2.9 Nature of gaseous radioactive effluent
Primary gaseous Gaseous effluent from Secondary gaseous effluent ventilation effluent
Gaseous waste Condenser processing vacuum system
Nuclear auxiliary building ventilation Nuclear auxiliary system, Nuclear auxiliary building Controlled safeguard building ventilation, building ventilation system Containment sweep ventilation system, ventilation system Effluent treatment building ventilation system
Discharge to stack
286. These three categories are described in the following paragraphs.
2.2.2.1. Gaseous effluent from the primary system 287. Gaseous effluent from the primary system comes from the degassing of the primary effluent by the CSTS degassers and degassing in the gas caps of the tanks containing primary coolant or primary waste (CVCS, CSTS and reactor coolant system tanks, and some NVDS tanks). It comprises mainly hydrogen, nitrogen and gaseous fission and activation products and is therefore radioactive.
288. Nitrogen purging is used to maintain low levels of hydrogen and oxygen. Hydrogenated or aerated effluent is therefore not found in the gaseous discharge from the EPR unit's primary system. Primary gaseous effluent is directly discharged into the GWPS.
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2.2.2.2. Gaseous effluent from ventilation 289. This effluent comes from the ventilation extraction systems in the rooms that may be contaminated or present an iodine risk in the NAB, the fuel building, the safeguard buildings, the reactor building, the operational service centre, the access building and the ETB.
2.2.2.3. Gaseous effluent from the secondary system 290. Gaseous effluent from the secondary system comes from vapour that is extracted when the condenser is evacuated.
291. This air may be slightly contaminated, particularly with tritium if there is a leak between the primary and secondary systems at the steam generator tubes.
292. The effluent is collected by the condenser vacuum system then sent to the NAB ventilation system, where it passes through a HEPA filter before being discharged via the stack.
2.2.3. Treatment of gaseous radioactive effluent 293. A diagram with an overview of the GWPS is shown in Figure 2.10.
294. Different types of gaseous effluent are treated by different systems, depending on their nature. Primary gaseous effluent is treated by the GWPS. Ventilation air is filtered by the respective building ventilation systems such as the NAB ventilation system, fuel building ventilation system, controlled safeguard building ventilation system, containment sweep ventilation system (high and low flow rate), operational service centre ventilation system, access building ventilation system and the ETB ventilation systems.
295. These treatment systems are described in the following paragraphs.
2.2.3.1. Treatment of gaseous effluent from the primary system 296. The treatment of primary gaseous effluent includes, the regulation of oxygen and hydrogen levels, the purging of the tanks and the recycling of the treated gases.
297. Primary gaseous effluent is treated by the GWPS. This system compensates for the variations in the free volume of the purged tanks and confines the gases by keeping the pressure below atmospheric pressure. It limits the hydrogen concentration through recombination with oxygen, manages the excess gas produced and retains the radioactive noble gases for a period of decay.
298. The effluent is collected and the hydrogen is then recombined. The recombined effluent is compressed using two compressors so that it can either be redistributed to purge the facilities, or sent to the retarding beds and discharged via the stack.
299. There are three retarding beds, whose purpose it is to retain the noble gases for a period of decay; xenon is retained for 40 days and krypton for 40 hours. Discharging via the stack is automatic and depends on the pressure in the retarding beds. When discharged, the gaseous effluent is passed through a HEPA filter.
300. The following diagram summarises the processes for discharging gaseous effluent from the primary system.
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Figure 2.10 Processes for discharging primary gaseous effluent
2.2.3.2. Treatment of gaseous effluent from ventilation 301. Figure 2.11 summarises the treatment of gaseous effluent from ventilation.
Figure 2.11 Treatment of gaseous effluent from ventilation
Ventilation system of the NAB 302. The gaseous effluent from ventilation extracted from the rooms in the NAB, the safeguard building (excluding incidents), fuel building and from the purging of the reactor when the unit is shut down (high flow rate containment sweep ventilation system), is treated by the NAB
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ventilation system. A NAB ventilation system operates for each unit and each will include extraction equipment connected to the stack with pre-filters and HEPA filters which can be switched to iodine traps that are bypassed in normal operation.
303. The extraction system connected to the stack comprises:
six filtration trains with a unit flow rate of 20,000m3 h-1: three for output from the nuclear auxiliary ventilation, two for fuel building ventilation and one for safeguard building ventilation; one 25,000m3 h-1 filtration train for the high flow containment sweep ventilation system; four extraction fans; four iodine traps, each with its own heater; and four booster fans to make up the additional pressure loss.
304. The ventilation system also discharges treated primary gaseous effluent from the GWPS via the stack. The NAB ventilation system includes iodine traps which are used if high iodine levels are detected in the gaseous effluent.
Ventilation system of the fuel building 305. The fuel building ventilation system manages the supply, extraction, filtration, monitoring and discharge of ventilation air from the fuel building. This ventilation air goes to the NAB ventilation system for treatment prior to discharge via the stack.
Ventilation system of the controlled area in the safeguard buildings 306. In normal operating conditions, extraction and discharging are performed by the NAB ventilation system. The controlled safeguard building ventilation system is only used in accident situations, such as incidents involving the loss of primary coolant or fuel handling accidents in the fuel building. It has two filtration trains, consisting of pre-filters, HEPA filters and iodine traps.
Ventilation system of the reactor building Low flow rate containment sweep ventilation system: 307. This system conditions, extracts and filters the air used to purge the containment. It operates whether the unit is shut down or operational so that personnel can access the reactor building. It has two filtration trains, consisting of pre-filters, HEPA filters and iodine traps.
High flow rate containment sweep ventilation system: 308. During unit shutdown, this system conditions and discharges the air that ventilates the containment via the stack. The air flowing through the system also filters through an iodine trap in the NAB ventilation system.
Ventilation system of the controlled area in the operational service centre 309. The controlled area in the operational service centre comprises laboratories and changing rooms. The system has two filtration trains with pre-filters and HEPA filters.
Ventilation system of the effluent treatment building 310. This system conditions, extracts and filters the ventilation air from the ETB. It comprises a network of extraction ducts connected to the stack, with pre-filters, HEPA filters and an iodine trap which is bypassed under normal operating conditions.
Ventilation system of the controlled area in the access building 311. The system includes two filtration trains with pre-filters and HEPA filters.
2.2.3.3. Treatment of gaseous effluent from the secondary system 312. Figure 2.12 summarises the treatment of gaseous effluent from the secondary system.
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Figure 2.12 Treatment of gaseous effluent from the secondary system
Secondary Condenser HEPA filtration (nuclear auxiliary Discharge to gaseous effluent vacuum building ventilation system) stack
313. Gaseous effluent from the secondary system is collected by the condenser vacuum system and then sent to the NAB ventilation system. It is discharged to the stack after passing through a HEPA filter.
2.2.3.4. Discharge of gaseous radioactive effluent 314. Gaseous radioactive waste is discharged to one of the stacks located on the NAB for each unit, depending on the origin of the effluent (see Figure 2.13 and Figure 2.14.). The rate of extraction to the stack must remain above a minimum value and the effluent is checked by measuring the different categories of nuclides. Details of the requirements associated with the inspection and discharge of gaseous waste are provided in Chapter 3.
Figure 2.13 Unit 1 stack discharges
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Figure 2.14 Unit 2 stack discharges
2.2.3.5. Summary of optimisation measures for limiting the impact of gaseous radioactive waste from EPR 315. The EPR facilitates a reduction of gaseous radioactive effluent discharges to the environment when compared to previous generation reactors. This is achieved by a newly designed GWPS, the absence of pneumatic valves in the reactor building, the absence of CSTS intermediate tanks and improved ventilation and filtering. These features are described in more detail below.
Gaseous waste processing system 316. The EPR's design provides a significant reduction in gaseous radioactive discharge due to its GWPS. In particular, this system offers the advantage of being able to treat gaseous effluent and to operate in quasi-closed loop mode when in normal operation. The main improvements which allow this are:
the sharing of the CSTS and reactor boron water make-up system tank covers, which restrict the volume of gaseous discharge in normal operating conditions (constant gaseous balance during water movement); continuous nitrogen flushing of the tank covers, lowering the hydrogen content and standardising the gas treatment, whether its composition is hydrogen or oxygen dominant; recycling of gases, which restricts the volume of gaseous discharge in normal operating conditions; recombination of hydrogen; decay of noble gases with a short half-life (especially xenons and kryptons) on retarding beds (charcoal beds); and automatic discharge via the stack as soon as a threshold pressure is reached (this threshold may be modified (setpoint) according to the volumes of gas to be treated, allowing the system's storage capacity to be adapted).
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Absence of pneumatic valves in the reactor building 317. The absence of pneumatic valves in the reactor building reduces the gaseous discharge from this building, limiting the amount of waste during maintenance operations and the commissioning of its containment sweep ventilation system.
Reduction of gaseous tritium discharge 318. Unlike the 1300MWe plant series, the EPR design applies the design measures adopted for the French N4 unit, in that the CSTS has no intermediate tanks (in the case of the 1300MWe plant, 80% of the gaseous tritium discharge comes from these tanks). Consequently, in the case of the EPR, the majority of gaseous tritium discharge comes from the evaporation of the fuel pool.
An improved ventilation/filtering system 319. All of the ventilation systems for the rooms in the NAB, safeguard building and the fuel building have improved iodine filtration. As regards the 1300MWe plant series, only some of the NAB rooms can be switched to iodine traps after filtration through a HEPA filter. For the EPR, all of the rooms which are divided into "cells" (two for the fuel building, three for the NAB, one for the safeguard building) associated with the ventilation systems, the ventilation air filters through HEPA filters and iodine traps, may be used if iodine is detected.
2.3. LIQUID WASTE TREATMENT
2.3.1. Nature of liquid radioactive effluent 320. Liquid radioactive effluent includes:
Activated corrosion products, the majority of which (iron, nickel and cobalt) are released as a result of the corrosion of the steam generator pipes. These products circulate and are deposited in the reactor's primary system. Corrosion products are present in the primary coolant in soluble or particulate form. As they pass through the reactor core, they are activated by neutrons. The activated corrosion products formed are mainly cobalt-58 (from nickel-58), cobalt-60 (from cobalt-59), silver-110 (from silver-109), manganese-54 (from iron-54), and antimony-124 (from antimony-123). They may go into suspension when the water is physically or chemically changed, for instance, when the unit is shut down. These activated corrosion products are mostly retained in the resins in the CVCS. Activation products from chemicals in the primary coolant and secondary neutron sources. These are carbon-14 (produced from oxygen-17 in the molecules of the primary cooling water and from any dissolved nitrogen-14), and tritium (produced from beryllium-9, boron-10 and lithium-6). These two activation products are generated in proportion to the energy produced. They are not retained by the resins in the CVCS. Volatile fission products (caesium-134, caesium-137 and iodine-131), usually in soluble form in the primary cooling water, may originate from fuel "leaks" due to the fuel elements being insufficiently sealed (cladding defects). As the primary cooling system is purified continuously, these radioactive fission products are retained in the resins in the CVCS. Tritium produced by ternary fission is almost entirely retained in the fuel cladding, even if this is defective.
321. The liquid effluent treatment devices (filters, demineralisers, evaporators and degassers) help to limit the radioactivity released into the environment.
2.3.2. Sources of liquid radioactive effluent 322. Liquid radioactive effluent is divided into three categories, as shown in Figure 2.15 and detailed in Section 2.3.2.1 to Section 2.3.2.4.
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Figure 2.15 Sources of liquid radioactive effluent
Drainage water Primary liquid Spent liquid from the turbine waste waste halls
Other gas Residual Chemical Rest of sec. Floor drainage systems drainage effluent system blowdowns 1 and 2 3
Coolant storage Steam and treatment generator blow system down system
No Liquid waste processing No Recyclable waste Recyclable waste system
Yes Yes
Reactor boron Liquid radwaste monitoring and discharge system Site liquid waste and water make- Condenser (additional liquid waste discharge system) on site discharge system up system
Discharge from site
2.3.2.1. Primary liquid effluent 323. Primary liquid effluent includes liquid that has leaked or been drained from the primary coolant water which is not chemically polluted and water from the systems containing primary coolant, which is discharged when the concentration of boric acid in the primary water is changed.
324. It consists of water containing dissolved hydrogen (when the reactor is in operation) or dissolved oxygen (when the unit is shut down), which is borated and contaminated by fission and activation products.
325. The liquid effluent is sent to the CSTS where it is decontaminated and degassed by the evaporator-degasser. After treatment in the evaporator-degasser, they may be reused as make-up water and boric acid in the primary system. Primary effluent that cannot be recycled is sent either to the on-site LRMDS tanks before discharging (distillates only), or to the LWPS.
2.3.2.2. Spent liquid effluent 326. This is divided into three categories described below.
327. Residual drainage: this consists of polluted primary coolant as a result of equipment bleeds and leaks after rinsing. It is not recycled owing to its low boron concentration and potential pollution (inappropriate chemical properties, matter suspension rates too high) in the event of decontamination. Generally speaking, its pollution level means that it may be treated in a different way from chemical drainage.
328. Chemical effluent or drainage: this is produced in the NAB, and consists of water that is more polluted than the residual drainage water. It comes from the nuclear sampling system laboratory and the CSTS.
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329. Floor drainage: divided into three categories:
Floor drainage 1: this is potentially contaminated and comes from leaks from equipment carrying primary coolant and from washing the floors. The sumps are installed in areas with rooms that contain equipment transporting primary coolant. Floor drainage 2: this is potentially uncontaminated and comes from leaks, floor washing and the bleeding of equipment (feedwater or component cooling water system). The sumps are installed in controlled areas with rooms that do not contain equipment transporting primary coolant. Floor drainage 3: this effluent is solely produced in uncontrolled zones. It is usually uncontaminated and comes from leaks, floor washing and the bleeding of equipment (feedwater or component cooling water system).
330. With the exception of Floor drainage 3 effluent, which is sent to the SiteLWDS tanks, the spent liquid effluent is either sent to the LWPS, or treated in a specific way depending on its nature. This usually means demineralisation for residual drainage, evaporation for chemical effluent and filtration for floor drainage. After treatment, the waste is collected in on-site storage tanks before being discharged.
2.3.2.3. Steam generator blowdown water 331. Blowdown water from the steam generators is largely made up of secondary feedwater, which may be polluted with tritium if there is a leak between the primary and secondary systems at the steam generator pipes.
332. It is sent to the SGBS, where it is filtered and demineralised, then recycled to the condenser. When recycling is not possible, the blowdown is sent to the LRMDS tanks before discharge.
2.3.2.4. Turbine hall drainage water 333. Water drained from the turbine halls comes from leakage and from blowing down and emptying the secondary system, with the exception of blowdown from the steam generators. It is sent directly to the SiteLWDS tanks used for turbine hall drainage water.
2.3.3. Treatment of liquid radioactive effluent 334. A diagram providing an overview of the liquid effluent treatment and storage system is shown in Figure 2.15.
335. Liquid effluent is treated by different systems, depending on its nature. Primary liquid effluent is treated in the CSTS. The spent liquid effluent collected by the NVDS is treated by the LWPS installed in the ETB. The turbine hall drainage water is either treated by the SGBS, or directed to the SiteLWDS tanks.
2.3.3.1. Treatment of primary liquid effluent 336. An overview of the treatment of primary liquid effluent is shown in Figure 2.16.
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Figure 2.16 Treatment of primary liquid effluent
337. Primary liquid effluent is treated by the CSTS. The main function of this system is to recycle the boron and water in the primary system after the primary effluent has been treated.
338. In conjunction with the GWPS for gaseous effluent, the CSTS treats all primary liquid effluent, whether it contains dissolved hydrogen or dissolved oxygen.
339. The installation comprises:
six tanks that may be used for demineralised water, distillates or primary coolant; a system for purification by demineralisation; an evaporation and degassing station; and a degasser on the CSTS discharge.
340. The entire CSTS is installed in the NAB. The system is described below.
341. The filtration-decontamination system comprises:
a mixed-bed demineraliser containing resins that reduce the activity of the primary effluent; a filter, which prevents fine particles of resin escaping into the rest of the treatment system; and a feed line to the downstream evaporation and degassing station.
342. The evaporator separates the primary coolant into concentrates and distillates. The degasser treats the evaporation distillates and the demineralised water to produce make-up water.
343. The degasser on the CSTS discharge, extracts the gases produced.
2.3.3.2. Treatment of spent liquid effluent 344. Figure 2.17 below provides an overview of the LWPS.
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Figure 2.17 Treatment of spent liquid effluent
345. The spent liquid effluent is treated in the LWPS. The purpose of this system, designed for two units, is to limit the activity of spent effluent before it is transferred to the LRMDS tanks prior to discharging, while adapting treatment to each category of spent effluent.
346. The spent liquid effluent is collected depending on its source and characteristics. It is then stored in one of three sets, of two tanks. These are the residual drainage tanks, the chemical effluent tanks and the floor drainage tanks.
347. Each set of two tanks has a mixing system, so that the contents of the tanks can be homogenised for sampling. The type of effluent treatment is determined based on the results from this sampling.
Demineralisation for active effluent that has little chemical pollution (residual drainage). Evaporation for active effluent that is chemically polluted (chemical effluent). Filtration for effluent containing little activity (floor drainage).
Residual drainage 348. Residual drainage water is sent from the residual drainage storage tanks to the demineralisation plant which consists of:
an initial filtration to remove suspended solids from the spent effluent; three demineralisers containing resins to reduce the activity of the spent effluent; and a second filter, which prevents fine particles of resin escaping into the rest of the treatment system.
349. There are two successive stages in the demineralisation process:
Recirculation: the spent effluent treated in the demineralisers is returned to the front tank from which it originated. Processing through the open system: once the activity has been checked, the treated spent effluent is sent to the LRMDS tanks.
350. The residual effluent may also be treated through evaporation.
Chemical effluent 351. Chemical effluent is sent from the tanks where it is stored to the evaporation plant. This comprises an evaporator to separate the spent effluent into distillates (with low activity and little chemical pollution) and concentrates (containing most of the activity and chemical pollution) and a storage tank for distillates.
352. The concentrates resulting from evaporation are sent to the solid waste treatment system.
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353. The evaporation distillates may be sent, depending on the activity measured during sampling, for re-input to the evaporation system, for additional treatment or to the LRMDS tanks before being discharged.
Floor drainage 354. Floor drainage water from potentially contaminated areas, is sent from the front tanks where it is stored, to the filtration plant in the LWPS. This system comprises a filter for the removal of suspended solids. The filtered effluent is then sent to LRMDS tanks before being discharged. Floor drainage water may also be treated through evaporation and/or demineralisation.
2.3.3.3. Treatment of steam generator blowdown water 355. Figure 2.18 below provides an overview of the treatment of blowdown water from the steam generators.
Figure 2.18 Treatment of steam generator blowdown water
356. Steam generator blowdown water is treated by the SGBS. One system is used for each unit to purify the blowdown water before it is recycled.
357. The purification plant for the steam generator blowdown water comprises:
two parallel filters that remove some of the suspended solids from the blowdown water; and two parallel demineralisation lines, each with two demineralisers containing resins, plus a second filter that prevents fine particles of resin escaping into the rest of the treatment system.
358. After purification, the blowdown water is sent to the condenser where it is recycled. It may also be sent to the discharge storage tanks before being discharged if the tritium level in the secondary system is lowered, or if the condenser is not available.
359. If the SGBS is not available, the blowdown water can be sent directly to the discharge storage tanks before discharge.
2.3.3.4. Treatment of turbine hall drainage water 360. The turbine hall drainage water (secondary system leaks, blowdown water and drainage, aside from steam generator blowdown water) is sent directly to the turbine hall drainage water storage tanks (CILWDS) (see Figure 2.19).
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Figure 2.19 Treatment of turbine hall drainage water
2.3.4. Storage and release of radioactive liquid effluent 361. Liquid effluent undergoes different treatment depending on its source. Either primary effluent treatment, spent effluent treatment or turbine hall drainage water treatment.
362. All of the effluent collected in the LRMDS is passed through a filter.
363. The different types of effluent are sent to three specific types of tank for temporary storage and checking before discharge:
the LRMDS tanks for primary and spent effluent; the SiteLWDS tanks for water collected from the turbine hall; and the ExLWDS tanks.
364. The liquid radioactive effluent is pre-diluted with cooling water from one of the two units, in one of the pre-discharge ponds before it is released offshore via one of the underwater discharge tunnels. The effluent is checked in accordance with the procedures used for the different categories of nuclides before and during discharge. These checks are carried out in the discharge storage tanks.
365. Details of the requirements associated with the inspection and discharge of liquid waste are provided in Chapter 4.
2.3.5. Summary of optimisation measures for limiting the impact of liquid radioactive waste
2.3.5.1. Reducing the production of liquid chemical and radioactive effluent 366. The EPR installation incorporates design improvements that aim to reduce the production of liquid radioactive effluent. These include:
A choice of materials that limit the release of substances which, through the process, cause the generation of radioactive elements (the reduction of stellites, for example, which cause the release of cobalt). Reinforced sealing requirements for active parts (pumps and valves) and the recovery of primary leaks.
367. The production of tritium is an intrinsic feature of pressurised water reactors. In order to minimise the amount of spent fuel generated, the reactor can be operated with a high fuel burn-up, which enables more energy to be produced using the same quantity of fuel. However, higher burn-up leads to an increase in tritium in liquid waste. To limit the increase in tritium, the fuel rods are clad with a zirconium alloy to limit the diffusion of tritium through the cladding and lithium in the primary coolant is enriched with lithium-7. The following additional measures have also been taken:
An increased number of gadolinium rods absorbing the neutrons in the reactor core, which enables the use of a lower boron (a source of tritium on activation) concentration of the primary water. This represents an annual cost of two electrical production days at full power. Optimisation of the primary chemistry by boron-lithium coordination, releasing an optimal lithium content to prevent the high concentrations that generate tritium. Reduced mass of beryllium in the secondary neutron sources, therefore reducing the
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amount of tritium generated by activation.
2.3.5.2. Design of optimal effluent sorting and treatment systems 368. The liquid radioactive effluent sorting and treatment systems are designed to minimise activity in the liquid effluent and the discharge of liquid radioactive waste into the sea.
369. The design efforts aim to optimise the sorting and treatment systems for:
Optimal recycling of primary water in the process. In the EPR, the aerated primary water can be treated and recycled, particularly due to the design of the GWPS for the processing of aerated gas emissions. This has the effect of reducing the volume and activity of the liquid effluent in relation to the current units. Optimal selective collection of the different types of effluent. In the EPR, more advanced sorting of the drains that collect leaks from equipment, bleeds and floor washing water, means that treatment can be better adapted to the chemical and radiological characteristics of the different types of effluent. This enables a marked reduction in the activities discharged without increasing the production of solid radioactive waste (filters, concentrates and resins). 2.3.5.3. Design of storage and discharge systems adapted to the site 370. The systems for storing and discharging the liquid radioactive effluent are designed to check and quantify the activity of the effluent before it is discharged and to minimise the impact of liquid radioactive effluent on the environment by achieving optimal dilution.
371. The operator can also optimise the use of the storage tanks available, by using the control reservoirs to reduce the activity of the effluent by deliberately extending the holding time before the waste is discharged. In this way, advantage can be taken of radioactive decay, particularly for short-lived radionuclides such as iodine-131 and cobalt-58.
2.4. SOLID WASTE TREATMENT 372. Solid radioactive waste from each unit is collected and sent to the shared ETB for treatment. The ETB is shared between the two units and adjoins the NAB of Unit 1. Therefore the waste from Unit 2 is pre-conditioned in the radioactive waste treatment building of Unit 2 to allow transfer by road to the ETB.
373. The waste consists of active resins from the CVCS, FPC(P)S and CSTS demineralisers, SGBS resins (low activity) and irradiated waste, such as effluent from the CVCS, CSTS and the CSTS filters.
374. The site system collects and sorts the solid waste produced by the EPR unit, provides buffer storage for the containers, manages the radioactive decay of radioactive effluent and partially or completely treats waste for transfer or disposal off-site.
375. Solid waste from the nuclear island and ETB during normal operation, is sent to the solid waste treatment system and then conditioned for transfer off-site to a final disposal facility.
376. Radioactive waste produced by the power plant is divided into two categories:
Process waste is associated with the operation of the plant. This is produced by the treatment of liquids with the aim of, either limiting contamination deposits and reducing their activity to prevent staff irradiation, or reducing the activity of the discharged effluent, whether gas or liquid. The process waste produced by the treatment of gaseous effluent mainly consists of filters and iodine traps. From the treatment of liquids, the waste includes filters, concentrates and ion exchange resins. Operational waste is produced as a result of maintenance work (repair work, reworking, replacement of active materials, etc.). This mainly consists of compactable materials.
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377. Waste treatment and processing described in the sections below represents the reference case and may be further optimised in the detailed implementation at Hinkley Point C.
2.4.1. Waste treatment in the NAB 378. The NAB’s waste treatment system is operated when spent resins are transferred from the unit to one of the solid waste treatment system resin storage tanks via the solid waste treatment system header. These resins include:
the active resins contained in the CVCS, FPC(P)S and CSTS demineralisers; and the inactive resins in the SGBS demineraliser.
379. The system also discharges spent filter cartridges from the production unit to the solid waste treatment system in the ETB for encapsulation.
380. Operational waste (e.g. contaminated paper, clothing) produced in the NAB, is stored in the NAB before being transferred to the ETB for processing.
381. Operational waste from Unit 2 will be pre-conditioned in vinyl bags and containers prior to transfer by road to the shared ETB.
2.4.1.1. Transfer of filter cartridge holders 382. Spent filter cartridges are transferred from the NAB to the ETB by means of a handling machine, which discharges the spent filter and replaces it with a new one in the same operation (barrel system). The handling machine is also used to replace spent filters from the LWPS. The filters are handled in a bay, which is common to the NAB and the ETB and drumming of the filters is also carried out in the ETB.
383. The handling machine also collects drip-off from the spent filter cartridges. This drip-off is then discharged to the NVDS.
384. The spent filter cartridges from Unit 2 are pre-conditioned in concrete drums (type C1 or C4) and closed with a temporary biological plug before transfer by road to the ETB.
2.4.1.2. Transfer of resins from the NAB to the ETB 385. The spent resins from the CVCS, CSTS, FPC(P)S and SGBS, are transferred by flushing to one of the ETB solid waste treatment system resin storage tanks.
386. Liquid radioactive wastes including active ion exchange resins (from demineralisers in the CVCS, FPC(P)S and sometimes CSTS) generated in Unit 2, are transferred directly to the shared ETB via underground galleries.
387. Low activity ion exchange resins generated in Unit 2 are those from the demineralisers of the SGBS and sometimes from the CSTS if monitoring shows activity is low. These resins will be packaged in metallic boxes prior to transfer by road to the shared ETB.
2.4.2. Waste treatment in the ETB 388. EPR waste will be stored in the ETB before it is disposed of. The methods used, enable the waste to be conditioned so that it can be safely transported off-site in accordance with current transportation regulations and so that it can be stored safely and permanently at a surface site away from the plant, in accordance with current regulations. The process also involves the encapsulation of water filters, resins and concentrates. ILW will be sent to the ILW interim storage facility on-site for storage and LLW will be sent off-site for disposal at the appropriate disposal facility as discussed in Chapter 5.
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389. This ETB solid waste treatment system provides the following functions:
Selective collection of all radioactive solid waste produced by the EPR units including spent resins, evaporator concentrates and filter cartridges. Buffer storage for waste (tanks and metal drums). Conditioning of solid waste for transfer or disposal off-site. Intermediate storage of full and empty waste containers.
2.4.2.1. Packaging of spent resins 390. The active resins are evacuated from each demineraliser by flushing into one of the two temporary storage tanks, depending on their level of activity, via the network of pipes and valves in the ETB solid waste treatment system. Spent resins can also come from the LWPS demineralisers. These are also evacuated by flushing via a network of pipes and valves specific to the solid waste treatment system.
391. The active resins are then packaged in C1 pre-cast concrete casks after decay storage, if required.
392. The resins are collected from the tanks and transferred for encapsulation. The active resins are conditioned and drummed in a portable treatment unit which can be moved from one site to another. This is known as the Mercure portable encapsulation unit and is the proposed process for Hinkley Point C.
393. The Mercure process involves mixing the resins with an epoxy polymer inside the concrete container. The epoxy polymer then hardens during the polymerisation process. An initial steel plug is inserted after mixing. The packages are then permanently sealed with a concrete plug in the ETB storage room.
394. The Mercure machine has an “open tunnel” type design and comprises:
a station for loading and identifying the containers; a station for weighing the containers; a station for packaging the waste and inserting/removing the steel plug; and a station for sealing the cover (cold welding on the shell) and discharging the containers.(10)
395. The machine is located in a controlled area and is connected to the permanent LWPS by flexible tubing. It has a pump with a diaphragm, which feeds the spent resins into a hopper, where they are weighed and their activity is measured by gamma spectrometry. The waste is then immediately transferred to a concrete container.
396. The machine is also connected by flexible tubing to a mobile tank located outside the building, which is used for storing the epoxy resin and hardener.
397. Figure 2.20 summarises the spent resin packaging process.
(10) This is a reference process which may alter.
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Figure 2.20 Flow diagram showing the active resin packaging process
2.4.2.2. Packaging of evaporator concentrates 398. Concentrates from evaporation are collected by the LWPS evaporators and are temporarily stored in one of the storage tanks. The most likely treatment for this waste is encapsulation followed by disposal to an off-site facility.
399. Given the presence of high concentrations of boric acid, all pipes carrying concentrates are insulated and electrically marked out to prevent the risk of crystallisation.
400. The concentrates will be collected from the tanks and transferred to the encapsulation unit by pumps. It is anticipated that a mobile machine would be installed in the ETB primarily comprising a locking module, which includes:
a cement hopper; a mechanical arm for handling the packs; a concentrate dosing system; a metal-clad, sealed mixer; and a carriage for transporting the containers.
401. The machine will be connected to the permanent solid waste treatment system installations by flexible tubing, which supplies it with fluid and power. The concentrates from the LWPS evaporators are solidified in a cement and sand based mortar, which is prepared in pre- measured bags in a suitable disposal container. Slaked lime is added to neutralise the sodium borates.
402. Figure 2.21 below summarises the concentrate packaging process.
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Figure 2.21 Flow diagram showing the concentrate packaging process
2.4.2.3. Packaging of spent filters 403. Spent filters from the units and LWPS are transported upwards by the solid waste treatment system’s handling machine, through a transfer tube leading to the encapsulation cell. The filter then travels down the transfer tube and is deposited into a concrete container for packaging.
404. The plugging formulation is then used to plug the spent filters. Mortar is poured into the container through a vibrating channel. The shell is then sent to a buffer area where it is stored during the drying period.
405. Following the drying period, the containers are plugged with concrete which has exactly the same composition and characteristics as the container.
406. Figure 2.22 summarises the water filter packaging process.
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Figure 2.22 Flow diagram showing the water filter packaging process
Supply of concrete NAB Filters, Units 1 shells & 2 & ETB
Transfer by handling machine
Temporary storage Transfer tube
Packed in concrete container
Temporary storage in ETB room
Encapsulation Insertion of temporary cover No Yes
Concrete encapsulation
Drying
Plugging in the storage room
Drying
Storage before dispatch
Dispatch
2.4.2.4. Storage of concrete containers 407. Once filled, concrete containers packaged in the ETB cell are stored for approximately one month in the ETB’s main storage room for drying.
408. The storage capacity of this room is 96 containers, which covers the drying period required for filter waste, equal to a minimum of one month. This is calculated by taking into account the number of packages produced during one mobile encapsulation machine program.
2.5. CONTAINMENT
2.5.1. Description of radioactive product containment barriers 409. The risk of spreading radioactive products, in gaseous or liquid form, is prevented by an in- depth defence procedure, involving the interposition of three static barriers between the radioactive material contained in the reactor and the environment.
The fuel and its cladding, which in normal conditions, retains fission products and almost all of the tritium resulting from fission. The primary system and the systems transporting the primary coolant, which retain the fission and activation products.
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The reactor building, which contains the gaseous radioactive effluent released into the containment and retains leaks of liquid from the systems transporting the primary coolant.
410. These measures are complemented by a dynamic containment function, performed by the ventilation systems in the rooms and buildings that house contaminated fluids (the reactor building, NAB, fuel building, ETB and safeguard building).
2.5.1.1. Measures implemented to preserve the integrity of the fuel cladding 411. The fuel rod cladding provides the first containment barrier. Its role is to contain radioactive substances (actinides and fission products). The integrity of the cladding is assured in Category 1 and 2 situations (normal operation and predictable transient conditions) due to the absence of a boiling crisis. The use of niobium alloy cladding, combined with the implementation of the new in-core instrumentation (based on sensors distributed evenly inside the core), help to reduce the risk of cladding fractures due to pellet-cladding interaction or stress corrosion.
2.5.1.2. Measures implemented to preserve the integrity of the primary system 412. The entire primary cooling system acts as a barrier against radioactive product leaks. The shell is designed, manufactured, assembled and tested to ensure that there is an extremely low probability of sealing defects, fast propagating cracks and significant ruptures.
413. Codes and regulations are used for its design, taking into account the most extreme operating conditions and the most unfavourable transient conditions (static and dynamic loads, normal or exceptional thermal cycles, vibrations, etc.), in order to guarantee its resistance to the occurrence of targeted damage.
414. Materials are chosen based on experience and the tests carried out, with a particular focus on irradiation and inter-granular corrosion. Pressurisation is carried out with sufficient temperature margins so as not to risk the spreading of fragile ruptures (depending on irradiation, changes in the steel transition temperature are monitored by test specimens placed inside the vessel).
415. Support structures are installed for the large pieces of primary equipment to limit the stresses exerted on the primary pipework during normal operation, or in the event of an incident or accident, and to protect against the risks of cracking or ruptures in the pipework.
416. Protection against primary overpressure is provided by safety devices. In the event of an accident, this function is performed by the pressuriser safety valves in addition to automatic reactor shutdown. In the event of a severe accident, these measures are complemented by a depressurisation line, which is specially designed for these situations.
417. The chemical quality of the primary water is defined, to prevent the corrosion of the steels that constitute the primary shell. Boric acid and lithium hydroxide dissolved in the water controlled by the CVCS, therefore provide protection against corrosion and the deposit of corrosive products on both the fuel cladding and in the primary cooling system.
418. With reference in particular to the primary pumps, leaks along the pump shafts are avoided due to a specific sealing system comprising:
the CVCS, which performs injections to ensure that the seals are leak proof and that leaks are contained; the component cooling water system, which cools the coolant fluid as if flows toward the seals; and a sealing system for use during shutdown, which can be activated when the pumps are stopped.
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2.5.1.3. Measures implemented to preserve the integrity of the containment 419. The EPR containment consists of a double enclosure: the outer enclosure is made of reinforced concrete and the inner enclosure is made of pre-stressed concrete. The sealing of the unit is strengthened by:
a metal liner, which covers the inner enclosure; and a leak recovery system for the inner enclosure, known as the containment annulus ventilation system, which filters potential leaks before discharge to the NAB stack.
420. Based on experience acquired at the design stage of previous enclosures, the design of the EPR containment is based on its resistance to an absolute dimensioning pressure. This pressure is used to dimension the entire civil engineering structure and in particular, the pre- stressing system. It includes all of the pressures reached in accidental situations, taken into account at the reactor's design stage, whether these result from benchmark transients, incidents or accidents (Plant Condition Categories (PCC) 2 to 4), or multiple failures and hypothetical accidents involving core meltdown (Risk Reduction Categories (RRC) A and B). Based on results from corresponding studies, the internal containment of the EPR is designed to ensure containment up to a pressure of 6.5 bars.
421. Furthermore, in addition to this containment, some systems and buildings also assist the containment function. The installation is designed so that all of the routes through the containment lead to the peripheral buildings. These buildings and systems are subject to sealing requirements for the situations in which they might be used. With reference to the equipment access buffers in situations when the primary system is open, resealing times are established to ensure rapid recovery of the containment in the event of an accident.
422. Finally, the base mats of the lower parts of the nuclear buildings, provide a barrier to protect the water tables from contamination by the leakage of radioactive liquid.
2.5.1.4. Measures implemented to ensure dynamic containment 423. Rooms and buildings containing contaminated liquids are equipped with ventilation systems. These systems are installed to graduate the pressure levels, by forcing circulation from the less contaminated areas to the more contaminated areas, by maintaining a constant negative pressure inside the areas that can be contaminated, by blowing and extracting air. The systems allow any risk of dissemination in normal operation to be prevented.
424. The air extracted by ventilation is treated by HEPA filters before being discharged to the NAB stack. The ventilation automatically passes over iodine traps in the event of an accident causing the release of radioactive material (see the description of ventilation systems in Section 2.2).
2.5.2. Monitoring of containment barrier seals
2.5.2.1. Measures implemented to monitor and limit spreading outside the fuel cladding 425. The monitoring of the fuel assembly seals in the reactor, is based on radiological activity measurements taken from the primary coolant during reactor operation. This allows any rupture affecting the fuel, to be detected and ensures that follow-up procedures are carried out. The containment of leaks from the first barrier is ensured by the primary cooling system which provides the second barrier.
426. The measures implemented to control the core's reactivity, help to protect the integrity of the fuel cladding.
427. When the fuel is handled during unit shutdown, the fuel assembly seals are monitored by measuring the activity in the reactor building and the fuel handling bay. Potential leaks are contained by enclosing the shell of the corresponding buildings and carrying out the associated ventilation procedures.
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2.5.2.2. Measures implemented to monitor and limit spreading outside the reactor coolant pressure boundary 428. The primary cooling system's water inventory is monitored via the water levels in the reactor pressure vessel and pressuriser. Significant primary leaks are detected by comparing the amount of primary water made-up and discharged (CVCS and collection systems).
429. Changes in the primary inventory are also monitored by measuring the water level in the tanks or the bleed, vent and drainage sumps. These monitoring methods are supplemented in the reactor building and the rooms containing auxiliary systems, by activity measurement sensors used for the ambient air and sump water. Furthermore, special monitoring of the steam generator pipe seals and other heat exchangers is carried out by examining the activity in the cooling systems.
430. With regard to the long-term resistance of the pressurised primary shell, irradiation of the reactor vessel's steel is monitored during operation. This parameter allows ageing to be checked by examining the irradiation of the steel in relation to the risk of the vessel rupturing.
431. The entire primary shell is also subjected to an important inspection programme, which aims to monitor any changes in the materials and structures, particularly with regard to the welding. The programme consists of periodic inspections and hydraulic tests.
432. If an isolable leak is detected in the primary shell, the two production isolation devices installed on all of the systems connected to the primary system are sealed. These devices isolate the primary system, even in the case of failure to seal one of the two devices due to a mechanical, electrical or instrumentation and control fault. In the case of a non-isolable primary leak in the reactor building, containment is assured by the enclosure and the base mat.
2.5.2.3. Measures implemented to monitor and limit spreading outside the containment 433. The containment is monitored regularly by performing resistance and global tests to check that it is sealed to withstand pressure levels higher than or equal to the accident pressure. The inter-containment space is equipped with an activity measurement system and a leakage recovery system, with switchable filters on iodine traps for use in the event of an accident.
434. The containment itself is monitored during normal operation by performing radiological checks on the incoming and outgoing equipment, personnel and waste packages when they leave the nuclear area. Environmental radioactivity monitoring in the vicinity of the facility is also undertaken and ensures that any increased levels of radioactivity in the environment are detected.
435. In accident situations, the containment is automatically isolated at all of the enclosure routes that are not required for the management of accidents. The atmosphere in the containment is checked for fission products and hydrogen. A permanent hydrogen recombination system limits the hydrogen concentration and guarantees the containment’s resistance.
436. In the event of liquid leaks in the nuclear island buildings, the radioactive effluent is recovered by sumps, retention pits and retention tanks. In terms of the reactor building, the base mat is sealed by the metal liner which covers the inner containment below the base mat and above the common raft. The core catcher provides the containment function of the base mat, even in severe accident situations.
437. Finally, a leak in the safety injection system or containment heat removal system, which transports contaminated water outside the reactor building to cool the core or the containment in accident situations, is taken into account in the design by installing contaminated waste re-injection systems in the reactor building.
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2.6. DECOMMISSIONING AND DISMANTLING 438. The EPR has a design life of 60 years before decommissioning. At the end of its operational life, the last spent fuel will be removed and transferred to join the previously discharged spent fuel, in the spent fuel interim storage facility for continued cooling and awaiting the availability of the GDF. It is currently planned that the spent fuel will be transferred to a deep GDF for disposal.
2.6.1. Design for decommissioning 439. The EPR has been designed with maintenance and decommissioning in mind, enabling radiation doses to workers and radioactive waste quantities to be minimised when decommissioning takes place. The design incorporates a number of features to achieve this objective including:
choice of construction materials to minimise activation - low cobalt steels are to be used wherever possible; optimisation of neutron shielding - neutron shielding is utilised between the core and reactor vessel to reduce irradiation of the steel and reactor compartment; optimisation of access routes to nuclear areas - the layout of the primary circuit plant takes account of the handling and access routes for decommissioning; reactor systems design - designed to minimise activation products and circuit contamination; ease of removal of major process components - major components can be removed as a single item for size reduction in purpose built facilities elsewhere; submerged disassembly of reactor pressure vessel; modular thermal insulation - the design facilitates easy removal minimising worker dose; fuel cladding integrity - improved fuel cladding reduces contamination of the circuit with fission products; careful control of primary circuit chemistry to minimise level of activity in the primary circuit; design for decontamination - plant design facilitates decontamination during decommissioning; prevention of contamination spread - containment, ventilation and segregation are utilised to prevent contamination spread; and minimisation of hazardous materials - the use of materials which would result in the creation of hazardous waste during decommissioning is minimised as far as possible.
440. In summary, the design of the EPR includes measures which will:
minimise the activity level of irradiated components; reduce worker dose during decommissioning; permit easy decontamination; minimise the spread of contamination; facilitate the access of personnel and machines for decommissioning and the removal of waste from the reactor building; minimise the volume of radioactive waste; reduce the operator intervention time; and minimise the toxicity of the waste.
2.6.2. Outline of Regulatory provisions 441. The 2008 Nuclear White Paper(11) sets out the UK Government’s policy, that the owners of new nuclear power stations must set aside funds over the operating life of the power station, to cover the full costs of decommissioning and their full share of waste and spent fuel management and disposal costs. This includes the costs of providing safe, secure,
(11) Meeting the Energy Challenge - A White Paper on Nuclear Power, Cm7296, Department for Business Enterprise and Regulatory Reform, 2008. (Note that BERR became the Department for Business Innovation and Skills in 2009)
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environmentally acceptable interim storage for spent fuel and ILW until a GDF is ready to accept this material.
442. The costs for decommissioning, waste and spent fuel management and disposal would be funded through a Funded Decommissioning Programme (FDP), approved by the Secretary of State, which must be in place before the operator uses the site by virtue of the site licence. This ensures that EDF Energy sets aside funds over the operating life of the power station to cover these costs in full. The FDP is periodically reviewed, revised and developed in detail, in readiness for the closure of the site. The FDP describes the totality of the decommissioning process, covering all facilities on the site, commencing up to five years prior to closure and concluding with the removal of all facilities and remediation of the site as necessary to make it available for reuse.
443. Before decommissioning can take place, there is a requirement for the operator to obtain consent from the HSE under the Nuclear Reactors (Environmental Impact Assessment for Decommissioning) Regulations 1999 (EIADR 99). This requires the submission of an Environmental Impact Assessment specific to decommissioning activities and a period of public consultation. This will demonstrate that decommissioning will take place with no unacceptable environmental impact.
444. Current UK policy is set out in a statement published in September 2004, The Decommissioning of the UK Nuclear Industry’s Facilities(12). Key aspects include:
i. Each operator is expected to produce and maintain a decommissioning strategy and plans for its sites. ii. Decommissioning operations should be carried out as soon as reasonably practicable, taking all relevant factors into account as provided for in the relevant operator’s strategy and plan. iii. Sites of decommissioned nuclear facilities may represent a potentially valuable resource. The future use of the site, once decommissioning operations have been safely completed, could therefore be a significant factor in determining decommissioning operations. iv. The use of best available techniques (BAT) to minimise the volumes of radioactive wastes which are created, particularly the volume of ILW. Wherever possible wastes should not be created during decommissioning until an appropriate management solution is, or will shortly be, available for use. v. Any new facility covered by this policy should be designed and built so as to minimise decommissioning and associated waste management operations and costs.
445. Regulation of the decommissioning of a nuclear facility is carried out under a Nuclear Site Licence and relevant environmental permits.
(12) Statement of the UK Government and devolved administrations’ on the decommissioning of nuclear facilities, September 2004
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3. RELEASE FROM THE INSTALLATION OF AIRBORNE RADIOACTIVE EFFLUENTS IN NORMAL CONDITIONS
3.1. AUTHORISATION PROCEDURE IN FORCE
3.1.1. Description of the current procedure 446. Discharges of radioactive substances from nuclear facilities are authorised under the Environmental Permitting (England and Wales) Regulations 2010 (as amended). The Environment Agency of England and Wales is the body that is responsible for the application and enforcement of these regulations at nuclear sites.
447. The Regulations provide a definition of radioactive waste. No discharges or disposals of radioactive waste are permitted unless they are authorised within an Environmental Permit granted under the Regulations. The process for applying for and granting a permit for a nuclear site is described below.
Application. The person or organisation wishing to dispose of radioactive waste completes the application forms and supplies additional information that describes the sources of waste; quantities to be discharged; proposed limits; arrangements for minimising, managing and monitoring the wastes; environmental impacts and anything else requested by the Environment Agency. On receipt of the application, the Environment Agency will determine whether it is duly made and if so, make it publicly available (by placing copies on the Environment Agency’s and local authority’s ‘public register’). It will also invite comments from the public on the applications for new permits and major variations, and will consider any responses during the determination. Determination. The Environment Agency will assess the application and the additional information that has been supplied. It will decide whether or not to grant an Environmental Permit and if so, will decide upon the limits and conditions that will apply. A copy of the standard template with typical conditions and guidance are available from the Environment Agency’s website(13). The Environment Agency will prepare a draft permit and an Explanatory Document that explains the rationale for granting the permit and its limits and conditions. The limits and conditions in a permit are set such that compliance will ensure that the radiation exposure that results from the proposed disposals and discharges of radioactive waste, will be well below the dose limits and dose constraints specified in the Basic Safety Standards Directive. Consultation on Draft Decision. The Environment Agency will consult the public and key partner organisations (for example the ONR and Food Standards Agency). The Environment Agency will review the responses and make any amendments to the Environmental Permit that it considers to be appropriate in the light of those responses. Ministerial Review. Under the Regulations, Government Ministers at the Department of Energy and Climate Change and the Department of Health, have the power to call in a permit application for their own determination if they so choose. Review and Issue of Permit. The Environmental Permit and Decision Document will be completed. Internal approval will be obtained and the documents will be issued to the applicant. Typically, the permit will come in to force 28 days after the permit has been issued. Appeal. The applicant has the right of appeal in the event that it disagrees with the Environment Agency’s decision to refuse an application, or with any of the limits or conditions contained within a permit.
448. The permit holder is required to comply with all of the limits and conditions attached to it. The Environment Agency will undertake compliance assessment activities to satisfy itself that permits are being complied with and to demonstrate that dose limits and constraints specified in the Basic Safety Standard Directive are being met. Typical compliance assessment activities are:
(13) http://www.environment-agency.gov.uk/static/documents/Business/GEHO0410BSHS.pdf
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Inspection. Regulators carry out a programme of planned and reactive inspections which may be announced or unannounced. Monitoring. The Environment Agency, with its partner organisation the Food Standards Agency, undertakes a programme of monitoring to determine the environmental impact and doses to humans from permitted discharges. The results of this programme contribute to the annual Radioactivity in Food and the Environment (RIFE) report which is published by the UK’s environment agencies and the Food Standards Agency. This programme is used as an independent check on the environmental monitoring programmes that permit holders are required to carry out. Enforcement. The Environment Agency will take action, as described in its enforcement policy, to address any non-compliances or potential non-compliances with the limits and conditions of the Environmental Permit. Review. The Environment Agency will review on a periodic basis, the limits and conditions contained within an Environmental Permit. Where appropriate, the Environment Agency will amend the permit to reflect developments in policy, strategy, waste management techniques or new information about impacts. Surrender. An operator may apply to surrender the permit for all the regulated activities or partial surrender covering only the regulated activities that have ceased. The Environment Agency will determine such an application. If an operator wishes to reduce the area of the premises permitted, then it must apply for partial surrender to do so.
3.1.2. Waste discharge limits and associated requirements
3.1.2.1. Annual limits for gaseous waste 449. The proposed limits for gaseous discharges of radioactive waste from Hinkley Point C are presented in Table 3.1. These limits represent the maximum discharges that will be permitted over a rolling 12 month period. These proposed discharge limits have been determined from:
data on effluent discharges from the EDF Group 1300MWe PWR’s and the associated limit values, with consideration for operating margins covering standard contingencies relating to plant operations (unscheduled system drainage for maintenance, temporary faults with abatement plant, etc.); the discharges produced by the EPR in relation to the 1300MWe units and in particular the production of effluent that has no cost-effective abatement technique (tritium, carbon-14); the types of fuel envisaged for the EPR; and system design improvements reducing the amount of discharge as described in Chapter 2.
Table 3.1 Proposed limits for gaseous radioactive discharges from Hinkley Point C
Proposed gaseous annual Category discharge limits (GBq) Tritium 6,000 Carbon-14 1,400 Iodine-131 0.4 Noble gases 45,000 Other radionuclides 0.24
450. Operational experience has shown that discharges of actinides to the environment in gaseous waste are not expected. Therefore a specific limit will not be set for the discharge of alpha emitting radionuclides.
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3.1.2.2. Conditions relating to the disposal of gaseous waste 451. The permit will also apply a number of conditions to the disposal of gaseous waste in addition to the limits. These are summarised thematically below.
Management. Management systems, organisational structures and resources shall be provided to identify and minimise risks; to demonstrate compliance with the limits and conditions of the permit; to make and retain records; to make the permit available to those having duties under it; and to appoint and consult with qualified expects on matters related to compliance. Operations. The scope and geographical extent of permitted activities is defined. BAT shall be applied to minimise the radioactivity in the waste, minimise the activity discharged and minimise the impacts from discharges. Optimisation techniques and disposal systems shall be maintained and monitored to demonstrate their continued effectiveness. Disposal. The types, routes and limits for disposals are identified. Sampling, measurements, tests, surveys and calculations shall be undertaken to demonstrate compliance with the limitations and conditions of the permit. BAT shall be selected. All defined techniques shall be appropriately commissioned, maintained and calibrated. MCERTS (the Environment Agency’s Monitoring Certification Scheme) certification/accreditation is required for monitoring of disposals and is expected to be in place prior to operation of Hinkley Point C. Records of all monitoring shall be maintained and access to outlets for independent monitoring shall be provided. The operator will also undertake an environmental radioactivity monitoring programme in the vicinity of the facility. The Environment Agency may specify monitoring requirements. Information. Records to demonstrate compliance with the limitations and conditions of the permit shall be generated and retained on-site. Records of disposal shall be sent to the Environment Agency. An environment case to demonstrate that people and the environment are being protected, shall be maintained throughout the activity’s lifecycle. Information shall be provided if requested by the Environment Agency. The Environment Agency shall be notified of any disposal of radioactive waste that is, has or could result in a breach of the limitations and conditions contained in the permit.
3.2. TECHNICAL ASPECTS 452. Feedback from the Fukushima event known at the time of writing reveals that there are no implications for the release of radioactive effluents from the Hinkley Point C installation in normal conditions.
3.2.1. Annual discharges foreseen 453. The robust application of the waste minimisation and waste management techniques presented in Section 3.2.3 will ensure routine annual discharges at levels below the annual limits proposed in Section 3.1.2.1. Operational feedback from EDF’s existing fleet of reactors has identified a number of factors that impact on the environmental performance of the plant and can result in elevated discharges. These factors include unplanned shutdowns and the failure of fuel within the reactor core. The proposed limits include contingencies that take account of these factors. The performance of the reactors and the associated waste management systems are carefully monitored by the operator and Regulators to ensure that any event which results in an increase in discharges is promptly identified and addressed.
3.2.2. Origins of the radioactive effluents, their composition and physico-chemical forms 454. The main sources of gaseous radioactive waste are:
activated corrosion products from corrosion of metal surfaces in contact with the primary coolant; activation products from chemicals in the primary coolant and secondary neutron sources; and
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volatile fission products from small defects that may occur in the fuel cladding.
455. Gaseous radioactive discharges may either be periodic or continuous.
Periodic discharge. Gaseous waste from planned time bound tasks. These discharges include the gaseous waste from sweeping, (which enables personnel to access the reactor building when the unit is in operation), from iodine trap tests for the various ventilation systems and from calibration checks on the activity measuring channels.
Continuous discharge. Gaseous waste that is continuously discharged via the main stacks. These are gaseous wastes collected by the building ventilation systems but excludes the removal of air from the reactor building associated with sweeping.
456. The different types of radioactive gaseous waste that are created by the power generation process, are described in detail in Section 2.2. They are divided into the following five radionuclides or radionuclide groups:
Tritium; Carbon-14; Iodine-131(14); Noble gases; and Other radionuclides.
457. The breakdown of the total activity of noble gases and other radionuclides is based on an analysis of data for the actual discharge from the French nuclear power plants most similar to the EPR unit. The anticipated distribution of the radionuclides for each of these groups, expressed in percentage terms by activity is presented in Table 3.2 and Table 3.3. This takes account of the fact that the first barrier (the fuel rod cladding) cannot ensure complete leak resistance and explains the presence of fission products such as isotopes of iodine and isotopes of caesium in the gaseous waste.
Table 3.2 Distribution by activity of noble gases discharges in gaseous form
Radionuclide Ratio (%) Krypton-85 13.90 Xenon-133 63.10 Xenon-135 19.80 Argon-41 2.9 Xenon-131m 0.30 Total noble gases 100.00
Table 3.3 Distribution by activity of fission and activation products discharged in gaseous form
Radionuclide Ratio (%) Cobalt-58 25.50 Cobalt-60 30.10 Caesium-134 23.40 Caesium-137 21.00 Total fission/activation products 100.00
(14) Representative of other iodine isotopes
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3.2.3. Management of these effluents, methods and paths of release
3.2.3.1. Waste minimisation techniques 458. The liquid and gaseous waste collection systems contribute to the containment of radioactive substances and limits releases into the environment. Gaseous waste is segregated at source and depending on its characteristics, is passed to one of a number of treatment systems. The radioactive gaseous processing system is described and discussed in detail in Chapter 2.
459. An approach has been developed to minimise radioactive gaseous discharges from Hinkley Point C’s operating activities. This approach is based on the design of the plant and the operational practices that will be implemented. BAT will be used to minimise gaseous discharges at source and to abate any wastes that are unavoidably created. The techniques aim to balance worker doses and costs incurred during treatment in the plant with public doses from discharges. Systems and equipment are managed and used in a manner so as to minimise the environmental impacts of discharges. All discharges are monitored and recorded to demonstrate that they fall within the permitted limits.
460. A number of techniques have been applied to the design of Hinkley Point C which will ensure that discharges of gaseous radioactive waste are minimised. These centre on minimisation at source, containment within plant and recycling of process gases. Outgassing of the primary coolant takes places when liquids pass from the primary coolant system into other liquid processes. A range of design features ensure abatement of the gases generated by this and the associated discharges. In particular, Hinkley Point C’s primary circuit GWPS utilises the best current methods that were developed for the German Konvoï reactor design. The key feature of this design is the recovery of purge gas (nitrogen) which is compressed and recycled into the system to ultimately minimise discharges and retain short-lived radioactive gases to allow decay. The use of nitrogen is a safety benefit during the processing of the hydrogen off-gas in the GWPS.
3.2.3.2. Discharge of gaseous radioactive waste via the nuclear auxiliary building stack 461. The devices for treating or partitioning gaseous wastes, shown in Chapter 2 (filters, iodine traps, recombination, delay beds) help to limit the activity discharged into the environment. Following treatment, gaseous radioactive effluent is directed to the NAB stack for controlled discharge. The following checks are performed to ensure compliance with the required limit values and to monitor performance:
The activity discharged every year, for the five categories of radionuclides. The emission rate at the discharge point, for tritium, noble gases, iodine-131 and the other activation or fission products emitting beta or gamma radiation is expected to be used as an alarm level to monitor plant performance. The activity concentration in the receiving environment, for tritium and particulate activity.
462. Gaseous waste is produced from the primary cooling circuit which is also processed by the GWPS. The discharge rate for this waste varies between approximately 0.2m3 h-1 and 125m3 h-1. This waste is monitored in the same way as for continuous discharges.
463. Periodic discharge can only be released once the checks described in Section 3.3.1.1 have been performed and the discharge rate has been determined (in compliance with the emission rate values). The Hinkley Point C power station can only release one batch of periodic discharge from either of its units at any one time. Periodic discharges can only be released if the volumetric flow rate via the stack is greater than 120,000m3 h-1 and if the total beta activity concentration discharged via the stack does not exceed the pre-determined threshold set by plant operating procedures.
464. In addition, any periodic discharges other than those from the reactor building, can only be discharged if the iodine extraction systems are operating on the relevant ventilation systems.
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465. Any periodic discharge associated with sweeping ventilation in the reactor building is immediately filtered on an iodine trap using the low flow containment sweeping ventilation system when the reactor is shut down.
466. If, during periodic discharge one of these conditions is not met, the discharge procedure is interrupted and can only be resumed once these conditions are once again satisfied.
3.2.3.3. Discharge of gaseous radioactive waste from supporting facilities 467. The Hinkley Point C site will also contain supporting facilities for the storage of spent fuel, the storage of solid radioactive waste and the laundering of work wear. These facilities will have gaseous waste management systems that will, if required, discharge gaseous radioactive waste to the environment.
468. The hot facilities include the hot laundry, the hot workshop and hot warehouse. The ventilation system is designed so that the air flows from less contaminated to most contaminated area. The decontamination facilities are kept at negative pressure which provides containment. Air from the fume cupboards and hoods of the decontamination area is sent to an air washer before leaving the room by the general ducts networks. The exhaust of the hot facilities ventilation system is connected to the exhaust stack of NAB via the ETB ventilation system.
469. The interim storage facilities for ILW and spent fuel will contain separate ventilation systems. The routine discharges from these facilities will be very low. During detailed design it will be determined if ventilation systems require filtration. The design and operation of the ventilation systems will take full account of the need to apply and demonstrate BAT in accordance with the environmental permit conditions.
470. Consideration has been given to discharges from these facilities and the contribution that they will make to the total discharges from Hinkley Point C. The conclusions of these considerations are that the overall contribution will be low and that the site’s total discharges will be dominated by those from the two EPR units. The key arguments developed to support these conclusions are:
Tritium. The main source of gaseous tritium is from the evaporation of tritiated water from the reactor building’s fuel pool. The water becomes tritiated by the diffusion of tritium through the fuel casing. There is not expected to be a significant contribution to gaseous tritium discharges from the spent fuel interim storage facility pool because this fuel has previously been stored for up to 10 years (minimum 3 years) in the reactor building pool and will have cooled considerably. Therefore the activity of the interim storage facility pool will be lower than the reactor building pool and there will be less evaporation of tritiated water. Carbon-14. Carbon-14 is produced in the primary cooling circuit as a result of activation. This effect will not occur in either the spent fuel interim storage facility or the ILW interim storage facility. Iodine-131. Iodine-131 and other isotopes of iodine are produced as a result of fission and are typically released as a consequence of fuel failures. No additional iodine will be produced once the fuel has been removed from the reactor. The majority of isotopes of iodine are expected to have decayed completely by the time the fuel is transferred from the reactor building fuel pool to the spent fuel interim storage facility pool. Noble gases. Noble gases are produced as a result of fission and are typically released as a consequence of fuel failures. Negligible noble gases will be produced once the fuel has been removed from the reactor. The majority of noble gases, other than krypton-85, are expected to have decayed completely by the time the fuel is transferred from the reactor building fuel pool to the spent fuel interim storage facility pool. Other radionuclides. These comprise other fission or activation products emitting beta or gamma radiation. The majority of these discharges have been observed to take place from fuel failures during shutdown and during operational maintenance activities. The quiescent nature of fuel and waste storage is expected to reduce the potential for fuel failure and therefore limit the migration of these radionuclides into the gaseous waste
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stream.
3.2.3.4. Discharge of gaseous radioactive waste by other means 471. The EPR unit has no areas that are likely to be contaminated during normal operation which are not connected to a stack.
472. However, a very small proportion of waste, consisting of steam discharge from the secondary cooling system, is released via the atmospheric dump system. Any micro-leaks in the steam generator pipes, cause tritium to be transferred to the secondary system. Where such leaks exist, the activity concentration due to tritium in the steam from the steam generators, is no more than a few thousand becquerels per litre. Some normal operating conditions can therefore lead to the discharge of steam with low levels of tritium into the atmosphere via the atmospheric dump system.
3.3. MONITORING OF DISCHARGES
3.3.1. Sampling, measurement and analysis of discharges 473. Operators on nuclear facilities in England and Wales are required to demonstrate that they are applying BAT for sampling, measurement and analysis of discharges. The full details of the monitoring arrangements are not yet available for Hinkley Point C. Fully developed monitoring practices will be developed at a later stage to ensure consideration of the most relevant and best practices closer to the start of operations at Hinkley Point C. Monitoring arrangements will be fit for purpose and it is expected that the arrangements will be consistent with UK custom and practice.
474. To provide indicative information of the system and approach anticipated at Hinkley Point C, details are provided based on the Sizewell B PWR operated in the UK by EDF Energy. It is expected that Hinkley Point C will have systems in place that are, as a minimum, of comparable standard to Sizewell B. This will ensure that operational practices in the UK are maintained, that consistency with Environment Agency guidance is achieved and the requirement to demonstrate the application of BAT is fulfilled.
3.3.1.1. Sampling, measurement and analysis techniques 475. All of the sampling and analyses performed on the GWPS, at the level of the reactor building and the stack, will be the responsibility of the waste laboratory located on the Hinkley Point C site.
476. The sampling devices are not, as yet, precisely defined. They will be, at least, equivalent to counterparts deployed at the UK’s current PWR at Sizewell B for the measurement of tritium, carbon-14, noble gases, iodine-131 and other radionuclides.
477. Before any periodic discharge is made, effluent from the reactor building is subject to the following analyses, which are carried out on representative samples.
Measurement of noble gas activity. Measurement of iodine-131 activity. Measurement of beta particulate activity.
478. These analyses enable the composition of gaseous waste to be determined, to establish whether discharge is possible (in compliance with annual activity limits and associated notification levels).
479. Furthermore, all of the gaseous waste discharged via the stack will be subject to the following weekly analyses:
Flow rate measurement in the stack.
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Overall beta activity measurement in the stack. The measurement of tritium activity, based on continuous sampling. The measurement of carbon-14, based on continuous sampling. An analysis of noble gases. The measurement of overall gamma activity and an analysis of iodine-131, based on continuous sampling. An overall beta activity measurement and a gamma spectrometry analysis of the other radionuclides, based on continuous sampling. Analysis of particulate samples for the presence of peaks related to americium, to demonstrate the absence of actinides.
480. These analyses enable assessment of the activities discharged and to determine the emission rates from the stack, therefore ensuring compliance with the annual activity limits and emission rates for the stack.
3.3.1.2. Determination of discharges 481. This section presents a summary of the evaluation of the discharge activity.
Discharged activity 482. For each category of radionuclides, the activity discharged is the product of the activity concentration for that category of radionuclides and the volume of air released into the stack during an agreed period.
483. The amount of discharged activity is assessed at a frequency agreed with or specified by the Environment Agency.
484. The volume of air released via the stack during a period, is the product of the average flow rate in the stack and the duration of that period.
Activity concentration 485. The activity concentration for each of the radionuclide categories is determined as follows:
For tritium, activity concentration is based on the tritium activity measured from the sample and the volume of effluent flowing through the sampling device throughout the duration of the sampling time. For carbon-14, activity concentration is based on the carbon-14 activity measured from the sample and the volume of effluent flowing through the sampling device throughout the duration of the sampling time. For noble gases, activity concentration is derived from activity concentration in the stack and the volume of effluent sampled. For iodine-131 and other radionuclides, activity concentration is based on the spectrometric analysis results and the volume of effluent flowing through the sampling device during the period.
3.3.2. Principal features of monitoring equipment
3.3.2.1. Detection limits for the measuring techniques used 486. The sampling devices will be at least equivalent to counterparts used at the UK’s current PWR at Sizewell B. These sampling devices have undergone an appraisal process to demonstrate that they represent BAT for industrial measurements. The appraisal includes an assessment of detection limits corresponding to each technique and consideration of how these detection limits may vary according to the operating modes of the EPR.
487. The detection limits as shown in Table 3.4, are applied to Sizewell B and are provided for information purposes. Comparable performance is expected at Hinkley Point C.
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Table 3.4 Detection limits for measurements of gaseous radioactive discharges from Sizewell B (for information purposes)
Type of Detection Radionuclides Measurement Type of discharge sampling limit (Bq m-3) Main unit vent stack 2 Radwaste building 2 Tritium Gas bubbler Liquid scintillation HVAC(15) stack Gaseous radioactive 500 waste system Main unit vent stack 0.005 Activated Radwaste building Gamma 0.005 Iodine-131 carbon HVAC stack spectrometry cartridge Gaseous radioactive 0.05 waste system Main unit vent stack 0.001 Gamma Radwaste building Other 0.001 Paper filter spectrometry and HVAC stack radionuclides scintillation probe Gaseous radioactive 0.1 waste system Main unit vent stack 1,000 Radwaste building Ionisation 1,000 Noble gases Overall beta HVAC stack Chamber Gaseous radioactive 10,000 waste system Main unit vent stack 0.5 Radwaste building Precipitation and 0.5 Carbon-14 Gas bubbler HVAC stack liquid scintillation Gaseous radioactive 200 waste system
3.3.2.2. Checks on equipment and measuring devices 488. The condition of all pipes transferring gaseous radioactive waste between the various buildings will be subject to routine inspection.
489. Installation start-up devices such as iodine traps are duplicated by a manual control. These devices are subject to a maintenance programme and their operation is checked routinely.
490. The operation of the equipment and associated alarms located in the stack will be checked once a month. Calibration of these devices will be carried out periodically.
3.3.3. Alarm levels and intervention actions 491. Instruments that continuously monitor the airborne activity of a gaseous effluent stream, feature alarms both for recording faults and for high levels of activity of specific radionuclides or groups of radionuclides, such as noble gases. It is standard practice for such instruments to have one or two ‘high’ alarm levels, with the lower alarm triggering a notification locally and to a control room, and potentially a higher alarm level linked to a plant actuation such as the closure of an isolation valve or the diversion of effluent through additional abatement plant.
(15) Heating, ventilation and air conditioning
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492. Gaseous effluent waste streams from the NAB stacks are also continuously sampled using media that is changed after a pre-determined period. This media, from periodic emissions monitoring, is normally analysed in a laboratory and the results are compared to pre- determined thresholds.
493. Responses to the alarms or readings above the thresholds will be embedded in site procedures, typically as follows:
Check, as far as is reasonably practicable, that the alarm or reading is genuine and is not due to an instrument fault. For continuous emissions monitors, compare trends from other instruments monitoring the same or influent streams to narrow the potential origin of the increased activity to a particular plant area. Check tasks or plant areas that could have contributed to the elevated reading, considering work that if inadequately conceived or executed could lead to an increased discharge. Obtain advice from Qualified Experts or other suitably qualified and experienced persons as directed by procedures. Compare the activity release rate or activity concentration in the effluent stream with pre- determined thresholds listed in site procedures. o A lower threshold, set to minimise discharges and to indicate an upward trend. If exceeded it will necessitate a change of periodic emission monitoring samples within a reasonable time period and repeat monitoring at a suitable frequency to establish the cause of the elevated values and to confirm when readings have returned below the threshold. Additional abatement plant may be put into service. o A higher threshold, set to indicate an abnormal plant occurrence that could threaten compliance with discharge limits if discharges were to continue above the threshold for an extended period. If exceeded it will necessitate an immediate and then regular change of periodic emission monitoring samples and immediate implementation of additional abatement or termination of the discharge, if practical and safe to do so. Take precautionary actions where appropriate, for example termination of work in an area that could be giving rise to the elevated discharge activity.
3.4. EVALUATION OF TRANSFER TO MAN
3.4.1. Models and parameter values used to calculate the consequences of the releases 494. An assessment was undertaken of four reference groups to assess the consequences of the release of airborne radioactive effluents in normal conditions. The assessment was based on the discharge limits proposed in Table 3.5. The four reference groups assessed were:
1. Local reference group in the vicinity of the Hinkley Point site. 2. Channel Islands reference group. 3. Nearest Member State reference group (in France). 4. Reference group in the Republic of Ireland (although not the closest).
495. It is acknowledged that the Channels Islands are not a Member State of the European Union. However it has been decided to include relevant data for information and completeness.
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Table 3.5 Reference group location data
Reference group Distance to site Bearing (deg. from north) Local <2km - Channel Islands 178km 158º Nearest Member State (in France) 186km 152º Republic of Ireland 246km 298º
496. The local reference group is the same as that determined in the Environmental Permit application for Hinkley Point C.
497. Hinkley Point site data, considered for the assessment of the dispersion of gaseous discharges, are outlined in Table 3.6.
Table 3.6 Assumptions and parameters for atmospheric discharges
Parameter Value Stack diameter (m) 3 Release rate (m s-1) 9.6 Emission rate (m3 h-1) 244,290 Ambient temperature of discharge gases (°C) 15 Station orientation Turbine hall facing sea Physical Stack Height (m) 70 Effective Release Height (m) 23.3
3.4.1.1. EPR stack height 498. The height of the main stack on the reactors is determined by a number of different criteria, such as civil engineering, environment (atmospheric dispersion) and impact on the visual landscape. The ADMS dispersion model is used to predict dispersion factors and deposition factors for routine releases for a range of stack heights. The main stack will have a physical height of 70m. Assessments have been undertaken to determine this is the optimum height with respect to off-site impacts. For stack heights more than 70m, the dispersion and deposition factors do not decrease significantly.
499. The ADMS model takes into account the plume rise and the building wake. Thus, the real stack height entered is used to establish the optimum physical stack height. PC-CREAM98, which is used to assess the dose, does not take into account the plume rise and the building wake, thus the “effective” stack height is entered in this code instead of the real stack height. Therefore to account for the impacts of build wakes etc. the physical stack height is reduced using the 2/3 reduction (Simmonds et al 1995). This results in an “effective” stack height of 23.3m. This is generally considered to be a conservative method for estimating the effective stack height.
3.4.1.2. Atmospheric dispersion of the effluents 500. Two models were used to evaluate the consequences of releases of airborne radioactive effluents from the proposed Hinkley Point C facility. For the assessment of consequences to the local reference group in the vicinity of the site, PC-CREAM98 was used. For the assessment of consequences to the nearest Member State(s) the atmospheric dispersion model described in National Radiological Protection Board (NRPB)-R123 (Jones, 1981a), was used to determine the atmospheric concentrations and ground deposition values. The methodology used in PC-CREAM (Simmonds et al 1995) was then used to calculate the doses.
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501. Both models are based on Gaussian plume models. The long range model described in Jones (1981a) is adapted to account for changes in wind direction and atmospheric stability during the passage of the plume over extended distances.
Meteorological data 502. Meteorological data have been obtained from the UK Meteorological Office for the Hinkley Point site for the years 1999 to 2008. These data, used in the assessment of local impacts, are formatted for use in PC-CREAM and are presented in Table 3.7 and Figure 3.1. Figure 3.1 shows the direction from which the wind is blowing. The assessment methodology includes the local site specific meteorological data presented below and therefore impact of wind direction on the local reference group is implicitly included in the assessment.
Table 3.7 Hinkley Point meteorological data PC-CREAM format for years 1999 to 2008
Stability category A B C-rain* D-rain * E F/G C+rain ** D+rain ** Total Wind angle 345-15 0.00E+00 2.63E-04 2.32E-02 2.32E-02 8.01E-03 1.11E-03 1.21E-02 2.90E-02 9.7E-02 15-45 1.14E-05 4.00E-04 1.93E-02 2.79E-02 6.99E-03 1.20E-03 1.13E-02 2.90E-02 9.6E-02 45-75 1.14E-05 2.40E-04 1.56E-02 2.90E-02 4.73E-03 9.59E-04 1.23E-02 2.95E-02 9.2E-02 75-105 1.14E-05 4.80E-04 3.03E-02 4.94E-02 5.90E-03 1.12E-03 1.71E-02 4.17E-02 1.5E-01 105-135 0.00E+00 3.71E-03 7.17E-02 5.94E-02 8.96E-03 1.39E-03 2.31E-02 2.31E-02 1.9E-01 135-165 1.14E-05 2.60E-03 2.11E-02 8.56E-03 3.29E-03 1.14E-03 6.75E-03 3.38E-03 4.7E-02 165-195 0.00E+00 1.95E-03 1.76E-02 2.31E-03 2.06E-03 9.94E-04 5.52E-03 2.38E-03 3.3E-02 195-225 2.28E-05 1.56E-03 2.84E-02 7.66E-03 3.39E-03 9.14E-04 9.86E-03 5.49E-03 5.7E-02 225-255 0.00E+00 1.28E-03 3.52E-02 1.21E-02 3.72E-03 7.42E-04 1.44E-02 6.63E-03 7.4E-02 255-285 1.14E-05 9.82E-04 2.44E-02 6.41E-03 2.91E-03 8.34E-04 9.74E-03 3.85E-03 4.9E-02 285-315 0.00E+00 6.62E-04 2.14E-02 8.06E-03 4.36E-03 9.25E-04 1.01E-02 7.79E-03 5.3E-02 315-345 0.00E+00 4.57E-04 2.01E-02 1.05E-02 4.91E-03 9.94E-04 1.05E-02 1.47E-02 6.2E-02 Total 8.0E-05 1.5E-02 3.3E-01 2.4E-01 5.9E-02 1.2E-02 1.4E-01 2.0E-01 1.0E+00 * This includes the meteorological conditions which are in the C stability class and which have a precipitation rate equal to 0. It is the same for the D stability class. ** This includes the meteorological conditions which are in the C stability class and which have a precipitation rate greater than 0. It is the same for the D stability class.
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Figure 3.1 Windrose for Hinkley Point Site
503. The meteorological data for the long range model, based on a representative windrose for the UK for use in long range dispersion calculations (taken from Jones, 1981a), is presented in Table 3.8. The table presents the frequency with which each wind speed group occurs within the relevant sector. The wind blows into those sectors 10.74% and 4.24% of the time for French/Channel Islands and the Republic of Ireland respectively.
Table 3.8 Wind speed and frequency parameters used in long range model
Percentage frequency of time wind is at a given wind speed into the Speed (m s-1) relevant sector Representative Range French and Channel Islands* Republic of Ireland value <5 2.5 18% 41% 5-10 7.5 33% 50% 10-15 12.5 22% 9% >15 20 27% 0% Total 100% 100% * The same sector applies to both Channel Islands and French reference groups and is used above.
504. The probability of dry and wet weather stopping, used in the long range model are presented in Table 3.9.
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Table 3.9 Other meteorological parameters used in long range model
Parameters Value Probability of dry weather stopping (s-1) 4.6E-06 Probability of wet weather stopping (s-1) 4.6E-05 Fraction of time wet 9.0E-02
Surface roughness 505. The local model assumes a surface roughness of 0.3m over the land, based on typical agricultural terrain found in the region and 0.01m over the sea. The long range model used a single value to account for the passage of the plume of 0.1m as recommended by Jones 1981a.
3.4.1.3. Ground deposition and re-suspension 506. The atmospheric concentrations are modified to account for deposition of material from dry and wet deposition mechanisms. This is done automatically within PC-CREAM98 for the local reference group assessment. The long range model used the approach described in NRPB-R122 (Jones, 1981b). The key parameter values are described Table 3.10.
Table 3.10 Assumptions/parameters ground deposition
Parameters Value Tritium 5E-03 Carbon-14 0E+00 Deposition velocity (m s-1) Particulates 1E-03 Iodine-131 1E-02 Noble Gases 0E+00 Washout coefficient (s-1) 1E-04
507. Depletion of the plume by the process of dry deposition was determined in the long range model from interpolation of NRPB-R123 Table 7. A value of 0.71 was determined for the French and Channel Islands groups and a value of 0.65 for the Republic of Ireland representative group.
508. Re-suspension of material deposited onto the ground is automatically determined for the local reference group within PC-CREAM98 methodology (Mayall et al 1997)(16). The long range assessment model uses a re-suspension factor of 1E-06 m-1, taken from NRPB-W1 (2002).
3.4.1.4. Food chains, inhalation, external exposure 509. The following exposure pathways were considered in assessing the impact from airborne discharges:
Inhalation of activity, including inhalation of material re-suspended; External dose from activity in the air; External dose from activity deposited on the ground; and Ingestion of foods contaminated with activity.
(16) Mayall A, Cabianca T, Attwood C, Fayers C, Smith J, Penfold J, Steadman D, Martin G, Morris T, Simmonds JR (1997). PC-CREAM. Installing and using the PC system for assessing the radiological impact of routine releases. EUR 17791 EN, NRPB-SR296.
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510. The following foods were assessed:
Milk; Milk products; Cow meat; Offal; Sheep meat; Green vegetables; Root vegetables; and Fruit.
Local reference group food intake data 511. Two sources of information were used to produce the habit data and food intake used in the assessment of the local reference group: 2006 CEFAS report (2007)(17) and the NRPB-W41 (2003)(18). In order to perform the Hinkley Point site specific assessment, local data from the CEFAS report are preferred to national habit data whenever possible. Therefore, details of input data and calculations based on the CEFAS report are given below.
512. The 97.5th percentile and mean ingestion rates are based on all consumption data in the local habit survey data.
513. In the CEFAS report, there are a number of individuals who consume large quantities of locally grown foodstuffs. The assumption that food is 100% locally produced is therefore applied. This is a conservative assumption.
514. Consumption values calculated for the local reference group are based on the “Top Two”(19) approach using CEFAS data, are presented in Appendix 1.
515. For child and infant consumption parameters(20), essentially the same method is used to determine consumption rates. However, in cases where there are no instances of a food group being consumed by children and infants, relevant data for these two age groups is derived from adult consumption patterns.
516. The “Top Two” foodstuffs determined, are taken forward for use in the full calculations of dose from terrestrial food ingestion pathways. They are:
Adults: root vegetables and milk products. Children: milk and milk products. Infants: milk and milk products.
517. Average and 97.5th percentile ingestion rates for all foodstuffs and other parameters are presented in Table 3.11.
(17) Centre for Environment, Fisheries & Aquaculture Science (2007) Radiological Habits Survey Report for the Hinkley Point Area 2006 (18) National Radiological Protection Board (2003) Generalised Habit Data for Radiological Assessments (NRPB-W41) (19) It is recommended in NRPB-R271 (Robinson, 1994) that for general assessments, the consumption habits of a population should be modelled using the “Top Two” method. All of the foods are considered at critical rates (97.5th percentile). Those two foods that contribute the highest dose are kept at this critical rate and the rest of the foods are then assessed at mean rates. (20) For the purposes of the assessment children are assumed to be 10 years old and infants 1 year old.
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Table 3.11 Food intake data for the local reference group
Parameter Adult Child Infant Fraction of food locally produced (kg person-1 y-1) 1.0 Ingestion of green vegetables 33.4 23.3 4.8 * Ingestion of root vegetables 121.5 26.6 12.3 * Ingestion of fruit 16.2 6.6 7.5 Ingestion of sheep meat 7.7 3.1 0.9 Ingestion of cow meat 26.9 14.3(21) 6.0 Ingestion of milk 136.6 207.4 * 276.5 * Ingestion of milk products 36.3 27.2 * 27.2 * * values calculated using the ratio derived using NRPB –W41 data
Food intake data for Member State reference groups 518. Data from European Commission report RP153 were used in the assessment of dose to the reference groups in France and the Channel Islands. These data are derived from FAOSTAT (2003)(22). Data for the Republic of Ireland was based on the North/South Ireland Food Consumption Survey (Irish Universities Nutrition Alliance, 2001) for adults. Data for children and infants was not available, therefore this data was scaled on the ratio between adult:children and adult:infant based on the Channel Islands data and applied to the Republic of Ireland data.
Table 3.12 Food intake data for Member State reference groups (kg y-1)
Consumption Age Green Root Sheep Offal Cow Milk Fruit Milk* data group Veg* veg Meat (Cow) Meat products Adult 80 60 15 3 1 15 240 20 Channel + Child 35 50 15 1.5 1 10 240 15 Islands Infant 5 15 7.5 0.6 0.4 3 320 15 Adult 125 32.2 12 1.7 3.1 19 272 73.6 France Child 55 27 12 1 3 13 272 55 Infant 8 8 6 0 1 4 363 55 Adult 31 69 26 10 7 26 178 43 Republic of Child 13 57 26 5 7 17 178 33 Ireland Infant 2 17 13 2 3 5 237 33 * Green vegetables and milk consumed at higher than average (97.5th percentile) rates + Channel Islands data based on UK
Foodchain models 519. The PC-CREAM98 models, ASSESSOR and FARMLAND, were used to determine local reference group doses.
520. To calculate the doses in the long range model the same dose assessment methodology was used to determine consequences to the local reference group. Using PC-CREAM08(23), the FARMLAND model was run for a unit release of each radionuclide to calculate the activity concentration in food per unit deposition. These data are presented in Table 3.13.
(21) As there is only 1 child observation who consumes cow meat, the terms ‘mean’ and ‘97.5th percentile’ are not strictly correct in this case. However, these data are consistent with the NRPB-W41 value. (22) Food and Agricultural Organisations of the United Nations, Statistical Database-Food Balance Sheets 2003, HTTP://FAOSTAT.FAO.ORG/SITE/502/DEFAULT.ASPX (23) PC-CREAM08 was not available when the dose assessment for local groups begun in support of the planning process. For consistency with previously published information, the same values are presented.
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Table 3.13 Food transfer factors used in long range assessment (in Bq kg-1 per Bq m-2 s-1)
Radionuclide Green veg Root veg Fruit Sheep Offal Beef Milk Milk products Tritium* 100 100 100 87.5 87.5 87.5 113 113 Carbon-14* 533 533 533 800 800 800 267 267 Argon-41 0 0 0 0 0 0 0 0 Krypton-85 0 0 0 0 0 0 0 0 Xenon-131m 0 0 0 0 0 0 0 0 Xenon-133 0 0 0 0 0 0 0 0 Xenon-135 0 0 0 0 0 0 0 0 Iodine-131 41200 8610 29200 31700 24700 24700 58200 58200 Iodine-133 6180 46.6 588 679 1110 1110 3790 3790 Cobalt-58 91300 256 14800 1160 69300 693 2560 2560 Cobalt-60 115000 5100 24600 4340 292000 2920 3550 3550 Caesium-134 131000 119000 344000 1540000 793000 793000 159000 159000 Caesium-137 150000 133000 364000 1910000 915000 915000 179000 179000 * Tritium and carbon-14 are determined using a specific activity model, as described in Simmonds et al (1995). The unit for these radionuclides in the above table is Bq kg-1 per Bq m-3.
521. Doses from ingestion of contaminated foods are calculated the same way as in PC-CREAM98, summarised below:
522. [Eqn 1] Dose(ing)k CFfn IR fk DPUI kn f n Where: -1 Dose(ing)k = Dose from ingestion to age group k, in Sv y , summed across all foods and all radionuclides. -1 CFfn = Activity concentration of radionuclide, n, in food, f, in Bq kg -1 IRfk = Intake rate of food, f, for age group k in kg y -1 DPUIkn = Dose coefficient from radionuclide, n, for age group k, in Sv Bq
523. The concentration in food, CFfn, is the product of the food transfer factor and ground deposition rate.
Inhalation and exposure data 524. The most exposed members of the public for gaseous discharges are assumed to be a farming family living at one of the nearest dwellings. The adults are assumed to spend time each year outdoors, working on the land adjacent to the site, the children and infants also spend time outdoors. The habit data of the farming family are presented in Table 3.14.
Table 3.14 Farming family habit data
Parameter Adult Child Infant Source Cloud shielding factor 0.2 0.2 0.2 Simmonds et al 1995 Ground shielding factor 0.1 0.1 0.1 Simmonds et al 1995 Occupancy (h y-1) 8760 8760 8760 NRPB-W41 Fraction of time indoors 0.5 0.8 0.9 NRPB-W41 Breathing rate (m3 h-1) 1.12 0.64 0.22 NRPB-W41
525. Cloud shielding factor represents the shielding afforded by a masonry and tiled house from external irradiation of material in the routine gaseous discharge.
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526. Ground shielding factor represents the shielding afforded by a masonry house from external irradiation of material deposited on the ground from routine discharges
527. The PC-CREAM98 model, ASSESSOR, used in the local reference group assessment, calculates doses from inhalation (including re-suspension), external exposure from immersion in the plume and external exposure from deposited material.
528. The long range model uses a similar approach described below.
Inhalation dose 529. [Eqn 2] Dose(inh)k CAn Occk BRk DPUI kn n Where: Dose(inh)k = Dose from inhalation to age group k, in Sv y-1, summed across all radionuclides. CAn = Concentration of radionuclide, n, in air, in Bq m-3 Occk = Occupancy for age group k in hours per year BRk = Breathing rate for age group k in m3 h-1 DPUIkn = Dose per unit intake from radionuclide, n, for age group k, in Sv Bq-1
Resuspension dose 530. [Eqn 3] Dose(res)k CGn Rf Occk BRk DPUI kn n Where: -1 Dose(res)k = Dose from inhalation from re-suspension to age group k, in Sv y , summed across all radionuclides -2 CGn = Concentration of radionuclide, n, on the ground, in Bq m Rf = Re-suspension factor (constant) 10-6 m-1 Occk = Occupancy for age group k in hours per year 3 -1 BRk = Breathing rate for age group k in m h -1 DPUIkn = Dose per unit intake from radionuclide, n, for age group k, in Sv Bq
Immersion Dose 531. [Eqn 4] Dose(imm)k Findk Csf Occk Findk CAn DPUEn n
Where: -1 Dose(imm)k = Dose from immersion in the plume to age group k, in Sv y , summed across all radionuclides -3 CAn = Concentration of radionuclide, n, in air, in Bq m
Occk = Occupancy for age group, k, in hours per year for age group, k Csf =Cloud shielding factor Findk = Fraction of time indoors in hours per year for age group, k -1 -3 DPUEn = Dose per unit exposure from radionuclide, n, in Sv s per Bq m
External dose from deposited activity 532. [Eqn 5] Dose(ext)k Findk Gsf Occk Findk CGn DPUEn n
Where: Dose(imm)k = Dose from external exposure from activity deposited on the ground to age group k, in Sv y-1, summed across all radionuclides -2 CGn = Concentration of radionuclide, n, on the ground, in Bq m Occk = Occupancy for age group, k, in hours per year for age group, k Gsf =Ground shielding factor Findk = Fraction of time indoors in hours per year for age group, k -1 -2 DPUEn = Dose per unit exposure from radionuclide, n, in Sv s per Bq m
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Total committed effective dose 533. Total committed effective doses(24) for each age group are the summation of equations [1], [2], [3], [4] and [5] described above.
534. [Eqn 6] Dose(total)k Dose(ing) Dose(inh) Dose(res) Dose(imm) Dose(ext)
Where: -1 Dose(total)k = Dose through all pathways and radionuclides to age group, k, in Sv y .
3.4.1.5. Other parameters used in the calculation
Dose coefficients 535. Inhalation and ingestion dose coefficients are taken from International Commission on Radiological Protection (ICRP) Publication 72 (1996). The absorption type is also described. The specific values used are presented in Table 3.15. To account for the exposure pathway by which tritium is absorbed through the skin, the same approach has been adopted as used in the PC-CREAM methodology by applying a tritium skin multiplier to the inhalation dose coefficient; a value of 1.5, consistent with PC-CREAM08, was used.
Table 3.15 Inhalation and ingestion dose coefficients (Sv Bq-1)
Adult Child Infant Radionuclide Absorption Type(25) Inhalation Ingestion Inhalation Ingestion Inhalation Ingestion Tritium V 1.80E-11 1.80E-11 2.30E-11 2.30E-11 4.80E-11 4.80E-11 Carbon-14 V 5.80E-10 5.80E-10 7.90E-10 8.00E-10 1.60E-09 1.60E-09 Iodine-131 V 2.00E-08 2.20E-08 4.80E-08 5.20E-08 1.60E-07 1.80E-07 Iodine-133 V 4.00E-09 4.30E-09 9.70E-09 1.00E-08 4.10E-08 4.40E-08 Cobalt-58 M 1.60E-09 7.40E-10 2.40E-09 1.70E-09 6.50E-09 4.40E-09 Cobalt-60 M 1.00E-08 3.40E-09 1.50E-08 1.10E-08 3.40E-08 2.70E-08 Caesium-134 M 6.60E-09 1.90E-08 5.30E-09 1.40E-08 7.30E-09 1.60E-08 Caesium-137 F 4.60E-09 1.30E-08 3.70E-09 1.00E-08 5.40E-09 1.20E-08
536. Dose per unit exposure factors were taken from US EPA Federal Guidance Report 12 using the whole body effective dose coefficients (e) found in Table I.3(26). These are presented below in Table 3.16.
(24) For brevity, the term effective dose is used in place of committed effective dose. For the purposes of this document they are considered to be the same. (25) A term introduced in ICRP’s revised respiratory tract model (ICRP-66) to replace the clearance classes in ICRP-30. The absorption types are Type F for fast absorption, formerly called Class D; Type M for moderate absorption, formerly called Class W; and Type S for slow absorption, formerly called Class Y. the term V relates to vapour and very fast absorption. (26) http://www.ornl.gov/~wlj/fgr12tab.htm
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Table 3.16 External exposure factors (dose rate per unit activity concentration in air)
Cloud shine Ground shine Radionuclide (Sv s-1 per Bq m-3) (Sv s-1 per Bq m-2) Tritium 0 0 Carbon-14 2.60E-18 1.27E-20 Argon-41 6.14E-14 0 Krypton-85 2.40E-16 0 Xenon-131m 3.49E-16 0 Xenon-133 1.33E-15 0 Xenon-135 1.10E-14 0 Iodine-131 1.69E-14 3.64E-16 Iodine-133 2.76E-14 6.17E-16 Cobalt-58 4.44E-14 9.25E-16 Cobalt-60 1.19E-13 2.30E-15 Caesium-134 7.06E-14 1.48E-15 Caesium-137 2.70E-14 5.82E-16
3.4.2. Evaluation of activity concentration and exposure levels associated with discharge limits
3.4.2.1. Annual average concentrations of activity in the atmosphere near the ground and surface contamination levels for the most exposed areas
Methodology and environmental concentrations for local receptors 537. Air concentration and surface concentration values taken from the ADMS modelling are presented below in Table 3.17.
Table 3.17 Local environmental concentrations from Hinkley Point C discharges
Surface (27) Air concentration Radionuclides -3 concentration (Bq m ) (Bq m-2) Tritium 1.64E-02 3.40E+03 Cobalt-58 1.76E-07 1.38E-02 Cobalt-60 2.08E-07 1.63E-02 Iodine-131 9.50E-07 3.48E-01 Iodine-133 1.14E-06 4.17E-01 Caesium-134 1.62E-07 1.27E-02 Caesium-137 1.45E-07 1.14E-02
Methodology and environmental concentrations for other Member States 538. The air concentration and surface concentration (ground deposition) values were taken from the long range model described above and presented below in Table 3.18.
(27) Non-depositing radionuclides are not included
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Table 3.18 Environmental concentrations in other Member States from Hinkley Point C using the long range model
Channel Islands France Ireland Radionuclide Air Surface Air Surface Air Surface concentration concentration concentration concentration concentration concentration (Bq m-3) (Bq m-2) (Bq m-3) (Bq m-2) (Bq m-3) (Bq m-2) Tritium 4.72E-04 8.85E+01 3.20E-04 6.37E+01 2.04E-04 4.89E+01 Cobalt-58 3.41E-09 2.51E-04 3.25E-09 2.38E-04 1.35E-09 2.14E-04 Cobalt-60 4.03E-09 2.97E-04 3.85E-09 2.81E-04 1.59E-09 2.53E-04 Iodine-131 2.00E-08 7.15E-03 1.91E-08 6.81E-03 7.85E-09 3.48E-03 Iodine-133 8.03E-09 2.87E-03 7.29E-09 2.60E-03 2.08E-09 9.19E-04 Caesium-134 3.14E-09 2.31E-04 2.99E-09 2.19E-04 1.24E-09 1.97E-04 Caesium-137 2.81E-09 2.07E-04 2.68E-09 1.96E-04 1.11E-09 1.76E-04
3.4.2.2. Radiation doses
Local reference group 539. The annual committed effective dose to the local reference group (farming family), assuming CEFAS ingestion rates, is 2.4, 2.2 and 4.4μSv y-1 for adults, children and infants respectively.
540. A dose breakdown by exposure pathway is presented in Table 3.19 for the three age groups.
Table 3.19 Annual dose to the local reference group by exposure pathway (µSv y-1)
Age Inhalation Cloud Deposited Cloud Deposited group Inhalation Foodstuffs Total resusp. gamma gamma beta Beta Adult 1.3E-01 2.6E-06 3.2E-02 1.9E-02 1.7E-03 1.1E-03 2.2E+00 2.4E+00 Child 1.0E-01 3.3E-06 1.9E-02 9.4E-03 1.7E-03 4.4E-04 2.1E+00 2.2E+00 Infant 7.2E-02 3.7E-06 1.5E-02 6.4E-03 1.7E-03 2.2E-04 4.3E+00 4.4E+00
541. Infants within the reference group receive the greatest dose, dominated by ingestion of carbon-14 in milk (47%) and milk products (29%).
Radiation doses in other Member States 542. The annual effective doses to the reference group living in the Channel Islands are 9.6, 10 and 16nSv y-1 for adults, children and infants respectively.
543. A dose breakdown by exposure pathway is presented in Table 3.20 for the three age groups.
544. Infants receive the greatest dose, dominated by ingestion of carbon-14 in milk (66%).
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Table 3.20 Annual effective dose to reference group in the Channel Islands by exposure pathway (µSv y-1)
Age External group Inhalation Resuspension External cloud Foodstuffs Total ground Adult 1.2E-03 2.5E-07 1.7E-04 8.4E-05 8.1E-03 9.6E-03 Child 1.2E-03 2.4E-07 8.6E-05 3.4E-05 8.9E-03 1.0E-02 Infant 9.6E-04 1.9E-07 6.0E-05 1.7E-05 1.5E-02 1.6E-02
545. The annual effective doses to the reference group living in France are 9.0, 9.6 and 15nSv y-1 for adults, children and infants respectively.
546. A dose breakdown by exposure pathway is presented in Table 3.21 for the three age groups.
547. As was the case with receptors in the Channel Islands, infants receive the greatest dose, dominated by ingestion of carbon-14 in milk (66%).
Table 3.21 Annual effective dose to reference group in France by exposure pathway (µSv y-1)
Age External group Inhalation Resuspension External cloud Foodstuffs Total ground Adult 1.1E-03 2.3E-07 1.6E-04 7.9E-05 7.6E-03 9.0E-03 Child 1.1E-03 2.2E-07 8.1E-05 3.2E-05 8.3E-03 9.6E-03 Infant 9.0E-04 1.8E-07 5.6E-05 1.6E-05 1.4E-02 1.5E-02
548. The annual effective doses to the reference group living in Republic of Ireland are 3.7, 4.0 and 5.9nSv y-1 for adults, children and infants respectively.
549. A dose breakdown by exposure pathway is presented in Table 3.22 for the three age groups.
550. As was the case with the reference groups in the other groups, infants receive the greatest dose, dominated by ingestion of carbon-14 in milk (52%).
Table 3.22 Annual effective dose to reference group in Republic of Ireland by exposure pathway (µSv y-1)
Age External External group Inhalation Resuspension. Foodstuffs Total cloud ground Adult 4.9E-04 1.4E-07 6.4E-05 4.4E-05 3.1E-03 3.7E-03 Child 4.7E-04 1.3E-07 3.3E-05 1.8E-05 3.5E-03 4.0E-03 Infant 3.8E-04 1.1E-07 2.3E-05 8.8E-06 5.5E-03 5.9E-03
3.4.2.3. Evaluation of activity concentration and exposure levels against measured data and dose limits
Annual average concentrations of activity in the atmosphere near the ground 551. The installations listed in Section 1.1.3 and Section 3.5 are discharging in the atmosphere and are contributors, together with other installations not in the UK territory, of the overall atmospheric activity on the reference groups considered in Section 1.1.4.
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552. Environmental monitoring programmes are regularly recording the effect of all the aforementioned discharges, on a range of reference groups all around Europe. This section provides a comparison between the existing radiological background and the contribution expected from the Hinkley Point C discharges.
553. French authorities have developed a national network of measurements of radioactivity in the environment(28). A portal for accessing data in near real time was developed by the Authority for Nuclear Safety and the Institute of Radioprotection and Nuclear Safety.
554. The atmospheric monitoring station nearest to the French receptor was identified in the commune of Jobourg.
555. An evaluation between the modelled contribution from Hinkley Point C and the recorded concentrations recently measured in the area of the receptor show the current levels are from three to six orders of magnitude higher than the modelled contribution from Hinkley Point C (see Table 3.23 below).
Table 3.23 Comparison of modelled environmental air concentrations in France from Hinkley Point C discharges with monitored readings
Radionuclide Hinkley Point C contribution (Bq m-3) Background (Bq m-3) Tritium 3.20E-04 7.60E-01 Carbon-14 7.46E-05 5.00E-02 Krypton-85 5.14E-04 8.40E+01 Iodine-131 1.91E-08 <1.00E-03 Cobalt-60 3.85E-09 <1.00E-03 Caesium-134 2.99E-09 <1.00E-03 Caesium-137 2.68E-09 <1.00E-03
Effective doses to adults, children and infants, taking account of the existing background
556. Article 13 of the European Commission Basic Safety Standards (1996) (BSS) establishes a public dose limit of 1mSv y-1. The Environmental Permitting Regulations (England and Wales) 2010 enshrines the pertinent requirements of the BSS within English and Welsh law. The Environmental Permitting Regulations also describes the UK application of dose constraints in relation to the planning stage of radiation protection, by ensuring that the maximum doses to an individual from a single source do not exceed 300µSv y-1. The dose assessment results are lower than the constraint by a factor of nearly 70 for the local Hinkley Point C receptor and by at least four orders of magnitude for the reference groups in France, Republic of Ireland and the Channel Islands.
557. The contribution from the gaseous discharges at the proposed limits is well below the UK source constraint of 300µSv y-1. Even if it were assumed that the fishing family (marine reference group) and farming family (atmospheric reference group) shared the same habits, the level of exposure would still be considerably below the source constraint. The values are also below the public dose limit of 1mSv y-1 locally (and at the nearest Member State). The annual average radiation dose in the UK is 2.7mSv of which approximately 85% is from natural sources. Therefore doses to the UK reference groups are very small.
(28) http://www.mesure-radioactivite.fr/public/
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3.5. RADIOACTIVE DISCHARGES TO ATMOSPHERE FROM OTHER INSTALLATIONS 558. The current UK Strategy for Radioactive Discharges(29) sets out the national plan for adhering to the objectives of the OSPAR Convention with respect to radioactive discharges to the marine environment of the north-east Atlantic Ocean. In addition, the Strategy sets out the national plan for progressive reduction of aerial discharges and outlines the UK Government’s commitment to the progressive reduction of:
radioactive discharges and discharge limits; human exposure to ionising radiation arising from radioactive discharges; and concentrations of radionuclides in the marine environment resulting from radioactive discharges.
559. The Strategy acknowledges the need to take account of what is practicable and focuses on the delivery of the OSPAR Convention’s objectives through the application and use of BAT.
560. In the UK, new build reactors such as Hinkley Point C will be replacing the existing suite of reactors. With the relatively low discharges from the EPR units, the levels of discharges will progressively be reduced in comparison to current levels. The UK Government has considered that on the basis of the low levels of discharges from current Light Water Reactors (LWR) in the UK and abroad, such a programme, on a purely illustrative basis, would not prevent the UK from achieving the objective of the OSPAR Radioactive Substances Strategy.
561. Discharges to the environment from Hinkley Point C will be optimised in accordance with the application of BAT and thus prevented or kept as low as is practicable. Given the ongoing reduction of both liquid and gaseous discharges from the UK nuclear industry, due to the closing and decommissioning of facilities, there will be an overall downward trend in discharges. This is consistent with the national progressive reduction of discharges and in line with the current UK’s Strategy for Radioactive Discharges.
3.5.1. Source of aerial discharges 562. This sub-section illustrates the aerial discharges from other sources other than Hinkley Point C to the closest Member State for aerial discharges, France. It should be noted that the facilities listed are only the ones on UK territory excluding nuclear sites such as the power plants in Flamanville and the fuel reprocessing plant in La Hague, which are relatively close to the receptor considered.
563. Annual discharges from principle sources in the UK are reported annually in the RIFE Report. The table below reports the principal 2008 discharges from nuclear establishments as reported in the appendix of the 2008 RIFE Report (RIFE-14). The establishments listed in the table are the ones identified in Section 1.1.3 and summarised in the following Table 3.24. Figure 3.2 shows the location of these sites.
564. In view of the small additive effect of the discharges from the new installation on neighbouring Member States, in addition to the known effects of other discharges in the same area, no specific plans are needed for coordination of discharges between these installations.
(29) UK Strategy for Radioactive Discharges, Department of Energy and Climate Change, July 2009
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Table 3.24 Discharges of gaseous radioactive wastes
Discharges during 2008 Establishment Radioactivity (TBq) Alpha Nil Tritium 4.75 Winfrith (Inutec Ltd) Carbon-14 6.07 10-6 Other Nil Alpha 6 10-11 Winfrith Research Sites Restoration Ltd Tritium 0.048 (UKAEA) Carbon-14 9.28 10-4 Other 1.2 10-8 Beta 4.99 10-7 Hinkley Point A Tritium 0.113 Carbon-14 7.31 10-4 Tritium 1.48 Carbon-14 1.11 Sulphur-35 0.12 Hinkley Point B Argon-41 8.85 Cobalt-60 8.03 10-6 Iodine-131 7.03 10-6 Beta/gamma 1.57 10-8 Tritium 4.46 10-4 Devonport Royal Dockyard Ltd Carbon-14 2.49 10-4 Argon-41 10-5
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Figure 3.2 Other nuclear establishments contributing to air discharges
4°0'0"W 3°0'0"W 2°0'0"W
E%, Hinkley Point A and B
51°0'0"N - 51°0'0"N
Winfrith and Inutec Ltd %,
Devonport Royal Dockyard Ltd %,
50°0'0"N 50°0'0"N 4°0'0"W 3°0'0"W 2°0'0"W
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4. RELEASE FROM THE INSTALLATION OF LIQUID RADIOACTIVE EFFLUENTS IN NORMAL CONDITIONS
4.1. AUTHORISATION PROCEDURE IN FORCE
4.1.1. Description of the current procedure 565. The current procedure is described in Section 3.1.1 of this submission.
4.1.2. Waste discharge limits and associated requirements
4.1.2.1. Annual limits for liquid waste 566. The proposed limits for liquid discharges of radioactive waste from Hinkley Point C are presented in Table 4.1. These limits represent the maximum discharges that will be permitted over a rolling 12 month period. These proposed discharge limits have been determined from:
feedback on effluent discharges from EDF Group 1300MWe production units and the associated limit values, with consideration for operating margins covering standard contingencies relating to plant operations (unscheduled system drainage for maintenance, temporary faults with abatement plant, etc.); the discharges produced by the EPR in relation to the 1300MWe units and in particular the production of effluent that has no cost-effective abatement technique (tritium, carbon-14); the types of fuel envisaged for the EPR; and system design improvements reducing the amount of discharge which are presented in Section 2.3.5.
Table 4.1 Proposed limits for liquid radioactive discharges from Hinkley Point C
Proposed liquid discharge Category limit (GBq) Tritium 200,000 Carbon-14 190 Caesium-137 1.9 Other radionuclides 18.1
567. Operational experience has shown that discharges of actinides to the environment represent a fraction (0.2%) of the annual average exposure of a member of the public in the UK. Discharges could arise from the occasional failure of fuel cladding or from traces of uranium on the surface of the fuel resulting from the manufacturing process. However, the levels of activity discharged would be very low and difficult to detect, whilst other fission products are more readily detectable. Consequently a specific limit is unlikely to be set for the liquid discharge of alpha emitting radionuclides. Any discharges would be kept As Low As Reasonably Achievable (ALARA) using Best Available Techniques (BAT).
4.1.2.2. Conditions relating to the disposal of aqueous waste 568. The Environmental Permit will also apply a number of conditions to the disposal of liquid waste in addition to the limits. These are summarised thematically as:
Management. Management systems organisational structures and resources shall be provided to identify and minimise risks; to demonstrate compliance with the limitations and conditions of the permit; to make and retain records; to make the permit available to those having duties under it; and to appoint and consult with qualified expects on matters related to compliance.
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Operations. The scope and geographical extent of permitted activities is defined. BAT shall be applied to minimise the radioactivity in the waste, minimise the activity discharged, minimise the impacts from discharges and to remove entrained matter from any waste discharged. Optimisation techniques and disposal systems shall be maintained and monitored to demonstrate their continued effectiveness. Disposal. The types, routes and limits for disposals are identified. Sampling, measurements, tests, surveys and calculations shall be undertaken to demonstrate compliance with the limitations and conditions of the permit. BAT shall be selected. All defined techniques shall be appropriately commissioned, maintained and calibrated. MCERTS certification/accreditation is required for monitoring of disposals. Records of all monitoring shall be maintained and access to outlets for independent monitoring shall be provided. The Environment Agency may specify monitoring requirements. Information. Records to demonstrate compliance with the limitations and conditions of the permit shall be generated and retained on-site. Records of disposal shall be sent to the Environment Agency. An environment case to demonstrate that people and the environment are being protected shall be maintained throughout the activity’s lifecycle. Information shall be provided if requested by the Environment Agency. The Environment Agency shall be notified of any disposal of radioactive waste that is, has or could result in a breach of the limitations and conditions contained in the permit.
4.2. TECHNICAL ASPECTS 569. Feedback from the Fukushima event known at the time of writing reveals that there are no implications for the release of radioactive effluents from the Hinkley Point C installation in normal conditions.
4.2.1. Annual discharges foreseen 570. The robust application of the waste minimisation and waste management techniques presented in Section 4.2.3 will ensure routine annual discharges at levels below the annual limits proposed in Section 4.1.2. Operational feedback from EDF’s existing fleet of reactors has identified a number of factors that impact on the environmental performance of the plant and can result in elevated discharges. These factors include unplanned shutdowns and the failure of fuel within the reactor core. The proposed limits include contingencies that take account of these factors. The performance of the reactors and the associated waste management systems are carefully monitored by the operator and Regulators to ensure that any event which results in an increase in discharges is promptly identified and addressed.
4.2.2. Origins of the radioactive liquid effluents, their composition and physico- chemical forms 571. The main sources of liquid radioactive wastes are:
activated corrosion products, from corrosion in the steam generator pipes; activation products from chemicals in the primary coolant and secondary neutron sources; and volatile fission products, from small defects that may occur in the fuel cladding.
572. These different types of radioactive liquid effluent resulting from the production process are described in detail in Section 2.3. They are divided into four radionuclide categories: tritium, carbon-14, caesium-137 and other radionuclides.
573. The total activity relating to other radionuclides is divided between the various isotopes of other fission and activation products emitting beta or gamma radiation, based on an analysis of information on the actual discharge from EDF’s existing fleet. The presence of fission products such as isotopes of caesium is due to the occasional failure of fuel cladding.
574. It is expected that tritium will be discharged as tritiated water and will not be organically bound. Waste management systems and equipment are used to minimise the presence of particulate matter, gases and non-aqueous liquids in the aqueous discharges.
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Table 4.2 Distribution by activity of other radionuclides discharged in liquid form
Radionuclide Ratio (%) Silver-110m 5.70 Cobalt-58 20.70 Cobalt-60 30.00 Caesium-134 5.60 Caesium-137 9.45 Manganese-54 2.70 Antimony-124 4.90 Antimony-125 8.15 Nickel-63 9.60 Tellurium-123m 2.60 Other 0.60 Total fission/activation products 100
4.2.3. Management of the effluents, methods and paths of release
4.2.3.1. Waste minimisation techniques 575. Efforts have been focused at the design stage on reducing/minimising the production of effluents at source, taking account of operational feedback and experience from the experience of EDF and Areva. A number of recommendations from the Electric Power Research Institute have also been taken into account, such as the improvement of collection and effluent treatment systems.
576. The approach and the processes implemented, constitute BAT for the management of liquid radioactive effluents for Hinkley Point C. The reduction of the radiation exposure of the public from discharges (and other environmental impacts) to as low as reasonably achievable (optimisation) through the use of BAT, requires the balancing of a number of factors including worker exposures and financial costs.
577. The strategy for the reduction of liquid radioactive effluent at source using BAT is based on:
high integrity fuel cladding; the choice of materials, for example reducing stellites (cobalt based alloys) in the primary circuit will reduce the presence of the activation product cobalt-60; the reduction of the mass of beryllium in the secondary neutron sources, to reduce the generation of tritium by activation; the use of gadolinium oxide in some of the fuel pellets to reduce the use of boron in the primary coolant for reactivity control, thereby reducing the generation of tritium by activation; reinforced leak tightness requirements for active parts (pumps and valves) and the recovery of primary coolant leaks; the choice of a specific primary coolant chemistry control regime which minimises corrosion processes and the build-up of activated particulate material which contributes to the radioactivity present in effluent; optimal recycling of borated primary coolant; and enhanced coolant clean-up systems containing filters and demineralisers to remove any activation prone particulate matter.
578. In addition, operational and management controls add to optimisation via the rigorous management of effluent, which minimises the volumes and activities of waste at source and controls and optimises treatment and decisions regarding their discharge. These
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management procedures are used at all stages of plant start-up, operation at power, unit shutdown and refuelling.
4.2.3.2. Discharge of liquid radioactive waste from EPR units 579. The systems and equipment for treating liquid waste are presented in Section 2.3. These include filters, demineralisers and evaporators, which help to minimise the radioactivity discharged into the environment whilst also limiting the production of secondary solid waste. Following treatment, liquid radioactive effluent is directed to the LRMDS tanks, ExLWDS tanks or the SiteLWDS tanks for sampling and analysis prior to discharge.
580. The LRMDS tanks are successively filled with liquid effluent from the nuclear island and the SGBS. The SiteLWDS tanks receive effluent collected from the turbine hall, which is normally of very low activity. The ExLWDS tanks can collect any effluent from either the nuclear island or the secondary system, if needed. These tanks are only used in exceptional circumstances.
581. To ensure that the effluent is optimally diluted in the receiving environment to comply with chemical emission rate restrictions and to minimise the radiological impact on the environment and members of the public, it is pre-diluted with cooling water in the Unit 1 or Unit 2 pre-outfall pond.
582. This effluent is sampled and analysed to ensure compliance with the annual permitted limits for the four categories of radionuclides.
583. Liquid radioactive effluent can only be discharged after appropriate checks have been performed to ensure compliance with these limits. In addition, two LRMDS tanks or ExLWDS tanks cannot be discharged at the same time.
584. If, during the discharge of liquid radioactive waste from the LRMDS and ExLWDS tanks, the overall gamma activity concentration, which is measured continuously on the discharge line, exceeds the alarm threshold set at a pre-determined level, the discharge in progress will automatically stop.
585. If specific situations are observed during discharge, such as non-compliance with discharge conditions or the unavailability of monitoring devices on the discharge line, the discharge process is interrupted, the discharged activity is evaluated and analyses are carried out to determine the cause of the specific situation.
4.2.3.3. Discharge of liquid radioactive waste from supporting facilities 586. The Hinkley Point C site will also contain supporting facilities for the storage of spent fuel, the storage of solid radioactive waste and the laundering of work wear. These facilities will have liquid waste management systems that will discharge liquid radioactive waste to the environment via the site’s permitted disposal system.
587. Consideration has been given to discharges from these facilities and the contribution that they will make to the total discharges from Hinkley Point C. The conclusions of these considerations are that the overall contribution will be low and that the site’s total discharges will be dominated by those from the two EPR units. The key arguments developed to support these conclusions are:
Tritium. Tritium is generated in the primary coolant as a result of neutron activation from power generation and secondary neutron sources. Neither of these effects will take place in supporting facilities. Tritium generation from the spent fuel interim storage facility’s fuel pool will be negligible compared to tritium generated in the reactor. Some tritium may migrate into the pool through the fuel cladding. This would be at a considerably lower level than in the reactor buildings fuel pool because the fuel will be much cooler. Fuel pool water will be continuously recirculated and only discharged in exceptional circumstances.
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Carbon-14. Carbon-14 is generated in the primary coolant as a result of neutron activation and is not expected in significant quantities in the supporting facilities. Caesium-137. Caesium-137 is a fission product and as such it is an indicator of fuel integrity. It is relatively mobile in the fuel, highly soluble in the reactor cooling water and can be readily measured at very low levels of detection. It has a relatively long-half life so that as well as giving indications of fuel leaks in the primary circuit, it can also assist in providing indication of the longer term performance of the down-stream abatement systems. As such, the quantity of caesium-137 in liquid discharges is regarded as a good indicator of fuel and plant performance. The store will have a closed looped cooling and purification system which includes techniques to remove radioactivity from the pool water. This will result in very limited discharges of pool water into the site liquid effluent system. Other radionuclides. Although fuel in the store will be cooler than fuel stored in the spent fuel pools associated with units 1 and 2, some radioactivity will continue to enter pond water from fuel defects and from contamination present on the external surfaces of fuel. The store will have a closed looped cooling and purification system which includes techniques to remove radioactivity from the pool water. This will result in very limited discharges of pool water into the site liquid effluent system.
4.3. MONITORING OF DISCHARGES 588. Operators on nuclear facilities in England and Wales are required to demonstrate that they are applying BAT for sampling, measurement and analysis of discharges. The full details of the monitoring arrangements are not yet available for Hinkley Point C. Fully developed monitoring practices will be developed at a later stage to ensure consideration of the most relevant and best practices closer to the start of operations at Hinkley Point C. Monitoring arrangements will be fit for purpose and it is expected that the arrangements will be consistent with UK custom and practice.
589. It is expected that Hinkley Point C will have systems in place that are, as a minimum, of comparable standard to Sizewell B. This will ensure that operational practices in the UK are maintained, that consistency with Environment Agency guidance is achieved and the requirement to demonstrate the application of BAT is fulfilled.
590. Batch samples will be taken from liquid waste storage tanks and will undergo basic measurements and assessment on-site. This will be followed by more detailed analysis of samples collected by proportional samplers over a pre-defined period.
4.3.1. Monitoring of liquid radioactive effluent from the LRMDS tanks and ExLWDS storage tanks
4.3.1.1. Nature of checks 591. All samples taken from the LRMDS tanks and ExLWDS tanks are analysed in the site’s laboratory. A sampling device enables a representative sample to be taken of the waste to be discharged from the LRMDS tank or ExLWDS tank. This sample is obtained after the tank has been mixed. Tritium analysis is carried out by liquid scintillation and the sample's composition is determined by gamma spectrometry.
Analyses performed prior to discharge 592. Some of the radioactive characteristics of the effluent from the LRMDS tanks and ExLWDS tanks, need to be determined before it is discharged into the sea. At Sizewell B a representative one-off sample is taken from each tank and the following analyses are performed:
Measurement of tritium activity. Gamma spectrometry analysis.
593. These analyses enable determination of the composition of the effluent to be discharged and to establish whether discharge is permissible.
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594. Similar analyses as those described above for Sizewell B are anticipated for Hinkley Point C, noting that no limit is set for Sizewell B on the liquid discharges of carbon-14. BAT will be applied in the monitoring of the grouping of radionuclides that are proposed for the setting of discharge limits for Hinkley Point C. This will meet, at least, Sizewell B current standards.
Analyses performed after discharge 595. Additional radioactive characteristics of the liquid effluent from the LRMDS tanks and ExLWDS tanks are determined after the waste has been discharged from the tanks. These post-discharge checks are carried out on a representative sample of all of the discharges from the tanks over a predetermined period.
596. For each predetermined period, a representative sample of all of the discharge from the LRMDS tanks or ExLWDS tanks is prepared. At Sizewell B the following analyses are undertaken:
Tritium. Distillation followed by liquid scintillation. Beta emitting isotopes (excluding tritium). Evaporation to dryness followed by dissolution and liquid scintillation for low, medium and high energy beta emitters. Gamma isotopes including caesium-137 and cobalt-60. Undertaken weekly by gamma spectrometry. Nickel-63. Undertaken annually at the National Nuclear Laboratory by radiochemical separation. Strontium-90. Undertaken annually at the National Nuclear Laboratory. Americium-241. Undertaken quarterly and annually at the National Nuclear Laboratory by radiochemical separation followed by alpha spectrometry. Plutonium-239 and plutonium-240. Undertaken annually at the National Nuclear Laboratory.
597. Similar analyses as those described above for Sizewell B are anticipated for Hinkley Point C. It is noted that neither Sizewell B nor Hinkley Point C are limited on discharges of actinides. Routine discharge samples would not include an analysis for radionuclides such as americium-241 or plutonium-239. The period bulk sampling is used to confirm that additional scrutiny of these radionuclides is not required.
4.3.1.2. Discharge monitoring 598. Discharge from the LRMDS tanks and ExLWDS tanks, is monitored from a liquid waste monitoring and discharge system control room, by means of transmitted discharge rate measurements, level measurements and two alarm thresholds (low level and high level). For the LRMDS tanks, this monitoring is also carried out from the main control room, by means of transmitted level measurements for each tank and the associated ‘full tank’ and ‘empty tank’ alarms.
4.3.1.3. Calculation of discharged activity 599. A summary of the evaluation of discharged activity is presented below. The formulae representing the calculations used at Sizewell B are presented in Table 4.3. All elements are recorded in registers, which are sent to the Environment Agency as required by the site’s Environmental Permit.
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Table 4.3 Determination of activity discharged and activity concentrations for each radionuclide category – liquid radioactive waste
Discharged Discharged activity Activity concentration in the tank volume All During each During each Act.RN sample radionuclides discharge: discharge: Act.Con.RN =
Act.RN discharged = VR = Q x T Vsample Act.Con. x V RN R Special case of SiteLWDS tanks:
If the overall beta activity concentration measured is Over the year : lower than 10Bq l-1 and if the monthly representative
Act.RN year = sample analysis does not reveal significant activity, then only the discharged tritium activity is calculated.
Act.RN alldischarge Otherwise, the effluent contained in the SiteLWDS year tanks is treated and calculated in the same way as the radioactive effluent from the LRMDS and ExLWDS tanks.
Where: RN: radionuclide (category) Q: discharge rate in the LRMDS line (m3 s-1) Act.Con.: activity concentration (Bq m-3) Act.: activity (Bq) T: discharge time(s) 3 VR: volume discharged into the discharge pond (m ) V sample: volume of sample (m3)
Discharged activity 600. For each category of radionuclide, the activity discharged is determined as being the product of the activity concentration of that category of radionuclide, considered in terms of the volume of effluent discharged via the discharge line during the discharge period.
Activity concentration 601. The activity concentration for each category is determined as follows:
For tritium and carbon-14, from the activity measured in the respective sample taken before discharge and the volume of this sample. For caesium-137 and other radionuclides, from the spectrometric analysis results, based on the sample taken before discharge and the volume of this sample.
4.3.1.4. Detection limits for the measuring techniques used 602. The sampling equipment and analytical methods used at Hinkley Point C will be at least equivalent to counterparts used at the UK’s current PWR at Sizewell B. These equipment and methods have undergone an appraisal process to demonstrate that they represent BAT for industrial measurements. The appraisal includes an assessment of detection limits corresponding to each technique and consideration of how these detection limits may vary according to the operating modes of the EPR.
603. The following detection limits applied to Sizewell B are provided for information purposes.
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Table 4.4 Detection limits for measurements of liquid radioactive waste
Radionuclides Type of sample Measurement Detection limit Batch sample from LRMDS and Distillation and liquid Tritium 20Bq l-1 ExLWDS tanks scintillation Batch sample from LRMDS and High resolution gamma Caesium-137 6Bq l-1 ExLWDS tanks spectrometry Evaporation to dryness, Batch sample from LRMDS and dissolution of precipitate Other radionuclides 10Bq l-1 ExLWDS tanks and triple-channel scintillation analysis Cobalt-60 Weekly bulk sample Gamma spectrometry 10Bq l-1 Quarterly bulk sample Radiochemical separation 5Bq l-1 Americium-241 Annual bulk sample Radiochemical separation 0.2Bq l-1 Strontium-90 Annual bulk sample Radiochemical separation 2Bq l-1 Plutonium-239 and Annual bulk sample Radiochemical separation 0.2Bq l-1 Plutonium-240
4.3.1.5. Checks on equipment and measuring devices 604. The leak resistance of each of the liquid radioactive effluent storage tanks (LRMDS tanks and ExLWDS tanks) is subject to a periodic check. The discharge pipework for these tanks undergoes a complete inspection routinely to visually check the absence of leaks.
605. The correct operation of the devices and associated alarms located on the pipework (for flow rate and activity measurement) is checked routinely.
606. The various sampling devices and measuring equipment are maintained and checked routinely to ensure that they are in good working order. The measuring equipment is also calibrated routinely.
4.3.2. Monitoring of turbine hall drainage water
4.3.2.1. Nature of checks 607. At Sizewell B discharges of effluent from the SiteLWDS tanks are monitored using an on-line gamma detector. In the event that activity is detected or that the detector is not in service, a sample is taken for analysis in the laboratory. A sampling device enables a representative sample to be taken of the contents to be discharged from the SiteLWDS tank. This sample is obtained after the tank has been mixed. Tritium analysis is carried out on this sample by liquid scintillation and the overall beta activity measurement is taken using an alpha-beta counter.
608. A representative sample of all of the discharge from the tanks over a pre-determined period is also prepared. The composition of the samples is determined by gamma spectrometry.
Analyses performed prior to discharge 609. Some of the radioactive characteristics for the water drained from the conventional island liquid waste discharge tanks in the turbine halls, need to be checked before discharge into the sea. A representative one-off sample is taken from each tank and the following analyses are performed before discharge:
Measurement of tritium activity. Measurement of overall alpha/beta activity.
610. These analyses enable calculation of the overall alpha/beta and tritium activity concentrations and to therefore determine whether the effluent contained in the conventional
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island liquid waste discharge tanks can be discharged without the controls associated with the LRMDS tanks and ExLWDS tanks. Furthermore, the tritium analysis also enables assessment of the tritium activity discharged via this route and to therefore ensure that the annual tritium limits are adhered to.
611. If the overall activity concentration exceeds a predetermined value, the waste contained in the conventional island liquid waste discharge tanks is managed in exactly the same way as the LRMDS and ExLWDS effluents.
Analyses performed after discharge 612. The other radioactive characteristics for the water drained from the conventional island liquid waste discharge tanks in the turbine halls are determined after discharge from the tanks. The analysis is performed after the discharge because the effluent is not considered to be radioactive. These post-discharge assessments are carried out on a representative sample of all of the discharge from the tanks over a pre-determined period. The representative sample is checked for beta/gamma activity and tritium. It is also analysed by gamma spectrometry.
4.3.2.2. Inline (realtime) discharge monitoring 613. Discharge from the SiteLWDS tanks is monitored from a liquid waste monitoring and discharge system control room, by means of transmitted discharge rate measurements, level measurements and two alarm thresholds (low level and high level). This monitoring is also carried out from the main control room, by means of transmitted level measurements for each tank and the associated ‘full tank’ and ‘empty tank’ alarms.
4.3.2.3. Calculation of discharged activity 614. A summary of the evaluation of discharged activity and discharge rates is presented below. The formulae representing the calculations are presented in Table 4.3. All elements are recorded in registers, which are sent to the Environment Agency as required by the site’s Environmental Permit.
615. The tritium activity discharged is determined as being the product of the tritium activity concentration and the volume of effluent discharged via the discharge line during the discharge period. The tritium activity concentration is determined from the tritium activity measured in the sample taken before discharge and the volume of this sample.
616. If the overall beta activity measured before discharge is lower than a specified threshold and if the monthly analysis of the representative sample does not reveal significant activity, then the activity in the tank is considered to be zero and is therefore not calculated. Otherwise, the effluent contained in the SiteLWDS tank is inspected, discharged and calculated in the same way as the radioactive effluent from the LRMDS tanks or ExLWDS tanks.
4.3.3. Alarm levels and intervention actions 617. Instruments that continuously monitor the activity of a liquid effluent stream, feature alarms both for recording faults and for elevated levels of activity of gamma emitting radionuclides. It is standard practice for such instruments to have one or two ‘high’ alarm levels, with the lower alarm triggering a notification locally and to a control room, and potentially a higher alarm level linked to a plant actuation such as the closure of a liquid effluent isolation valve.
618. Conservative decision making will be applied and precautionary actions taken where appropriate, for example, termination of an effluent discharge during the investigation of the elevated reading. Such investigation will involve comparison of the elevated reading, with the activity in the sample taken from the tank prior to discharge.
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4.4. EVALUATION OF TRANSFER TO MAN 619. An assessment was undertaken to assess the consequences of a marine discharge of radioactive effluents under normal conditions at the closest Member State to Hinkley Point C. The closest reference group to the site, is approximately 200km to the northwest of Hinkley Point C on the east coast of the Republic of Ireland. For the purposes of this assessment it has been assumed that the reference group is likely to be a fishing family. The impacts in the vicinity of Hinkley Point C from marine discharges are also presented.
620. Using a standard modelling approach to ensure that accumulation in the marine environment is taken into account, it was assumed that discharges occur continuously for 50 years. The activity concentrations determined at the end of this period are then used as the basis for the dose assessment. This approach was used for both the local and long range assessment. Where possible data has been used in the assessment based on local factors.
621. All modelling and assessments have been undertaken using the annual limits that are proposed in Section 4.1.2.1.
4.4.1. Models and parameter values used to calculate the consequences of the releases 622. For the local reference group, the effect of water borne radioactive elements was calculated for Hinkley Point C using PC-CREAM98(30).
623. To assess the effects of liquid radioactive effluents from the site to the east of the Republic of Ireland, the modelling package devised by the Health Protection Agencies Radiation Protection Division PC-CREAM08 was used.
624. The PC-CREAM modelling package uses two steps to calculate water borne radioactive effects on an identified group. The first is known as DORIS and the second is known as ASSESSOR.
DORIS 625. DORIS is a marine dispersion model, based closely on the MARINA II model, which was developed based on the findings of research carried out by the EC (2003)(31).
626. The DORIS model is used to predict activity concentrations in sea water, sediments and marine biota for unit discharge rates from a specific site.
627. The activity concentrations from the DORIS model are then input to the ASSESSOR part of PC-CREAM which scales them by the discharge rates of interest (e.g. the proposed permitted limits) and combines them with habit data to calculate doses from ingestion of marine food, external exposure to beach sediments and inhalation of sea spray.
628. In the PC-CREAM08 version of DORIS, model documentation suggested that for discharges to the Irish Sea, the sediment distribution values for coastal waters (Kd values) for caesium and cobalt isotopes, should be different to the generic default values. This amendment was made in keeping with the advice presented in the model. All other parameters are as the generic default values for DORIS. The annual discharge rates used in DORIS are the same as those used in the application for the Environmental Permit application for Hinkley Point C.
629. The inputs to the DORIS model for the long range reference group are shown below in Table 4.5. These values are based on Kd Values used in PC-CREAM08 are shown in Table 4.6.
(30) PC-CREAM08 was not available when the dose assessment process for local groups was started in support of the planning process. To ensure consistency with work already undertaken and presented, the results from PC-CREAM98 for exposure in the vicinity of the facility are retained. (31) MARINA II. Update of the MARINA project on the radiological exposure of European Community from radioactivity in North European marine waters, Radiation Protection 132. EC, Luxembourg.
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Table 4.5 DORIS – discharge and output related parameters
Parameters Input Discharge Rates (GBq y-1) Radionuclide Tritium 200000 Carbon-14 190 Caesium-137 1.89 Silver-110m* 1.14 Manganese-54* 0.54 Antimony-124* 0.98 Antimony-125* 1.63 Tellurium-123m* 0.52 Chromium-51* 0.12 Cobalt-58* 4.14 Cobalt-60* 6.0 Caesium-134* 1.12 Nickel-63* 1.92 Iodine-131+ 0.1 Output Effective doses in the 50th year of
continuous release Output Compartments Irish Sea South Bristol Channel Celtic Sea * These other radionuclides are described in Table 4.2 and their respective proportions applied to the limit for other radionuclides presented in Table 4.1. + Based on the findings of the UK Environment Agency in their GDA process of the UK EPR reactors, it is not currently proposed to limit isotopes of iodine discharges in liquid because of the low activity discharged and low radiological impact. However for completeness it is included in the dose assessment.
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Table 4.6 DORIS – Sediment distribution coefficient (Kd) and concentration factors
Deep Kd Coastal Fish Crustaceans Molluscs Seaweed (Bq t-1 Kd(Bq t-1 concentration concentration concentration concentration Radionuclide per per factor (Bq t-1 factor (Bq t-1 factor (Bq t-1 factor (Bq t-1 Bq m-3) Bq m-3) per Bq m-3) per Bq m-3) per Bq m-3) per Bq m-3) Antimony-124 5.00E+02 1.00E+03 4.00E+02 2.50E+01 2.00E+01 2.00E+01 Antimony-125 5.00E+02 1.00E+03 4.00E+02 2.50E+01 2.00E+01 2.00E+01 Caesium-134 2.00E+03 2.30E+02 1.00E+02 3.00E+01 3.00E+01 5.00E+01 Caesium-137 2.00E+03 2.30E+02 1.00E+02 3.00E+01 3.00E+01 5.00E+01 Carbon-14 2.00E+03 2.00E+03 2.00E+04 2.00E+04 2.00E+04 1.00E+04 Chromium-51 5.00E+04 5.00E+04 2.00E+02 5.00E+02 8.00E+02 2.00E+03 Cobalt-58 1.00E+07 2.50E+03 1.00E+03 1.00E+04 5.00E+03 1.00E+04 Cobalt-60 1.00E+07 2.50E+03 1.00E+03 1.00E+04 5.00E+03 1.00E+04 Tritium 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Manganese-54 2.00E+08 2.00E+05 4.00E+02 5.00E+02 5.00E+04 6.00E+03 Nickel-63 1.00E+06 1.00E+05 1.00E+03 1.00E+03 2.00E+03 2.00E+03 Tellurium-125m 1.00E+03 1.00E+03 1.00E+03 1.00E+03 1.00E+03 1.00E+04 (Antimony-125) Tellurium-123m 1.00E+03 1.00E+03 1.00E+03 1.00E+03 1.00E+03 1.00E+04 Tellurium-123 1.00E+03 1.00E+03 1.00E+03 1.00E+03 1.00E+03 1.00E+04 (Tellurium-123m)
ASSESSOR 630. Once activity concentrations have been calculated in the environmental media using DORIS, they can be used in ASSESSOR, which calculates the effective dose.
631. ASSESSOR considers the following pathways for the marine environment.
Inhalation of radionuclides from sea spray. External gamma dose from radionuclides in sediment. External beta dose from radionuclides in sediment. External gamma dose from radionuclides in fishing gear. External beta dose from radionuclides in fishing gear. Consumption of radionuclides in seaweed, fish, molluscs and crustaceans.
632. Calculating the external beta and gamma exposures to radionuclides in contaminated fishing gear, is done using an empirical formulae developed by Hunt (1984) and the mean energy of the beta and gamma radiation (ICRP 1983). The same mean gamma energies are also used to calculate the external gamma dose above aquatic sediments.
633. Default dose coefficients from ASSESSOR, taken from ICRP (1996) were used in the assessment.
4.4.1.1. Habit data – Member State reference group (Republic of Ireland) 634. The consumption data used in ASSESSOR was taken from published data for ingestion rates of fish, crustaceans, molluscs and seaweed for adults and children in the Republic of Ireland reference group(32). These data were considered to be the best available as they are based on studies of habits along the coast of north-east of the Republic of Ireland and representative of coastal communities living in the Republic of Ireland.
(32) CEFAS, An assessment of aquatic radiation pathways in Ireland, 2008 Environmental Report RL 16/08, Clyne, Garrod, Jeffs and Jenkinson 2008
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635. All ingestion rates are assumed to be at the 97.5th percentile of the distribution of ingestion rates. Children are assumed to be aged 10 years and infants aged 1 year. It is assumed that all seafood has been caught in the regional compartment adjacent to the site (Irish Sea South).
636. Breathing rates are based on the generalised breathing rates presented in NRPB-W41.
637. It has been assumed that the activities carried out in inter-tidal areas, is a combination of the time spent undertaking the following activities: oyster farming; dog walking; walking; water sports; angling and winkle collecting. These activities have been observed at a variety of inter-tidal environments along the northeast Republic of Ireland coast. For the purposes of this assessment, the combination of these activities undertaken on sand, sand & stone and mud & sand, has been utilised to establish the number of hours per year spent in the inter- tidal zone.
638. As the activity is located in the inter-tidal zone it has been assumed for the purposes of modelling that they are located 1m from the sea, as a worst case.
Table 4.7 Habit data for the Republic of Ireland marine reference group
Inter-tidal Handling fishing gear, Fish Crustaceans Mollusc Seaweed activities catch and sediment (kg y-1) (kg y-1) (kg y-1) (kg y-1) (h y-1) (h y-1) Adult 42 18 35 0.5 1,970 4,100 Child 5 4.4 0 0 239 15 Infant 0 0 0 0 140 0
4.4.1.2. Habit data – local reference group 639. The habit data for the local reference group (see Table 4.8) is based on a fishing family living in the local vicinity as this group is considered to be the most exposed members of the public for liquid discharges. In the fishing family, the adults spend time fishing near the coast and the children and infants spend time playing on the coast.
640. The two sources of information of habit and food intake data are the CEFAS Radiological Habits Survey (2007)(33) and the NRPB (now Health Protection Agency – Radiation Protection Division) publication NRPB-W41(34). Local data from the CEFAS report are preferred to national habit data whenever possible and should lead to a more realistic impact assessment.
641. Occupancy times for the assessment of dose from external exposure for adults are based on the 97.5th percentile of the distribution of observed rates and are 1,560 hours per year for handling fishing gear and 400 hours per year for recreational beach activities resulting in a total of 1,960 hours per year to adults. Values for children and infants are 175 and 10 hours per year respectively.
642. The following pathways for the fishing family are considered for releases of liquid discharges to the marine environment.
External exposure to radiation from radioactivity in beach sediments. External exposure to radiation from handling contaminated fishing gear (adults only). Inhalation of sea spray when on the coast. Consumption of sea fish, crustacea and molluscs caught locally.
(33) CEFAS (2007) Radiological Habits Survey Report for the Hinkley Point Area 2006 (34) NRPB (2003) Generalised Habit Data for Radiological Assessments (NRPB-W41)
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Table 4.8 Habit data for local reference group
Intertidal Handling fishing gear, Fish Crustacea Molluscs Seaweed activities catch and sediment (kg y-1) ns (kg y-1) (kg y-1) (kg y-1) (h y-1) (h y-1) Adult 47.2 15.3 1.9 0 1,960 1,560 Child 11.8 1.5 0 0 175 0 Infant 2.6(35) 0 0 0 10 0
4.4.2. Evaluation of activity concentrations and external dose rates at the proposed discharge limits 643. Table 4.9 shows the activity concentrations in unfiltered seawater and sediment calculated by DORIS for coastal waters local to the site (the local model compartment), which are used in the assessment of dose to the local reference group.
Table 4.9 Activity concentrations calculated by DORIS for Bristol Channel (local compartment) (Bq kg-1)
Unfiltered Filtered Suspended Seabed Nuclides Fish Crustaceans Mollusc Seaweed seawater seawater sediment sediment Carbon-14 1.90E+00 1.90E+00 1.90E+00 9.50E-05 9.50E-05 9.48E-05 1.90E-01 1.22E+00 Cobalt-58 5.51E-05 5.51E-04 2.74E-04 6.04E-07 6.04E-07 5.51E-08 5.51E-01 1.60E-04 Cobalt-60 2.39E-04 2.39E-03 1.19E-03 2.63E-06 2.63E-06 2.39E-07 2.39E+00 1.35E-02 Chromium-51 1.07E-06 2.66E-06 4.27E-06 5.60E-09 5.60E-09 5.33E-09 2.66E-04 1.07E-05 Caesium-134 4.69E-05 1.41E-05 1.41E-05 4.69E-07 4.69E-07 4.69E-07 9.37E-04 1.40E-04 Caesium-137 9.36E-05 2.82E-05 2.82E-05 9.37E-07 9.37E-07 9.36E-07 1.87E-03 1.03E-03 Tritium 9.77E-02 9.77E-02 9.77E-02 9.77E-02 9.77E-02 9.77E-02 9.77E-02 1.20E+00 Manganese-54 7.45E-08 9.34E-08 9.34E-06 3.75E-08 3.75E-08 1.87E-10 3.74E-02 2.68E-03 Nickel-63 2.53E-04 2.53E-04 5.07E-04 5.07E-07 5.07E-07 2.53E-07 2.53E-01 2.40E-01 Antimony-124 5.13E-05 3.20E-06 2.56E-06 1.28E-07 1.28E-07 1.28E-07 6.40E-05 1.23E-05 Antimony-125 2.79E-04 1.74E-05 1.39E-05 6.98E-07 6.98E-07 6.96E-07 3.49E-04 9.19E-04 Tellurium-123m 5.20E-04 5.20E-04 5.20E-04 5.20E-07 5.20E-07 5.20E-07 5.20E-04 9.55E-04
644. Table 4.10 and Table 4.11 show the activity concentrations calculated by DORIS for the output compartments and for each pathway for the Member State reference group.
(35) As there is only 1 infant observation it should be noted that using the term ‘97.5th percentile’ is not strictly correct in this case.
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Table 4.10 Activity concentrations calculated by DORIS for Irish Sea South (Bq kg-1)
Unfiltered Filtered Suspended Seabed Nuclides Fish Crustaceans Mollusc Seaweed seawater seawater sediment sediment Carbon-14 1.99E-02 1.99E-02 1.99E-02 9.96E-07 9.96E-07 9.94E-07 1.99E-03 4.08E-02 Cobalt-58 7.20E-08 7.20E-07 3.61E-07 7.95E-10 7.95E-10 7.20E-11 7.20E-04 7.00E-07 Cobalt-60 1.58E-06 1.58E-05 7.92E-06 1.74E-08 1.74E-08 1.58E-09 1.58E-02 2.98E-04 Chromium-51 3.86E-10 9.67E-10 1.55E-09 2.03E-12 2.03E-12 1.93E-12 9.67E-08 1.30E-08 Caesium-134 2.96E-07 8.86E-08 8.86E-08 2.96E-09 2.96E-09 2.96E-09 5.90E-06 2.93E-06 Caesium-137 9.43E-07 2.84E-07 2.84E-07 9.45E-09 9.45E-09 9.43E-09 1.89E-05 3.44E-05 Tritium 9.20E-04 9.20E-04 9.20E-04 9.20E-04 9.20E-04 9.20E-04 9.20E-04 3.76E-02 Manganese-54 4.50E-11 5.62E-11 5.62E-09 2.26E-11 2.26E-11 1.12E-13 2.25E-05 5.37E-06 Nickel-63 8.03E-07 8.03E-07 1.61E-06 1.61E-09 1.61E-09 8.03E-10 8.03E-04 2.25E-03 Antimony-124 5.91E-08 3.69E-09 2.96E-09 1.48E-10 1.48E-10 1.48E-10 7.39E-08 4.75E-08 Antimony-125 1.87E-06 1.17E-07 9.34E-08 4.66E-09 4.66E-09 4.66E-09 2.33E-06 2.05E-05 Tellurium-123m 4.71E-06 4.71E-06 4.71E-06 4.71E-09 4.71E-09 4.71E-09 4.71E-06 2.17E-05
Table 4.11 Activity concentrations calculated by DORIS for Celtic Sea (Bq kg-1)
Unfiltered Filtered Suspended Seabed Nuclides Fish Crustaceans Mollusc Seaweed seawater seawater sediment sediment Carbon-14 2.11E-02 2.11E-02 2.11E-02 1.06E-06 1.06E-06 1.05E-06 2.11E-03 4.43E-03 Cobalt-58 3.22E-07 3.22E-06 1.61E-06 3.54E-09 3.54E-09 3.22E-10 3.22E-03 3.13E-07 Cobalt-60 2.07E-06 2.07E-05 1.03E-05 2.27E-08 2.27E-08 2.07E-09 2.07E-02 3.90E-05 Chromium-51 4.40E-09 1.10E-08 1.76E-08 2.32E-11 2.32E-11 2.21E-11 1.10E-06 1.48E-08 Caesium-134 4.18E-07 1.25E-07 1.25E-07 4.18E-09 4.18E-09 4.18E-09 8.34E-06 4.14E-07 Caesium-137 1.00E-06 3.01E-07 3.01E-07 1.00E-08 1.00E-08 1.00E-08 2.00E-05 3.65E-06 Tritium 1.00E-03 1.00E-03 1.00E-03 1.00E-03 1.00E-03 1.00E-03 1.00E-03 4.11E-03 Manganese-54 4.41E-10 5.51E-10 5.51E-08 2.21E-10 2.21E-10 1.10E-12 2.20E-04 5.26E-06 Nickel-63 2.40E-06 2.40E-06 4.82E-06 4.82E-09 4.82E-09 2.40E-09 2.40E-03 7.45E-04 Antimony-124 2.91E-07 1.82E-08 1.46E-08 7.29E-10 7.29E-10 7.28E-10 3.65E-07 2.34E-08 Antimony-125 2.54E-06 1.59E-07 1.27E-07 6.37E-09 6.37E-09 6.37E-09 3.19E-06 2.80E-06 Tellurium-123m 5.53E-06 5.53E-06 5.53E-06 5.54E-09 5.54E-09 5.53E-09 5.53E-06 2.93E-06
4.4.2.1. Local reference group dose results 645. The committed effective dose for adults, children and infants has been calculated for marine discharges at the local reference group, a local fishing family, presented in Table 4.12.
Table 4.12 Effective doses to the local reference group from liquid discharges at the proposed limits
External marine pathways Internal marine pathways Total marine pathways dose dose dose (µSv y-1) (µSv y-1) (µSv y-1) Adult 1.5E-03 1.1E+00 1.1E+00 Child 1.4E-04 3.1E-01 3.1E-01 Infant 7.8E-06 1.2E-02 1.2E-02
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646. Of these doses, adult members of the fishing family receive the highest dose. The dose to the fishing family (all age groups) from liquid discharges is dominated by ingestion pathways which account for almost 100% of the dose. Carbon-14 is the dominant radionuclide accounting for almost 100% of the dose.
647. Article 13 of the European Commission BSS (1996) establishes a public dose limit of 1mSv y-1. The Environmental Permitting Regulations (England and Wales) 2010 enshrines the pertinent requirements of the BSS within English law. The Environmental Permitting Regulations also describes the UK application of dose constraints in relation to the planning stage of radiation protection, by ensuring that the maximum doses to an individual from a single source does not exceed 300µSv y-1. The contribution from the liquid discharges at the proposed limits is well below the UK source constraint of 300µSv y-1. Even if it were assumed that the fishing family (marine reference group) and farming family (atmospheric reference group) shared the same habits the level of exposure would still be considerably below the source constraint. The values are also below the public dose limit of 1mSv y-1 locally (and at the nearest Member State). The annual average radiation dose in the UK is 2.7mSv of which 85% is from natural sources. Therefore the critical group dose from Hinkley Point C discharges is less than the average dose from natural background.
4.4.2.2. Long range reference group dose results 648. The effective dose for adults, children and infants has been calculated for liquid discharges at the closest Member State, the Republic of Ireland and are presented in Table 4.13.
Table 4.13 Effective doses to the nearest Member State reference group from liquid discharges at the proposed limits
External marine pathways Internal marine pathways Total marine pathways dose dose dose (µSv y-1) (µSv y-1) (µSv y-1) Adult 7.0E-03 1.0E-01 1.1E-01 Child 8.0E-04 1.4E-02 1.5E-02 Infant 4.7E-04 6.3E-10 4.7E-04
649. The dose to the long range reference group is also dominated by the dose from ingestion pathways. This is mainly from the ingestion of fish; 41% for adults and 50% for children. The dose is dominated by the dose from carbon-14, for example 94% of the dose to adults and children is from this radionuclide, with the rest of the dose (6%) from cobalt-60.
650. The assessment indicates that the main exposure pathway for infants is from the external dose pathway. This is mainly from radiation exposure from radioactivity in beach sediments, with most of the dose from gamma radiation. This result is consistent with the results of CEFAS habit surveys which indicate that infants do not generally consume seafood. For infants, 90% of the dose is from cobalt-60, with the rest of the dose coming from a mix of radionuclides.
4.5. RADIOACTIVE DISCHARGES INTO THE SAME RECEIVING WATERS FROM OTHER INSTALLATIONS 651. The UK has committed to the objectives of the OSPAR Convention 1992 and the OSPAR Commission’s Radioactive Substances Strategy(36).
652. A key interim objective of the OSPAR Radioactive Substances Strategy is that by the year 2020 discharges of radioactive substances are reduced to levels where the additional concentrations in the marine environment above historic levels, resulting from such discharges, are close to zero.
(36) 2003 Strategies of the OSPAR Commission for the Protection of the Marine Environment of the North-East Atlantic, Radioactive Substances Strategy, (Reference number: 2003-21)
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653. The current UK Strategy for Radioactive Discharges(37) sets out the national plan for adhering to the objectives of the OSPAR Convention and outlines the UK Government’s commitment to the progressive reduction of:
radioactive discharges and discharge limits; human exposure to ionising radiation arising from radioactive discharges; and concentrations of radionuclides in the marine environment resulting from radioactive discharges.
654. The UK Strategy acknowledges the need to take account of what is practicable and focuses on the delivery of the Convention’s objectives through the application and use of BAT.
655. In the UK, new reactors such as Hinkley Point C will be replacing the existing suite of reactors. With no current proposals for reprocessing of spent fuel from the new reactors and the relatively low discharges from the EPR units, the levels of discharges will progressively be reduced in comparison to current levels. The UK Government has considered that on the basis of the low levels of discharges from current LWRs in the UK and abroad, such a programme on a purely illustrative basis, would not prevent the UK from achieving the objective of the OSPAR Radioactive Substances Strategy.
656. Discharges to the environment from Hinkley Point C will be optimised in accordance with the application of BAT and thus kept as low as is practicable. Given the ongoing reduction of liquid and gaseous discharges from the UK nuclear industry, due to the closing and decommissioning of facilities, there will be an overall downward trend in discharges. This is consistent with the national progressive reduction of discharges and in line with the current UK Strategy for Radioactive Discharges.
657. Annual discharges from principle sources in the UK are reported annually in the RIFE reports. Table 4.14 reports the principal 2008 marine discharges from nuclear establishments as reported in the appendix of the 2008 RIFE Report (RIFE - 14). The establishments listed in the table are the ones identified in Section 1.1.3. Figure 4.1 shows the location of these sites.
658. In view of the small additive effect of the discharges from the new installation on neighbouring Member States, in addition to the known effects of other discharges in the same area, no specific plans are needed for coordination of discharges between these installations.
659. The Radiological Protection Institute of Ireland carries out a comprehensive environmental monitoring programme and reports its results in a yearly report(38). Cahore, in County Wexford, is the nearest sampling location (about 40km to the north) to the location of the reference group.
(37) UK Strategy for Radioactive Discharges, Department of Energy and Climate Change, July 2009 (38) RPII (2008), Radioactivity Monitoring of the Irish Environment 2007, RPII-08/02
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Table 4.14 Discharges of liquid radioactive wastes from other nuclear sites
Discharges during 2008 Establishment Radioactivity (TBq) Tritium 0.294 Hinkley Point A Station Caesium-137 0.11 Other radionuclides 0.37 Tritium 77.9 Sulphur-35 0.15 Hinkley Point B Station Cobalt-60 2.28 10-4 Caesium-137 0.00419 Other radionuclides 0.00354 Tritium 3.0 Wylfa Other radionuclides 0.012 Tritium 14.4 Carbon-14 0.0648 Cardiff (GE Healthcare) Phosphorus-32/33 3.7 10-8 Iodine-125 7.4 10-7 Others Nil Tritium 4.89 10-5 Berkeley Caesium-137 3.75 10-4 Other radionuclides 0.00125 Tritium 0.184 Oldbury Caesium-137 0.309 Other radionuclides 0.127
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Figure 4.1 Other nuclear sites contributing to marine discharges
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5. DISPOSAL OF SOLID RADIOACTIVE WASTE FROM THE INSTALLATION 660. This chapter addresses the management and disposal of solid radioactive waste produced by the operation and eventual decommissioning of the EPR reactors at Hinkley Point C. The solid wastes that will arise are:
solid waste arisings (excluding fuel) during operations (process and operational); decommissioning solid waste; solid waste generated during interim storage of spent fuel after reactor decommissioning; and spent fuel if it has been classified as waste.
5.1. SOLID RADIOACTIVE WASTE
5.1.1. Categories of solid radioactive waste including spent fuel and estimated amounts
5.1.1.1. Categories of solid radioactive waste 661. Any waste material contaminated with or incorporating radioactivity above certain thresholds defined in legislation and for which no further use is envisaged, is designated as radioactive waste.
662. A description of the solid radioactive waste categories used in the UK, that are applicable to Hinkley Point C are presented in Table 5.1.
663. Spent fuel from new nuclear power stations is not categorised as waste because it still contains uranium and plutonium which could potentially be separated out through reprocessing and used to make new fuel.
664. The 2008 Government White Paper, Meeting the Energy Challenge A White Paper on Nuclear Power(39), concluded that in the absence of any proposals from the industry, any new nuclear power stations that might be built in the UK, should proceed on the basis that spent fuel would not be reprocessed and that plans for, and financing of, waste management should proceed on this basis. This is the approach being followed at Hinkley Point C. A description of spent fuel is therefore also set out in Table 5.1 below.
(39) Meeting the Energy Challenge, A White Paper on Nuclear Power, Cm7296, January 2008, Department for Business Enterprise and Regulatory Reform
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Table 5.1 Categories of solid radioactive waste
Waste type Description Spent fuel Nuclear fuel that has been irradiated and permanently removed from a reactor core. Spent fuel from currently operating nuclear power stations is not categorised as waste, because it still contains uranium and plutonium which could potentially be separated out through reprocessing and used to make new fuel, however it is not expected that the fuel from Hinkley Point C will be reprocessed. Due to the long half-life of the nuclides contained within spent fuel and the associated high levels of radioactivity, the management of spent fuel is a key issue for the design of nuclear power stations. High Level Waste Waste containing high concentrations of alpha/ beta/gamma emitting radionuclides. (HLW) HLW only arises from nuclear fuel reprocessing operations and therefore would not be generated during operations at Hinkley Point C. HLW generated during reprocessing of spent fuel requires remote handling (due to the radiation levels) and cooling (due to the heat produced) for many years. In the UK, HLW is defined as waste in which the temperature may rise significantly as energy is released by radioactive decay, so this factor has to be taken into account in designing storage or disposal facilities. ILW Radioactivity levels exceeding the upper boundaries for low-level wastes, but which do not require heating to be taken into account in the design of storage or disposal facilities. LLW Radioactive waste having a radioactive content not exceeding four giga Becquerels per tonne (GBq te-1) of alpha or 12 GBq te-1 of beta/gamma activity. A sub-set of this category is VLLW - waste with maximum concentrations of 4MBq te-1 of total activity may be disposed of to specified landfill sites. For waste containing tritium, the concentration limit for tritium is 40MBq te-1. Some radioactive substances may be classified as exempt from regulation (released from the requirements of the BSS Directive) if they comply with clearance levels set by the Environment Agency.
5.1.1.2. Solid LLW arising during operations
Solid LLW generated during operations 665. Table 5.2 provides the annual estimated production of raw (untreated) LLW for two EPR units. The volume and activity of LLW requiring disposal from Hinkley Point C, will be minimised by the use of the Waste Hierarchy and the application of BAT.
666. Two broad categories of LLW will be generated from the operation of the Hinkley Point C reactors and auxiliary facilities:
LLW generated through operation of systems and processes used to ensure safe operation of the power station, or to minimise discharges of radioactivity to the environment (process wastes); and LLW generated during maintenance and refuelling operations (dry active waste (DAW)).
667. The types of LLW that will be generated at Hinkley Point C are described below.
668. SGBS ion exchange resins. Ion exchange beds are utilised in the SGBS to control the chemistry of the secondary circuit and to address potential leakages of activation and fission products from the primary coolant circuit. In recycling the SGBS blowdown water from the EPR secondary circuit, the blowdown water is purified by the use of two parallel filters for the removal of suspended solids and two parallel demineralisation lines which use ion exchange resins to perform the demineralisation. The resins that comprise this waste stream consist of balls or grains (diameter ranges between 0.3-1.2mm) of organic resins with polystyrenic, phenolic, acrylic or formophenolic skeleton (cationic resins strongly acid, anionic resins strongly basic and mixed bed).
669. Wet sludge (sumps, tanks). During the operation of the Hinkley Point C EPR units, particulates will settle as sludges in various buffer and storage tanks associated with the auxiliary water circuits (e.g. LWPS, LRMDS). These are variously contaminated with a range
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of fission and activated corrosion products. This sludge is periodically cleaned out and removed for treatment prior to disposal. The physical form of this waste stream is described as a sludge consisting of settled metal oxide particulate.
670. Evaporator concentrates. The EPR proposes to make use of evaporation for the minimisation of radioactive liquid effluents arising from the LWPS. Evaporation of effluents results in the production of a sludge-like concentrate that will contain the bulk of the radioactivity initially present in aqueous effluent streams as activated metal oxides. The concentrates from PWR units that utilise evaporation typically contain on average 40,000ppm of boron, with a total salinity of 300g l-1 and are liable to crystallise if concentrated further. However, it is expected that the concentrates from Hinkley Point C will contain much lower levels of boron.
671. Water filters from effluent treatment. Corrosion occurring in the steam generators results in particulate corrosion products becoming mobile in the primary circuit, which can then become activated on passing through the reactor core. Filters are used to capture and hence minimise such particulate material in the water auxiliary circuits. Spent filter cartridges arise from the treatment lines of a number of water auxiliary circuits including the CVCS, LWPS, and the FPC(P)S. Water filters are withdrawn from operation on the basis of clogging and/or dose rate and then treated as waste. The physical form of this waste stream consists of filter cartridges that are composed principally of stainless steel supports with glass fibre filter media and some organic materials. The amount of particulate radioactive material (metallic oxides) trapped on each filter can vary. A small percentage of these filters are anticipated to be LLW at the time of generation.
672. Air and water filters. In addition to the water filters serving the primary system described above, LLW water filters will be generated from filtration of low active effluent (LWPS, SGBS). The low active effluent water filters are composed principally of stainless style supports with a filter medium and some organic materials. The amount of particulate radioactive material (metallic oxides) trapped on each filter can vary. All radiation controlled areas of the of the NAB, fuel building, safeguard buildings, reactor building, operational service centre, access building and ETB are served by dedicated ventilation systems. The extract from these systems is subject to a number of airborne activity abatement techniques, including the use of HEPA filtration, before discharge to the environment. The HEPA filters remove particulate material, mainly to ensure doses to workers are As Low As Reasonably Practicable (ALARP), but also ensures that the doses to members of the public associated with discharges of particulates to atmosphere are minimised. The abatement systems will therefore produce a number of spent HEPA filters over the course of reactor operations.
673. Dry active wastes. DAW corresponds to pre-compacted operational wastes and other DAW. The waste is generated through routine and maintenance operations in the nuclear island and will consist of contaminated personal protection equipment, monitoring swabs, plastic, clothing, contaminated tools, segregated pieces of metal, glassware, rubble and other process consumables. Mainly arising during outages these wastes are collected into plastic bags. The bags are put into an appropriate 200 litre drum or container. Small quantities of higher dose rate DAW will arise from routine and maintenance operations on the EPR units. Wastes of higher dose rate that are categorised as LLW will be safely managed, packaged and where possible, sent for off-site disposal at an appropriate facility.
674. Oils and solvents. Oils used in the reactor coolant pumps and in the lubrication of various components, such as process pumps, have the potential to become radiologically contaminated during normal service. Contaminated liquids such as chemical cleaning solutions, used liquid scintillation cocktails from laboratory sampling and solvents used as decontamination agents, also arise and will be included within this waste stream where appropriate to do so.
675. Metal scraps and other metallic wastes (Dose rate <2mSv h-1). During maintenance operations a variety of metals wastes can be generated, arising from the replacement of engineering components. The redundant metal components or equipment used may become contaminated and require disposal as radioactive waste.
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Solid LLW volume estimates 676. The raw waste volume estimate is based on a review of the waste arisings from existing French nuclear reactors of similar power rating to the EPR. It is assumed at present that Hinkley Point C, with two units, will produce double the arisings predicted for one unit, even though some facilities will be shared. The sharing of facilities, such as the waste treatment facilities, may result in some reduction of operational arisings. However, at this stage it is not possible to make accurate predictions of reductions, so the data in Table 5.2 are considered to present an upper estimate of solid LLW arisings.
677. The first EPR unit at Hinkley Point C is not expected to be fully operational until 2018. In the preceding 6 months, commissioning of the active systems will take place. This will result in the generation of small volumes of solid waste being generated. It is expected that these will fall well within the envelope of the expected arisings during full operations.
678. During initial active operations, particularly during the first cycles of operation it is anticipated that the solid waste arisings will increase beyond those generated during commissioning, but will still be below the levels anticipated when all systems and processes are operating at capacity. Consequently LLW transfers and disposals would be expected to be well within the proposed limits.
679. It will take some time for the processes that result in the generation of wastes to start in full and for waste volumes generated to be sufficient to make up full shipments for disposal. Waste will initially be stored in the buffer storage area until the volume required for transfer is reached.
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Table 5.2 Estimated annual volumes of solid low level radioactive wastes produced during operation of two EPR units at Hinkley Point C
Estimated raw waste volume Waste type Preferred waste arrangement annual (m3) Package as required to meet Conditions for Acceptance and SGBS ion exchange resins 15 transfer for disposal as VLLW Condition/package as required to meet Conditions for Wet sludge (from sumps, tanks) 1 Acceptance and transfer for disposal to the national Low Level Waste Repository (LLWR) Condition/package as required to meet Conditions for LLW cartridge filters from auxiliary circuit treatment 0.10 Acceptance and transfer for disposal to LLWR Condition/package as required to meet Conditions for Evaporator concentrates 6 Acceptance and transfer for disposal to LLWR Transfer for incineration (water filters) Air and water filters 8 Transfer for high force compaction (air filters) and onward disposal to LLWR Non-combustible 25 Transfer for high force compaction and onward disposal to LLWR DAW (excluding metals) Combustible 75 Package and transfer for off-site incineration Waste oils and solvents 4 Package and transfer for off-site incineration Metal scraps and metallic waste 12 Package and transfer for metals treatment
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5.1.1.3. Solid ILW arisings during operations
Solid ILW generated during operations
680. Routine operation of the Hinkley Point C reactors and their associated auxiliary systems will generate ILW. The majority of ILW will arise from the treatment of liquids and gases in order to reduce worker doses and discharges of radioactivity to the environment e.g. ion exchange resins.
681. In addition to the process wastes, a variety of ILW streams may be generated as a result of maintenance work carried out during reactor operation and work performed during reactor outages.
682. The ILW streams that are anticipated to arise from normal operation and maintenance of the two units at Hinkley Point C are described below.
683. Ion exchange resins. Ion exchange beds are used to capture and minimise soluble radioactive material. This material results from corrosion in the primary circuit (mainly in the steam generators and activation of chemicals in the primary circuit) and also in the CVCS and FPC(P)S. The ion exchange resins in the beds are periodically changed to optimise their performance. Additional volumes of ILW ion exchange resins may arise from the maintenance of water quality and the abatement of liquid discharges from the reactor spent fuel pool and spent fuel interim storage facility.
684. Spent cartridge filters. This waste consists of filters used in the clean-up of primary circuit water and water from the LWPS and FPC(P)S. There are several designs of filters depending on the abatement required. A proportion of the filters generated, would fall into the ILW category at the time of generation. After a period of decay storage it may be possible to reclassify a proportion of these wastes as LLW.
685. Wet sludges. During operation, particulates will settle as sludges in storage tanks associated with the auxiliary water circuits e.g. LWPS. These are variously contaminated with a range of fission and activated corrosion products. This sludge will be periodically cleaned out and removed for treatment prior to disposal.
686. Dry active waste (>2mSv h-1). This comprises a range of materials, including activated core components, contaminated metal, plastics, cloth, glassware and rubble, arising from operations during planned shutdown periods. This type of waste will include activated core components, such as those associated with reactor control rods. Some activated components generated during maintenance operations may be temporarily placed into the reactor fuel pools to allow for a period of radioactive decay until decommissioning and would be treated as a decommissioning waste in order to minimise dose to workers. After a period of decay storage it may be possible to reclassify a proportion of DAW as LLW.
Solid ILW volume estimates 687. The baseline processing strategy for the Hinkley Point C ILW streams is summarised in Table 5.3. The proposed baseline set out in Table 5.3 is the reference case for ILW processing which has been used to demonstrate that a suitable strategy can be implemented to manage the waste streams. As these wastes will not be generated until after the start of reactor operations, the detailed processing and packaging arrangements may be modified to further optimise the waste requiring storage.
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Table 5.3 Estimated annual volumes of solid intermediate level radioactive wastes produced during operation of two EPR units at Hinkley Point C
Anticipated Lifetime annual raw (60yr) raw Anticipated processing strategy ILW stream Waste description waste volume waste volume (m3) (m3) Organic resins that arise from the clean-up Polymer immobilisation in Concrete C1 casks. Followed by interim storage ILW ion of primary circuit water, water from the on-site awaiting availability of a GDF. exchange 6 360 effluent processing systems and the resins reactor fuel pools. Filters from the clean-up of primary circuit Cement grouted in Concrete C1 casks. Followed by interim storage on-site water and water from the LWPS and awaiting availability of a GDF. ILW spent FPC(P)S. The filters consist of a stainless 5 300 cartridge steel support, with a glass fibre or organic filters filter media. Other designs of filters, typically with lower Cement grouted in Concrete C4 casks. Followed by interim storage on-site 5 300 activity. awaiting availability of a GDF. A range of materials, including activated Cement grouted in Concrete C1 casks. Followed by interim storage on-site core components, contaminated metal, awaiting availability of a GDF. DAW plastics, cloth, glassware and rubble 2 120 >2mSv h-1 Note: Activated core components with heat generation levels above the ILW arising from operations during planned categorisation would be transferred to the reactor fuel pool where they would shutdown periods. be held for a period of delay storage before processing. ILW wet Sludge arising from cleaning the bottoms Cement grouted in Concrete C1 casks. Followed by interim storage on-site 2 120 sludge of liquid waste tanks and various sumps. awaiting availability of a GDF. Totals 20m3 1,200m3
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5.1.1.4. Spent fuel arisings
Spent fuel generated during operations
688. Heat generated through radioactive decay means that spent fuel removed from a reactor must be cooled for an initial period before it can be placed into interim storage. For Hinkley Point C, fuel assemblies removed from the reactor would be cooled underwater in the fuel building fuel pool for around three years.
689. Following this initial storage period in the fuel building fuel pool, the spent fuel assemblies would be transferred to the separate spent fuel interim storage facility, where they will be safely stored until a UK GDF is available for transfer and the spent fuel is ready for final disposal.
Spent fuel volume estimates 690. The reactor core contains the nuclear fuel in which the fission reaction occurs. The remainder of the active core structure serves either to support the fuel, control the chain reaction or to channel the coolant.
691. A maximum of 90 spent fuel assemblies will be removed every 18 months of operation from each reactor. With time included for planned outages for maintenance over the 60 years operation, a total of approximately 3,400 assemblies per reactor unit are expected to be generated. Through the lifetime of the two reactors at Hinkley Point C a total of around 6,800 fuel assemblies will be generated.
692. The dimensions of one fuel assembly are 0.214m x 0.214m x 4.859m, so the volume associated with the lifetime total of 6,800 fuel assemblies will be 1,513m3.
693. No disposal facility currently exists in the UK for spent fuel and none will be available when the Hinkley Point C reactors start operating. The strategy for spent fuel management is therefore to store the spent fuel on-site pending availability of a GDF. In order to ensure sufficient capacity is in place, a facility sized to store the lifetime volume of two reactors will be constructed.
694. The radioactivity of spent nuclear fuel falls to about one hundredth of its original levels within a year and to one thousandth of its original levels within 40 years. This characteristic makes interim storage an important element of spent fuel management. Further details on expected packaged spent fuel volumes and future disposability are detailed in Section 5.1.2.6.
5.1.1.5. Decommissioning wastes 695. The full range of waste minimisation methods, will be used to reduce the amount of waste produced during decommissioning to as low a level as possible, including decontamination, volume and size reduction and appropriate segregation of the waste to enable:
the maximisation of materials recycling; minimal production of waste which is difficult to dispose of, particularly long-lived, high activity waste and chemically hazardous waste; minimal production of 'secondary' waste (equipment used for the decommissioning phase and contaminated during the operations); and the maximum use of exempt crushed construction material for backfilling voidage and the minimum importation of clean material onto the site.
696. By treatment of the surface of contaminated material, the amount of waste, which has to be provided for final disposal, can be reduced substantially. In particular, the use of chemical cleaning or blasting of the surface and melting of metallic material can increase the amount of material suitable for unrestricted or restricted release. The use of these methods will have to be balanced against possible liquid and gaseous discharges arising from their use.
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697. Estimates of the volume and characteristics of radioactive waste generated during decommissioning are being developed to underpin the FDP.
698. Estimates of the quantities and characteristics of decommissioning ILW have been developed based on modelling of the neutron flux projected, anticipated lifetime power generation and material composition data for the core of an EPR reactor. The activation calculations used the highest total flux experienced by each component, to derive its total inventory. Therefore, the inventories for decommissioning ILW are considered to be upper bound estimates (maximum package inventories).
699. In order to manage these wastes it is planned that a decommissioning waste processing facility will be constructed on the site during the plant decommissioning phase. This facility is required to process the decommissioning LLW and ILW resulting from decommissioning of contaminated or activated plant in the nuclear island. It is anticipated that a single facility will be constructed in the turbine hall of Unit 1, to serve the management of the decommissioning wastes from both units.
5.1.2. Processing and packaging
5.1.2.1. Processing and packaging arrangements for LLW 700. Detailed arrangements for radioactive waste management will be covered in operating procedures required to demonstrate compliance with UK regulatory requirements. For LLW, these instructions are anticipated to cover minimisation, segregation, characterisation/assessment, accumulation, packaging, labelling, record keeping and consignment for transfer/disposal. Arrangements will be in place, to demonstrate that the volume and activity of LLW requiring disposal, will be minimised by the use of the Waste Hierarchy and the application of BAT.
701. The design of the EPR units incorporates a number of measures aimed at minimising the volume of solid waste requiring disposal. This includes the facilitation of segregation and the incorporation of volume reduction techniques.
702. LLW generated during the operational period from both the reactors and the associated auxiliary plant, will be transferred to the site solid waste treatment facilities within the ETB. This facility is designed to manage waste through segregation and application of suitable treatments in preparation for disposal. LLW will be processed and packaged as required to meet the Conditions for Acceptance of the appropriate off-site disposal facility.
703. LLW will be safely transferred from different locations in the radiation controlled area, to the ETB. Waste will be collected and stored according to waste activity categorisation at dedicated locations in the ETB and placed into a temporary buffer store prior to treatment. The waste will then be separated on the basis of the treatment method and will be stored in these areas until sufficient quantities have accumulated for a treatment campaign to start or for shipment off-site.
5.1.2.2. LLW on-site treatment 704. The treatment of solid waste is determined (once it has been monitored and assayed) generally by the categorisation of the waste together with its physical and chemical characteristics. Once categorised the waste will be packaged (and conditioned if necessary) and transferred off-site to the most appropriate facility for its treatment (such as super- compaction, melting or incineration) or disposal.
Segregation. Solid wastes will, as far as is practicable, be segregated and sorted at source to minimise secondary handling. Waste streams that generate mixed wastes will be sorted in a dedicated unit within the ETB, to optimise their subsequent management and disposal. The benefits associated with the segregation of waste need to be balanced with the detriments associated with increased operator exposure. The segregation of the waste into different waste groups will be carried out on the basis of different physical and
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chemical properties, e.g. combustible, non-combustible and compactable waste, and non-compactable waste. Shredding. Bulky solid combustible and compactable waste may be size reduced by shredding in the ETB prior to further treatment. The waste is size reduced by the use of a rotating blade assembly. The shredded material then falls through a duct into a compactable drum located directly below the shredder. Once full, the drum will be returned to the storage area and temporarily stored until a sufficient volume of waste for treatment or disposal is collected. Low force compaction. A low force compactor in the ETB will be used on-site to assist in the volume reduction of appropriate wastes prior to transfer off-site for disposal. Conditioning of LLW for disposal. Some LLW, e.g. LLW sludges and LLW resin, may require processing within the ETB either by dewatering, drying, or encapsulation in a mortar matrix within the waste disposal package, prior to transfer from the site in order to meet the Conditions for Acceptance for the proposed disposal site. Handling and transfer of final packages. Following treatment, the waste will be placed in an appropriate container for transport or disposal. After being sealed, the containers will be checked for the presence of external contamination prior to transfer out of the ETB. Waste containers awaiting transfer off-site, will be placed in buffer stores and transferred into transportation containers prior to loading onto the transportation vehicle. Disposal of LLW. A key consideration for the choice of preferred disposal route, has been the commitment to demonstrate best use of existing UK LLW management assets. Therefore direct disposal to LLWR is seen as the least desirable option and where a reasonably practicable alternative disposal route exists, e.g. incineration or metal melting, this has been chosen as the preferred option. This approach is consistent with the national strategy for LLW and Hinkley Point C will aim to utilise alternative disposal routes as available. This will contribute to the minimisation of disposal of wastes to LLWR and maximise its remaining operational lifetime.
5.1.2.3. Packaging and processing arrangements for ILW 705. The strategy is for ILW to be retrieved, conditioned and packaged on-site on a campaign basis throughout the operational phase. Waste processing will result in a passively safe package ready for interim storage. The passively safe packages will be stored in the sites ILW interim storage facility for the duration of operations. The stored ILW packages will be removed from the ILW store when a GDF is available to accept new build waste for final disposal.
Storage of waste for re-categorisation 706. The radioactivity of a proportion of the ILW that will be generated during operation of the Hinkley Point C EPRs would be dominated at the time of arising by relatively short lived radionuclides including cobalt-60 (half-life of 5.27 years), caesium-137 (half-life of 30.2 years) and iron-55 (half-life of 2.7 years). After a period of interim storage, the radioactivity of some of this waste will have reduced to such levels that the waste will no longer be classified as ILW. This waste will therefore be managed and disposed of as LLW.
Reference case for ILW processing 707. The proposed strategy for ILW conditioning and packaging at Hinkley Point C is termed the reference case and it assumes that operational ILW will be conditioned and treated using the same procedures as applied during the operation of existing PWR’s in EDF’s French fleet, with due consideration of UK specific requirements.
708. The reference case includes two types of cylindrical pre-cast concrete casks, designated C1 and C4, to be utilised for all operational ILW. Both of these casks can include internal mild steel shielding of variable thickness to provide shielding against different concentrations of gamma emitting radionuclides. The C1 Cask is 1.4m in diameter, 1.3m high and has a 0.15m thick concrete shield wall. The C4 Cask has the same dimensions apart from the diameter which is 1.1m. In the reference case scenario, the operational ILW will be immobilised within the casks using polymer immobilisation or cement grout, prior to being placed into the on-site ILW interim storage facility until a UK GDF is available.
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5.1.2.4. On-site ILW management 709. Arrangements and requirements for radioactive waste management cover minimisation, segregation, quantitative assessment, packaging, labelling, record keeping and consignment for transfer/disposal. Processes will be established and implemented for the packaging of radioactive wastes and these arrangements will be reviewed periodically and adequate records maintained.
710. The management arrangements apply to all activities, interactions and aspects that can affect the quality of the waste package product, including:
waste characterisation; container design; container manufacture; wasteform development; process development; plant specification and design; Letter of Compliance submissions and advice actions; plant commissioning and operation; raw materials storage; waste package interim storage and monitoring; control of non-conforming packages; change control and continual improvement of waste package design, processing plant and interim storage; and package records and their long-term retention.
On-site ILW processing and packaging 711. ILW generated during operation will be conditioned in the ETB. The ETB is used for the processing of all ILW that will be generated by the operation of the EPR and includes facilities for safe handling, treatment, conditioning, buffer storage, packaging and monitoring of wastes prior to transfer of packages to the ILW interim storage facility.
712. The key solid waste management function at the ETB, is the conditioning process (cementation and resin encapsulation) for the treatment or solidification of ILW, which will ensure that the waste is in a passively safe final form to be transferred from the ETB, to the ILW interim storage facility. The waste package itself will also be compliant with the requirements of the Nuclear Decommissioning Authority’s Radioactive Waste Management Directorate.
713. ILW cementation. Cementation through the use of specially formulated grouts, provides a means to immobilise radioactive material, that is either solid or in various forms of sludge. At Hinkley Point C, it is anticipated that all ILW wastes, other than ion exchange resins, will be conditioned utilising a cementation process. The waste is placed into containers and grout is then added into this container and allowed to set. The container, with the now monolithic block of concrete and waste, is then suitable for storage and disposal.
714. Similarly in the case of sludges, the current packaging assumption is that the waste will be placed in a container and a grouting mix, in powder form, is then added. The two are mixed inside the container and left to set, leaving a similar type of product as in the case of solids, which can be managed in a similar way.
715. ILW epoxy resin encapsulation. Ion exchange resins consist of small beads used to remove radioactivity from contaminated liquids. The radioactive ions in the liquid are absorbed onto the resin by the chemical process of ion exchange. The resins retain the activity and the cleaned liquids can then be safely disposed of. When the ability of the resins to absorb more radioactive ions is exhausted, they become radioactive waste.
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716. It is proposed that spent ion exchange resins will be processed by in-drum solidification utilising a polymer solidification process. The process is established as a technique for treating ILW ion exchange resins in the UK, at the Magnox site at Trawsfynydd, and in France using mobile processing units. In France, since 1996, the mobile processing units of the type proposed for use at Hinkley Point C, have completed 57 encapsulation campaigns and produced a total of 5,333 packages through the processing of 2,034m3 of resins.
5.1.2.5. Disposability of ILW from Hinkley Point C 717. Before conditioning and packaging of ILW, regulatory arrangements require that sites produce an ILW conditioning proposal. This would include a demonstration that, following conditioning, the waste would be compatible with existing, or future planned management and disposal options. This requires that a Letter of Compliance is obtained for the packaging proposal. The Letter of Compliance process is the mechanism that the Nuclear Decommissioning Authority’s Radioactive Waste Management Directorate utilises, to provide confidence that a waste package can be accepted at a future GDF.
718. The overall objective of the Letter of Compliance assessment process, is to give confidence to all stakeholders that the future management of waste packages has been taken into account as an integral part of their development and manufacture. This is achieved by the site operator working with the Radioactive Waste Management Directorate to demonstrate that the waste packages produced by a proposed packaging process, will be compliant with the generic waste package specification and compatible with plans for transportation and emplacement in the planned future geological repository.
719. In cases where the assessment has concluded that the waste package is compliant with the repository concept and underpinning assessments, the Radioactive Waste Management Directorate is prepared to confirm this by the issue of a Letter of Compliance.
720. As part of the GDA process, the opinion of the Radioactive Waste Management Directorate was sought on the likely acceptability for disposal in a GDF, of packaged ILW generated by the EPR. The Radioactive Waste Management Directorate was asked for its views on a number of different waste packages, including those that would be produced by implementing the GDA reference strategy for on-site ILW management. The Radioactive Waste Management Directorate indicated that, in principle, any of the proposed waste packages would be acceptable for disposal. Work will be ongoing with the Radioactive Waste Management Directorate through the Letter of Compliance process to ensure that packaged ILW from Hinkley Point C would be acceptable for disposal in a GDF.
5.1.2.6. Packaging (encapsulation) of spent fuel for disposal 721. The Radioactive Waste Management Directorate is developing disposal concepts for HLW and spent fuel, undertaking work on several related areas.
722. In relation to disposal, the Radioactive Waste Management Directorate has developed a reference concept based on the Swedish KBS-3V method. This concept is known as the UK Reference HLW and Spent Fuel Repository Concept. The concept was developed in order to demonstrate the viability of geological disposal of HLW and spent fuel in the UK.
723. Under this concept, spent fuel would be over-packed before disposal into durable, corrosion resistant disposal canisters, manufactured from suitable materials, which would provide long term containment for the radionuclides contained within the spent fuel. This process is known as encapsulation.
724. There are two basic options for encapsulation of spent fuel:
packaging into disposal containers at the nuclear power station site; or packaging into disposal containers at a central location.
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725. In the event that a national, or regional, facility for encapsulation of spent fuel is unavailable at the time of store emptying, it is possible that encapsulation into disposal containers could occur on-site at Hinkley Point C. This would require the construction of a new facility to undertake the process. The facility and operations would be required to be compliant with the Nuclear Site Licence and the sites Environmental Permits with regard to safety and radioactive waste discharges.
Disposal of spent fuel to GDF 726. The final packaged volume of spent fuel generated during operation of the two reactors at Hinkley Point C, that will require disposal to the GDF is estimated as follows. Assuming a total of 6,800 fuel assemblies with the envelope volume of a canister capable of accommodating four fuel assemblies being 3.33m3, the packaged volume of the waste consisting of a total of 1,700 canisters is therefore 5,661m3.
727. As with ILW, the Radioactive Waste Management Directorate has undertaken a disposability assessment for the spent fuel expected to arise from the operation of an EPR. This assessed the implications of the disposal of the proposed spent fuel disposal packages, against the waste package standards and specifications developed by the Radioactive Waste Management Directorate and the supporting safety assessments for a GDF. The safety of transport operations, handling and emplacement at a GDF and the longer term performance of the system have been considered, together with the implications for the size and design of a GDF.
728. The disposability assessment performed has provided confidence that spent fuel generated during operation of Hinkley Point C will be acceptable for disposal in the planned UK GDF.
5.1.3. Storage arrangements
5.1.3.1. Interim on-site storage of solid LLW 729. The strategy for LLW is, that waste generated throughout nuclear power plant operations and decommissioning, will be disposed of as soon as reasonably practicable following treatment to minimise volume and perform appropriate conditioning or packaging.
730. LLW generated during the operational period from both the reactors and the associated auxiliary plant, will be transferred to the ETB until such time as sufficient quantities have accumulated for a treatment campaign to start or for shipment off-site. The process is described in more detail above in Section 5.1.2.1.
5.1.3.2. Interim on-site storage of solid ILW 731. There is currently no ILW disposal facility in the UK. The GDF is not expected to be available for disposal of wastes for a number of years after Hinkley Point C starts operations. The strategy for ILW management at Hinkley Point C, is therefore to process the waste and store the waste on-site, according to the principles of passive safety(40), pending availability of the GDF. In order to ensure sufficient capacity is in place, a facility sized to store the lifetime volume of two reactors will be constructed.
732. The key requirement of the interim storage facility will be to provide protection for the waste packages from potential external corrosion caused by atmospheric salts, temperature and humidity levels, which could have a long-term impact on the integrity of the package and eventual acceptance of the package at the GDF. In terms of containment of radioactivity and prevention of releases which could impact upon the environment, a number of barriers and environmental controls are provided:
The conditioned wasteform is the primary barrier, e.g. the cemented or epoxy matrix. The waste container is the secondary barrier, e.g. the concrete package.
(40) HSE document, Safety Assessment Principles for Nuclear Facilities, 2006 Edition, Rev 1
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Control of the store environment is important in maintaining integrity of the waste container to ensure compliance with Letter of Compliance requirements. The store structure is the final layer of weather/atmospheric protection for the waste package and also is an important element in the physical security of the waste.
733. The store will be designed to align with the timeline for GDF availability dates. The store will require appropriate maintenance and various levels of in-service refurbishment. The facilities on-site, including the ILW interim storage facility, will be subject to periodic safety case reviews throughout the operational life of the store ensuring any necessary and timely improvements will be made.
734. The lifespan of the ILW interim storage facility is expected to be up to 100 years, but would be capable of extension through refurbishment.
735. The facility is designed to receive and store packages of ILW waste arising from the planned 60 years of operation of the two units on the Hinkley Point C site.
Facility design 736. The final design of the ILW interim storage facility has not been completed, but it is anticipated that it would consist of areas performing the following functions:
receipt and dispatch area; interim storage space for all operational ILW until a GDF becomes available; package inspection area; and storage of ILW that would become LLW following a period of decay storage.
737. The facility will also require a number of auxiliary systems and facilities, such as electrical power unit, ventilation system unit and maintenance area.
738. The facility will be designed, constructed and operated to minimise radiation doses to workers and the public. The facility will include the following safety functions:
the waste packaging and the storage facility design will provide containment for radioactive material; the facility will minimise the radiation exposure of workers and the public through the provision of shielding; and during detailed design, it will be determined if a filtered ventilation system is required.
739. Further measures would also be implemented to prevent the risk of a loss of containment from a waste package including:
minimising waste package handling operations and where practicable minimising the lift height of packages, where package movements cannot be avoided; inspection and monitoring of the waste packages in the storage hall to allow early intervention if any package defect is identified; and the waste packages are designed to be robust against impact and or being dropped during package movement operations.
5.1.3.3. Interim on-site storage of spent fuel 740. The operational life of the spent fuel interim storage facility is assumed to be 100 years, although it would be capable of extension beyond that if necessary, subject to any required refurbishment or replacement of equipment. The life would be determined by the lifetime of the power station, the availability of alternative off-site disposal or storage facility and the time required for the heat rate within the fuel to reduce before disposal. The bounding case from the Radioactive Waste Management Directorate, is that the fuel might require 100 years of interim storage after the fuel has been removed from the reactor, although this figure is acknowledged to be conservative and is likely to be reduced in future. This will allow
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safe and secure storage to be maintained until a GDF, or alternative disposal route has been established.
741. The options available for on-site interim storage of spent fuel have been reviewed and determined that for the site specific circumstances at Hinkley Point C, wet interim storage within an engineered pool is the preferred approach.
742. In reaching this proposed storage decision, consideration has been given to possible future developments regarding UK spent fuel management. Whilst the preferred option will deliver a safe and secure solution, there may be alternative options available in the long term that means spent fuel does not need to be stored for long periods on site. The wet interim storage option proposed for Hinkley Point C, is one that is flexible enough to be adapted to such future changes, should they occur.
743. Wet storage of spent fuel has been used widely in the UK and internationally and has been licensed previously. It is considered to be both safe and environmentally acceptable for use for spent fuel generated from operation of Hinkley Point C. The use of wet interim storage of spent fuel is capable of providing Hinkley Point C with a safe, secure and technically flexible solution until such time that the spent fuel is suitable for transfer and a UK GDF, or other off- site management facility, is available.
Facility design 744. The design of the spent fuel interim storage facility will:
ensure safe operations (e.g. by preventing a criticality incident and maintaining effective containment); provide radiological protection to the public, workers and the environment at all times in compliance with dose limits and ensuring that all doses are ALARP and discharges to the environment are demonstrated to be minimised in accordance with BAT; ensure cooling to maintain spent fuel integrity; and maintain spent fuel in a condition appropriate for transport and final disposal.
5.2. RADIOLOGICAL RISK TO THE ENVIRONMENT
5.2.1. Assessment of risks to the environment 745. A set of core principles underpin the waste management strategy to be implemented at Hinkley Point C. These are as follows:
i. Minimise the generation of wastes, as far as is reasonably practicable, through application of the waste hierarchy and minimise the use of harmful substances as far as is reasonably practicable. ii. Protect the environment and people by minimising the presence of harmful substances in any discharges and disposals. iii. Document, retain and record appropriate information relating to the management and disposal of wastes. iv. Safely store wastes in robust and adequate containment to prevent leakages which would otherwise result in the generation of additional waste, contaminated land, groundwater or contamination of the broader environment. v. Provide access to Suitably Qualified and Experienced Persons to ensure that anyone involved in activities that impact on the generation and/or disposal of wastes will be provided with adequate and appropriate training, information and advice. vi. To keep the strategy up to date by maintaining consistency with Government Policy, regulatory requirements, the availability of waste storage and disposal facilities, advances in waste management technologies and any internal changes in operating conditions. vii. Review operational experience to enable ongoing development and continual improvement in waste management practices, to support the key principles of application of the waste hierarchy and the use of BAT.
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746. In addition to the core principles for waste management, a number of key elements form the basis of the strategy for the management of radioactive wastes:
viii. To maintain radiation doses to the workforce and the general public from radioactive waste management operations, including disposal, within legal limits and As Low As Reasonably Achievable (ALARA). ix. To develop and maintain adequate safety and environment cases where necessary, for all waste management activities including handling, accumulation and storage of wastes on-site and ultimately disposal off-site. x. To dispose of all radioactive wastes as soon as reasonably practicable via established routes that have been identified in the relevant safety/environment cases. xi. To store safely all radioactive wastes for which a disposal route has yet to be established in the relevant safety/environment cases.
5.2.2. Precautions taken
5.2.2.1. Precautions taken to contain and shield the waste
Operational waste excluding fuel 747. The radioactive waste from plant operation (filters, resins, gloves, personal protective equipment, etc.), is packaged to ensure the containment of radioactive materials, before transfer to approved centres. The containment of this waste is managed and monitored at every stage of the treatment process.
748. The NAB and the ETB, dedicated to the collecting and sorting of the radioactive waste, have thick concrete walls and roofs, allowing radiation to be greatly attenuated.
749. External exposure of the public is monitored using dose rate measurements, to check that all of these provisions limit the dose received by the population at the site boundary, to a value much lower than the statutory level under normal operating conditions. Before these measurements are taken, the dose rate is checked by the plant radiation monitoring system in the various nuclear rooms inside the installation itself in order to protect the site's workers.
750. For LLW, the site operator will follow a strategy of early disposal using commercially available routes, unless clear justifiable reasons can be given for not pursuing early disposal. Conditioning, packaging and interim storage of ILW at Hinkley Point C and the passively safe nature of the ILW treatment on-site, are detailed above in Section 5.1.3.2.
Spent fuel 751. A proportion of the fuel in the reactor is replaced during each scheduled shutdown as dictated by the refuelling cycle. The spent fuel assemblies that have been removed are transferred to the fuel building fuel pool, where they are stored underwater for cooling; the fuel building fuel pool has a capacity to hold spent fuel assemblies corresponding to at least 10 years of operation. During this period the shorter lived radio-isotopes undergo decay which significantly reduces the radioactivity associated with the spent fuel. Following this initial pool storage period, the spent fuel will be transferred to the on-site spent fuel interim storage facility - a more detailed description of this can be found above in Section 5.1.3.3.
752. The spent fuel interim storage facility will have a range of safety features to maintain the safety of spent fuel. The design and operation of the facility will be required to be compliant with the Nuclear Site Licence conditions, with regard to the safety of workers and the public.
753. A brief outline of the key safety features of wet storage in the spent fuel interim storage facility is set out below:
The significant water volume within the pool provides a variety of safety functions. It slows down the rate of any water temperature increase and reduces the significance of any loss of water, so that the water make-up system would easily maintain the water level
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in the event of losses. In the highly unlikely case of a total loss of the pond cooling, there is a lengthy ‘grace period’ before evaporation of the water would lead to fuel uncovering, which would allow the operator time to react to put the installation in a safe state. In the event that there were any radioactivity releases from within the fuel, the pool water provides a medium within which the activity is held and ultimately removed, so mitigating any release into the environment. The facility will be designed to be resistant to movement by events such as earthquakes and other external events. The spent fuel pool would be equipped with water cooling systems (i.e. pumps and heat exchangers). Clean-up systems would also be provided to maintain water quality. The water chemistry is also controlled to minimise corrosion of fuel assemblies. The spent fuel pool will be designed with appropriate containment systems and have leak detection and collection systems. Wet storage allows the monitoring of water parameters (temperature, radioactivity, pH and chemical composition) and ventilation parameters. These features permit the rapid detection of changes and therefore allow mitigation measures to be implemented if required. The assemblies in wet storage are accessible and the storage area visible. The water provides effective shielding against radiation emitted by the spent fuel. Thus spent fuel inspection in wet storage is therefore possible without retrieval.
5.3. ARRANGEMENTS FOR THE MOVEMENT AND DESTINATIONS OF THE DIFFERENT WASTE CATEGORIES TRANSFERRED OFF-SITE
5.3.1. Destinations of waste packages transferred off-site 754. All disposals of solid radioactive waste (apart from those radiological wastes which meet the criteria for exemption) will be permitted by the Environment Agency in the form of an Environmental Permit granted under the Environmental Permitting (England and Wales) Regulations 2010.
755. The strategy for LLW disposal has considered the commitment to demonstrate best use of existing UK LLW management assets. Therefore direct disposal to LLWR is seen as the least desirable option and where a reasonably practicable alternative disposal route exists, e.g. incineration or metal melting, this has been chosen as the preferred option. This approach is consistent with the UK national strategy for LLW(41). This will contribute to the minimisation of disposal of wastes to the LLWR and maximise its remaining operational lifetime.
756. The strategy for LLW, is that waste generated throughout nuclear power plant operations and decommissioning, will be disposed of as soon as reasonably practicable, following treatment to minimise volume and perform appropriate conditioning or packaging. The ultimate disposal of the wastes is expected to be via one of the following main routes depending on the radioactivity level of the waste produced, its physical characteristics and its chemical properties:
treatment of metals, ultimately for recycling, via commercially available routes subject to meeting the relevant Conditions for Acceptance; incineration of combustible wastes using commercially available routes, subject to meeting relevant Conditions for Acceptance (there would be no on-site incineration of wastes); use of appropriate authorised disposal facility for exempt wastes and VLLW disposal (notably for soil, rubble and aggregates) where no reuse or recycling options are viable, subject to meeting relevant Conditions for Acceptance; transfer of suitable LLW for super-compaction prior to disposal at LLWR to minimise
(41) UK Strategy for the Management of Solid Low Level Radioactive Waste from the Nuclear Industry, Nuclear Decommissioning Authority, August 2010.
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disposal volume; and disposal of LLW directly to LLWR will be utilised only where the above alternatives are not practicable.
757. Conditions and limits for the transfer of LLW will be set by the Environment Agency in the Hinkley Point C Environmental Permit issued under the Environmental Permitting Regulations 2010.
758. No off-site movements of ILW or spent fuel are anticipated until the GDF is available (unless a suitable alternative interim storage, or management, arrangement becomes available during the lifetime of Hinkley Point C) and appropriate permissions have been granted by the UK competent authorities.
5.3.1.1. Off-Site metal recycling facility operations 759. Where the metallic waste generated by operational maintenance work cannot be adequately decontaminated on-site, the waste will be transferred to an off-site commercial Metals Recycling Facility e.g. Studsvik Metal Recycling Facility at Lillyhall, Cumbria. The volume of metallic waste requiring disposal could be reduced by up to 95% using metal recycling techniques.
760. Once transferred to the recycling facility, a range of industrial cutting and cleaning techniques are applied. The metallic waste is decontaminated and cleaned using methods such as dry grit blasting so that the resulting materials can either be recycled in the UK or potentially sent to a facility for further cleaning by melting.
5.3.1.2. Off-site incineration operations 761. LLW will be segregated within the ETB to separate combustible waste from non-combustible. Combustible waste suitable for incineration, will be transferred to an off-site commercial incinerator and incinerated in a specially engineered kiln up to around 1000°C. Any gases produced during incineration are treated and filtered prior to emission into the atmosphere and will conform to international standards and national emissions regulations.
762. Incineration of combustible wastes is applied to both radioactive and other wastes in the UK. In the case of radioactive waste, incineration is used for the treatment of LLW from nuclear power plants, fuel production facilities, research centres (such as biomedical research), the medical sector and waste treatment facilities.
763. Modern incineration systems are well engineered and designed to burn the waste efficiently whilst producing minimum emissions. Ash remaining following incineration is disposed of as appropriate.
5.3.1.3. Off-site super compaction facility operations 764. Suitable LLW will be transferred off-site to a super-compaction facility to minimise its volume. In this process drums or boxes of waste are compacted under high pressure of up to 2,000 tonnes per square metre. Following super compaction, the drums are transferred onward to LLWR for disposal.
5.3.1.4. VLLW operations 765. High-volume VLLW can be disposed of to specified approved landfill sites. The waste will be subject to controls on its disposal which are specified by the environmental regulators.
5.3.1.5. LLWR operations 766. LLW unsuitable for disposal via the above disposal routes, but which meets the Conditions for Acceptance for LLWR, will be packaged on-site and transferred directly for disposal to LLWR in approved transport packages e.g. half height ISO containers (HHISO).
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5.3.1.6. Disposability of LLW 767. For all LLW, with the exception of oils and solvents, acceptance for disposal has been agreed in principle with LLWR during the GDA process. This has now been updated for Hinkley Point C LLW/VLLW and disposability in principle has been confirmed by LLWR for the volume and activity levels presented within this submission.
768. In order to demonstrate the acceptability of the potential non-LLWR disposal routes for Hinkley Point C, disposability in principle for the appropriate waste streams, has been obtained for incineration, VLLW landfill, and the segregated services provided by LLWR (metal recycling and super compaction).
5.3.2. Movement of LLW and VLLW packages 769. Following treatment, conditioning and packaging of the waste within the ETB, the containers will be monitored for the presence of external contamination and placed into buffer storage prior to transfer out of the ETB. ILW will be transferred to the ILW interim storage facility while VLLW and LLW will be transferred outside of the nuclear island to a transit area for temporary storage.
770. LLW and VLLW containers leaving the transit area will be packaged into appropriate transport containers and transferred off-site to a number of locations including off-site treatment facilities and VLLW and LLW disposal facilities.
771. Prior to shipping, the data that have been collected for the waste package, will be verified for completeness and the appropriate transportation, storage and/or disposal documentation prepared. Waste packages will not be transported until this paperwork has been verified to meet the required standards.
772. All radioactive waste despatched from the site will comply with relevant UK and international legislation including the relevant requirements of The Radioactive Material (Road Transport) (Amendment) Regulations 2003(42), thus fulfilling the requirements of the International Atomic Energy Agency (IAEA) regulations for the transport of radioactive materials. Each consignment is monitored for contamination and for external radiation before leaving the site.
773. Radioactive waste is transported in specially designed packages. The design of these packages must satisfy regulatory requirements administered by the Radioactive Materials Transport section of the Department for Transport (which is scheduled to become part of the Office of Nuclear Regulation in April 2011). The packages must provide adequate protection to operators and members of the public during normal and accident conditions.
774. Following thorough verification, some wastes will be exported from the ETB as exempt radioactive waste (see Section 5.4). These wastes can be disposed of using conventional routes, for example for recycling.
5.4. CRITERIA FOR CONTAMINATED MATERIALS TO BE RELEASED FROM THE REQUIREMENTS OF THE BASIC SAFETY STANDARDS
5.4.1. National strategy, criteria and procedures for the release of contaminated and activated materials
5.4.1.1. The BSS Directive 775. The BSS are to be applied to all practices involving risk from ionising radiation and are contained in the Council Directive 96/29/EURATOM of 13 May 1996.
(42) SI 2009 No. 1348 – The Carriage of Dangerous Goods and Use of Transportable Pressure Equipment Regulations 2009
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776. The disposal, recycling or reuse of radioactive substances or materials containing radioactive substances arising from any practice, is subject to the requirement of reporting and prior authorisation.
777. However, the disposal, recycling or reuse of such substances or materials may be released from the requirements of this Directive provided they comply with clearance levels established by national competent authorities. These clearance levels shall follow the basic criteria used in Annex I of the Directive and shall take into account any other technical guidance.
5.4.1.2. Related UK legislation 778. The statutory instruments which define radioactive waste and which provides the mechanism for authorising its disposal are the Environmental Permitting (England and Wales) Regulations 2010. These have replaced the Radioactive Substances Act 1993 in England and Wales.
779. There are a number of Exemption Orders which, subject to meeting certain criteria, allow the disposal of radioactive wastes without an Environmental Permit. Some of these Exemption Orders apply conditions which must be met by the waste owner when undertaking disposals.
780. Proposals for a revised exemptions regime across the UK were set out in a document entitled “Proposals for A Future Exemptions Regime under The Radioactive Substances Act 1993 and The Environmental Permitting Regulations 2010”. The aim of the new regime is to provide a modern approach to the regulation of ubiquitous and low risk radioactive materials and waste.
5.4.1.3. General criteria 781. The criteria for the definition of radioactive material and waste are given in the Environmental Permitting (England and Wales) Regulations 2010 and are based on the type of radioisotope.
782. If an isotope originates from nuclear fission, or neutron or radiation bombardment and is not naturally occurring, then its presence in a substance will confer the status of radioactive material/waste to the substance, irrespective of its activity concentration.
783. Conversely, if an isotope is listed in the table reported in Schedule 23 (Radioactive substances activities) of the Environmental Permitting (England and Wales) Regulations 2010, then this isotope will confer the status of radioactive material/waste only if its specific activity in the substance expressed in becquerels per gram is above the corresponding value reported in the table. The table is reproduced below.
Table 5.4 Specified elements
Becquerels per gram (Bq g-1) Element Solid Liquid Gas or Vapour Actinium 0.37 7.40E-02 2.59E-06 Lead 0.74 3.70E-03 1.11E-04 Polonium 0.37 2.59E-02 2.22E-04 Protoactinium 0.37 3.33E-02 1.11E-06 Radium 0.37 3.70E-04 3.70E-05 Radon - - 3.70E-02 Thorium 2.59 3.70E-02 2.22E-05 Uranium 11.1 0.74 7.40E-05
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784. A key Exemption Order is The Radioactive Substances (Substances of Low Activity) Exemption Order 1986 No. 1002, as amended by the Statutory Instrument 1992 No. 647. The key requirements of this Exemption Order are described below:
a) a solid, other than a closed source, which is substantially insoluble in water, the activity of which, when it becomes waste, does not exceed 0.4Bq g-1 of mass; b) an organic liquid which is radioactive solely because of the presence of carbon-14, or tritium (or both), the activity of which, when it becomes waste, does not exceed 4Bq ml-1 (amended in 1992); or c) a gas containing one or more radionuclides none of which, nor the decay products of which, has a half life greater than 100 seconds.
5.4.2. Clearance levels established by competent authorities for disposal, recycling and reuse 785. The clearance levels established by competent authorities in the UK are presented in Section 5.4.1.3.
5.4.3. Envisaged types and amounts of released materials 786. Non-radioactive solid wastes arise during the operation and maintenance of the process plant (e.g. maintenance of pipes and equipment), and as the result of a number of routine activities). The annual quantities that are anticipated are presented in Table 5.5.
Table 5.5 Annual quantities of released materials
Waste type Tonnes Inert waste and commercial waste 940 Hazardous (non-radioactive) waste 200 Total arisings (annual) 1,140
787. A proportion of the solid low level waste referred to in Section 5.1.1.2 could be disposed of, subject to the appropriate clearance and release regimes with this waste.
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6. UNPLANNED RELEASES OF RADIOACTIVE EFFLUENTS 788. As stated in Chapter 0 of this submission, the Office of Nuclear Regulation’s (ONR) Chief Nuclear Inspector, Dr Mike Weightman, published an Interim Report on 18th May 2011 investigating the implications of the Fukushima event for the UK’s nuclear fleet and highlighting any lessons learned. The final version of Dr Weightman’s report is expected in autumn 2011.
789. The Interim Report’s analysis highlights that the direct causes of the accident, a magnitude 9 earthquake and the associated 14 metre high tsunami, are far beyond the most extreme natural events that the UK would be expected to experience. Furthermore it sets out the differences between the Japanese approach to safety assessment and that employed in the UK. In particular, it notes the disparity between the Japanese deterministic methodologies and the UK’s probabilistic approach, under which safety must be demonstrated against extreme natural hazards with a return period of 1 in 10,000 years.
790. The Interim Report draws a number of important conclusions relevant to the Hinkley Point C project:
“To date, the consideration of the known circumstances of the Fukushima accident has not revealed any gaps in scope or depth of the Safety Assessment Principles for nuclear facilities in the UK.” “Our consideration of the events in Japan, and the possible lessons for the UK, has not revealed any significant weaknesses in the UK licensing regime.” “In considering the direct causes of the Fukushima accident we see no reason for curtailing the operation of nuclear power plants or other nuclear facilities in the UK. Once further work is completed any proposed improvements will be considered and implemented on a case by case basis, in line with our normal regulatory approach.” “More generally, in the course of our examination of the events in Japan, we have not seen any significant defects in the UK’s approach to nuclear regulation. This reinforces the way in which we have been able to develop an effective approach to regulating nuclear new build through a system of Generic Design Assessment (GDA) and specific nuclear site licensing, and construction consents.” “There is no need to change the present siting strategies for new nuclear power stations in the UK.”
791. The Interim Report also draws a number of recommendations relevant to this submission. All recommendations that can be applied to Hinkley Point C will be taken into account at the appropriate time as the project develops.
Weightman considers both the location of sites in areas subject to particularly onerous natural hazards, and the ability to take precautionary countermeasures such as evacuation. The Interim Report notes that for sites with a flooding risk, detailed consideration may require changes to plant layout and the provision of particular protection against flooding. The UK nuclear industry should ensure that all extreme hazards are considered in taking account of Weightman’s recommendations, particularly for plant layout and design of safety-related plant. There is no need to depart from a multi-plant site concept given the design measures in new reactors being considered for deployment in the UK and adequate demonstration in design and operational safety cases. The UK nuclear industry should learn from concrete and other structural responses in earthquake once detailed information becomes available. The UK nuclear industry should review the dependence on off-site infrastructures and sites’ self-sufficiency in extreme conditions. The UK nuclear industry should learn from the Fukushima event in light of the provision of on-site of emergency control, instrumentation and communications, the time length and physical spread of the disruption encountered, and the environment on-site associated with a severe accident.
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792. The conclusions and recommendations presented in Dr Weightman’s Interim Report are in line with the findings and lessons learned presented in the preliminary summary report of the IAEA Fact Finding Mission.
793. These conclusions and recommendations allow us to confirm that, to date, the consideration of the known circumstances of the Fukushima accident:
has not revealed any gaps in the flood defence and precautionary countermeasures proposed for Hinkley Point C; has not revealed any gaps in the consideration of extreme hazards in the safety case and the design of the plant; has confirmed that the safety related aspects of the design and the site have been and will be assessed in a robust regulatory context; does not foreclose the progress of the Hinkley Point C project in the current UK regulatory environment; demonstrates that NNB Generation Company Ltd is responding to the Interim Report’s recommendations. This will lead to incorporation of lessons learnt into the Hinkley Point C project as it develops.
6.1. REVIEW OF ACCIDENTS OF INTERNAL AND EXTERNAL ORIGIN WHICH COULD RESULT IN UNPLANNED RELEASES OF RADIOACTIVE SUBSTANCES
6.1.1. Plant safety principles 794. Nuclear safety includes all of the technical provisions and organisational measures relating to the design, construction, commissioning, operation and dismantling of installations that involve a source of ionising radiation and the transportation of radioactive materials, intended to prevent accidents, to keep risks ALARP and to limit the effects of any exposure to radiation i.e.:
to ensure normal plant operation, while keeping the radiological impact for workers and the public ALARA and in any event, below the limits prescribed by the Regulations; to prevent incidents and accidents; and to limit the consequences of any possible incident or accident by taking measures to control radiation risks to ensure that no individual bears an unacceptable risk of harm.
795. In terms of normal operation, the methods implemented for the UK EPR and the resulting impacts are described in detail in Chapter 3, Chapter 4 and Chapter 5.
796. With reference to risk management, the measures put in place during the plant design, construction, commissioning, operation and dismantling stages, cover:
risk prevention to reduce the probability of occurrence; monitoring and detection of operating anomalies; and limiting consequences with the aim of making residual risks acceptable with regard to personnel, the public and the environment.
797. Nuclear risks are controlled using measures which ensure that the plant can be managed in any situation, by assessing a comprehensive range of potential faults and demonstrating adequate protection in terms of:
Reactivity control. Removal of residual thermal power. Containment of radioactive substances.
798. In order to guarantee a high level of safety, a large number of independent measures are implemented. This collection of measures results from the application of the ‘defence in
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depth’ concept, which involves systematically taking technical or human failures into account and providing several levels of protection against potentially significant faults and failures.
799. The UK EPR safety process, which has been implemented at the design stage, is based on defence in depth over five levels.
Level 1 is a combination of design, quality assurance and control margins aimed at preventing the occurrence of abnormal operating conditions or plant failures. Level 2 consists of the implementation of protection devices which make it possible to detect and correct the effects of deviations from normal operation or the effects of system failures. This defence level is aimed at ensuring the integrity of fuel cladding and that of the primary cooling system so as to prevent accidents. Level 3 consists of safeguard systems, protection devices and operating procedures which make it possible to control the consequences of accidents that may occur, so as to contain radioactive material and prevent the occurrence of severe accidents. Level 4 comprises measures aimed at preserving containment integrity and controlling severe accidents. Level 5 includes, in the event of the failure of previous levels of defence, all measures for protecting the public against the effects of significant radiological releases.
800. A systematic, comprehensive analysis of all potential faults is carried out to verify that even in these situations, defined safety objectives are met and consequences for the environment and populations are minimised so far as is reasonably practicable and remain below the thresholds prescribed by national and international authorities.
801. With regard to reducing the potential consequences of incidents and accidents, safety performance is improved in the following four main ways:
By accounting for, and reducing the frequency of initiating events (which cause transients, incidents or accidents) liable to occur during the different states which the reactor may encounter during operation (including full power and shutdown states, and states with the core completely unloaded in the spent fuel pool). Taking internal hazards into account on a deterministic basis in accordance with design principles similar to those used for simple initiating events, enhances the defence in depth approach. By taking into account external hazards at high severity levels, whether the hazards are of human origin (aircraft crashes, explosions etc.) or of natural origin (earthquakes, extreme temperatures, flooding etc.). In addition to their direct effects, these hazards are studied from the point of view of the damage they may cause on non-protected structures and equipment, inside or outside the plant. By taking severe accidents (such as a core melt accident) into account at the design stage and implementing physical measures to ensure "practical elimination" of events and sequences that could have a significant radiological impact on the environment during the power plant’s service life. For events which cannot be prevented by design, the probability of environmental releases is minimised by strengthening the containment, including conditions which could lead to containment bypass. By use of Probabilistic Safety Analysis (PSA) at the concept design phase, to confirm the design approach and identify the multiple failure sequences that should be considered in the design basis, so as to prevent core meltdown accidents.
802. The safety assessment performed in the safety report for Hinkley Point C is based on the well established deterministic methods, augmented by probabilistic methods using appropriate numerical targets and analysis. The main deterministic method is adoption of a defence in depth approach and the concept of independent physical barriers against the escape of radioactivity. The principal quantitative safety targets are outlined below and are based upon UK practice:
The collective dose to workers shall be ALARP. Doses to workers during normal operation of the plant will not exceed UK statutory limits. The dose limits to be applied, shall be those specified in the Statutory Instrument currently in force under the relevant Parliamentary Act.
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Doses to the public during normal operation of the plant will not exceed UK statutory limits. The dose limits to be applied shall be those specified in the Statutory Instrument currently in force under the relevant Parliamentary Act. An annual whole body effective dose for individual employees and contracted workers due to normal operation of a reactor unit shall not exceed 10mSv. The maximum dose to an individual off-site (member of the public) due to normal operation of a reactor single unit shall not exceed 0.3mSv y-1 and shall not exceed 0.5mSv y-1 from the operation of all facilities on the site. The risk of an individual worker fatality due to exposure to radiation from an on-site accident (all facilities) will be below 1x10-6y-1 and/or demonstrated as ALARP. The risk of fatality of any person off-site (member of the public) due to exposure to radiation from on-site accidents (all facilities) will be below 1x10-6y-1 and/or demonstrated as ALARP. The total predicted frequency of accidents (from all facilities) resulting in more than 100 fatalities (either immediate or delayed) of members of the public will be below 1x10-7y-1 and/or demonstrated as ALARP.
803. The primary purpose of the safety report that is being prepared for Hinkley Point C is to demonstrate that people and the public are protected from the harmful effects of ionising radiation. A series of fundamental safety principles are applied to the design, construction, commissioning and operation of nuclear facilities at Hinkley Point C.
6.1.2. Development of the safety report for Hinkley Point C 804. A safety report will be produced for Hinkley Point C encompassing all nuclear facilities on site. These will demonstrate that the plant safety principles that are presented in Section 6.1.1 have been applied and that the safety targets have been met. Hinkley Point C has a phased approach to the development of safety cases which allows the identification, assessment and mitigation of hazards and their associated risks. The phases of the safety case are aligned with the key phases of the design, construction, commissioning and operation of facilities on the site and are summarised as:
Pre-Construction Safety Report (PCSR). A series of reports that are prepared during detailed design and are submitted prior to the construction of key safety related structures. Pre-Commissioning Safety Report (PCmSR). Prepared during construction of the facility and submitted before the start of plant and process commissioning (non-active commissioning). Pre-Operation Safety Report (POSR). Prepared during non-active commissioning and submitted before the start of nuclear operations (active commissioning). Station Safety Report. Documents the Safety Case throughout the operational phase of the site.
805. The safety reports that will be produced for Hinkley Point C will fully accord with the requirements of the ONR. The ONR is the regulatory body with responsibility for regulating the nuclear power industry. Each nuclear site licence includes the requirement to produce and manage safety reports. The safety reports will be subjected to rigorous and comprehensive internal assessment by the site operator.
806. The ONR will then assess the Safety Reports as part of the licensing process for the project. This independent scrutiny examines both plant design and management systems. It provides additional confidence that the claims made within the safety reports are valid and that the safety targets referenced above will be met. Once the safety report has been implemented the ONR will undertake a programme of inspection and review. This programme is designed to demonstrate that the plant and systems are performing as expected and that the claims that are made within the safety reports remain valid. Deviations are the subject of corrective action with could include enforcement activities.
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6.1.3. Design scope 807. The object of the design scope is to define the events taken into account in the design basis and to categorise them. The PSA is a confirmatory step which is used to support the robustness of the design.
808. The initiating events considered are of different types and are dealt with differently.
809. In terms of the design process, the overall approach is the same:
Definition of the design basis list of plant-based faults, internal and external hazards, and events/sequences with consideration of combinations. Quantification of the event/sequence impacts, the results being used for the design of systems and structures and/or the demonstration that the safety requirements are met. Design verification which completes the safety analysis by providing a further demonstration that the safety requirements are met. It invariably includes the use of PSA and in some cases a deterministic verification is carried out. This step can result in design feedback.
6.1.4. Internal faults – design basis analysis 810. The safety approach applied to the EPR requires consideration of a number of representative internal faults and enveloping conditions, which could occur during normal operation and various associated reactor states. The relating initiating events are grouped together in four categories based on their estimated frequency of occurrence and their impact.
811. On this basis, events are grouped into four PCC’s as follows:
PCC-1 which includes all normal operating conditions characterised by initiating events whose estimated frequency of occurrence is greater than 1 per year. PCC-2 which includes design basis transients, characterised by initiating events with an estimated frequency of occurrence in the range of 10-2 to 1 per year. PCC-3 which includes all design basis incidents characterised by initiating events with an estimated frequency of occurrence is within the range of 10-4 to 10-2 per year. PCC-4 which includes all design basis accidents characterised by initiating events with a frequency of occurrence is within the range of 10-6 to 10-4 per year.
812. Identification of these events and their classification by category determines the design of systems intended to control them, preventing unacceptable consequences for the plant or the environment. The transients of categories 2, 3 and 4 for a single EPR unit are listed in the following tables. The transients for the spent fuel and ILW interim storage facilities will be developed during the design of these facilities.
813. The PCC design basis transients consider a number operating conditions or ‘states’ which are summarised below.
State A. Power states and hot and intermediate shutdown (P>130bar). In these shutdown states, all the necessary automatic reactor protection functions are available as in the power state. State B. Intermediate shutdown above 120°C (P<130bar). State B covers all shutdown states during normal plant operation, where primary heat is removed by the steam generators. State C. Intermediate and cold shutdown with safety injection system/residual heat removal. State D. Cold shutdown with the reactor cooling system open so that the steam generators cannot be used for decay heat removal. State E. Cold shutdown with the reactor cavity flooded for refuelling. State F. Cold shutdown with the core fully unloaded.
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6.1.4.1. Category 2 events (PCC-2): design basis transients 814. The Category 2 transients studied in the PCSR are listed in Table 6.1.
Table 6.1 Category 2 events (PCC-2): design basis transients
Design basis transients with internal causes(43) Main feed water system malfunction causing a reduction in feed water temperature Main feed water system malfunction resulting in an increase in the feed water flow rate Excessive increase in secondary steam flow Spurious turbine trip Loss of condenser vacuum Short-term loss of off-site power (≤ 2 hours) Loss of normal feed water flow (loss of all main feed water system pumps, start-up and shutdown pump) Partial loss of core coolant flow (loss of one reactor coolant pump) Uncontrolled rod cluster control assembly bank withdrawal at power Uncontrolled rod cluster control assembly bank withdrawal from hot zero power conditions Rod cluster control assembly misalignment up to rod drop, without control system action Start-up of an inactive reactor coolant loop at an incorrect temperature CVCS malfunction resulting in boron concentration in the reactor coolant CVCS malfunction causing increase or decrease in reactor coolant inventory Primary side pressure transients (spurious operations of pressuriser sprays or heaters) Uncontrolled reactor cooling system level drop (states C, D) Loss of one cooling train of the safety injection system/residual heat removal system in residual heat removal mode (states C, D) Loss of one train of the fuel pool cooling system or of a supporting system Spurious reactor trip (state A)
6.1.4.2. Category 3 events (PCC-3): benchmark incidents 815. The Category 3 incidents studied in the PCSR are listed in Table 6.2.
(43) Where the status of the reactor is not shown, the event is assumed to be analysed for an initial power state
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Table 6.2 Category 3 events (PCC-3): benchmark incidents
Design basis incidents with internal causes(44) Small steam or feed water system piping failure (DN<50), including break of connecting lines to a steam generator (DN<50)(45) (states A, B) Long-term loss of off-site power (>2 hours) Inadvertent opening of a pressuriser safety valve (state A) Inadvertent opening of a steam generator relief train or a safety valve (state A) Small break Loss Of Coolant Accident (LOCA) (≤ DN 50), including a break on the extra boration system injection line (states A, B) Steam Generator Tube Rupture (1 tube) Inadvertent closure of one or all main steam isolation valves Inadvertent loading of a fuel assembly in an incorrect position Forced reduction of reactor coolant flow (4 pumps) Leak in the gaseous or liquid waste processing systems Uncontrolled rod cluster control assembly bank withdrawal (states B, C, D) Uncontrolled single control rod withdrawal Long term loss off-site power, fuel pool cooling aspects (state A) Loss of one train of the fuel pool cooling system or supporting system (state F) Isolable piping failure on a system connected to the fuel pool
6.1.4.3. Category 4 events (PCC-4): benchmark accidents 816. The Category 4 accidents studied in the PCSR are listed in Table 6.3.
Table 6.3 Category 4 events (PCC-4): benchmark accidents
Design basis accidents with internal causes(46) Long term loss of off-site power in state C (>2 hours) Main steam line break Feed water system pipe break Inadvertent opening of a steam generator relief train or of a safety valve (state B) Spectrum rod cluster control assembly ejections Intermediate or large break LOCA (up to surge line break – states A, B) Small break LOCA (≤ DN 50), including a break on an extra boration system injection line (states C, D) Reactor coolant pump seizure (locked rotor) Reactor coolant pump shaft break Steam Generator Tube Rupture (2 tubes in 1 steam generator) Fuel handling accident Boron dilution due to a non-isolable rupture of a heat exchanger tube Rupture of systems containing radioactivity in the nuclear auxiliary building Isolatable safety injection system break (≤ DN 250) in residual heat removal mode (states C, D) Non-isolatable small break (≤ DN 50) or isolatable safety injection system break residual heat removal mode (≤ DN 250), fuel pond drainage aspects (state E)
(44) Where the status of the reactor is not shown, the event is assumed to be analysed for an initial power state (45) DN - Normal diameter in mm (46) Where the status of the reactor is not shown, the event is assumed to be analysed for an initial power state
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6.1.5. Multiple failure accidents under risk reduction category A (RRC-A) 817. In addition to examining the incidental and accidental situations with simple initiating events, the scope of the analysis is extended to situations involving multiple failures based on probabilistic safety evaluation. The purpose of studying operating conditions involving multiple failures is to define specific measures, which may be manual actions intended to limit the risks of core melt associated with these scenarios. The accidental transients associated with RRC-A multiple failures, studied in the PCSR are listed in Table 6.4.
Table 6.4 Category RRC-A internal accidental transients
Accidental transients with internal causes Anticipated transient without scram through the mechanical blocking of the rods (state A) Anticipated transient without scram due to the failure of the protection system (state A) Total loss of off-site power and failure of the four main diesel generators (state A) Total loss of the water supply to the steam generators (state A) Total loss of the cooling chain and failure of the stand still seal system leading to a loss of primary coolant (state A) LOCA up to 20cm2 and failure of the protection system for the safety injection signal activation (state A) LOCA up to 20cm2 without medium head safety injection (state A) LOCA up to 20cm2 without low head safety injection (state A) Uncontrolled level drop and failure of the protection system for the activation of the safety injection system (state Cb and D) Total loss of the cooling chain (state D) Total loss of the ultimate heat sink for 100 hours (states A to C) LOCA without medium head safety injection (states C and D) LOCA outside containment on safety injection system and residual heat removal system train (states C and D) LOCA outside containment on safety injection system and residual heat removal system train and failure of the automatic isolation signal (states C and D) Non-isolatable homogenous boron dilution outside the volume control tank and failure of the operator's actions (states Cb and D) Loss of the two main trains of the spent fuel pool cooling system during core refueling (state F)
6.1.6. Core melt accidents under category RRC-B 818. The purpose of some specific safety improvements made to the EPR is to reduce the risk of core melt accidents involving perforation of the reactor vessel, to one tenth of that associated with the existing reactors, for which the risk is already extremely low. The risk of such an event occurring is estimated overall for the EPR using a probabilistic approach, at:
1 in 100,000 per reactor per year when taking into account all the reactor states and all types of event (internal events, internal and external hazards). 1 in 1,000,000 per reactor per year when taking into account internal events only, i.e. with internal and external hazards excluded.
819. The practical measures contributing to this reduction in risk are, for example:
The physical separation of important safety systems into four compartments, to increase the reliability of these systems. The installation of a borated water tank inside the reactor building. Improvements to protection measures for the main external hazards (aircraft crashes, earthquakes, extreme temperatures, etc.). Improved diversification of support systems (diversified main diesel generators, diversified ultimate heat sink, etc.).
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An optimised human-machine interface, based on information from the most recent unit commissioned in France.
820. A core melt accident is still taken into account at the design stage even though the probability is extremely low. Measures are employed to prevent any premature failure of the containment, manage low pressure meltdown scenarios and reduce any associated impacts. The specific aim is to eliminate the need to evacuate any population beyond the immediate vicinity of the power station, even in the event of a severe accident. This aim is achieved by the presence of a molten fuel spreading area, installed below the reactor vessel to enable cooling and a metal lined containment installed to minimise accidental releases to the environment.
821. The low pressure core meltdown scenarios studied in the PCSR cover all water loss events which could lead to exposure of the core and subsequent damage. Since safety injection is unavailable, the progress of an accident is dictated by the size of the break of the reactor coolant system if there is one and the heat removal capacity of the steam generators.
6.1.7. Consideration of additional safety related scenarios 822. The scenarios considered above relate to the presence of a single EPR unit. Hinkley Point C will have two EPR units, a spent fuel interim storage facility and an ILW interim storage facility. The presence of these additional nuclear facilities has been considered to determine the degree to which they influence the identification and selection of reference accidents for the purpose of this submission. The current assumption is that, in terms of design basis accidents, any release from an accident in these additional facilities and that any release associated with an interaction between these facilities in an accident scenario, is bounded by the RRC-B reference accident for a single EPR unit. This assumption has been developed using the arguments presented in Section 6.1.7.1, Section 6.1.7.2 and Section 6.1.7.3.
6.1.7.1. Presence of two EPR units on the site 823. The inventory of nuclear fuel, spent nuclear fuel and waste for two units will be double that for a single unit. The first of a series of PCSRs (PCSR1) prepared for Hinkley Point C does not identify any design basis accidents that relate to both units. The interactions between the two EPR units have been explored during preliminary probabilistic safety assessments that were undertaken to support the preparation of PCSR1. These assessments show that interactions between the two EPR units do not make a significant contribution to the overall risk associated with the Hinkley Point C site and that the numerical targets identified in Section 6.1.1 will be met.
824. Interaction between the two units will be subjected to additional safety analysis and risk assessment during the further development of the safety report. These developments of the safety report will provide a comprehensive demonstration that the numerical targets will be met.
6.1.7.2. Presence of interim storage facilities for spent fuel and ILW 825. The spent fuel and ILW interim storage facilities are in the early stages of development. The proposed inventories for these facilities are presented in Chapter 5 of this document.
826. Safety reports for these facilities have yet to be prepared. The safety and design approach and assumptions to be applied for these facilities and their auxiliaries, will be consistent with those developed for the EPR unit, including the ALARP requirement. Design basis accidents will be identified and assessed during the production of the safety case. Application of comparable design principles to the ones detailed in Section 6.1.1 during the design of the facilities and the development of their associated safety cases is expected to identify design basis accidents with potential consequences that are broadly similar to those for a single EPR unit. The current assumption is that accidental releases from these facilities will be enveloped by those identified for an EPR unit. Key considerations developed to support this assumption in relation to a spent fuel interim storage facility are:
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Radioactive decay. The fuel will have spent up to ten years, but usually around three years, in the fuel pool associated with the nuclear reactor in which it was used. This will result in a substantial reduction of some of the radionuclides present in the spent fuel. In particular isotopes of iodine and noble gases, except krypton-85, will have decayed almost completely. The amount of radioactivity released in the event of a fuel handling accident in this building is expected to be considerably less than for fuel that has just been removed from one of the reactors. Cooling. The fuel will have cooled considerably during the time spent in the fuel building pool. This will limit the impacts of any effects that are driven by heat. Reactivity. The quiescent nature of storage ensures that there is no opportunity for the generation of further fission and activation products by a nuclear reaction. Storage conditions. The fuel is stored in water that is continuously circulated. Any excess heat is removed. There are also opportunities to remove radionuclides that may be present in the water.
6.1.7.3. Interactions between facilities in accident scenarios 827. The methodology adopted for the preparation and evolution of the safety report for Hinkley Point C is not expected to result in the identification of additional design basis accidents related to multiple facilities on the site. Interactions have been explored during preliminary probabilistic safety assessments.
828. The Hinkley Point C safety report will be updated in due course to include the findings of safety reports for the two interim storage facilities. Comprehensive safety analysis and risk assessment will be undertaken at this stage to demonstrate that the numerical targets identified in Section 6.1.1 will be met.
6.2. REFERENCE ACCIDENTS TAKEN INTO CONSIDERATION 829. Representative operating conditions, from the point of view of radiological consequences, are selected from the design basis accidents studied in the PCSR. These are based on the initial conditions and the limit conditions such as the discharge type, path, height and operating modes.
830. For atmospheric discharges, the different possible locations of leaks inside the plant were considered (containment, safeguard buildings, fuel building, nuclear auxiliary building, main steam and feed water systems, steam generators, etc.) in order to select representative cases.
831. These considerations have resulted in the selection of three PCC-4 design basis accidents and one RRC-B accident as the reference accidents. The accidents have been selected on the basis of their radiological consequences. The scenarios and associated assumptions related to each accident are presented below.
832. The reference accidents selected reflect the design basis that is defined for the EPR in the UK regulatory context. The RRC-B accident is considered bounding for accidents that inform the design basis. The reference accidents selected are therefore considered to remain appropriate following the conclusions of Dr Weightman’s Interim Report.
833. For liquid waste, the precautions taken to ensure total containment in the event of an accident are described in Section 2.5. Accidental releases of radioactive liquid waste into aquatic environments are therefore excluded from any further assessment.
834. The radiological consequences of the design basis accidents studied below, therefore relate only to accidental atmospheric radioactive discharges.
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6.2.1. Category 4 (PCC-4) accidents
6.2.1.1. Main hypotheses 835. The main hypotheses used to evaluate the radiological consequences of accidents are as follows:
Activities taken into account in the primary circuit and secondary circuit water Primary circuit water activity
836. The activity selected for primary circuit water is based on the maximum values adopted for the technical specifications for all French nuclear power plants in operation, equal to:
primary circuit activity in stable operation: 20GBq t-1 equivalent to iodine-131(47); and primary circuit activity after power transient (iodine spiking): 150GBq t-1 equivalent to iodine-131.
Secondary circuit water activity
837. The maximum water activities on the secondary side of the steam generators are calculated based on the following hypotheses:
Primary circuit water activity in the unit, corresponding to the maximum values specified in the technical operating specifications. A primary-secondary leakage rate of 20l h-1. A drainage rate from all of the steam generators corresponding to the plant operating at nominal power. Drive factors in the steam from the steam generators corresponding to the values detailed below.
Drive factors in steam from the steam generators 838. The drive factors taken into account for the transfer of activity during the steam generator steam phase are as follows:
All noble gases present in the steam generator water are assumed to be transferred in the gaseous phase. For other radionuclides, ‘healthy’ steam generators are distinguished from damaged steam generators (due to a steam generator tube rupture, for example). For “healthy” steam generators, the drive factor taken into account is 0.25%. For damaged steam generators, the drive factor taken into account is 1%.
Release of activity in the event of cladding failure 839. During some accidents (particularly LOCAs), the fuel cladding is subject to a thermohydraulic transient, which can cause it to fail. The fission product activity, accumulated as a result of pellet/cladding gap, can then be released into the primary system.
840. The release rates for the proportion of fission products in the fuel rod inventory assumed to be discharged into the system when the cladding fails are presented here. Information is provided for both uranium oxide and mixed-oxide (MOX) fuel. The use of MOX fuel in the UK EPR has not been justified by the UK Government. It is included here because it is used in assessments to provide a more conservative estimate of impact.
(47) Iodine-131 equivalent = I131 + I132 ÷ 30 + I133 ÷ 4 + I134 ÷ 50 + I135 ÷ 10
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841. The overall release rates selected, based on the considered fuel assemblies, are as follows:
Table 6.5 Activity release rate in the event of cladding failure
Selected burn-up for the evaluation Selected burn-up for the evaluation Elements of discharge – UO2 fuel of discharge – MOX fuel ≤47 GWd tU-1 > 47 GWd tU-1 ≤ 33 GWd tU-1 > 33 GWd tU-1 Krypton-85 8% 25% 8% 50% Other noble gases 2% 8% 2% 15% Bromine, rubidium, 2% 8% 2% 15% iodines, caesium
Deposition of fission products 842. The physical laws on aerosol and molecular iodine deposits in the containment take into account an exponential decay law, the deposition constants of which are equal to 0.035h-1 and 0.014h-1 respectively.
Leakage rate from the containment vessel 843. The overall leakage rate through the containment inside the EPR (which has a metal liner) is 0.3% volume per day at the design pressure (5.5bar).
Filter performance 844. The retention performances of the extraction filters used to reduce radioactive waste are as follows:
845. High efficiency filters:
Noble gases 0% Aerosols (including particulate iodine) 99.9% All other substances 0%
846. High efficiency filters and iodine traps:
Noble gases 0% Iodine in organic form 99% Elementary iodine 99.9% Aerosols (including particulate iodine) 99.9%
6.2.1.2. Specific hypotheses relating to the accidents under consideration 847. In addition to the general hypotheses described above, particular hypotheses relating to the accidents examined are presented below.
Large primary circuit break during operation at nominal power 848. The Category 4 LOCA is defined as a rupture in the safety injection line in the nozzle of the primary coolant system’s cold leg.
849. During this accident, it is assumed that the reactor core becomes uncovered, with a 10% break in the fuel rod assembly. The discharge considered for this accident is a result, on the one hand, of leaks in the containment vessel and on the other hand, of leaks assumed to occur in the reactor core cooling systems outside the containment vessel, in the ventilated and filtered safeguard auxiliary buildings.
Rupture of two steam generator pipes 850. The radiological consequences of this accident are the result of the release of activity to atmosphere, via the atmospheric steam dump valves on a faulty steam generator. The
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activity is due to contamination of the secondary circuit by the primary circuit through a break in the steam generator tubes.
851. The peak of activity in the primary circuit (iodine spiking), caused by the transfer of activity due to pellet/cladding gap in the primary coolant following automatic reactor shutdown, is considered for the assessment of the radiological consequences of this accident. To create a more conservative scenario, it is assumed that the iodine spiking is fully developed when emergency shutdown occurs.
Fuel handling accident 852. The accident examined involves a fuel assembly with a maximum irradiation, dropped into the spent fuel pit in the fuel building. All fuel rods in the damaged assembly are assumed to be broken.
853. The cooling time for the damaged assembly is 60 hours, corresponding to the minimum time required between reactor shutdown and the start of fuel handling.
854. The release rates considered are those provided in Section 6.2.1.1.
855. It is assumed that the radioactive isotopes released from the pit in the hall are distributed immediately and evenly throughout the entire free volume of the hall.
856. The automatic closure command for the main air extraction system’s cut-off devices enables the activity to be contained by switching to reduced ventilation and iodine traps. This command is activated by means of a high activity measurement signal, which is located on the operating floor for the irradiated fuel pit.
6.2.2. RRC-B operating conditions - core melt accident 857. This extreme type of accident, which was not considered in the design of the existing nuclear reactors, is taken into account in the design measures specific to the EPR. The associated radiological consequences are analysed to ensure compliance with objectives in terms of population protection.
858. The design of the EPR is such that the risk of a core meltdown is extremely low. As part of the implementation of an improved defence in depth philosophy, low pressure core melt accidents constituting RRC-B operating conditions, have been addressed by means of specific design features that aim to ensure that the integrity of the containment is maintained and the release of radioactive products outside the plant remains within prescribed limits.
859. An examination of RRC-B operating conditions shows that, given the design features adopted, the following radiological objectives associated with these situations are met:
Limited sheltering; No need for emergency evacuation outside the immediate vicinity of the plant; No permanent relocation; and No long term restrictions on the consumption of foodstuffs.
860. The dose levels to be assumed for these different protective measures are as follows:
Short-term measures: o Requirement for sheltering: 10mSv (effective dose); o Evacuation: 50mSv (effective dose); and o Distribution of iodine tablets: 100mSv (equivalent dose to the thyroid).
Medium and long term measures: o Relocation: 10mSv per month for prolonged exposure (dose rate due to ground contamination) or 1Sv (effective dose).
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861. Any restrictions concerning consumption of foodstuffs produced in the vicinity of the plant are governed by relevant European marketing regulations applicable in the event of a nuclear accident or other radiological emergency.
862. Exposures beyond the local population are further discussed in Section 6.3.1.6.
6.2.2.1. Source term 863. A benchmark source term has been defined, based on reasonably conservative disconnection hypothesis which are independent of the accident scenario. The main hypotheses are as follows:
A 100% core melt is assumed. The radionuclide release rates, in terms of the radiological consequences on populations (noble gases, iodine and caesium) have been maximised (100% release into the containment vessel). The quantity of suspended aerosols in the containment falls due to natural deposition. The effectiveness of the containment spray systems has not been taken into consideration. Iodine is mainly released into the containment in the form of aerosols. A fraction of suspended organic iodine equal to 0.15% is taken into account from the beginning of the accident. This value has been used to bound the quantity measured in the long term phase of the accident at Three Mile Island.
864. During a RRC-B type accident, the integrity of the EPR containment is guaranteed by specific provisions. This special design justifies the use of hypotheses developed for PCC accidents to assess discharge into the environment:
An internal containment leakage rate of 0.3% volume per day (maximum internal containment leakage rate at its absolute pressure and design temperature). Filtration, downstream from the ventilation, which allows 99.9% of the aerosols and elementary iodine and 99% of the organic iodine to be retained. Noble gases are not filtered.
865. Given the conservative assumptions made, this source term covers RRC-B type accident sequences.
6.2.2.2. Fraction discharged 866. The EPR source term, calculated using these hypotheses, is shown in Table 6.6 below. It is expressed as a percentage of the fractions discharged, in comparison with the initial core activity for a certain number of radionuclides (total activity discharged, taking radioactive decay into account).
Table 6.6 Fraction of radionuclides discharged
Source Term Radionuclides (% initial core inventory) Xenon-133 1.50E+00 Iodine-131 6.10E-01 Caesium-137 7.00E-06 Tellurium-132 5.10E-06 Strontium-90 1.30E-06 Ruthenium-106 2.60E-07 Cerium-141 2.60E-07 Plutonium-241 4.60E-08
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6.3. EVALUATION OF THE RADIOLOGICAL CONSEQUENCES OF THE REFERENCE ACCIDENT(S)
6.3.1. Release to atmosphere
867. An assessment of the radiological consequences of the following design basis accidents (PCC-4) is presented:
Fuel handling accident; Steam generator tube rupture; and LOCA.
868. These three design basis accidents are considered, as they present the most significant consequences to the local reference group in terms of radiological impact.
869. In addition, a severe accident scenario (RRC-B), based on a core melt accident is also assessed.
870. The assessment considers the releases to atmosphere to reference groups in the vicinity of the facility, the Channel Islands and to the nearest member state, France. The same locations are used as identified in the routine gaseous releases assessment. The local reference group is assumed to be 2km from the facility. The French reference group is 186km from the site on a bearing of 152º from north. The Channel Islands reference group is 178km from the Hinkley Point C site on a bearing of 158º from north.
6.3.1.1. Assumptions used to calculate the releases to atmosphere 871. The assessment of impacts to the local representative group in the vicinity of the facility, uses the full radionuclide inventory of the assessed releases. The long range assessment takes a simplified approach assessing only key radionuclides. These key radionuclides include the activity contributions of the other radionuclides but they are modelled as a single radionuclide. Table 6.7 identifies the key radionuclides used to represent radionuclide groups. The long range assessment also makes the assumption of a uniform release rate over the duration of the release.
Table 6.7 Key radionuclides and other radionuclides considered
Assessed radionuclide Other radionuclides included in group
Krypton-85 Other isotopes of Krypton (e.g. Krypton-85m, Krypton-87, Krypton-88) Xenon-133 Including Xenon-133m Xenon-135 Including Xenon-135m and Xenon-138 Iodine-131 Other isotopes of tellurium iodine (e.g. Iodine-132, Iodine-134, Iodine-135) Iodine-133 - Other long lived beta gamma radionuclides (e.g. Strontium-90, Caesium-138, Caesium-137 Barium-140, Lanthanum-140, Cerium-141, Cerium-143, Praseodymium-143, Cerium-144) Other actinides (e.g. Plutonium-238, Plutonium-239, Neptunium-239, Curium-242, Alpha Curium-244)
6.3.1.2. Release duration 872. The design basis accidents have different release durations from one hour up to seven days. The simple long range model conservatively assumes that the weather conditions remain constant for the duration of the release and for the time the long range reference group is affected by the plume. For events that last more than a few hours it is unlikely that the
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weather patterns will remain constant and in reality changes will result in enhanced dispersion that would result in lower impacts that those presented.
873. The RRC-B could result in a longer duration release, of the order of approximately one month, although the majority of particulate activity is released in the first 48 hours. Therefore the modelling for the severe accident is based on a release duration of 48 hours. This is more realistic in terms of the underlying assumption of constant weather conditions for the duration of the release.
6.3.1.3. Amounts and physico-chemical forms of those radionuclides which are significant from the point of view of health 874. In the event of an accident it is likely that only after the event, will the physico-chemical forms of those radionuclides released actually be determined. However, for the purposes of these assessments, it has been assumed that isotopes of iodine are released in their elemental form. This is a conservative assumption because this results in the use of a higher dose per unit intake value associated with isotopes of iodine, inhaled in an organic vapour form. The accident scenario, presented in the PCSR, assumes a mixture of elemental and aerosol of isotopes of iodine. Other radionuclides released excluding noble gases, are assumed to be in a particulate form. Table 6.8 presents the lung classes(48) used in the assessment for all accidents.
Table 6.8 Lung class used in assessment for key radionuclides
Radionuclide Lung class Iodine-131 V Iodine-133 V Cobalt-58 M Cobalt-60 M Caesium-134 M Caesium-137 F Plutonium-239 M
875. Table 6.9 describes the quantities of significant radionuclides released in each of the accidents assessed.
Table 6.9 Amounts of significant radionuclides assessed in each scenario
Source term released to environment (Bq) Assessed Fuel handling Steam generator radionuclide LOCA RRC-B accident tube rupture Krypton-85 1.64E+14 8.41E+11 7.77E+12 5.70E+16 Xenon-133 2.03E+16 1.08E+14 3.45E+14 1.47E+17 Xenon-135 6.07E+14 2.54E+13 1.79E+13 1.04E+15 Iodine-131 3.20E+10 1.86E+11 3.78E+10 2.92E+12 Iodine-133 3.05E+09 3.56E+11 2.11E+10 1.48E+12 Caesium-137 1.23E+09 4.24E+10 6.15E+09 4.47E+10 Alpha 0.00E+00 0.00E+00 0.00E+00 4.08E+06 TOTAL 2.11E+16 1.35E+14 3.71E+14 2.05E+17
(48) Lung classes, developed by ICRP, are indicative of the rate of clearance of inhaled activity from the pulmonary region of the lung, very fast or vapour (V), fast (F), medium (M) and slow (S) are used to represent the clearance rate.
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6.3.1.4. Models and parameter values used 876. This section outlines the models and parameters used in the assessment of consequences from the defined accident scenarios. Due to the issues associated with dispersion of activity over long distances, such as changes in wind direction and weather conditions, a different model was adopted to that used in the local assessment.
877. Data for the local assessment is taken from information already presented to the UK Regulators as part of the GDA process.
Short range model used for local reference group 878. The short range model is based on that used in the PCSR developed in support of the UK GDA process. The local reference group is assumed to be 2km from the facility.
Local meteorological conditions 879. Depending on the prevailing weather conditions, the dispersion and the deposition of the released radionuclides is subject to a wide variability. Consequently the results of the assessment are sensitive to the meteorological conditions assumed.
880. A probabilistic approach is used. The real weather situations of a representative year (hourly documented weather data including stability, rainfall, wind speed, etc.) from the UK Meteorological Office are used to calculate the doses. This is the same basic data used in the assessment of routine discharges but broken down to consider shorter time periods. Wind direction is included in the probabilistic assessment. Using this approach, the value which covers 95% of the cases is judged to be adequately conservative.
881. The atmospheric dispersion is calculated using a Gaussian model. The models CORRA and ASTRAL are used to assess the initial and longer term impacts respectively.
Deposition data 882. Deposition and washout factors in the model are used to calculate the amount of radioactive substances that would be deposited during dry weather or during precipitation respectively. The fallout and washout factors are not only a function of atmospheric parameters, e.g. wind velocity or precipitation rate, they also depend on the physico-chemical form of the radionuclides. Deposition velocities and the washout coefficients used in the assessment are presented in Table 6.10.
Table 6.10 Deposition parameters used in the local assessment model
Deposition velocity Substance group Washout coefficient (s-1) (m s-1) Elemental iodine 1.0E-02 7.0E-05 Organically bound iodine 1.0E-04 7.0E-07 Aerosols 1.5E-03 7.0E-05
Exposure pathways 883. The following exposure pathways are considered in the dose calculation for design basis accidents.
Gamma radiation from the passing plume. Inhalation of radioactive substances by persons affected by the plume for the time during which the plume passes. Gamma radiation from radioactive substances deposited on to the ground surface. Ingestion of foodstuff contaminated by radionuclides.
884. The exposure period for the radioactive substances deposited on the ground surfaces, as well as the ingestion of foodstuffs, is assumed to be the whole life of the individual. This is 50 years for adults. Due to the changing dietary habits of children, ingestion doses are not calculated beyond the first year. Given that most of the dose is delivered in the first year this
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is seen as being a reasonable approach. The committed effective dose from the intake of radionuclides related to internal radiation exposure, due to inhalation and ingestion, is assumed to be 70 years for infants and 50 years for adults.
885. It is assumed that local food consumption restrictions are in place 24 hours after the beginning of the accident, within a radius of 2km from the release point. It is assumed that the food produced in this area is not used for the first year after the accident. Outside this area no mitigation measures are assumed. Therefore for the ingestion pathway the dose at 2km distance may be greater than the dose at 1km distance.
886. The assessment of food doses is based on the ASTRAL code developed by Institut de Radioprotection et de Sûrete Nucléairé (IRSN)(49) and used by EDF in their assessments. The ASTRAL spreadsheets provide dose per unit deposition for each radionuclide, for adult and infants summed across the assessed foods groups. The ASTRAL foodchain factors are presented in Table 6.11. These are based on consumption of the foods at the rates presented in Annex A, Table A.1. Further information on the ASTRAL food chain model is presented in Annex A.
Table 6.11 Dose per unit deposition via food ingestion (from ASTRAL code) (mSv Bq-1 m-2)
Radionuclide Adult Infant Iodine-131 6.80E-09 1.80E-08 Iodine-133 6.80E-09 1.80E-08 Cobalt-58 4.70E-10 1.30E-09 Cobalt-60 3.90E-09 1.30E-08 Caesium-134 4.30E-08 1.30E-08 Caesium-137 3.30E-08 1.10E-08 Alpha (Plutonium-239) 1.10E-07 9.20E-08
887. In addition, the dose from exposure of the thyroid by the inhalation of radioiodine is assessed.
Dose coefficients 888. The dose coefficients are taken from the ICRP publication 72(50). These factors are consistent with the dose coefficients given in the Council Directive 96/29/Euratom.
Long range model used in the assessment of impacts to reference groups in the Channel Islands and France 889. The long range model used to estimate the consequences to the nearest Member State was based on NRPB-R124(51). This makes some simple but conservative assumptions regarding the travel of a short term release over longer distances, where traditional Gaussian plume models are not appropriate. The model used is a deterministic one, which assumes the wind is blowing towards the assessed critical groups for the duration of the release. In addition, it assumes a constant boundary layer height and wind speed. Parameters used in the assessment are detailed in Table 6.12.
(49) IRSN/DPRE/SERLAB/03-16 Report (September 2003): "Equations et paramètres du logiciel ASTRAL V2.1." (Equations and parameters in the ASTRAL V2.1 software). (50) ICRP72, 1996 (51) Jones, 1981a – NRPB-R124
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Table 6.12 Long range atmospheric dispersion modelling parameters
Parameter/description Value Mixing layer depth 1000m Wind speed 8 s-1
890. Depletion of the plume by radioactive decay and deposition (wet and dry) were included, based on the approach presented in NRPB-R122(52). The parameter values are the same as those used in the long range routine release model presented in Section 3.4.
891. The output of the model is time integrated air concentration (in Bq s m-3) and for depositing radionuclides, surface deposition density on the ground (in Bq m-2).
Exposure pathways 892. As for local reference group.
893. Given the low doses in the vicinity of the facility from the ingestion of foodstuffs, a proportionate approach was taken in determining ingestion doses in the nearest Member States. The dose from the ingestion of contaminated foods was determined by taking the ratio of surface concentration value in the Member State, to the surface concentration value in the vicinity of the facility, multiplied by the local ingestion dose. Surface concentration values are presented in Table 6.14.
Occupancy data 894. The long range model uses the same French and Channel Islands groups as defined in Section 3.4. The same habit data, breathing rates, fraction of time indoors also applies to these representative groups, with the exception that it is assumed there is 100% occupancy and food consumption rates are based on the ASTRAL code.
Dose coefficients 895. The inhalation and ingestion dose coefficients are taken from the ICRP publication 72 using the lung class indicated in Table 1 of the ICRP report. These factors are consistent with the dose factors given in the Council Directive 96/29/Euratom. External dose per unit exposure factors are taken from US EPA Federal Guidance Report 12.
6.3.1.5. Expected levels of radioactive contamination of foodstuffs which might be exported to other affected Member States 896. The assessment of food doses is based on the ASTRAL code developed by IRSN and used by EDF in their assessments. Calculations for the local reference group have conservatively assumed that all the food consumed by the reference group is sourced locally. Data presented in Chapter 1 indicates that the majority of foodstuffs are distributed locally and that there are no foodstuffs produced in the vicinity of Hinkley Point C that are exclusively for export to other Member States and not consumed by the reference group.
6.3.1.6. Maximum time integrated air concentrations and total surface concentrations 897. The maximum time integrated concentrations and the total surface concentrations resulting from the four scenarios for the three groups, for air concentration and surface contamination are presented below in Table 6.13 and Table 6.14 respectively.
(52) Jones, 1981b – NRPB-R122
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Table 6.13 Maximum time integrated air concentrations
Reference group Scenario Local area (at 2km) Channel Islands Member State (Bq s m-3) (Bq s m-3) (France) (Bq s m-3) Fuel handling fault 3.2E+11 5.0E+06 4.8E+06 Steam generator 2 tube rupture 7.2E+09 4.2E+05 4.0E+05 LOCA 5.6E+09 8.8E+04 8.4E+04 Core melt 1.8E+12 1.7E+08 1.7E+08
Table 6.14 Total surface concentrations
Reference group
Scenario -2 Channel Islands Member State Local area (Bq m ) -2 -2 (Bq m ) (France) (Bq m ) Fuel handling fault 1.8E+03 1.2E+00 1.2E+00 Steam generator 2 tube rupture 1.5E+05 1.6E+02 1.5E+02 LOCA 1.1E+04 1.8E+00 1.7E+00 Core melt 9.7E+04 4.2E+02 4.0E+02
6.3.1.7. Corresponding maximum doses 898. The corresponding maximum committed effective doses in microsieverts, to the three identified groups resulting from the four scenarios, are presented below in Table 6.15, Table 6.16, Table 6.17 and Table 6.18 respectively.
899. The results provide a breakdown in terms of pathway (inhalation including inhalation of re-suspended materials, external irradiation from immersion in the plume, external irradiation from material deposited on the ground and ingestion of contaminated foodstuffs).
900. The results include the impacts from the initial passage of the plume and accrued during the first pass of the plume, which includes the inhalation and immersion pathways. The longer term dose is the summation of re-suspension, external and ingestion pathways. The assessment also includes an assessment of thyroid dose from the inhalation of radioiodines.
Table 6.15 Maximum committed effective doses to an adult member of reference groups by pathway for a fuel handling accident
Reference group - adult Pathway Member State Local area (µSv) Channel Islands (µSv) (France) (µSv) Inhalation 1.1E+00 7.3E-05 7.1E-05 Immersion 5.8E+02 4.5E-03 4.3E-03 External 1.5E+01 2.3E-04 2.2E-04 Ingestion 1.8E+01 9.2E-04 8.8E-04 Total 6.1E+02 5.8E-03 5.5E-03 First Pass 5.8E+02 4.6E-03 4.4E-03 Longer term 3.2E+01 1.2E-03 1.1E-03 Short term thyroid dose 2.0E+01 7.8E-04 7.5E-04
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901. First pass infant doses for the fuel handling accident are 5.8E+02µSv, 2.2E-03µSv and 2.1E-03µSv and for the local, Channel Islands and Member State reference groups respectively. Corresponding thyroid doses are 3.0E+01µSv, 1.5E-03µSv and 1.5E-03µSv for the local, Channel Islands and Member State reference groups respectively.
Table 6.16 Maximum committed effective doses to an adult member of reference groups by pathway for the steam generator tubes rupture accident
Reference group - adult Pathway Member State Local area (µSv) Channel Islands (µSv) (France) (µSv) Inhalation 2.4E+01 6.6E-03 6.3E-03 Immersion 1.0E+02 6.6E-04 6.3E-04 External 1.3E+03 1.0E-01 1.0E-01 Ingestion 7.2E+02 7.5E-01 7.2E-01 Total 2.2E+03 8.6E-01 8.3E-01 First Pass 1.2E+02 4.3E-03 4.1E-03 Longer term 2.0E+03 8.5E-01 8.2E-01 Short term thyroid dose 3.7E+02 6.7E-02 6.5E-02
902. First pass infant doses for the steam generator tube rupture are 1.3E+02µSv, 7.1E-03µSv and 6.8E-03µSv for the local, Channel Islands and Member State reference groups respectively. Corresponding thyroid doses are 6.4E+02µSv, 1.4E-01µSv and 1.3E-01µSv for the local, Channel Islands and Member State reference groups respectively.
Table 6.17 Maximum committed effective doses to an adult member of reference groups by pathway for the LOCA
Reference group - adult Pathway Channel Islands Member State (France) Local area (µSv) (µSv) (µSv) Inhalation 1.2E+01 9.1E-05 8.7E-05 Immersion 1.4E+01 8.5E-05 8.2E-05 External 6.7E+01 1.1E-03 1.1E-03 Ingestion 4.4E+01 5.6E-04 5.3E-04 Total 1.4E+02 1.8E-03 1.8E-03 First Pass 2.7E+01 1.4E-04 1.3E-04 Longer term 1.1E+02 1.7E-03 1.6E-03 Short term thyroid dose 1.9E+01 9.4E-04 9.0E-04
903. First pass infant doses for the LOCA are 2.1E+01µSv, 1.3E-04µSv and 1.3E-04µSv for the local, Channel Islands and Member State reference groups respectively. Corresponding thyroid doses are 3.1E+01µSv, 1.9E-03µSv and 1.8E-03µSv for the local, Channel Islands and Member State reference groups respectively.
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Table 6.18 Maximum committed effective doses to an adult member of reference groups by pathway for a core melt accident
Reference group - adult Pathway Member State Local area (µSv) Channel Islands (µSv) (France) (µSv) Inhalation 1.3E+02 2.4E-02 2.3E-02 Immersion 2.3E+03 1.1E-01 1.0E-01 External 2.9E+02 3.2E-02 3.1E-02 Ingestion 3.5E+02 4.7E-01 4.5E-01 Total 3.1E+03 6.3E-01 6.1E-01 First Pass 2.4E+03 1.2E-01 1.2E-01 Longer term 6.4E+02 5.1E-01 4.9E-01 Short term thyroid dose 2.4E+03 2.5E-01 2.4E-01
904. First pass infant doses are 2.5E+03µSv, 7.6E-02µSv and 7.3E-02µSv for the local, Channel Islands and Member State reference groups respectively. Corresponding thyroid doses are 3.7E+03µSv, 5.0E-01µSv and 4.8E-01µSv for the local, Channel Islands and Member State reference groups respectively.
Results 905. The maximum effective doses arising from a single initiating event from a design basis accident, arises from the rupture of steam generator tube accident scenario. This results in a total effective dose to a reference person in the nearest Member State of 8.2E-01µSv to an adult. The dose to the Channel Islands reference person is slightly higher. The same scenario resulted in the highest thyroid dose as well of 1.3E-01µSv to an infant in the reference group assessed in the nearest Member State.
906. The highest dose from the severe accident scenario (RRC-B), results in a total effective dose to a reference person in the nearest Member State of 6.1E-01µSv to an adult. The dose to the Channel Islands reference person is slightly higher. The same scenario resulted in the highest thyroid dose as well of 4.8E-01µSv to an infant in the reference group assessed in the nearest Member State. The increased source term associated with the severe accident is offset by the increased dispersion associated with such a long release.
6.3.2. Release into an aquatic environment 907. Given the precautions, described in Chapter 2 that are taken to ensure that the containment remains sealed, no design basis accident has been identified that would cause the continuous discharge of waste into the aquatic environment. Indeed the base mats of the lower part of the nuclear buildings provide a barrier to protect the environment from contamination due to radioactive liquid spills or leaks.
908. In the event of liquid leaks in the nuclear island buildings, the radioactive effluent is recovered by sumps, retention pits and retention tanks. With regard to the reactor building, the base mats are sealed by specific measures, which are a membrane below the base mat, a coating on the lowest floors, a sealed lining in the fuel pool and the sealing of the molten core spreading area. All of these measures protect the containment function of the base mat, even in a severe accident situation. Molten core cooling is also monitored to follow up the impacts on the containment.
909. Finally, a leak in the safety injection or the residual heat removal systems, both of which transport contaminated water outside the reactor building to cool the core or the containment in accident situations, is taken into account in the design by installing contaminated waste re- injection systems in the reactor building.
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7. EMERGENCY PLANS, AGREEMENTS WITH OTHER MEMBER STATES
7.1. INTERVENTION LEVELS ESTABLISHED FOR DIFFERENT TYPES OF COUNTERMEASURES 910. Actions in an emergency situation will be based on well considered pre-determined and accepted principles. A key element of this for a nuclear emergency, is the response necessary for likely levels of radiation doses for those persons on the site, those dealing with the emergency and the members of the public. In the UK the Health Protection Agency is the independent body with the responsibility for specifying and giving advice on emergency reference levels for the public. The Health Protection Agency also gives other guidance for return and relocation.
911. The Emergency Reference Levels (ERLs) are levels of 'dose saved' at which it is justifiable to introduce countermeasures. In recommending any ERL, the Health Protection Agency balances the risk from potential radiation exposures, against the risks which may be associated with the countermeasure. The ERLs are formulated in a two-tier system of dose levels of dose saved for the public. The lower levels of dose saved have been recommended as being, levels below which countermeasures should not, in general, be taken because the conventional risks and social disruption resulting from the countermeasures are likely to outweigh the benefits. The upper levels of dose saved have been recommended as being, those at which action should almost certainly be taken.
912. At values between these upper and lower bounds of ERL, the implementation of countermeasures is desirable but not essential and must be considered in the light of the situation at the time. The application of the ERL is aimed at ensuring that risks to the health of individuals are minimised. If the response is based on the ERLs, any resulting health effects would be small and would not subsequently be distinguished from the normal incidence of such effects. The ERLs are subject to a continuing review to reflect developments in the understanding of radiation risks.
913. In drawing up and developing emergency plans the ERLs, together with the predictions of the course of the accident and the likely effectiveness of the countermeasures, are used to define site-specific intervention levels of dose saved. The intervention levels of dose saved, expressed in directly measurable quantities, are used to provide advice on possible protective actions. The advice is given to the Police who carry the responsibility for instigating the necessary measures taking account of the local situation at the time.
914. The recommendations and advice provided by the Health Protection Agency cover the following countermeasures and actions:
Sheltering - the public would be advised to stay indoors, close doors and windows and follow advice given by local radio and television stations. Sheltering reduces the risk of exposure to direct radiation and the inhalation of radioactive material. The Health Protection Agency has specified the following radiation effective dose levels of dose saved for this counter measure:
Lower Emergency Reference Level 3mSv. Upper Emergency Reference Level 30mSv.
Taking of stable iodine tablets - potential consequences from postulated accidents at nuclear installations, are often dominated by the effects of radioactive iodine because of its relatively high volatility. Taking stable iodine tablets can significantly reduce the iodine uptake by the body and thus reduce the likely radiation dose. The Health Protection Agency has specified the following radiation equivalent dose levels of dose saved to the thyroid for the introduction of this counter measure:
Lower Emergency Reference Level 30mSv. Upper Emergency Reference Level 300mSv.
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Evacuation - this is an important counter measure as it removes the person from further exposure. It is however socially disruptive and incurs other risks. The Health Protection Agency have specified the following radiation dose levels of dose saved:
Lower emergency reference level 30mSv. Upper emergency reference level 300mSv(53).
7.1.1. Control of contaminated or potentially contaminated food supplies 915. The limits taken into account for judging the possibility of foodstuff restrictions, are taken from Regulations from the Council of the European Communities which specify intervention levels for radioactive contamination in marketed foods and animal feeds (here termed the Council Food Intervention Levels). These Council Food Intervention Levels are legally binding on the UK in the event of an accident.
7.1.2. Permanent relocation 916. It is considered that the criteria recommended by the ICRP, IAEA and the Article 31 Working Group for relocation, are appropriate for application in the UK. These criteria include a 10mSv per month criterion as an upper bound for determining the need for relocation and in practice, is likely to be most useful in the context of contamination dominated by radionuclides with half-lives of the order of days, such as iodine-131. In addition a 1Sv lifetime effective dose criterion, is recommended as an upper bound applicable to all forms of recovery countermeasures. These criteria should apply to the sum of doses received and/or committed, during both the emergency and recovery phases as a result of an accident(54).
7.2. UK EMERGENCY PLANNING 917. There are well established processes, guidance and legislation within the UK for the development of emergency arrangements for civil nuclear power stations. This aims to protect the public through robust response arrangements on and off-site, and the management of preparedness to ensure the arrangements are in a constant state of readiness. In the UK a Site Incident is defined as a hazardous condition which is confined in its effect to within the boundary of the nuclear licensed site. An Off-Site Nuclear Emergency is a hazardous condition which results, or is likely to result, in the need to consider urgent countermeasures to protect the public outside the site security fence from a radiological hazard.
918. The prevention and minimisation of radiation exposure from the accidental release of radioactive substances is the main focus of the emergency arrangements, however they will also provide the capability to respond to a full range of accidents on the site.
919. The adequacy of the arrangements will be developed and reviewed by:
assessing the potential consequences of an incident to ensure that the response is a proportionate approach to cover a range of events including worst case scenarios; developing intervention actions which will mitigate the event and reduce the harmful exposure to persons and the environment; ensuring personnel assigned roles in the emergency response and facilities and equipment used to support the response, are developed to fulfil their assigned function; ensuring that when support is required from external emergency services and agencies, consultation will provide alignment of actions to allow all organisations to fulfil relevant statutory obligations; and developing and evolving the current emergency response framework which would provide the model for Hinkley Point C to include utilising experience from past events national
(53) HSE, The UK’s National Report on Compliance with the Convention on Nuclear Safety Obligations, Revision 1, August 1998 (54) NRPB 1997, Intervention for Recovery after Accidents, Volume 8 No. 1, Pages 13-14
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and international guidance.
920. The UK’s emergency response framework includes:
focused site activities to remediate the plant fault; technical support to the site and advice to emergency services and agencies; coordination of local public protection and information; and support from central government to assess the impact nationally and internationally.
921. Consolidated Guidance has been produced by the Nuclear Emergency Planning Liaison Group. The purpose of the Consolidated Guidance is to bring together guidance into one document for general reference by planners and practitioners concerned with emergency response at nuclear sites(55). This guidance will be followed where applicable, when putting together the emergency plan for Hinkley Point C. The emergency response plan when developed, will take into consideration this guidance and will align with current UK arrangements for an off-site nuclear emergency response.
922. Hinkley Point C will work with and exchange information with local agencies and emergency response organisations to allow them to create their local and regional emergency response plans. The agencies and organisations can use the Nuclear Emergency Planning Liaison Group guidance to produce their emergency plan.
7.2.1. Regional level: authorities 923. The Civil Contingencies Act 2004 and Radiation (Emergency Preparedness and Public Information) Regulations (REPPIR), define a focus on local arrangements for civil protection, establishing a statutory framework of roles and responsibilities for local ‘category one’ responders. These duties include the duty to assess the risk of an emergency occurring and to maintain plans for the purposes of responding to an emergency(56).
924. Somerset County Council has produced and maintains the Hinkley Point Off-Site Plan in accordance with REPPIR. In the event of an accident, the designated off-site support facility will be set up in addition to site emergency centres, a company off-site support facility and strategic co-ordination centre (the local emergency management facility), and the Off-Site Plan will be activated. The plan includes five zones which are defined and explained in the paragraphs below.
7.2.2. Emergency planning zones 925. The ONR have designated an area in which detailed contingency plans should be produced for every licensed installation which allow for a rapid response to an emergency. This area is called the Detailed Emergency Planning Zone (DEPZ).
926. The DEPZ is currently set at 3.5km radius around Hinkley Point for which detailed plans are required, with outline planning required from the edge of the DEPZ, out to a distance of 15km. The area around Hinkley Point is divided in to 16 sectors with a series of concentric zones of increasing distance as follows:
Zone A: 0 - 3.5km DEPZ - Reference Accident Scenario; Zone B: 3.5 - 5km; Zone C: 5 - 10km - Extended Release Scenario; Zone D: 10 - 15km; and (57) Zone E: 15 - 40km - Food/Water Restrictions .
927. The current DEPZ for Hinkley Point was originally set for Hinkley Point A. This Magnox reactor was designed with reactor gas pipework which was transferred by ducts from the
(55) Nuclear Emergency Planning Liaison Group: Consolidated Guidance, Jan 2009, Chapter 1, Page 2 (56) Nuclear Emergency Planning Liaison Group: Consolidated Guidance, Jan 2009, Chapter 4, Page 3 (57) Somerset County Council, Hinkley Point Essential Services Off-Site Plan, April 2008, Page 5
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reactor building to the external boiler houses. If these pipes were breached, there may have been a widespread release to the atmosphere. Neither Hinkley Point B nor Hinkley Point C have this design feature.
928. The DEPZ for Hinkley Point is currently being reviewed as Hinkley Point A is no longer operational and is being decommissioned, thus the DEPZ for Hinkley Point could possibly be reduced. The future DEPZ for Hinkley Point is being discussed at the Nuclear Emergency Arrangements Forum(58) and is yet to be agreed.
7.2.3. National level: site operator 929. In the early phase of an off-site nuclear emergency, the operator of the Nuclear Licensed Site is responsible for advising the Police on the potential radiological hazard off-site and making recommendations on the counter measures which should be implemented to protect the public. When the designated off-site support facility is fully operational, it will take over these responsibilities, assessing all necessary technical data that has a bearing upon the radiological hazard to the public and giving clear expert advice based upon that technical assessment to enable informed and timely decisions on the need to take action to protect the public(59).
7.2.4. National level: authorities 930. The designated off-site support facility, once fully operational will provide a technical support service to the affected site and act as the focal point for routing advice and material assistance to the affected site. It will be responsible for the onward transmission of monitoring results and the outcome of radiological assessments to external agencies such as the Health Protection Agency and the Food Standards Agency, and for the supply of information to the Senior Management of the site operator. An appointed Government Liaison Officer will attend the designated off-site support facility and will provide a communication channel between the facility and the Department of Energy and Climate Change (DECC).
931. DECC aims to ensure it is fully equipped and prepared to respond, in the unlikely event of an emergency at a civil nuclear site in England and Wales. DECC has a Nuclear Emergency Briefing Room in its London office, which would become activated in the event of a nuclear emergency.
932. DECC are responsible for appointing a Government Technical Advisor. The Government Technical Advisor is a Deputy Chief Officer for the ONR, who ceases to work for the ONR, whilst working for DECC in the role of Government Technical Advisor. The Government Technical Advisor will maintain communications with the Government Liaison Officer in the designated off-site support facility.
933. Using the first hand information gathered from the facility, DECC would co-ordinate the response at a national level, briefing ministers and the UK's international partners, and issuing information nationally to the public and media(60). The Nuclear Emergency Briefing Room may also communicate with other government departments, for instance the Ministry of Defence, if further emergency response assistance were required.
(58) The Nuclear Emergency Arrangements Forum provides operators of nuclear licensed sites and the ONR with a best practice discussion forum relating, primarily, to the operators’ on-site emergency response planning, but also including the operators’ role in connection with the off-site response. (59) Nuclear Emergency Planning Liaison Group: Consolidated Guidance, Jan 2009, Chapter 7, Page 5 (60) http://www.decc.gov.uk/en/content/cms/what_we_do/uk_supply/energy_mix/nuclear/issues/emergency_plan/response/ response.aspx
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7.3. ARRANGEMENTS FOR THE EXCHANGE OF INFORMATION WITH OTHER MEMBER STATES
7.3.1. Cross-border communications and agreements with other member states 934. Several agreements and international conventions have been signed by the UK with regard to cross-border communication and mutual assistance.
7.3.1.1. International Atomic Energy Agency
Convention of 26 September 1986 on the early notification of a nuclear accident 935. This convention establishes a notification system for nuclear accidents which have the potential for international trans-boundary release, that could be of radiological safety significance for another State. It requires States to report the accident's time, location, radionuclide releases and other data essential for assessing the situation. Notification is to be made to affected States directly or through the IAEA and to the IAEA itself. Hinkley Point C will use this formal notification system to inform other Member States of an accident.
Convention of 26 September 1986 relating to assistance in the event of a nuclear accident or emergency radiological situation 936. This Convention requires that State Parties cooperate between themselves and with the IAEA to facilitate prompt assistance in the event of a nuclear accident or radiological emergency, to minimise its consequences and to protect life, property and the environment from the effects of radioactive releases.
7.3.1.2. European Atomic Energy Community 937. Council Decision 87/600/EURATOM on Community arrangements for the early exchange of information in the event of a radiological emergency, ensures that the Commission and Member States are promptly informed in the event of a radiological emergency, and sets out the information which is to be provided. The European Community Urgent Radiological Information Exchange (ECURIE) system provides an exchange platform for managing the graded response between Member States(61).
938. Therefore it is the responsibility of the DECC to notify European Union Member States via the notification systems above. Once this notification has taken place, responsibility for all overseas communication, is taken over by the Foreign and Commonwealth Office who are represented in the Nuclear Emergency Briefing Room.
939. In addition to the IAEA and ECURIE notification systems above, the UK has bilateral agreements with other Member States relating to the early exchanges of information in event of a major nuclear accident. These are:
Denmark 1987 bilateral agreement France 1983 amended 1993 Netherlands 1988 Memorandum of Understanding Norway 1987 bilateral agreement Former Soviet Union (now Russia) 1990 bilateral agreement Ireland 2004 bilateral agreement
7.4. EMERGENCY PLAN TESTING ARRANGEMENTS 940. A regular programme of emergency exercises is carried out in the UK to test procedures, facilities, systems and equipment and to enable everyone to practice their roles. There are three main types of regulatory exercise within the UK; Levels 1, 2 and 3.
(61) Handbook for the Implementation of EC Environmental Legislation, The Public Information (Radiological Emergency) Directive Section, Page 19
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941. Level 1 mainly concentrates on the operator’s actions in the event of an on-site emergency. They enable the operator to demonstrate the adequacy of its emergency arrangements to the ONR. The emergency services and other external organisations may be involved.
942. Level 2 tests the function and arrangements of each Strategic Co-ordination Centre (off-site facilities) every three years, under REPPIR. All the key agencies and Government departments which have a role in an emergency are expected to take part.
943. Level 3 is a national exercise. From the annual programme of local Level 2 off-site exercises, one is chosen as a national exercise to rehearse not only the functioning of the Strategic Coordination Centre but also the wider involvement of central Government, including the exercising of the various Government departments and agencies attending DECC’s Nuclear Emergency Briefing Room in London. The decision on which exercise should be selected as the national exercise, is made jointly between the licensees, the lead government department (DECC) and the Nuclear Emergency Planning Liaison Group in consultation with the ONR.
944. These exercises vary in the involvement of organisations locally and nationally. The site may choose to test the emergency arrangements more often than is required by regulation, for their own satisfaction. The ONR also plays a role in dealing with an emergency, by acting as liaison between the ONR, DECC and other Member States.
945. Since 1993, the Organisation for Economic Co-operation/Nuclear Energy Agency, International Nuclear Emergency Exercise series, organised by the Nuclear Energy Agency’s Working Party on Nuclear Emergency Matters, has addressed a range of objectives, scenarios and participants in order to facilitate improvements in emergency management systems, nationally and internationally. The last series, the International Nuclear Emergency Exercise 3 consequence management exercises held in 2005-2006, investigated arrangements for responding to widespread radiological contamination of the environment arising from an emergency situation.
946. Recognising the value of International Nuclear Emergency Exercise 3, in terms of the lessons and issues identified, the Working Party on Nuclear Emergency Matters has developed its new exercise, International Nuclear Emergency Exercise 4, to continue the exploration of the post-crisis response to an emergency situation, while addressing areas of current interest in emergency management, in particular the urban environment and the transition to recovery(62).
(62) Nuclear Energy Agency Committee on Radiation Protection and Public Health NEA/CRPPH/INEX(2009)6, Dec 2009
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8. ENVIRONMENTAL MONITORING
8.1. INTRODUCTION 947. There are statutory requirements to monitor the environment and foodstuffs around nuclear facilities in the UK. The specific requirements for the operator will be specified in the Environmental Permit and associated documentation. Monitoring and sampling regimes vary from site to site and year to year and are selected to be representative of existing exposure pathways. Knowledge of these pathways is informed by regular surveys of local people’s diets and lifestyle. This routine monitoring is supplemented by additional monitoring where necessary, for example in response to incidents or reports of unusual discharges of radioactive material.
948. The objectives of the environmental monitoring programme around Hinkley Point C are to:
assess reference group dose; monitor and review impact on wildlife; provide public and stakeholder reassurance (through the above); assess long term trends (indicator); and identify abnormal, fugitive and unauthorised releases.
949. These are consistent with the objectives outlined by the current UK Joint Agencies Draft guidance on Environmental Monitoring.
950. The main aim of environmental monitoring programmes is to monitor the environment and diet of local people living or working close to nuclear sites. Those people most exposed to radiation from the disposals of radioactive waste can be termed the ‘reference group’ for the purposes of dose assessment. It is assumed that if the most exposed group have a dose below the legal limit, then the exposure of all other members of the public should also be protected.
951. The UK environmental agencies have recently published draft joint guidance on environmental radiological monitoring programmes. It identifies that these programmes should be designed to meet the following generic principles:
952. Principle 1: Satisfy international requirements – Programmes should satisfy or be compatible with international requirements or guidance where available (e.g. IAEA Safety Standard on environmental and source monitoring, Article 35 of the Euratom Treaty).
953. Principle 2: Objective based – Programmes should be based on defined objectives and monitoring of different exposure pathways clearly linked to at least one objective.
954. Principle 3: Meet quality standards – Programmes should be undertaken to defined quality standards equivalent to ISO9001, ISO14001 and ISO17025.
955. Principle 4: Benefits exceed impacts – The benefits of the programme should exceed any significant environmental detriment (i.e. be environmentally sustainable).
956. Principle 5: Proportionate – The design and management of programmes should be proportionate to past, current and future potential impacts of discharges on humans and wildlife. Other considerations in determining the proportionality of the programme will be the cost, the environmental impact of undertaking the programme, the type of environment (including how dynamic it is), the likely behaviour of radionuclides in that environment (including half-life) and current state of knowledge. It will generally not be proportionate to require monitoring where the dose from a particular pathway is <0.001mSv y-1, but monitoring would be expected where the dose from a pathway exceeds 0.02mSv y-1 to ensure that a realistic dose assessment can be performed.
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957. Principle 6: Complementary – The regulators should ensure that their programmes and those of the operator are comprehensive and avoid unnecessary duplication.
958. Principle 7: Satisfy stakeholder concerns – Programmes should consider legitimate stakeholder concerns and expectations, as far as reasonably practicable.
959. Principle 8: Based on authorisations – Specific radionuclides should be selected for the monitoring programme, based on the source term (taking into account the magnitude of release and environmental impact) and radionuclides limited by the Environmental Permitting (England and Wales) Regulations 2010 or Radioactive Substances Act 1993 permits and/or authorisations, including those that could be released as fugitive emissions.
960. Principle 9: Optimised – Programmes should be optimised to achieve the maximum number of objectives from a minimum number of samples, ensuring that sufficient monitoring data of an acceptable quality are collected for all the objectives to be achieved.
961. Principle 10: Appropriate performance criteria – Performance criteria for the monitoring programme (in particular uncertainty criteria, limit of detection, analysis turnaround) should be designed to allow the objectives to be met, whilst ensuring proportionality (see Principle 5). Different objectives will have different performance criteria (e.g. for detecting abnormal releases a relatively quick analytical turnaround will be important, but a higher detection limit may be acceptable).
962. The development of a monitoring programme for Hinkley Point C will take full account of these principles in its design.
963. The new monitoring programme for Hinkley Point C will take account of local habit data and will seek to use and extend, where necessary, the existing Hinkley Point monitoring programme. This will ensure an efficient and comprehensive programme to monitor and protect the local environment and population.
964. To achieve this, it will be necessary to review the existing Hinkley Point arrangements and to understand and incorporate additional sampling requirements for Hinkley Point C, including sample type, frequency, location and analysis.
8.2. TERRESTRIAL ENVIRONMENT
8.2.1. Sample analysis 965. Laboratory analysis of samples varies according to the identity of the radionuclide. Gamma ray spectrometry is an effective method for detecting a wide range of nuclides that may be found in environmental samples.
966. Alternatively, radiochemical methods may be used which generally involve chemical separation techniques to quantify specific alpha and beta emitting radionuclides. These are sensitive but are much more labour intensive and are only used if the required information cannot be gathered using gamma ray spectrometry. Typically a laboratory will work to standard published procedures that have been well established for the assay of environmental media for their radioactive content(63). The laboratory may also participate in an accreditation scheme such as the United Kingdom Accreditation Scheme and participate in inter-comparison exercises to test and demonstrate the quality of its measurement and analytical capability.
967. The sampling strategy may be enhanced by using the Data Quality Objectives process. This process would ensure that environmental data are of the right type, quality and quantity to support defensible, confident decisions when planning a sampling programme.
(63) Sampling and measurement of radionuclides in the environment, Methodology Sub-Group to the Radioactivity Research and Environmental Monitoring Committee (RADREM), ISBN 0117522619, HMSO 1989
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968. The laboratory procedures will be developed to comply with the relevant standards including, ISO9001 and if reasonably practicable, British Standard EN 17025 “General requirements for the competence of testing and calibration laboratories”.
8.2.1.1. Milk 969. Milk is a convenient indicator of radioactivity in the environment as cows effectively sample a large area of pasture local to the power station.
970. Samples are typically screened for gamma emitting nuclides (e.g. iodine-131) on a monthly or quarterly basis, while carbon-14 analysis is carried out on annual bulk or quarterly samples(64). Appropriately selected farms would be sampled to ensure that the data are representative of the entire area.
8.2.1.2. Herbage 971. Herbage (mainly grass) can be collected from farms where milk is collected and also at sampling sites local to the power station. Appropriately selected areas would also be included in the survey.
972. Grass’ ability to concentrate radionuclides from the air and soil is similar to that of vegetables(63).
973. Samples of herbage are analysed quarterly by gamma ray spectrometry or radiochemical separation. Carbon-14 would be analysed from bulk samples of herbage collected.
8.2.1.3. Other foodstuffs 974. Herbage is generally considered to be the most useful indicator for the assessment of radioactivity in plants(3) but additional sampling may be undertaken by other agencies to ensure that there is no long-term accumulation of radioactivity in other foodstuffs.
975. Several local councils in the UK also have gamma ray spectrometry programmes to analyse leafy vegetables (sprouts, cabbage, broccoli and cauliflowers), fruit (mainly berries), root vegetables (onions, potatoes, carrots, turnips and parsnips), meat (beef, pork, lamb, venison and chicken) and eggs(65).
8.2.1.4. Soil cores 976. Soil cores would be taken at appropriately selected locations on a rotational basis so that each site is sampled at least once every five years.
8.2.1.5. Passive or Tacky Shades 977. Tacky Shades, otherwise known as dry cloth collectors, are suspended from poles above ground level. They collect airborne particulates. The Tacky Shades would generally be positioned around the site boundary and in an arc 1 to 2km from site covering all prevailing wind directions.
8.2.1.6. Thermo-Luminescent Dosimeters (TLDs) 978. TLDs are placed around the perimeter fence and in an arc around the site at a distance of about 2 to 4km. These are typically changed quarterly. The TLDs measure the external radiation dose over the period for which they are exposed.
(64) British Energy Generation Ltd, Sizewell ‘B’ Power Station Environmental Support Group Report, SZB/THR/042, 2005. (65) LARnet, Annual Report, 2005, NNC.
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8.2.2. Dose rate measurements 979. The environmental radiation dose rate would be measured quarterly at locations around the power station up to 10km from site. The locations would be divided into two zones; 0 to 2km and 2 to 10km from site.
980. Measurements of gamma dose rate in air are normally made at 1m above ground using a Mini Instruments environmental radiation meter type 680, with a compensated Geiger-Muller tube type MC-71or equivalent. These portable instruments are calibrated against recognised reference standards and the inherent instrument background is subtracted(66).
981. There are two quantities that can be presented as measures of external gamma dose rate; total gamma dose rate or terrestrial gamma dose rate. Total gamma dose rate includes all external radiation sources, while terrestrial gamma dose rate includes all sources except cosmic radiation. Both sets of gamma dose rates are usually reported.
8.3. MARINE ENVIRONMENT
8.3.1. Sample analysis 982. Marine samples are analysed in a similar fashion to terrestrial samples, using gamma ray spectrometry and radiochemical separation methods where required. The marine programme is likely to include sampling of fish and shellfish and analysis of silt and sediment from the shoreline and seabed.
8.3.1.1. Fish and shellfish 983. Fish and shellfish are both foodstuffs and shellfish in particular, are useful indicators of accumulations of radioactivity in the environment. Quarterly samples of locally caught fish are collected if they are available.
8.3.1.2. Silt and sediment 984. Samples will be collected quarterly from a number of locations around the site and analysed by gamma spectrometry and for gross beta radioactivity(64).
8.3.1.3. Seaweed 985. Seaweed will be collected along the coastline on a quarterly basis. The growing tips of the plants will be sampled as far as is reasonably practicable. The samples will be dried and thoroughly homogenised prior to gamma spectrometry and gross beta radioactivity.
8.3.2. Dose rate measurements 986. Environmental dose rate measurements will be taken at locations along the shoreline and over areas of sediment. This would be included as part of the terrestrial environment monitoring programme and measurements will be performed quarterly.
987. Beta-gamma contamination surveys may occasionally be carried out along the beach strandline. This high tide line is where deposited materials from the marine environment might be found. Strandline contamination monitoring is carried over a total survey length of approximately 2km, 1km either side of line of outfall, where there is safe access.
988. A beach monitoring probe, such as two thin end window or thin walled Geiger-Müller tubes and a 900 Mini Monitor, will be used. Equivalent probe/rate-meter combinations may also be used. Monitoring would be conducted just above the surface of the ground and would
(66) Routine measurement of gamma ray air kerma rate in the environment, Technical Guidance Note M5, HMIP ISBN 011751324 HMSO, 1995.
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concentrate on the most recent tide line, the most extreme high water mark and the wind blown debris above the extreme high water mark.
989. Fishing equipment will be monitored for contamination on an annual basis. Typically three pieces of fishing equipment used by local fishermen are monitored (where available and practicable). A beach monitoring probe, such as two thin end window or thin walled Geiger- Müller tubes and a 900 Mini Monitor, can be used. Equivalent probe/rate-meter combinations may also be used.
990. Beta and gamma contact dose rate measurements may be taken on all pieces of fishing equipment where contamination is detected during the annual contamination monitoring exercise.
8.4. DETECTION LIMITS 991. Gamma ray spectrometry can produce a large number of results at the limit of detection or at the minimum reporting level. In order to minimise the presentation of excessive information, results at these levels would only be reported when:
the radionuclide is one which is in the relevant authorisation; it has been analysed by radiochemical methods; it has been reported as being a positive value in the previous years; and a positive result is detected in any other sample.
992. Naturally occurring radionuclides measured by gamma spectrometry, are not usually reported unless they are intended to establish whether there is any enhancement above expected background levels.
8.5. EXISTING AND FUTURE MONITORING PROGRAMME 993. The geographical locations at which samples and radiation dose rate measurements are taken, are critical to the environmental monitoring programme. The locations should be evenly spread around the station and be at appropriate distances. The survey programme will provide representative data about the levels of radioactivity in the local area and ensure that locations where higher results might be found are sampled.
994. All sampling locations will be identified by a national Grid Reference. This will ensure that the sample is always taken from the same place such that representative trends can be compiled. If it is practicable and appropriate to continue data sets from sampling locations that have been used historically, this shall be done through the proposed programme.
995. Sampling frequencies currently defined in the environmental monitoring programme in place at Hinkley Point, are considered to be appropriate and are likely to be maintained for Hinkley Point C.
996. The following table summarises the monitoring arrangements (type of monitoring, type of sampling, location and frequency) in existence for Hinkley Point B.
997. The operator of Hinkley Point C will need to demonstrate that their environmental monitoring programme uses BAT. This is enforced through the Environmental Permit conditions. These requirements apply to the sampling, monitoring, measurement, analysis and assessment methods used in the environmental monitoring programme. The techniques used will be of comparable performance to those currently used.
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Table 8.1 Environmental monitoring in existence for Hinkley Point B
Sample/monitoring Location/frequency Sampling/monitoring type method DOSE RATE Quarterly deployment of TLDs located around site Spot dose rate MONITORING perimeter fence, at: measurement taken in accordance with Technical Guidance Note-M5. Location code Grid reference TLD 101 ST 2141 4579 TLD 102 ST 2123 4577 TLD 103 ST 2106 4576 TLD 104 ST 2092 4576 TLD 105 ST 2073 4581 TLD 106 ST 2072 4592 TLD 107 ST 2070 4607 TLD 108 ST 2078 4624 TLD 109 ST 2091 4629 TLD 110 ST 2098 4632 TLD 111 ST 2116 4634 TLD 112 ST 2128 4633 TLD 113 ST 2143 4628 TLD 114 ST 2152 4619 TLD 115 ST 2163 4605 TLD 116 ST 2158 4588
Quarterly deployment of TLDs in an arc around Hinkley Point, at distances of approximately 2 to 4km at the following locations:
Location code Grid reference Description TLD 201 ST 231 458 approx. 2km to E TLD 202 ST 229 452 approx. 2km to ESE TLD 203 ST 227 449 approx. km to ESE TLD 204 ST 228 438 approx. 3km to SE TLD 205 ST 228 428 approx. 4km to SSE TLD 206 ST 219 427 approx. 4km to SSE TLD 207 ST 211 427 approx. 4km to S TLD208 ST 204 428 approx. 4km to S TLD 209 ST 195 432 approx. 4km to SSW TLD 210 ST 193 439 approx. 3km to SW TLD 211 ST 175 435 approx. 4km to SW TLD 212 ST 167 442 approx. 5km WSW TLD 213 ST 172 451 approx. 4km WSW
Quarterly spot measurements of dose rate close to the site boundary at the following locations:
Grid Reference ST 203 461 ST 202 457
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Sample/monitoring Location/frequency Sampling/monitoring type method ST 206 456 ST 209 455 ST 212 456 ST 217 457 ST 217 461
Quarterly spot measurements of dose rate in an arc around Hinkley Point B at a distance of 5 to 10km from the site at the following grid reference locations:
NGR ST 145 442 approximately 7km to WSW NGR ST 163 422 approximately 7km to SW NGR ST 193 423 approximately 5km to SSW NGR ST 217 393 approximately 7km to S NGR ST 242 424 approximately 5km to SE NGR ST 259 450 approximately 5km to E NGR ST 279 468 approximately 7km to E
PASSIVE SHADES Deployment on or close to the site at the following locations:
A2 NGR ST 2080 A4 NGR ST 2120 4590 NGR ST 2120 4590 NGR ST 2102 4632 NGR ST 2076 4627 NGR ST 2074 4608
Deployment at the following locations in an arc around Hinkley Point, at distances of approximately 1 to 2km:
CT1 ST 197 458 approx. 1km to W CT2 ST 206 450 approx. 1km to S CT3 ST 214 450 approx. 1km to S CT4 ST 228 456 approx. 1.5km to E Monthly collection for analysis GRASS/HERBAGE Quarterly samples at the following grid reference locations: A sampling frame of known area should be placed on the area of grass to be NGR ST 218 460 approximately 0.7km to E sampled. NGR ST 198 440 approximately 1.5km to SSW Areas of poor vegetation NGR ST 214 435 approximately 2km to S cover or dominance of woody species should be NGR ST 232 448 approximately 3km ESE avoided. NGR ST 232 457 approximately 2km to ESE Grass should be trimmed to within 10mm of the soil surface (to prevent inclusion of soil). Leaf litter and soil should be avoided in the collection process. Grass should be
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Sample/monitoring Location/frequency Sampling/monitoring type method unwashed prior to analysis. For the measurement of carbon-14, samples from the listed locations may be bulked prior to analysis. SOIL Annual core samples at one of five locations where Prior to coring, overlying grass/herbage sampling undertaken, rotated so all vegetation should be cut locations are sampled once every five years. and the leaf litter and stones should be removed by gently scraping the NGR ST 218 460 approximately 0.7km to E surface. NGR ST 198 440 approximately 1.5km to SSW Five 5cm deep cores and NGR ST 214 435 approximately 2km to S five 30cm deep cores taken at each location with a NGR ST 232 448 approximately 3km ESE corer of known diameter. NGR ST 232 457 approximately 2km to ESE Each sample should be Location should be permanent pasture, as far as collected within a circle of reasonably practicable. about 10m diameter. Remaining vegetation should be carefully pared off with a knife. 5cm and 30cm cores bulked separately, dried, sieved to <2mm and thoroughly homogenised prior to sub- sampling for analysis. MILK Quarterly samples from the following farms: For the measurement of carbon-14 and strontium- 90, samples from the listed NGR ST 214 435 approximately 2km to S farms may be bulked prior NGR ST 232 448 approximately 3km to ESE to analysis. BEACH/ESTUARINE Quarterly spot measurements of dose rate at the following Spot dose rate DOSE RATE beach/estuarine locations: measurement taken in MONITORING accordance with Technical Guidance Note-M5. NGR ST 215 465 outfall NGR ST 228 463 0.8km east of outfall NGR ST 204 463 0.8km west of outfall NGR ST 234 549 off Stolford Common NGR ST 176 452 off Lilstock NGR ST 150 450 off Kilve NGR ST 302 490 off Burnham-On-Sea NGR ST 290 587 off Brean Down NGR ST 036 438 Blue Anchor Bay NGR ST 261423 off Combwich NGR ST 071 435 off Watchet STRANDLINE Six monthly strandline contamination monitoring. Total Beach monitoring probe CONTAMINATION survey length of 2km, 1km either side of line of outfall, such as two thin end MONITORING where there is safe access. Monitoring along strandline. window or thin walled Geiger-Müller tubes and a 900 Mini Monitor. Equivalent probe/ratemeter combinations may be used. Monitoring should be conducted just above the surface of the ground. Monitoring should
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Sample/monitoring Location/frequency Sampling/monitoring type method concentrate on the most recent tide-line, the extreme high water mark and the wind blown debris above the extreme high water mark. MONITORING OF Annual contamination monitoring of minimum of three Beach monitoring probe FISHING pieces of fishing equipment used by local fishermen such as two thin end EQUIPMENT (where available and practicable). window or thin walled Beta/gamma contact dose rate measurements on all Geiger-Müller tubes and a pieces of fishing equipment used by local fishermen where 900 Mini Monitor. contamination detected by contamination monitoring. Equivalent probe/ratemeter combinations may be used. SEDIMENT Quarterly samples from the following locations: Surface scrapes to a depth of 1cm. NGR ST 215 465 outfall Sediments to be dried, sieved to <250 microns (if NGR ST 228 463 0.8km east of outfall necessary) and NGR ST 204 463 0.8km west of outfall homogenised thoroughly prior to sub-sampling for NGR ST 234 459 off Stolford Common analysis. NGR ST 176452 off Lilstock NGR ST 150 450 off Kilve NGR ST 302 490 off Burnham-On-Sea NGR ST 036 438 Blue Anchor Bay NGR ST 290 587 off Brean Down NGR ST 261423 off Combwich NGR ST 071 435 off Watchet SEAWEED Quarterly samples of Fucus Versiculosus, approximately Growing edges of leaves to 1.0kg each, from the following locations: be sampled as far as is reasonably practicable. Samples dried and NGR ST 215 465 outfall thoroughly homogenised NGR ST 228 463 0.8km east of outfall prior to sub-sampling for analysis NGR ST 204 463 0.8km west of outfall NGR ST 234 549 off Stolford Common NGR ST 176 452 off Lilstock NGR ST 036 438 Blue Anchor Bay NGR ST 290 587 off Brean Down FISH Quarterly separate samples of locally caught free- Edible portions of fish sub- swimming and bottom feeding fish (as available). sampled for analysis. CRUSTACEA Quarterly samples of local crustacea; one or two genera Crustacea placed in boiling (eg shrimps, crabs, lobsters) (as available). water soon after collection (as per normal food preparation procedures) to prevent loss of radionuclides from digestive tract, edible portions sub- sampled for analysis.
8.6. MONITORING PROCEDURES 998. Hinkley Point Power Station has well documented and established procedures to cover all aspects of its current programme. Where additional monitoring is proposed, Hinkley Point C may adopt the comparable procedures, taking into account BAT.
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ANNEX A – ASTRAL FOOD CHAIN MODEL
999. The ASTRAL code (French acronym for "technical assistance for post-accident radiological protection") is designed to assess the impact of a nuclear accident on the environment and the populations concerned. Following the actual accident phase during which radionuclides are released to the environment, analyses and estimations must be performed to determine events in the following phase, called the "post-accident" phase. The radionuclides released are deposited on various surfaces in the environment. Their deposition on the entire soil area and all plant species, is the main input required to determine what becomes of the radionuclides deposited on the soil and vegetation cover. The code performs medium term assessment calculations over the first three years following deposition. The effect of a number of counter measures can also be taken into consideration.
1000. The calculation module is used to evaluate concentrations in agricultural products and derived foodstuffs. Concentration indexes are used to measure the impact on products, and human doses. It is described in the document ASTRAL V2.3 Code – Equations and Parameters(67).
1001. Calculation of the dose received via ingestion of contaminated foodstuffs is carried out in two steps:
An assessment of the contamination of the foodstuffs contaminated. An assessment of the dose itself.
Calculation of activity strength in foodstuffs Transfer to vegetable products 1002. Dry and/or wet deposition of radionuclides during the passage of the radioactive plume, is the origin of the contamination of vegetables. The assessment of the concentration in vegetables requires the modelling of the following processes:
Depositing of radionuclides on cultivated soil. Foliar transfer and radionuclide absorption via the roots of the vegetables. Temporal development of the concentration.
Direct deposition and foliar transfer 1003. Direct deposition on edible parts and foliar transfer, are the preponderant modes of contamination for ripening vegetables at the time of an accident.
1004. Foliar transfer results from the external parts of the vegetable intercepting contamination and involves the processes of leaf absorption and radionuclide transfer towards edible parts. Direct depositing on the part consumed is to be considered when the edible part is ripe at the time the activity is deposited.
1005. The impact of direct foliar transfer, principally depends on the stage of vegetable maturity, any precipitation accompanying the deposition, the type of vegetable and the radionuclide under consideration.
1006. An indirect foliar transfer may result from rainfall and wind, re-suspending radioactive particles deposited on the soil.
Root transfer 1007. Root transfer results from the incorporation of radionuclides in the soil and their becoming available to root systems and involves the processes of root absorption and transfer from the roots, to the edible parts of the vegetable.
(67) IRSN (2009), ASTRAL V2.3 Code – Equations and Parameters, Ph. Calmon and Ch. Mourlon, DOCUMENT DEI/SECRE – No. 2009 - 06
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1008. This type of transfer concerns vegetable roots, but also all plants since it becomes the preponderant contamination pathway after the first crop harvest.
Evolution over time of vegetable specific activity 1009. In a post-accidental situation, the evolution over time of vegetable specific activity must be determined. This evolution results from:
radioactive decay from the date of deposition to the date of harvest, then until the consumption date of the vegetable; concentration decay in the vegetable by biological elimination and by rainfall washing, following a transfer to foliage; and decay of activity available in the soil for root transfer due to vertical migration and the irreversible fixing of radionuclides.
1010. As a result of the above three processes, the specific concentration in a vegetable is modelled schematically as the product of three factors:
Deposition; transfer factor; and temporal evolution.
Transfer to animal products 1011. The contamination of animal products results from ingestion and inhalation of radioactive substances (inhalation being negligible after deposition) by the animal. The transfer mechanisms considered are incorporation and metabolising.
Incorporation 1012. Incorporation corresponds to the ingestion of contaminated products. This may be grass in pasture or harvested crops, such as maize silage, for example. The assessment of the quantity of ingested activity, requires knowledge of the feed calendar for each animal and the specific activity of each product at the time of its ingestion.
Metabolism 1013. Metabolism covers the transfer process of ingested radionuclides, to the parts of the animal consumed by man, as well as the biological elimination process of radionuclides by the animal.
Evolution over time 1014. Temporal evolution of the concentration in animal products takes into account:
daily ingestion of contaminated foodstuffs by the animal; activity decay in these foodstuffs from the deposition date; and concentration reduction in the animal by radioactive decay and biological elimination linked to metabolism.
1015. In order to take into account incorporation and the daily metabolism of the animal, it is necessary to integrate the transfers between the deposition date and the study date. The concentration in an animal product is thus schematically modelled as the product of three factors resulting from the above-described processes, integrated over the period under consideration:
incorporation; metabolism; and decay.
Calculation of ingested dose 1016. The effective dose received by ingestion is a function of individual diet, contaminated fractions of foodstuffs ingested and the specific activity of the foodstuff at the time of consumption.
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1017. The dietary regime is defined by the daily ingested quantity of each foodstuff. The specific activity of foodstuffs is deemed to vary by a step of time of one day. This assumption implies considering that the activity in the foodstuff is constant in every 24 hour period, which creates an upper bound.
1018. The dose received, ingesting foodstuffs contaminated by radionuclide ‘r’ is calculated according to the formula:
No days No. foods
Dose _ ingestionr Ratk fck Ark n FDr ingestion n k Where: -1 Ratk: ration of foodstuff k ingested daily by the individual (kg day ) fck: contaminated fraction of foodstuff ‘k’ (dimensionless) -1 Ark(n): specific activity of foodstuff k for radionuclide ‘r’ on day ‘n’ (Bq kg ) -1 FDr ingestion: ingestion dose factor for radionuclide ‘r’ (Sv Bq )
Food ration 1019. Diet varies with the age of the individual. The values shown below are derived from INSEE (National Institute of Statistics and Economic Studies – France)(68). Data for infants are derived using information on consumption of foods for infants(69).
Table A.1 Food consumption data – food consumption data used in Member State group assessment
Food (kg d-1) Adult 1 year infant Cow’s milk 0.094 0.000 Conserved milk 0.180 0.410 Butter 0.018 0.009 Fermented cheese 0.039 0.012 Flour 0.200 0.045 Beef 0.053 0.018 Chicken 0.066 0.018 Pork 0.067 0.026 Lamb 0.007 0.000 Egg 0.029 0.013 Potato 0.120 0.029 Leafy vegetable 0.087 0.037 Fruit 0.053 0.037 Root vegetable 0.043 0.037 Canned leafy vegetable 0.015 0.007 Canned fruit 0.021 0.032 Canned root vegetable 0.015 0.007
Contaminated fractions 1020. In the framework of a realistic approach, contaminated fractions generally lower than 100%, are retained to take into account the diverse provenance of foodstuffs.
(68) INSEE survey (1993): Consommation et lieux d'achats des produits alimentaires en 1991 (Purchasing sites and consumption of foodstuffs in 1991) (69) Archives de pédiatrie, Vol. 6 (1999): Consommation alimentaire des nourrissons et des enfants en bas âge en France en 1997 (Pediatric Archives, v. 6 (1999), Foodstuff consumption of infants and toddlers in France in 1997)
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1021. Taking into account the lack of any recent or precise data on the subject, the retained values are limited to the following order of magnitude (the ‘adult’ and ‘child’ values are identical):
Table A.2 Contaminated fraction of food
Product Contaminated fraction Milk 100% Vegetables (leaf, root, fruit) and potatoes 70% Meat (beef, lamb, pork, chicken) and eggs 50% Derived products (butter, cheese, flour, other milk products, canned vegetables) 10%
Calculation of ingested dose 1022. The modelling of agricultural soil in ASTRAL corresponds to a single parcel, whose characteristics are presumed to be vertically homogenous at any moment and at a given depth. The process of the transfer of radionuclides into the biosphere, is assessed with the aid of aggregated parameters.
Concentration in vegetable products 1023. ASTRAL software allows the calculation of concentrations in vegetables, as a function of time, while taking into account three distinct groups:
Arable crops, for which annual calendars are used. The first harvest is contaminated through the foliage if deposition takes place between ploughing and harvesting. After the first crop, only root transfer is calculated, taking into account radioactive decay available in the soil parcel. Market gardening, in which there are defined durations of soil occupation, during which cultures are likely to be subject to direct foliar transfer. The protection provided by greenhouses during the deposition is taken into account. Beyond the period of soil occupation, only root transfer is considered. The decay time takes into account biological elimination of contamination and rainfall washing of the plants. Pasture grass which is continually consumed by the animal and for which foliar and root transfer are considered simultaneously.
Concentration in animal products 1024. The concentrations in animal products are calculated by taking into account the feeding calendars and the actual contamination of plants at the date of consumption by the animal.
Calculation of ingested dose 1025. The ingested dose for humans is calculated according to the activity ingested daily, assessed by taking into account diet, associated contaminated fractions and foodstuff contamination on the consumption date.
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