Wylfa Newydd EP-RSR BAT Case Draft

Home , Winfrith

Wylfa Newydd Project Radioactive Substances Regulation – Environmental Permit Application: Appendices

Appendix C Wylfa Newydd EP-RSR Best Available Techniques (BAT) Case

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[This page is intentionally blank]

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Table of Contents

Table of Contents ...... i

Terms and Definitions ...... v

1 Introduction ...... 1

1.1 Definition of BAT ...... 1 1.2 Scope ...... 2 1.3 Regulatory Requirements ...... 2 1.4 Wylfa Newydd BAT Case Philosophy ...... 3 2 Methodology ...... 4

2.1 Approach to the Development of the Wylfa Newydd BAT Case ...... 4 2.2 Stage 1: Demonstrating the application of BAT ...... 5 2.2.1 Adoption of GDA material ...... 6 2.2.2 Forward actions ...... 7 2.3 Stage 2: Implementation of BAT ...... 7 2.4 Stage 3: Maintaining an optimised design and robust BAT Case ...... 8 2.5 Stage 4: Managing Decommissioning ...... 9 2.6 Management Arrangements ...... 10 2.7 Reporting Structure ...... 13 2.8 Demonstration of Best Available Techniques ...... 19 2.8.1 BAT Claims ...... 19 2.9 Reader’s Guide ...... 19 3 The BAT Claims ...... 20

3.1 Claim 1: Eliminate or Reduce the Generation of Radioactive Waste .. 20 3.1.1 Argument 1a: Design, Manufacture and Management of Fuel 21 3.1.2 Argument 1b: Reactivity Control ...... 28 3.1.3 Argument 1c: Efficiency of Fuel Use ...... 34 3.1.4 Argument 1d: Detection and Management of Failed Fuel ...... 38 3.1.5 Argument 1e: Commissioning, Start-up, Shutdown and Outage Procedures ...... 40 3.1.6 Argument 1f: Water Chemistry ...... 50

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.1.7 Argument 1g: Specification of Materials...... 59 3.1.8 Argument 1h: Recycling of Water to Prevent Discharges ...... 62 3.1.9 Argument 1i: Secondary Neutron Sources ...... 73 3.1.10 Argument 1j: Leak Tightness of Liquid, Gas and Mixed Phase Systems ...... 74 3.2 Claim 2 - Minimise the Radioactivity in Radioactive Waste Disposed to the Environment ...... 82 3.2.1 Argument 2a: Off-Gas Waste Treatment Facility ...... 82 3.2.2 Argument 2b: Charcoal Adsorbers for Noble Gases and Iodine ...... 88 3.2.3 Argument 2c: Heating, Ventilation and Air Conditioning System ...... 97 3.2.4 Argument 2d: Filtration of Airborne Particulate Matter ...... 103 3.2.5 Argument 2e: Optimisation of the Turbine Gland Seal ...... 105 3.2.6 Argument 2f: Configuration of Liquid Management Systems 109 3.2.7 Argument 2g: Sizing of Tanks, Vessels and Liquid Containment Systems ...... 118 3.2.8 Argument 2h: Demineralisers for Distillates from the HCW Evaporator ...... 119 3.2.9 Argument 2i: Evaporation of HCW ...... 121 3.2.10 Argument 2j: Radioactive Decay of Solid and Liquid Wastes 125 3.3 Claim 3 - Minimise the Volume of Radioactive Waste Disposed of to Other Premises ...... 130 3.3.1 Argument 3a: Design to Minimise the Volumes of Operational and Decommissioning Waste Arisings ...... 130 3.3.2 Argument 3b: Selection of Methods to Minimise Solid Waste Generation ...... 138 3.3.3 Argument 3c: Application of Volume Reduction Processes for Solid Waste ...... 147 3.3.4 Argument 3d: Solid Waste, Minimising the Quantity of Solidified HCW ...... 149 3.3.5 Argument 3e: Application of Decommissioning Techniques to Reduce the Activity and Volume of Decommissioning Waste 150 3.3.6 Argument 3f: Application of Waste Characterisation to Minimise the Volumes of Waste Sent for Disposal ...... 153 3.4 Claim 4 – Selecting the Optimal Disposal Routes for Wastes Transferred to Other Premises ...... 156

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.4.1 Argument 4a: Provision of Solid Waste Management Facilities ...... 156 3.4.2 Argument 4b: Optimal Disposal Route Selection ...... 167 3.4.3 Argument 4c: Agreement in Principle for Waste Routes - LAW ...... 170 3.4.4 Argument 4d: Disposability Assessments for Higher Activity Wastes ...... 171 3.4.5 Argument 4e: Compatibility of Existing UK Waste BAT Studies ...... 172 3.5 Claim 5 - Minimise the Impacts on the Environment and Members of the Public from Radioactive Waste that is Disposed of to the Environment ...... 174 3.5.1 Argument 5a: Gaseous Discharge System - Main Stacks .... 175 3.5.2 Argument 5b: Aqueous Waste Management ...... 178 3.6 Claim 6 – Horizon shall Apply BAT When Characterising and Quantifying Gaseous and Aqueous Radioactive Waste Discharges. 181 3.6.1 Argument 6a: Identification of those Parameters in the Gaseous and Aqueous Discharges to be Measured ...... 183 3.6.2 Argument 6b: Identification of Final Radioactive Discharge Locations and their Sampling Points ...... 186 3.6.3 Argument 6c: Sampling Methods and Analysis Techniques for Determination of Gaseous Radionuclides Activity Concentrations and Volumetric Flow Rates ...... 190 3.6.4 Argument 6d: Sampling methods and analysis techniques for determination of liquid radionuclide activity concentrations and volumetric flow rate ...... 200 3.6.5 Argument 6e: Space, capability and flexibility for maintenance activities and servicing ...... 204 3.6.6 Argument 6f: Independent Sampling Arrangements ...... 205 4 Radionuclide Specific BAT Route Map ...... 207

5 Forward Action Plan ...... 221

6 Conclusion ...... 223

7 References ...... 224

iii

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[This page is intentionally blank]

Wylfa Newydd Project – Best Available Techniques (BAT) Case

DfD Decontamination for Terms and Definitions decommissioning DPUR Dose Per Unit Release ABWR Advanced Boiling Water Reactor D/S Dryer/Separator ALARA As Low As Reasonably DZO Depleted Zinc Oxide Achievable EARWG Environment Agencies AO Air Operated Requirements Working Group ASTM American Society for Testing ECP Electrochemical Corrosion and Materials Potential BAT Best Available Technique EFPH Equivalent Full Power Hours

BSC Basis of Safety Case EMIT Examination, Inspection, Maintenance and Testing Bq Becquerel ENUSA Enusa Industrias Avanzandas BSW Biological Shield Wall EPC Engineering, Procurement and BWR Boiling Water Reactor Construction CAD Controlled Area Drain System EPRTM The EPRTM is a third CD Condensate Demineraliser generation pressurized water reactor designed by Areva NP CEM Continuous Emission Monitor and Electricité de France CF Condensate Filter EP-RSR Environmental Permitting CF/CD Condensate Filter Condensate Regulations 2016 - Radioactive Demineraliser Substances Regulation CP Corrosion Product EPR16 Environmental Permitting (England and Wales) CILC Crud-Induced Localised Regulations 2016 Corrosion FA Forward Action COD Commercial Operation Date FAC Flow Accelerated Corrosion CONW Concentrated Waste System FAP Forward Action Plan CORD Chemical Oxidation Reduction Decontamination FBS Fuel Business System COTS Commercial Off-The-Shelf FD Filter Demineraliser CST Condensate Storage Tank FDP Funded Decommissioning Programme CUW Reactor Water Clean-up System FHM Fuel Handling Machine CW Circulating Water FME Foreign Material Exclusion CWS Circulating Water System FP Fission Product DAC Design Acceptance FPC Fuel Pool Cooling and Clean- Confirmation up System DEFRA Department for Environment, FWP Feedwater Pump Food and Rural Affairs GAC Granulated Activated Carbon DF Decontamination Factor

Wylfa Newydd Project – Best Available Techniques (BAT) Case

GDA Generic Design Assessment LCW Low Chemical Impurities Waste GDF Geological Disposal Facility LCM Low Cobalt Material GNF Global Nuclear Fuel LDS Leak Detection System GSC Gland Steam Condenser LLI Long Lead Item GWd Gigawatt-day LLW Low Level Waste GWd/t Gigawatt-day/tonne LLWR Low Level Waste Repository HAT High Astronomical Tide LoC Letter of Compliance HAW Higher Activity Waste LPCP Low Pressure Condensate HCW High Chemical Impurities Pump Waste LPHD Low Pressure Heater Drain HEPA High Efficiency Particulate Air LPRM Local Power Range Monitor HFF Hollow Fibre Filter LT-SHC Low Temperature – Residual HLW High Level Waste Heat Removal Shutdown HMS Horizon Management System Cooling Horizon Horizon Nuclear Power Wylfa LTP Lower Tie Plate Limited LWR Light Water Reactor HOP Hydrazine, Oxalic Acid, MCR Main Control Room Potassium Permanganate MIDAS Multi-Inspection and Data HPCP High Pressure Condensate Acquisition System Pump MO Mechanically Operated HPHD High Pressure Heater Drain MOp Method of Operation HVAC Heating Ventilating and Air Conditioning MPS Missing Pellet Surface HWC Hydrogen Water Chemistry MSIV Main Steam Isolation Valves IAEA International Atomic Energy MUWC Make-up Water Condensate Agency System IGSCC Intergranular Stress Corrosion MUWP Make Up Water Purified Cracking System ILW Intermediate Level Waste NDA Nuclear Decommissioning Authority INPO Institute of Nuclear Power Operations NFWC Non-Fuel Waste Container IRAT Initial Radiological Assessment NMCA Noble Metal Chemical Addition Tool NRW Natural Resources Wales ISO International Standards NSSEP Nuclear Safety, Security and Organisation Environmental Principles IWS Integrated Waste Strategy NWC Normal Water Chemistry LAW Lower Activity Waste NZO Natural Zinc Oxide LAT Low Astronomical Tide OECD Organisation for Economic Co- Operation and Development

Wylfa Newydd Project – Best Available Techniques (BAT) Case

OD Outer Diameter RSR Radioactive Substances Regulation OG Off-Gas System RSW Reactor Building Service Water OLNC On-Line NobleChem™ RWM Radioactive Waste ONR Office for Nuclear Regulation Management Ltd OPEX Operational Experience Rw/B Radioactive W aste Building OSPAR Oslo and Paris Convention on S/B Service Building Protection of the Marine Environment of the North East S/C Suppression Chamber Atlantic S/P Suppression Pool PCI Pellet Cladding Interaction SCC Stress Corrosion Cracking PCmSR Pre-Commissioning Safety SCV Steel Containment Vessel Report SF Spent Fuel PCSR Pre-Construction Safety Report SFP Spent Fuel Storage Pool PCV Primary Containment Vessel SGTS Standby Gas Treatment PLR Primary Loop Recirculation System System SJAE Steam Jet Air Ejector POCO Post-Operational Clean Out SoDA Statement of Design POSR Pre-Operation Safety Report Acceptability ppb Parts per billion SPCU Suppression Pool Clean-up ppm Parts per million System PSD Process System Description SQEP Suitably Qualified and Experienced Person PST Power Suppression Testing SRV Safety Relief Valve PWR Pressurised Water Reactor SSC Structures, Systems and QA Quality Assurance Components R/B Reactor Building SSSS Separate Steam Seal System RCA Radiological Controlled Area Sv Sievert RCCV Reinforced Concrete T/B Turbine Building Containment Vessel T/D-RFP Turbine Driven Reactor RCW Reactor Building Cooling Water Feedwater Pump System TGS Turbine Gland Steam System REP Radioactive Substances Regulation – Environmental TGSCC Trans Granular Stress Principle Corrosion Cracking RGP Relevant Good Practice TMOL Thermal-Mechanical Operating Limits RHR Residual Heat Removal TOC Total Organic Carbon RIP Reactor Internal Pump T-OZON Toshiba Ozone Oxidizing RPV Reactor Pressure Vessel Decontamination for Nuclear RRS Reactor Recirculation System Power Plants

vii

Wylfa Newydd Project – Best Available Techniques (BAT) Case

TSW Turbine Building Services Water UK United Kingdom U.S. United States of America UT Ultrasonic Testing VLLW Very Low Level Waste WAC Waste Acceptance Criteria WMS Waste Management System WPS Waste Packaging Specification WSC Waste Service Contract WSILW Wet Solid Intermediate Level Waste WSLLW Wet Solid Low Level Waste WPM Wear Proof Material

Wylfa Newydd Project – Best Available Techniques (BAT) Case

1 Introduction This document is the Wylfa Newydd Best Available Techniques (BAT) Case (hereafter referred to as the Wylfa Newydd BAT Case). It summarises the claims, arguments and evidence that have been compiled to demonstrate that BAT has been applied to the design and future operation of the Wylfa Newydd Power Station. The claims, arguments and evidence set out in this document and the actions detailed in the Forward Action Plan (FAP) (Section 5 Forward Action Plan) are considered by Horizon Nuclear Power Wylfa Limited (Horizon) to comprise the BAT Case for the Wylfa Newydd Power Station. This Wylfa Newydd BAT Case forms part of Horizon’s Radioactive Substances Regulation (RSR) permit application [Ref-1], under Schedule 23 of the Environmental Permitting (England and Wales) Regulations 2016 (EPR16), for the disposal of radioactive waste from the Wylfa Newydd Power Station. The EPR16 RSR permit application will hereafter be referred to as the “EP-RSR application” and the permit will be referred to as the “EP-RSR permit”. This document addresses question 3 of the Natural Resources Wales (NRW) application form NRW-EP-RSR-B3 [Ref-1]: “3. Describe how you manage the production, discharge and disposal of radioactive waste to protect the environment and to optimise the protection of people”. This document also contains the information on the in-process monitoring arrangements that are installed to demonstrate that the plant systems are performing as expected. As a result, this document addresses some aspects of question 5 of the NRW-EP-RSR-B3 form [Ref-1]: “5. Provide a description of the sampling arrangements, techniques and systems for measurement and assessment of discharges and disposals of radioactive waste.” In addressing the questions above, this document demonstrates that the practice of generating electricity from twin United Kingdom (UK) Advanced Boiling Water Reactor (ABWR) units, is considered to be environmentally optimised at this stage of the project and that BAT is being applied. It is recognised that BAT is a continuous process, which will evolve and develop throughout the lifecycle of the Wylfa Newydd Power Station from initial design right through to decommissioning. Further details on Horizon’s approach to demonstrating the application of BAT can be found in the BAT Case Strategy [Ref-2].

1.1 Definition of BAT The purpose of BAT is to minimize and, as appropriate, eliminate any pollution caused by radioactive discharges. It is defined as [Ref-8]:

• Best is defined as the most effective in achieving a high level of protection of the public from exposure to ionising radiation assessed against the full range of detriments and benefits of further reductions.

• Available requires consideration of:

o Whether the techniques under consideration have been developed on a scale which allows implementation in the relevant industrial sector; and

Wylfa Newydd Project – Best Available Techniques (BAT) Case

o Whether the conditions mean that techniques are economically and technically viable, taking into consideration both the benefits and detriments.

• Techniques includes both the technology used and the way in which the installation is designed, built, maintained, operated and decommissioned. As set out within discharge permits, the application of BAT includes management regimes to ensure competence, maintenance, inspection, supervision and monitoring across all stages in the lifecycle of a facility.

1.2 Scope The scope of this report is to demonstrate and provide confidence that BAT is being applied to all lifecycle phases of the Wylfa Newydd Power Station as prescribed by the EP-RSR application. The Wylfa Newydd Power Station comprises twin UK ABWR units. This report demonstrates that the environmental performance associated with the practice of generating electricity from the Wylfa Newydd Power Station will be optimised, and that impacts from potentially harmful ionising radiation on members of the public and the environment will be As Low As is Reasonably Achievable (ALARA). The environmental performance of the Wylfa Newydd Power Station is assessed against normal operations, expected events (as defined in the Methodology for Expected Event Selection [Ref-3]) and, decommissioning activities. The Wylfa Newydd BAT Case emphasises the assessment of BAT for the lifecycle stages between design and operation. At the optimum time prior to decommissioning, a separate Wylfa Newydd Decommissioning BAT Case will be produced to demonstrate BAT for the decommissioning phase.

1.3 Regulatory Requirements Optimisation is a key principle used to protect people from the risks associated with exposure to potentially harmful ionising radiation. The Wylfa Newydd BAT Case specifically focuses on minimising dose to members of the public. Minimising the dose to ALARA through optimisation of the Power Station is achieved through the application of BAT as prescribed in a number of conditions within the EP-RSR permit [Ref-5]: • 2.3.1 – The operator shall use the best available techniques to minimise the activity of radioactive waste produced on the premises that will require to be disposed of on or from the premises. • 2.3.2 – The operator shall use the best available techniques in respect of the disposal of radioactive wastes pursuant to this permit to: a) Minimise the activity of gaseous and aqueous radioactive waste disposed of by discharge to the environment; b) Minimise the volume of radioactive waste disposed of by transfer to other premises; c) Dispose of radioactive waste at times, in a form, and in a manner so as to minimise the radiological effects on the environment and members of the public. • 2.3.3 – The operator shall use the best available techniques to:

a) Exclude all entrained solids, gases and non-aqueous liquids from radioactive aqueous waste prior to discharge to the environment;

Wylfa Newydd Project – Best Available Techniques (BAT) Case

b) Characterise, sort and segregate solid and non-aqueous liquid radioactive wastes, to facilitate their disposal by optimised disposal routes. • 3.1.3 – Subject to condition 3.1.1, the operator shall dispose of each form of solid and non-aqueous liquid radioactive waste by an optimised disposal route for that waste form. • 3.2.1 – The operator shall: a) use the best available techniques when taking such samples, conducting such measurements, tests, surveys, analysis and calculations, and carrying out such an environmental monitoring programme and retrospective dose assessment, unless particular techniques are specified in schedule 3 of this permit or in writing by the Environment Agency. The interpretation of the optimisation requirement and in particular the conditions within an EP-RSR permit that prescribe the application of BAT is aided by government and regulatory guidance documents. These documents along with the guidance provided by the Nuclear Industry have been used in developing the Wylfa Newydd BAT Case. These key documents include: • The Environmental Permitting (England and Wales) Regulations 2016 [Ref-4]. • Environment Agency, “How to comply with your environmental permit for radioactive substances on a nuclear licensed site” [Ref-5]. • Environment Agency, “RSR: Principles of optimisation in the management and disposal of radioactive waste”, Issue 2, April 2010, Environment Agency [Ref-6]. • Regulatory Guidance Series RSR 1: Radioactive Substances Regulation – Environmental Principles (REPs) (Natural Resource Wales, September 2014) [Ref-7]. • Best Available Techniques (BAT) for the Management of the Generation and Disposal of Radioactive Wastes, A Nuclear Industry Code of Practice (Nuclear Industry Safety Directors Forum, 2010) [Ref-8] and its addendum [Ref-9]. • Effluent Release Options from Nuclear Installations. Technical Background and Regulatory Aspects (Organisation for Economic Co-operation and Development, 2003) [Ref-10]. • Environmental Permitting Guidance Radioactive Substances Regulation for the Environmental Permitting (England and Wales) Regulations 2010 (DEFRA September 2011) [Ref-11]. Additionally, there are a number of REPs [Ref-7] which provide guidance on many aspects of operation and management of nuclear facilities. These REPs have been considered throughout the development of the Wylfa Newydd BAT Case and have been used to aid interpretation of what is required by those conditions of EPR16 which require the application of BAT. REPs considered are listed within the relevant claim.

1.4 Wylfa Newydd BAT Case Philosophy The ABWR reactor has evolved from previous Boiling Water Reactor (BWR) designs. The design has been assessed against UK legislation and requirements (through a Generic Design Assessment (GDA) process) and modified to meet UK requirements. This generic design is then further assessed within the context of a specific site (i.e. Wylfa Newydd) as part of the EP-RSR Application. The demonstration that the design and operation of the

Wylfa Newydd Power Station has been optimised through the application of BAT is provided in this Wylfa Newydd BAT Case. The approach to applying BAT to the design of the Wylfa Newydd Power Station, and to support the demonstration of BAT is based on the following principles [Ref-12]: • Evolution: Utilising previous experience to optimise design: • Safety, environment and operability have all influenced how the design of the Wylfa Newydd Power Station has evolved at each design iteration. The development of the Wylfa Newydd BAT Case allows credit to be taken for this process and to adopt the work undertaken as part of the GDA. • Integration: Ensuring that both the application of BAT and its implementation are an integral part of Horizon’s management systems. Integrating BAT within Horizon’s configuration control arrangements is essential, as the delivery of an optimised design is reliant upon managing change: • There are several considerations that must be taken into account when making decisions on the design and future operation of a nuclear power station. Some of these are directly attributed to the Office for Nuclear Regulation (ONR) requirements, for example, the reduction of worker dose to as low as is reasonably practicable, whereas others are less specific, such as ease of implementation, operability and decommissioning implications or technology maturity. Importantly, the demonstration of BAT has been integrated into the project programme and decision making process. • Opportunity: Prioritisation of forward actions: Recognising that the demonstration of BAT should cover the lifecycle of the plant, certain elements have been addressed at GDA, whereas others are managed at the site-specific level or during commissioning or operations. A FAP has therefore been developed to identify issues early and ensure they are addressed at a time that will allow the most benefit to be realised. These principles have been applied in the development of the Wylfa Newydd BAT Case and the ongoing optimisation of the design. Horizon has published the Nuclear Safety, Security and Environmental Principles (NSSEP) Policy [Ref-13] to set out expectations in the design and management arrangements for demonstrating BAT and achieving radiological environmental protection. The NSSEP’s environmental protection objective is for the deployment of BAT at all stages of the Power Station’s life cycle, to provide confidence that the Wylfa Newydd Power Station is operating in accordance with the limits and conditions set out in the EP-RSR Permit, to ensure that public dose is ALARA and that the impact on the environment has been minimised. Further opportunities to substantiate the BAT arguments can then be explored and the question asked “what more can be done?” subject to whether it is grossly disproportionate. 2 Methodology

2.1 Approach to the Development of the Wylfa Newydd BAT Case The approach that has been implemented in developing the Wylfa Newydd BAT Case uses an established methodology that is used throughout the nuclear industry, particularly on large new build projects. This approach recognises that when demonstrating the application of BAT considerable effort in terms of time, effort, and cost has already been expended in

Wylfa Newydd Project – Best Available Techniques (BAT) Case

developing, assessing and in many cases operating design options and that many processes have already been optimised. It is therefore important to be able to take account of how a design, technology, or management procedure has evolved over time and then to explore what more can be done. Horizon has developed management arrangements to direct and instruct the application, demonstration and implementation of BAT. The Management of EP-RSR BAT Process System Description (PSD) [Ref-14] is a management arrangement that sets out the activities that Horizon must undertake to demonstrate EP-RSR BAT. The PSD is a Horizon Management System (HMS) document. It is applicable to all aspects of the Wylfa Newydd Power Station project which have an effect on the production, management, control and monitoring of radioactive discharges and waste, and to the preparation of wastes for disposal to permitted sites. Supporting the PSD are the following three procedures: • EP-RSR BAT Evidence Management procedure [Ref-15] – This procedure instructs on the approach to collecting, evaluating and assessing Evidence used to support the development and substantiation of BAT arguments; • EP-RSR BAT Case Amendment Management procedure [Ref-16] – This procedure instructs on the approach for identifying changes impacting upon the Wylfa Newydd BAT Case and the process of amending the Wylfa Newydd BAT Case; • Optioneering Procedure [Ref-17] – This procedure instructs on the method for carrying out optioneering workshops and producing optioneering reports that identify an option as BAT. The UK ABWR has already been subject to detailed assessment as part of the GDA process. It is argued during GDA that many aspects of the UK ABWR design have been optimised and evidence has been provided to substantiate these arguments. However, not all aspects were considered or fully optimised and it was not possible to take account of site specific factors. It is also recognised that certain aspects of the design have changed as the generic design has evolved into the site specific design. It was therefore recognised that a site specific Wylfa Newydd BAT Case should both capture the relevant content of the UK ABWR GDA BAT Case [Ref-18] whilst continuing to support the demonstration that BAT is being applied. The approach to the application of BAT was undertaken in four stages: • Stage 1: Demonstrating the application of BAT • Stage 2: Implementation of BAT • Stage 3: Maintaining an optimised design and robust BAT Case • Stage 4: Managing decommissioning Each stage is summarised below and described in more detail in the BAT methodology.

2.2 Stage 1: Demonstrating the application of BAT The demonstration of BAT follows the claim, argument, evidence reporting structure. This structure allows the Wylfa Newydd BAT Case to be developed around a set of claims that are based on those conditions within the EP-RSR permit that require the application of BAT. A number of succinct arguments are then prepared that validate the claims. The arguments effectively tell the story as to how the design has evolved, recognising those aspects that have already been substantiated at GDA and what further assessment and future design

Wylfa Newydd Project – Best Available Techniques (BAT) Case

changes are still required. Each argument is substantiated using evidence to ensure that the Wylfa Newydd BAT Case is demonstrably robust.

2.2.1 Adoption of GDA material Information presented in Revision F of UK ABWR GDA BAT Case [Ref-18] will be ‘adopted’ by Horizon to form the basis of the Wylfa Newydd BAT Case. This is carried out in accordance with the “Adoption of GDA Documentation” PSD [Ref-19] (summarised in Figure 2.1-1). There are three levels of GDA documents. Level 1 and 2 documents support Claims and Arguments presented at GDA, these documents will be adopted by Horizon. Level 3 documents present Evidence in support of Claims and Arguments. Level 3 documents will not be formally adopted by Horizon but the information is reviewed by Suitably Qualified and Experienced Person(s) (SQEP) in the same way as for Level 1 and 2 documents. All Arguments within Claims 1, 2, 3, 4 and 5 have been incorporated into the Wylfa Newydd BAT Case. Formal “adoption” cannot take place until Design Acceptance Confirmation (DAC) and Statement of Design Acceptability (SoDA) have been issued on GDA. In the interim period ‘pre-adoption’ has taken place. [Ref-19]

GDA Process Step 4 DAC and SoDA Issued

Adoption Confirmation

DA Review Wylfa Newydd BAT Case

DA Queries and Observations Adopted Site returned Material Specific Material Hitachi-GE DA Reviewed GDA Level 1 Issued GDA and Level 2 Material

Level 1 and (together with associated DA Level 2 Queries and Observations) Material

Horizon Controlled Document Store

Figure 2.1-1: GDA Adoption Process

Wylfa Newydd Project – Best Available Techniques (BAT) Case

2.2.2 Forward actions Design decisions should be made at proportionate times in the project, that ensure maximum benefit is realised. This means considering when decisions for determining BAT need to be made on a case by case basis. Postponing a decision provides opportunity to take account of the technological developments that have occurred over time and use more detailed design information that may become available. The optimum time is one that balances the benefits for postponing, with project and regulatory requirements, for making a decision. If it is proportionate to postpone design decisions into the future (e.g. the selection of process equipment within a building), the Wylfa Newydd Power Station design will need to accommodate or be compatible with potential equipment or design features that have yet to be selected. Otherwise future decisions will be constrained by the foreclosure of options associated with other design decisions or the procurement of plant. In practice, to prevent the foreclosure of options it may be necessary to provide sufficient space or suitable services in rooms and buildings at the early stage of the design. Issues, uncertainties and gaps form an inherent part of a BAT Case and can be influential within the decision making processes used to support the development of the Claims and Arguments. Issues, uncertainties and gaps are be identified in the Wylfa Newydd BAT Case as Forward Actions (FA) and shall be addressed through appropriate management in the future. A forward action is defined as an action to resolve a significant issue, uncertainty or gap that is fundamental to the design or operation of the UK ABWR. Resolving the action is critical to demonstrating BAT by a future stage within the Power Station project. Forward actions are presented within the text where mentioned and are summarised in Section 5 (Forward Action Plan). Additionally, there are a number of Forward Commitments made within the EP-RSR application, some of which are referred to within the BAT case as they’re resolving is important in the demonstration of BAT. In such instances, reference to the applicable section of the EP-RSR application is made (e.g. “see Section 5 of the EP-RSR Application”).

2.3 Stage 2: Implementation of BAT The implementation of BAT is focused on the delivery of defined environmental protection functions that are required to optimise engineering controls in the design. Environmental protection functions encapsulate the performance requirements in the design and operation of Structures, Systems and Components (SSC) in the Power Station. The implementation of BAT is focused on defining the key inputs and environmental functional requirements required to optimise the design. The initial generic design was assessed to identify design elements which were already optimised, and to identify aspects where further information and/or design changes were required to fully optimise the design. These aspects have been addressed as the generic design has evolved into the site specific design. Where these aspects have been optimised, this has been presented with evidence in this document. Where design elements and/or information required to optimise the design are not yet available (for example some aspects will be decided on during commissioning) these have remained as a forward action. HMS arrangements will provide detailed guidance on defining environmental protection functions and the management of SSC that supports the demonstration BAT as identified in the Wylfa Newydd BAT Case and the sources of information referenced as Evidence. Satisfying environmental protection functions will support the demonstration of BAT and

Wylfa Newydd Project – Best Available Techniques (BAT) Case

adherence to the NSSEPs [Ref-13]. The HMS documents responsible for controlling the management of SSC fulfilling environmental protection functions are: • Management of Environmental Functions and SSC PSD [Ref-20]. • Environmental Functions and Categorisation procedure [Ref-21]. • Identification of Environmental SSC procedure [Ref-22]. The management arrangements stated above will direct the suitable specification of design codes and standards for the SSC. It also signposts to Quality Assurance (QA) arrangements to ensure the management and monitoring of quality in the procurement of SSC is commensurate with its role in demonstrating BAT.

2.4 Stage 3: Maintaining an optimised design and robust BAT Case Continuous improvement through the periodic review of engineered and management controls and the substantiation provided within this BAT Case, combined with the implementation of lessons learnt will be utilised to support the ongoing application of BAT and its demonstration throughout the lifecycle of the project. Within the project phases of development, construction and commissioning the issued Wylfa Newydd BAT Case will need to be reviewed and potentially updated in response to the deliverables listed below. • Issue of GDA BAT Case Revision G; • Issue of GDA Pre-Construction Safety Report (PCSR) Revision C; • Issue of GDA Assessment Findings and Minor Shortfalls; • Issue of Wylfa Newydd PCSR; • Issue of Wylfa Newydd Pre-Commissioning Safety Report (PCmSR); • Issue of Wylfa Newydd Pre-Operation Safety Report (POSR). The BAT Case team will be cognisant of GDA design change proposals and will consult with the Horizon Safety Case team on GDA and site specific changes to understand potential impacts to the Wylfa Newydd BAT Case. GDA Assessment Findings and Minor Shortfalls will be formally published following issue of DAC and SoDA. Those issued will be subject to the Horizon procedure on the Management of GDA Assessment Findings [Ref-23] which will detail how to screen, allocate ownership, schedule, control, undertake QA review and eventually close findings and shortfalls. Changes to the design and operation of the Power Station can affect the generation, management and monitoring of radioactive discharges and radioactive waste. Demonstration of BAT can be undermined if design changes are ill conceived or executed, poorly implemented or do not represent BAT. Changes as a result of GDA deliverables, site specific deliverables or other design and operational changes shall be assessed for their impact on the Wylfa Newydd BAT Case. Proposed changes to design documentation under Horizon configuration control will be assessed to determine their impact on nuclear safety, environment and security in accordance with the Design Change Control process [Ref-24]. The EP-RSR BAT Amendment Management procedure [Ref-16] provides guidance on how to determine if design or operation changes affect the content of the Wylfa Newydd BAT Case. When the Wylfa Newydd BAT Case requires updating to reflect design or operational changes to the

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Power Station, the Case shall be updated in accordance with Horizon’s Design Change Control process [Ref-24].

2.5 Stage 4: Managing Decommissioning A separate Wylfa Newydd Decommissioning BAT Case will be produced. Decommissioning is not anticipated to be performed for the UK ABWR until at least 60 years after commercial operation date (COD). It is expected that significant advancements in technology will occur in that time; therefore, it is likely that today’s decommissioning technologies that are BAT, will likely not be considered BAT at the time of decommissioning. Additionally, the site end state has not been established as this will be based upon the subsequent use of the site [Ref-25]. It is therefore appropriate to adopt a phased approach in determining exactly how the site will be decommissioned. Whilst in the development stage of the project, this document does not provide full details on decommissioning techniques. Designing for decommissioning has been considered in Topic reports [Ref-26] and [Ref-27] and the GDA PCSR [Ref-25] and these considerations have been discussed under relevant Arguments within this document. To support the demonstration of BAT during decommissioning, a number of management arrangements (Section 2.6 Management Arrangements) will be established at appropriate timescales within the project. The Decommissioning Strategy will outline key principles to ensure that BAT and current national policy and regulatory expectations are applied. These principles reflect those applied in the GDA PCSR [Ref-25] and are stated below: • Strategies and plans will be compliant with UK Government policies and legislation, including the policy aim of sustainable development. Strategies and plans should take account of the views of stakeholders. • The safety and protection of the public, the workforce and the environment are the key drivers and decommissioning should be managed in accordance with the ALARP principle to ensure an optimal level of protection taking into account all relevant factors. • The decommissioning plan should be developed to be “fit for purpose” in terms of the technology assumed, the organisational arrangements to be used and staff training and skill levels required. • The best appropriate scientific and technical knowledge should be used to inform the decommissioning plan, but technologies and techniques which are planned to be deployed should be readily available and easy to use. Simple and flexible solutions should be sought in preference to complex ones. • The decommissioning plan should focus on tasks and hazards, aiming for a progressive reduction in hazard, not on plant systems and should recognise that decommissioning is a project and not an on-going operational activity. • Strategies and plans need to recognise that the output of the decommissioning process is radioactive and non-radioactive waste for recycling or disposal (and potentially reuse). • BAT are to be applied to minimise volumes of radioactive wastes which are created, particularly ILW wastes. BAT will be used in the management of discharges from the

Wylfa Newydd Project – Best Available Techniques (BAT) Case

site, though it is recognised that adoption of the BAT solution to management of some wastes may result in short term increases in discharges of some radionuclides. • After reactor closure, the overall philosophy of the site will be changed from “safe, operational excellence and high availability”. Such a philosophy can be characterised by the following relevant aspects: • Decoupling of systems so that they are self-contained and independent, with minimal interactions. • The idea that providing new equipment and facilities may be preferable to modifying existing equipment and facilities. • Using appropriate “low-tech” and robust solutions rather than high-tech ones. • Recognising the changing hazards and risks on the site and reflecting these in the safety objectives and approaches. • Strategies and plans will be reviewed and updated on a regular basis and information relating to these plans, including their costs, schedule and implementation will be recorded and preserved. • The decommissioning fund accumulated during the reactors’ life must be adequate to pay for the planned decommissioning of the site and management of all waste arisings in line with the waste hierarchy. • The strategy should be developed so as to prevent or avoid risks wherever possible, and to allow mitigation of any residual risks. • Any new facility will be designed, built and operated so as to minimise decommissioning and associated waste management activities and related costs. • Decommissioning activities will be carried out as soon as reasonably practicable, taking into account all relevant factors. • Throughout the whole life-cycle of a facility the documents and records that might be required for decommissioning purposes should be identified, prepared, updated and retained.

2.6 Management Arrangements Recorded management arrangements are fundamental in complying with the EP-RSR Permit. Condition 1.1.1 (a) [Ref-5] requires the operator to manage and operate in accordance with a written management system. Management arrangements are a principle means of controlling activities that implement BAT and complement the use of engineering controls that have been identified as BAT. The Wylfa Newydd BAT Case uses management arrangements as an overarching term for the following types of document: • Management controls that are used with SSC so that the SSC fulfils its environmental protection role or function that represents BAT. • Management controls that direct activities relating to the management and monitoring of radioactive waste and discharges that represent BAT.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• HMS documents that provide instruction or guidance on processes within Horizon on the application and implementation of BAT. • Strategies, plans and technical reports that underpin decision making on design elements and operations that demonstrate BAT in the Power Station. The BAT principles of Evolution, Integration and Opportunity (Section 1.4 Wylfa Newydd BAT Case Philosophy) were applied to determine applicable management arrangements for demonstrating BAT. • Evolution – In addition to management arrangements recognised in the GDA BAT Case [Ref-18], there are management arrangements associated with site specific design features for controlling radioactive waste and discharges. • Integration – Demonstration of BAT comprises the use of management arrangements across different themes within the project that include chemistry, radioactive waste management, commissioning, radiation protection, decommissioning, management of SSC, conduct of maintenance, conduct of operations as well as training and competency. Management arrangements within themes are the responsibility of different functions and teams across Horizon. The Wylfa Newydd BAT Case provides oversight of deliverables. The BAT Case team will review them at the appropriate times to ensure they are compatible with the Wylfa Newydd BAT Case and that management arrangements will result in the demonstration of BAT. • Opportunity – Understanding when a management arrangement needs to be in place or training implemented to support operations or decision making. Horizon is a developing organisation, its management system and operating management controls are being produced to fulfil requirements of an EP-RSR permitted facility and a Nuclear Licensed Site. However, the majority of management arrangements for an operating Power Station are not required upon the receipt of the EP-RSR Permit. Horizon will adopt a phased approach in the development of management arrangements. They will be produced at a time that aligns with the maturity of the Power Station design and at an appropriate timeframe in advance of the project phase they will be needed for implementation. A proportionate approach is needed in defining the level of detail of future management controls within the Case. Whilst within the development phase of the project, this version of the Case provides a high level description of the key management arrangements that will enable project activities to maintain compliance with the EP-RSR permit. A route map will be developed to map key management arrangement deliverables against when they are needed to be produced in terms of Horizon’s project Hold Points. These have yet to be defined but high level and indicative hold points may be:

• Stage 1: First time procurement of safety related Long Lead Items (LLI).

• Stage 2: Start of deep excavation – Excavation beyond platform level (assume 13.5m).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• Stage 3: First nuclear construction – First irreversible construction works for base Slab on Nuclear Island. • Stage 4: First inactive commissioning – Back Energisation of Standby Auxiliary transformer and First System Level Testing. • Stage 5: First fuel to site – Arrival of Nuclear Fuel on Nuclear Licensed Site. • Stage 6: Start of fuel load – Moving First Fuel Assembly into the Reactor Pressure Vessel (RPV). • Stage 7: First criticality – First pull of multiple control rods to achieve criticality. • Stage 8: Commencement of commercial operations – Formal handover to operation (end of 100hr warranty run). Management arrangements that support the demonstration of BAT and represent Evidence within relevant Arguments of the BAT Case will be illustrated in the Route Map. Management arrangements are listed and grouped within the following themes: • Chemistry – Procedures for controlling water chemistry and for sample collection, testing and analytical reporting. • Radioactive waste management – Horizon Internal Waste Acceptance Criteria (WAC) for processing radioactive wastes, Discharge Criteria for radioactive discharges to the environment, waste quality plans, instructions on waste characterisation, radioactive waste package design standards and packing instructions, the radioactive waste inventory and record system, non-conformance process and clearance and exemption process. • Commissioning – Commissioning and testing instructions for equipment, performance of waste processing SSC, qualification of monitoring and assay equipment and procedures for establishing water chemistry prior to operation. • Radiation protection – Arrangements for access and egress to controlled areas, Health Physics procedures on collecting samples and carrying out monitoring and surveys. • Decommissioning – Funded Decommissioning Programme (FDP), reports on leaks and releases in a Corrective Action Programme and the Decommissioning Plan. • Management of plant and equipment – Procedures for identifying equipment performing roles in environmental protection, QA inspection process, Engineering Schedule and Maintenance Schedule. • Conduct of maintenance – Maintenance and surveillance procedures for in-service sampling equipment, filters and abatement systems. • Conduct of operations – Procedures for the normal operation of equipment and procedures for the operation of equipment in response to abnormal events. Training in the Power Station project on activities that support the delivery of BAT will be implemented by project hold points that are deemed appropriate and will be recorded in the Route Map. Key training areas identified at this phase in the project are outlined below:

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• General awareness training on BAT – Presentation material will include the structure of the Wylfa Newydd BAT Case, how to use HMS BAT arrangements, an overview of EP-RSR permit conditions that require the application of BAT, relevant REPs and other applicable guidance. Trainees will principally comprise designers and engineers (safety case, civil, mechanical, electrical, systems, configuration and quality). • Practitioners training on BAT – Presentation material provides detailed information on carrying out BAT assessments and optioneering to determine BAT options. The aim of the training is to support persons writing BAT reports, persons who are significant contributors to optioneering or are involved in making decisions relating to BAT options. Trainees will comprise radioactive waste specialists, environmental specialists, radiation protection specialists, project engineers and other appropriate lead engineers. • Radioactive waste packages and packing – Training will provide information on the approved radioactive waste packages in the Power Station and include practical exercises on packing them. Exercises will be phased, starting with trainees packing non-radioactive items and then being supervised packing radioactive waste items. Trainees will comprise operatives in waste management teams. • Users training on a radioactive waste inventory and records system – Tutoring on the computer based applications and software for managing the inventory and records of radioactive waste items. Trainees will comprise operatives and managers in waste management teams. • Radiological Controlled Area (RCA) access and egress controls – Training will include practice exercises in applying change procedures for crossing RCA barriers. Trainees will comprise all individuals with job roles that require access to the RCA. • Training on waste assay techniques and clearance – Training on the operation of instruments used in the non-destructive assay of radioactive waste items and packages. Trainees will comprise the operators of instruments. • Training on Health Physics instruments and equipment – Training on the use of Health Physics counters, probes and equipment deployed in radiation monitoring and contamination surveys, as well as procedures for collecting samples from stack air monitoring equipment. Trainees will comprise Health Physics staff.

2.7 Reporting Structure The claim, argument, evidence structure has been used to develop the Wylfa Newydd BAT Case. For the purposes of demonstrating the application of BAT a claim is defined as: • A clear statement as to what will be achieved; • Confirmation of which REPs have been taken into account when applying BAT; and • A demonstration of compliance with the requirements of those conditions in the EP-RSR permit that are subject to BAT.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Arguments are presented to demonstrate that a claim is valid. Arguments are a series of statements that are required to: • Demonstrate that the claims are valid; • Draw the evidence into a ‘story’; and • Identify uncertainties and assumptions. Important considerations for the preparation of arguments are: • One or more arguments must be established for each claim; • The contribution that each argument makes to fulfilling the claim must be determined; • The evidence that is important to the argument must be identified; and • The impact of uncertainties/assumptions must be described. Evidence is information available to support the demonstration that BAT is being applied. Evidence is required to: • Support the development of arguments; • Substantiate arguments; • Allow examination and challenge; and • Identify key gaps (uncertainties). The approach to develop the Wylfa Newydd BAT Case is illustrated in Figure 2.6-1 and the reporting structure itself is illustrated in Figure 2.6-2.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 2.6-1: Approach to the Reporting Structure of the Wylfa Newydd BAT Case

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[This page is intentionally blank]

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 2.6-2: Report Structure1

1 Note: Four arguments per claim, each with four evidence nodes is shown purely for illustrative purposes. Actual number of arguments and supporting evidence nodes varies per claim. © Horizon Nuclear Power Wylfa Limited 17

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[This page is intentionally blank]

Wylfa Newydd Project – Best Available Techniques (BAT) Case

2.8 Demonstration of Best Available Techniques

2.8.1 BAT Claims Six claims have been developed that set out what Horizon needs to achieve in terms of compliance with those conditions of an EP-RSR permit that require the application of BAT. The claims have been developed under the following titles that form the main structure of the report: Claim 1 – Eliminate or reduce the generation of radioactive waste. Claim 2 – Minimise the radioactivity in radioactive waste disposed of to the environment. Claim 3 – Minimise the volume of radioactive waste disposed of to other premises. Claim 4 – Selecting the optimal disposal routes for waste transferred to other premises. Claim 5 – Minimise the impacts on the environment and members of the public from radioactive waste that is disposed of to the environment. Claim 6 – Apply BAT when characterising and quantifying gaseous and aqueous radioactive waste discharges. Within this document, information that identifies and substantiates the engineering and management controls that contribute to the claims stated above are documented. Engineering controls are expressed as SSC which includes items of plant and equipment. Managerial controls are expressed as methods of operation (MOp) and encompass documented arrangements such as process descriptions, sub-process descriptions, procedures and guidance that are utilised in prescribing requirements and controlling operations. Each claim is validated by a number of arguments which are substantiated with evidence as described in Section 2.7 above. The substantiation is based on that which has already been demonstrated during the GDA stage, that which is being demonstrated on the existing Wylfa Newydd Power Station design and that which will be implemented through forward actions.

2.9 Reader’s Guide It is recommended that the content of the Claims, the Arguments and the FAP are read. The Evidence sections can be examined if required to understand the basis of the Argument and how they have been substantiated, these can be identified by the small grey text. As described in Section 2.2, the Claims, Arguments and Evidence in the GDA Demonstration of BAT report produced by Hitachi-GE has been used as the basis for developing the Wylfa Newydd BAT Case. An information marking system has been used to distinguish information which has been extracted from the GDA BAT Case, and what is new site specific information: • GDA text is denoted by an orange line in the right hand margin. • Site Specific text is denoted by a dark blue line in the right hand margin. Non-technical summaries of the most relevant elements of each argument are demonstrated by bold text. It is also important to note that, as explained in Section 3.2.1, text is being adopted from the UK ABWR GDA BAT Case and used in this document. As such, because the UK ABWR GDA

Wylfa Newydd Project – Best Available Techniques (BAT) Case

BAT Case reflects a one unit site, figures quoted in this document are, unless otherwise stated, assumed to also reflect a one unit site. 3 The BAT Claims

3.1 Claim 1: Eliminate or Reduce the Generation of Radioactive Waste The generation of radioactive waste during the operation of the Wylfa Newydd Power Station is undesirable due to the potentially harmful effects of exposure to members of the public and the environment and the time, trouble and cost incurred in its management. The two UK ABWRs that comprise the Wylfa Newydd Power Station have been designed to avoid the generation of radioactive waste at source. Where this has not been practicable, efforts have been made to minimise the activity and quantity of radioactive waste that will require subsequent management and disposal by permitted means. The Arguments presented in support of this Claim are considered to demonstrate compliance with the standard BAT conditions [Ref-5]: • EP-RSR Permit Condition 2.3.1 'The operator shall use the best available techniques to minimise the activity of radioactive waste produced on the premises that will require to be disposed of on or from the premises.' The Wylfa Newydd Power Station design contains a range of features that contribute to the substantiation of this Claim including: • The design, manufacture and management of nuclear fuel to minimise the potential for a release of Fission Products (FP) from the fuel into the steam circuit or cooling pool water; • The elimination or reduction of materials that are susceptible to activation at all stages of commissioning and operation; • The reduction of the generation of Spent Fuel (SF) and Higher Activity Waste (HAW) for a given energy output; • The reduction of the generation of Lower Activity Wastes (LAW) for a given energy output; • The prompt detection and management of failed fuel; and • The introduction of techniques to be used during commissioning, start-up and shutdown to minimise the incidence of Stress Corrosion Cracking (SCC) of key reactor components. In developing the Arguments presented to demonstrate the validity of Claim 1 the REPs have been taken into account [Ref-7]. The following REPs are considered to be relevant to this Claim: • Principle RSMDP3 ‘the best available techniques should be used to ensure that production of radioactive waste is prevented and where that is not practicable minimised with regard to activity and quantity.’ • Principle ENDP1 ‘The underpinning environmental aim for any facility should be that the design inherently protects people and the environment, consistent with the operational purpose of the facility.’

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.1.1 Argument 1a: Design, Manufacture and Management of Fuel The fuel assemblies present the largest source of radionuclides that are created as a result of nuclear fission in the reactor. Collectively these radionuclides are referred to as FPs. Any release of FPs from the fuel into the steam circuit or cooling pool water has the potential to generate radioactive waste that will ultimately require treatment and/or discharge to the environment. Ensuring that these FPs remain in the fuel and its cladding is a key element of the design and operation of each UK ABWR. It is assumed that the Wylfa Newydd Power Station will initially employ fuel products manufactured by Global Nuclear Fuels (GNF) (3.1.1.1 Evidence: Assumption of GNF Fuel). GNF is engaged in a long-standing and comprehensive programme of work to improve the performance of its products and to reduce the frequency of fuel failures. Fuel failures are typically small cracks in the fuel cladding which allow FPs to be released into the steam circuit. GNF collaborates closely with its customers to monitor the performance of its fuel and to understand the mechanisms that give rise to fuel failures. Comprehensive data are available on the performance of its fuel in reactors in Japan, the United States of America and Europe. Analysis of this data has been undertaken by GNF which concludes that GNF's improvement programme has significantly reduced the frequency of fuel failures within Light Water Reactors (LWRs) (3.1.1.2 Evidence: Analysis of Recent Fuel Failures). It has been assumed that, for at least the first fuel load, GNF’s GE14 fuel will be used. It is however noted that this will not be used for the entire life of the plant as GNF plan to phase this fuel out of production as its fuel products develop. GNF do have newer fuel products but these currently have very little Operational Experience (OPEX) data. It was therefore assumed, due to increased OPEX and due to it already being sufficiently developed that GE14 will be used for at least the first fuel load. Following this it is assumed that Horizon will move to newer and more developed fuel product. The data gathered from OPEX is fed back into the GNF fuel programme and is used to support the development of future enhancements. Developments to reduce fuel failures include: • The introduction and evolution of filters within the fuel assembly to remove debris that can damage the fuel. (3.1.1.3 Evidence: Debris Removal). • Quality control improvements to reduce failures at the Pellet Cladding Interaction (PCI) which can result in the cracking of the fuel cladding leading to a release of FPs from the fuel (3.1.1.4 Evidence: PCI Reduction). • Fuel manufacturing improvements (3.1.1.5 Evidence: Manufacturing Improvements). • The production of guidance to the users of GNF's fuel which clearly defines the operating parameters of the fuel and the means by which fuel failures can be minimised during operating cycles in the reactor (3.1.1.6 Evidence: Manufacturer's Guidance of Fuel Use). • The introduction of a pure zirconium liner to reduce SCC (3.1.1.7 Evidence: Selection of Fuel Cladding Materials). GNF's fuel cladding is manufactured from a zirconium alloy known as Zircaloy 2. This material is widely used in the nuclear industry and has been selected because it is transparent to neutrons, resistant to corrosion and is largely impermeable to the migration of FPs.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Any uranium on the external surfaces of the fuel, referred to as 'tramp uranium', has the potential to undergo nuclear fission and to generate FPs that will enter the steam circuit. GNF has developed QA processes that minimise the potential for the external surfaces of its fuel to become contaminated with uranium during manufacturing processes (3.1.1.8 Evidence: Manufacturing and QA Processes to Minimise Tramp Uranium). Hitachi-GE has developed fuel handling equipment that minimises the potential for damage during transportation, loading, unloading and storage of fuel and SF. OPEX and feedback from the operating fleet of BWRs in Japan, the United States of America and Europe has shown that there is a very low frequency of fuel damage associated with the management of fuel and SF outside of the reactor (3.1.1.9 Evidence: Fuel Handing Equipment - OPEX and Feedback). Creating FPs is an inherent result of nuclear fission in the operation of a Power Station. The design, manufacture and management of nuclear fuel has evolved to virtually eliminate the transfer of FPs from the fuel to the steam circuit and the cooling pool water. This will substantially reduce the amount of radioactivity in any waste that is unavoidably discharged to the environment from Wylfa Newydd Power Station. It will also minimise the quantity of secondary waste that is generated from the management and treatment of any FPs present in the gaseous and aqueous wastes.

3.1.1.1 Evidence: Assumption of GNF Fuel

It has been assumed that, for at least the first fuel load, GNF’s GE14 fuel will be used. It is however noted that this will not be used for the entire life of the plant as GNF plan to phase this fuel out of production as its fuel products develop. GNF do have a newer fuel products but these currently have very little OPEX data. It was therefore assumed, due to increased OPEX and due to it already being sufficiently developed that GE14 will be used for at least the first fuel load. Following this it is assumed that Horizon will move to newer and more developed fuel products.

3.1.1.2 Evidence: Analysis of Recent Fuel Failures

GNFs evolutionary product introduction strategy develops and implements new products and processes that deliver improved fuel performance. The introduction of these design changes has delivered a steady improvement in fuel reliability (Figure 3.1.1.1-1) whilst maintaining design and fabrication-related performance. In the past three decades, fuel reliability has improved by approximately three orders of magnitude. This is based on the instances of leakage of FPs from fuel rods being reduced from over five hundred rods per million operating, to below ten. In previous decades, most plants experienced at least one fuel failure during each cycle. Current practice is for most GNF fuelled plants to operate without a leaker for a number of years; with only a very small minority of GNF customers’ plants experiencing failures in any cycle.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.1.1.1-1 GNF Historical Fuel Reliability Performance

Fleet-wide, GENUSA (a GNF and Enusa Industrias Avanzandas (ENUSA) partnership) drove the improvement in reliability by systematically identifying and characterising failure mechanisms through poolside and hot cell examinations, and subsequently eliminating such failures through changes in design and fabrication of the fuel. The improvements that have been introduced as a result of this work include:

• Improved pellet fabrication in the 1970’s to eliminate primary hydride failures which resulted in the fuel cladding failing;

• Fuel duty operating recommendations in the 1970’s, followed by GE’s invention and patent of zirconium-lined barrier fuel in the early 1980’s, to mitigate PCI;

• Corrosion-resistant cladding, with a chemistry and microstructure specifically targeted to protect against Crud-Induced Localised Corrosion (CILC) failures;

• Improved cladding and welding fabrication and inspection techniques;

• Tightened pellet missing surface specifications to add margin to “duty-related” failures;

• A debris filter to reduce debris related damage to the fuel rods;

• Subsequent advanced debris filter designs to improve resistance to debris ingress; and

• Operating guidance to maximise the capacity factor while minimising the potential for “duty-related” failures associated with power increases after control rod withdrawals.

The current performance trends lead to the following conclusions:

• In the 10x10 fuel type which will be used in each of the UK ABWRs, the leading failure mechanism has been debris fretting which is where debris becomes entrained within the coolant and can damage the fuel assemblies. The only other significant failure mechanism observed has been several “duty-related” (PCI-type) failures, from a small number of plant manoeuvres in over 15 years

Wylfa Newydd Project – Best Available Techniques (BAT) Case

of operating experience. Almost no fabrication related defects are known to have occurred in the past ~15 years of production, which represents over 7 million fuel rods. (In 2010, a single failure occurred at a plant in Spain which may be related to a fabrication defect from the ENUSA facility. A missing pellet surface (MPS) issue, or some other undetected defect, may have been a factor in a failure in Olkiluoto-1 in May 2010.).

• Today’s fuel, with its increased performance capability, has the capability to be operated at increased power levels and for longer durations, both to achieve improved fuel cycle economics and to meet high-energy cycle demands. Power densities, capacity factors, cycle lengths and resultant cycle energies continue to increase and drive fuel duty. This provides operators with the opportunity to replace less fuel during an outage. Refuelling batch sizes approach 50% for some plants, exposing fresh fuel to conditions not typical in the past (e.g., higher duty and, in some applications, high control early in life).

• Advanced debris filters have performed well to date, with lower failure rates achieved relative to prior designs.

3.1.1.3 Evidence: Debris Removal

Debris can become entrained within the coolant and can damage the fuel assemblies which can result in a release of FPs. Debris fretting occurs when various types of debris in the coolant protrude through the lower tie plate (LTP) and cause through wall fretting of the cladding.

In 1990 GNF offered its first intra-bundle debris protection by introducing an LTP that has an entry hole that is one third the size used in the previous LTP design. This reduced the size of debris that could enter the bundle.

In 1996, GNF introduced a debris filter LTP that reduced the size of the debris that could enter the bundle by another factor of three. This filter was offered as an option on 9x9 and 10x10 products at that time. As part of GNF’s zero leaker initiative, it was integrated into the GE14 fuel type as a standard feature in 2001.

The next iteration of the design was to include a debris shield. The debris shield further reduced the size of the debris that could enter the bundle relative to the first generation 10x10 debris filter, with no pressure drop penalty. The debris shield was a perforated metal plate located at the top of the LTP and was held in place by the fuel rods. The design began operation in early 2005. Field experience with the debris shield included 8 reloads and ~2,250 bundles, with only one failure.

GNF then developed a next generation “DefenderTM” filter (Figure 3.1.1.3-1) that further reduced the size of debris, but specifically targeted wires or wire-like debris that have been associated with cladding perforations. The first reload began operation in a United States of America (U.S.) reactor in 2006; the experience as of January 2013 includes ~75 reloads and > 15,000 bundles. Generally, a debris filter has a potential impact on channel pressure drop. The Defender has adopted a non-line-of-sight design which efficiently captures wire debris during operation with no impact on channel pressure drop.

Table 3.1.1.3-1 provides the performance for the first generation 10x10 cast debris filter LTP vs. the more recent filter designs. To date, advanced filters have a failure rate about a factor of 5 to 6 better than the original standard filters.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.1.1.3-1 Development of Defender Filtration Technology

Table 3.1.1.3-1 Failure Rate of Bundles with Advanced Filters

Advanced Filters – Standard Debris Shield Defender either Debris Shield or Defender

Bundles ~22,900 ~2,250 ~15,200 ~17,500

Failure Rate (failed bundles per 2.9 0.5 0.5 0.5 thousand bundles)

3.1.1.4 Evidence: PCI Reduction

PCI is a phenomenon, which results from the dimensional changes experienced by the uranium dioxide pellet and zirconium cladding during in-reactor operation. When stresses build up between the uranium dioxide fuel pellets and the zirconium rod in which they are contained, in the presence of aggressive fission products (e.g. iodine), SCC can occur leading to premature failure of the fuel rods.

There are several features available to GNF that have been incorporated into the bundle design that contribute to the mitigation of PCI-type failures.

Zirconium-lined barrier fuel cladding was introduced in reload quantities in the mid-1980’s as a material solution to PCI-type failures. Power ramp tests (Figure 3.1.1.4-1) and reactor fleet trials demonstrated that a zirconium lined fuel rod was highly effective in mitigating PCI-type failures. GNF fuel designs have continued to employ barrier cladding since then. Barrier cladding provides significantly improved PCI resistance, especially when combined with other mitigation strategies such as core loading and bundle promotion practices. Core loading and bundle promotion practices consist of moving fuel bundles to different locations within the reactor core at each outage in order to improve the core’s efficiency and manage the burn up rates of the different fuel bundles. Because a small number of PCI-type failures have occurred in barrier fuel, operating guidelines have been implemented throughout the BWR fleet, mainly to avoid specific types of operation (for example, very long control intervals followed by step increases in power) that have been correlated with the failure events. Zirconium-lined barrier fuel cladding reduces the pellet area, but does not have a significant effect on uranium inventory. The barrier liner is thin, but is still effective enough for preventing PCI-type failure.

PCI-like failures have been correlated to fuel rods with chipped pellets (or areas of MPS) These defects can increase the probability of a PCI-type defect because there is an additional localised bending stress in the vicinity of the MPS, over and above the stress from the rod pull/power increase. Since the mid 1990’s GNF has adopted improved manufacturing and QA processes that reduce the probability of chipped pellets ending up in fuel rods. The recent introduction of larger chamfered pellet edges has further reduced the likelihood that pellets will be

Wylfa Newydd Project – Best Available Techniques (BAT) Case

damaged during the manufacturing process and improved pellet inspections have been implemented to identify and subsequently reject damaged pellets. Fuel pellet quality is high, and correspondingly, the fuel rod failure rate in the fleet due to PCI-type defects is at an all-time low level.

ABWR core design and operating strategies have been developed that minimise the risk of PCI related fuel failures. These strategies include:

• Control cell core loading pattern;

• Control rod sequence exchanges at regular operating intervals;

• Operating at a flat hot excess reactivity trajectory through the cycle that minimises control rod movement; and

• The use of Reactor Internal Pumps (RIP) to facilitate power manoeuvring.

As a result of applying the above improvements, from the mid-1990s to late 2003, PCI-type failures were observed only in a few legacy 8x8 bundles (and these were found to be in those plants that had not adopted the operational guidance suggestions). No PCI-type failures were observed in the 2 million 9x9 rods delivered from the GNF factories since the introduction of these products in 1990. In the period from late 2003 to early 2007, and in 2010 at one European plant (Olkiluoto-1), GNF experienced several PCI-type failure events in 10x10 fuel. All PCI-type failures in 10x10 barrier fuel occurred in a small number of reactor manoeuvres. The total failure rate in 10x10 barrier fuel due to PCI mechanisms is less than 4 ppm.

3.1.1.5 Evidence: Manufacturing Improvements

Over the past ten-year period, approximately 4,000,000 GNF 9x9 and 10x10 fuel rods have been fabricated and placed in operation. Within this population, GNF has identified a single fuel rod failure caused by a fabrication defect. GNF states that this performance is the result of the safeguards and process improvements that have been implemented since the identification of several factory-related failures in fuel fabricated during the 1980s. A tighter MPS specification was established in 1995 that eliminated the flaws associated with some PCI failures. Multiple inspections were established to assure compliance with the MPS specification to ensure that fuel pellets did not have missing surfaces which could then lead to fuel failures during operation. A state-of-the-art tubing inspection system, called Multi-Inspection and Data Acquisition System (MIDAS) was introduced that provided a thorough clad identification and Outer Diameter (OD) inspection process. It specifically addressed the OD flaws thought to be associated with some past duty-related failures. This system provides two independent flaw inspections: 100% of the OD by eddy current (OD only) for gross flaws as well as some geometries difficult to assess via ultrasonic; and >100% coverage of the OD and identification by ultrasonic, which is typically capable of identifying flaws in the order of 25 to 50 microns (one to two mils).

MIDAS also uses Ultrasonic Testing (UT) to measure critical dimensions along the full length of the fuel rod, since these directly affect such fuel performance behaviours as stress and bow. Fuel rods that do not meet specifications for dimensions or flaw criteria are automatically rejected. The MIDAS record for each rod is recorded in the Fuel Business System (FBS) and references to a barcode identifier laser-marked on each rod. FBS will not allow a reject tube to be processed into fuel rods and ultimately fuel bundles.

Again, learning from past experience, the GNF fuel rod process requires a 100% UT inspection of both the upper and lower end plug weld of the fuel assembly. All historical weld defects are targeted, including tungsten or other foreign material inclusions, inadequate bonding or penetration, and grain boundary separation. Weld records are tied to the fuel rod via its barcode and entered into GNF’s FBS. The system will lock out any weld rejects, precluding further processing.

Whilst GNF has been successful in delivering improvements in fuel design resulting in a reduction in the number of potential manufacturing-related failures, GNF recognises that it must continue to drive for manufacturing excellence to assure similar performance in the future. Several areas GNF is focused on are described below:

• Debris remains the number one cause of failures, and it is imperative that bundles delivered from GNF’s factory be free from debris. Many actions have been taken to assure success in this area,

Wylfa Newydd Project – Best Available Techniques (BAT) Case

including improved inspections and the establishment of debris exclusion zones in the bundle assembly and packing areas. An additional improvement GNF has implemented is the stainless steel fuel-shipping containers, replacing the wooden container, which often presented a risk of paint chips.

• GNF has deployed new grinding stations that also include, via a feedback system, the 100% OD inspection of each pellet. This system will 1) verify that each pellet meets dimensional specifications, 2) feed back to the grinder any necessary adjustments to maintain process control, and 3) inspect the pellets for OD surface defects. This represents a very significant upgrade to the overall pellet fabrication process.

3.1.1.6 Evidence: Manufacturer's Guidance of Fuel Use

GNF fuel is operated in accordance with a variety of design and licensing limits, such as Thermal-Mechanical Operating Limits (TMOL) for linear heat generation rate (kW/ft. or Watts/cm) as a function of exposure, and limits on exposure in terms of bundle average, rod average, or peak pellet, which can vary by the country in which the fuel is licensed. There are specifications for dry or wet storage prior to irradiation, including water quality. For PCI-type failure mitigation, power manoeuvring guidelines are provided which propose exposure-dependent threshold power levels above which power increases should occur at or below certain rates. These guidelines help to mitigate the tensile stress of the cladding (i.e., limiting the rate of pellet thermal expansion) and the release of embrittling FPs (iodine in particular) that promote SCC (i.e., limiting the rate of pellet temperature increase), both of which are key factors contributing to PCI-type failures. Implementation of these manoeuvring guidelines may limit the operating condition, but it has been shown to have a negligible impact on plant capacity factors in today’s BWR fleet.

3.1.1.7 Evidence: Selection of Fuel Cladding Materials

GNF offers a unique version of the zirconium alloy, Zircaloy-2, as its cladding material. Zircaloy-2 is widely used as a cladding material in BWRs, but GNF’s cladding has an inner liner of pure zirconium with additions of iron for corrosion resistance in order to serve as a buffer between the Zircaloy-2 and the swelling of the uranium pellet. The softer liner has been effective as a barrier for PCI since the early 1980’s. The outer cladding is annealed in order to achieve the final state of full recrystallisation.

In BWRs, a major performance deterrent for the cladding is nodular corrosion that occurs due to exposure to the reactor environment. GNF has developed a substantial database reflecting the performance of its cladding which is exposed to today’s modern water chemistries, including hydrogen and zinc injections as well as noble metal applications. The cladding may also be exposed to reactor water chemistry variations experienced within some plants. Based on this growing experience base, GNF has reached the following conclusions:

• Should the cladding be breached due to debris fretting at spacer grid locations, the corrosion resistant liner (mentioned above) is an effective solution to post-failure degradation. It appropriately balances the need for the cladding liner to provide both PCI protection as well as corrosion resistance.

• Modern cladding designs are required to have excellent corrosion resistance to high exposures even in today’s reactor environments, with modern water chemistries.

These conclusions have led GNF to develop and introduce the current Zircaloy-2 barrier cladding, known as Process 9 cladding, which combines the best features of previous designs, while taking steps at both the tube shell supplier and GNF’s Wilmington tube fabrication facility to reduce process variability.

Process 9 utilises modifications to the alloying elements in order to tighten variability in the ingot chemistry and enhance corrosion resistance. Tightening and biasing the alloying elements of iron, tin, nickel and chromium when compared to standard American Society for Testing and Materials (ASTM) Zircaloy-2 specifications has proven to demonstrate better performance in nodular corrosion resistance. In the raw material manufacturing process, a rapid quench from the beta metallurgical phase at the billet stage, known as hollow billet beta quench,

Wylfa Newydd Project – Best Available Techniques (BAT) Case

provides a smaller second phase particle size distribution and a more uniformly reproducible product. Tubing made from hollow beta-quenched billets has excellent corrosion resistance in laboratory corrosion tests and in the field. This was one of the earliest corrosion enhancing processes implemented.

3.1.1.8 Evidence: Manufacturing and QA Processes to Minimise Tramp Uranium

Tramp uranium is uranium or uranium dioxide dust that clings to the outside of the fuel elements and is insufficiently cleaned off during fabrication. Once in the reactor, it will undergo fission and its FPs readily enter the reactor coolant.

A number of measures have been introduced to minimise uranium oxide contamination on fuel rod surfaces. The handling process for unsealed uranium-dioxide pellets (e.g., pellet loading process and upper end plug weld process) has the potential to spread uranium oxide contamination onto the fuel rod surface. The unsealed uranium-dioxide is exhausted by air conditioning equipment to remove any uranium dust that becomes entrained in the air. In over 99% of cases, the measurement result for uranium contamination on the fuel rod surface is lower than the detection limit. QA measures ensure that tramp uranium remains below the QA thresholds. GNF’s QA measures for fuel production will be subject to auditing by an independent inspection agency to provide demonstrable evidence that they have been implemented for a suitable sample of product.

In order to minimise uranium contained in the Zircaloy raw material, GNF’s current material specification is for less than 3.5 ppm uranium in zirconium alloys. In practice the material certification reports indicate that it is usually reported as <0.5 ppm. The uranium content is driven by the input material for Zirconium processing, which is sand. A part of the processing from sand to crystals leaves the uranium in the silica waste stream; separation processing takes the majority of the residual uranium to the hafnium side. GNF’s Zircaloy tube shell suppliers report that the hafnium content in the feed has been significantly reduced since implementing the processing change with the intermediate step going to crystals.

Additional controls during Zircaloy processing include preventing the use any Zircaloy that has been irradiated. Use of returned Zircaloy-2 tubing from rods scrapped during manufacturing that had been loaded with pellets can also result in traces of pellet material if the recycled ingot includes material melted from the tubing. Material from these scrapped rods is prevented from entering the manufacturing process.

The low levels of impurities (low ppm levels) result in a very low level of off-gas FPs activity, even for initial cores with no fuel failures. GNF’s monitoring of off-gas activity data in the fleet supports the observation that uranium impurities in the cladding are at the lowest levels they have been in BWR history. Many cores today that have not had a fuel failure in a decade or more, have extremely low “tramp uranium based” off-gas activity levels, less than 3.7 MBq/s is routinely achieved, and some plants observe less than 1.85 MBq/s.

3.1.1.9 Evidence: Fuel Handing Equipment - OPEX and Feedback

Since 1974, Hitachi-GE has manufactured and supplied the Fuel Handling Machine (FHM) for BWR, ABWR and fuel reprocessing plants. The FHM is Category A Class 1 rated [Ref-28]. The total number of the manufactured and installed FHM’s is 20. No fuel damage or collision of fuel during fuel handling operations has occurred. OPEX validates the effectiveness of the FHM design and systems that preclude the dropping of a fuel assembly.

3.1.2 Argument 1b: Reactivity Control Fluids and materials that pass through the reactor core are exposed to an intense field of neutrons generated from nuclear fission. Interactions with neutrons in some instances result in the generation of activation products which require treatment and/or disposal as radioactive waste. The elimination or reduction of materials that are susceptible to activation is important to the minimisation of radioactive waste produced in the UK ABWRs at the Wylfa Newydd Power Station. A physical method, known as ‘recirculation flow control’ is used in the UK ABWR for controlling the rate of the nuclear reaction in the core (reactivity) (3.1.2.1 Evidence:

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Recirculation Flow Control). The ratio of water to steam is managed by controlling the recirculation flow rate in the core. The water acts as a neutron moderator and increases the rate of fission. Put simply, the greater the ratio of water to steam in the reactor core, the higher the reactivity will be and vice versa.

Each UK ABWR will also use burnable gadolinia (Gd2O3), a neutron poison, for reactivity control (3.1.2.2 Evidence: Gadolinia). The most effective burnable poisons must be depleted in one operating cycle so that no residual poison exists that could impact on the future fuel cycle length. It is also desirable that the positive reactivity from poison burn-up matches the almost linear decrease in fuel reactivity from the build-up of FPs and the depletion of uranium-235. Gadolinia which is a commonly used burnable poison in BWRs has been selected as the concentration can be controlled so that the poison is essentially depleted during the operating cycle. Control rods are used to absorb neutrons and by changing their location are used to control the reactivity in part or all of the reactor core. There are two types of control rod; boron carbide (B4C) rods and hafnium rods (3.1.2.3 Evidence: Introduction of Hafnium Control Rods). Each UK ABWR will use both boron carbide control rods and hafnium control rods in its reactor. Hafnium control rods have a longer operational life and are used not only for shutdown but for reactivity control during normal operation, which results in higher exposure of control rods. The operation of hafnium and boron control rods under normal reactor conditions will be controlled by operatives in the Main Control Room (MCR) using Unit Operating Procedures. The preference will be specified for using the hafnium control rods more frequently than the boron control rods. Boron carbide control rods are normally used only for shutdown. Hafnium control rods reduce the quantity of waste generated from the less frequent replacement and management of control rods at the end of their service life (3.1.2.3 Evidence: Introduction of Hafnium Control Rods). The use of the recirculation flow control method, burnable gadolinia and control rods eliminate any need for chemical agents to control the reactivity. In Pressurised Water Reactors (PWR) chemical agents such as boron are used to control reactivity and are a significant source of the tritium that is discharged to the environment. Reactivity control using ‘recirculation flow control’ compared with boron based water chemistry, eliminates a significant source of tritium. Optimising the generation of tritium is considered BAT as there are no large scale economic systems for the treatment of tritium and the practice across the nuclear industry world-wide is that tritium is managed by discharge to the environment following optimisation of its generation. Discharge data from plants operating in Japan shows that the amount of tritium disposed of to the environment from ABWRs is lower than PWRs (3.1.2.4 Evidence: Reactivity Control - OPEX and Feedback). The physical ‘recirculation flow control’ method that is used to control the reactivity of the UK ABWR core does not, unlike chemical control methods, generate tritium as a by-product. This will substantially reduce the quantity of tritium that will be unavoidably discharged to the environment from Wylfa Newydd.

3.1.2.1 Evidence: Recirculation Flow Control

Reactivity within each of the UK ABWRs is predominately controlled by changing the flow of water, a neutron moderator, through the active core. Simplistically, increasing the flow of water using the RIPs will increase the density of moderator in the reactor core consequently increasing reactivity. This increases reactor thermal power and therefore steam production and ultimately the electrical output of the power station. Conversely reducing the flow of water using the RIPs will decrease the density of moderator in the reactor core and therefore reduce reactivity, thermal power, steam generation and hence electrical output. In comparison the moderator

Wylfa Newydd Project – Best Available Techniques (BAT) Case

within a PWR is single phase which means that this method of reactivity control is not available to this reactor design. For PWRs reactivity is controlled by the addition of boron, a neutron absorber. The nuclear reactions involving boron and associated chemicals result in the generation of tritium in PWRs [Ref-29].

Recirculation flow rate is variable over a range from a minimum pump speed flow of 20% to the flow required to achieve rated core power. The flow control range allows automatic regulation of reactor power output without the requirement to move the control rods [Ref-30].

3.1.2.2 Evidence: Gadolinia

Following decades of manufacturing and OPEX gadolinium (III) oxide (Gd2O3), commonly referred to as gadolinia, has become the industry standard material for use as a burnable neutron absorber in BWRs. These burnable poisons are required for additional control during shutdown and for power output shaping particularly during early parts of the cycle. The most effective burnable poisons are those that become depleted in one operational cycle so that no residual poison exists that could impact on the future cycle length. It is also desirable that the positive radioactivity from poison burn-up matches the almost linear decrease in fuel reactivity from FP’s build-up and uranium-235 depletion.

The fuel rods are loaded with fuel pellets consisting of either uranium dioxide (UO2) or uranium dioxide with gadolinia (UO2-gadolinia). The location of the UO2-gadolinia pellets and the concentration of the gadolinia in these pellets, along with the enrichment of the UO2 in each pellet (both UO2 and UO2-gadolinia) is carefully chosen to provide the desired characteristics. Additionally, the gadolinia concentration in each UO2-gadolinia pellet is determined so that the poison is essentially depleted during the operating cycle. Gadolinia has been used in GE BWRs since the early 1970s, and has proven to be an effective and efficient burnable poison [Ref-30].

The concentration of the gadolinia contained in the uranium oxide and the number of rods that will contain gadolinia will be determined as part of optimising the reactor performance. This process will take account of relevant design policies and the reactor core performance requirements. The desired amount of gadolinia designed into a batch of fuel is dependent upon the fissile enrichment level, batch size and the operating cycle length.

The use of burnable poisons has the following benefits [Ref-31]:

• Reduces the number of control rods that are required for controlling reactivity during operations, which consequently reduces the number of control rods requiring disposal;

• The reduced requirement for control rod usage simplifies core operation and management for the operator; and

• Burnable poisons can be distributed more uniformly than control rods and are less disruptive to the core's power distribution.

Gadolinia was selected for use in the UK ABWR over alternative burnable poisons; notably boron, hafnium and europium for the following reasons [Ref-31]:

• Boron generates helium and tritium during the neutron absorption process. The build-up of these nuclides can cause an internal pressure rise and tritium can result in the hydride embrittlement of the fuel cladding.

• Hafnium and europium generate nuclides that also have a large absorption cross section. The negative reactivity of the burnable poison and its products does not decrease over time and therefore, such absorbers are unsuitable from the viewpoint of fuel economy.

• The residual gadolinia after burn-up is small and less than the other burnable poisons.

A disbenefit of gadolinia has a higher relative cost compared to boron. However, this is outweighed by the benefits presented above. The use of gadolinia also generates other radionuclides. However, the majority of these radionuclides have relatively short half-lives of less than one year (except Gd-152) and there is no significant radiological impact on waste categorisation, treatment and disposal when the fuel (including the

Wylfa Newydd Project – Best Available Techniques (BAT) Case

burnable poison) is disposed of. Gd-152 is considered to be a stable nuclide and has no impact on dose and disposal [Ref-31].

3.1.2.3 Evidence: Introduction of Hafnium Control Rods

The control rods and control-rod drive systems provide the functionality of controlling the power from the reactor and adjusting the power distribution across the reactor core. Power control is implemented by varying the depth of the control-rod insertion and the power distribution is adjusted by changing the location of the control rods.

The control rod contains neutron absorbers; either boron carbide powder filled stainless steel tube or hafnium metal. Figure 3.1.2.3-1 details the assembly of a boron carbide control rod. Hafnium control rods are similar in design but with hafnium replacing the boron carbide powder filled stainless steel tubes [Ref-32].

Table 3.1.2.3-1 shows the control rod specification for boron carbide type and hafnium type control rods.

Handle Stainless steel tube (Absorber Sheath Roller rod) Roller

End plug

Iron wool

Boron carbide Cooling hole B4C

Stainless steel Dimple tube Stainless (Absorber rod) steel ball

Cladding Sheath tube

Neutron absorbing material Fuel assembly Centre post

(Tie rod) Iron wool

End plug Connector

Detailed drawing of stainless steel tube (Absorber rod)

Bayonet coupling socket

Figure 3.1.2.3-1: Boron Carbide Control Rod Assembly [Ref-32]

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Table 3.1.2.3-1: Main Specification of Control Rods

Weight (kg) Active Absorber Neutron absorber Control rod [Ref-116] Length (m) Absorber Number

Type I 90 3.63 Boron carbide 72 cladding tubes powder. filled with boron carbide powder (per control rod).

Type II 110 3.63 Hafnium metal. 16 Hafnium tubes (per control rod).

Hafnium control rods will be used in the 29 control cells located in the reactor core. Boron carbide control rods will be used in the 176 shutdown cells. The purpose of the control cells is to control reactivity during power operations whilst the other control rod cells are used for shutdown operations [Ref-33]. An advantage of hafnium control rods is that they have a longer operational life than boron carbide control rods. The process of neutron adsorption in boron carbide control rods depletes the boron as illustrated in Figure 3.1.2.3-2. The boron reacts, emitting helium resulting in a reduction in the amount of boron-10.

B + n Li + He 10 7 4 Figure 3.1.2.3-2: Neutron→ Capture Reactions of Boron

Hafnium control rods are able to capture many neutrons because most of the isotopes produced by the neutron capture reactions still have the capability to capture neutrons, as shown in Figure 3.1.2.3-3. Thus, the nuclear life time of hafnium is longer than that of boron and as a result the hafnium control rods do not need replacing as frequently, resulting in the generation of less waste. Boron carbide control rods are still required for shutdown due to their large cross section of thermal neutron capture.

178Hfm 179Hfm 180Hfm

(n, γ) (n, γ) (n, γ) (n, γ) (n, γ) (n, γ)

174Hf 175Hf 176Hf 177Hf 178Hf 179Hf 180Hf ・・・・ 182W

175Lu

Figure 3.1.2.3-3 Neutron Capture Reactions of Hafnium

During power operation, about half of the 29 hafnium control rods located in control cells are used. The inserted control rods are exchanged at an interval determined by the operator typically following 3~4GWd/t exposure. The lifetime of a hafnium control rod is estimated to be 2 cycles with continuous irradiation, where 1 cycle is expected to be “operation + outage = 17 + 1 months”. Furthermore the control rod pattern (the location where the control rods are inserted) is changed regularly during the power operation. As a result, the number of

Wylfa Newydd Project – Best Available Techniques (BAT) Case

hafnium control rods replaced is estimated to be 29 per 4 cycles (72 months), which is nominally equal to 5 rods per year [Ref-33].

If boron carbide control rods were used as the control cell control rods instead of the hafnium rods during power operation, as they were in previous BWR designs, the replacement rate would be approximately double as the lifetime of a boron carbide control rod with continuous irradiation is estimated to be 1 cycle as displayed in Table 3.1.2.3-2. The use of hafnium rather than boron in the control cell control rods results in approximately 300 kg less waste per year that will require pre-treatment and disposal. Disposal of hafnium control rods has been assessed in the disposability assessment [Ref-34] and are considered to pose no disposal issues.

Table 3.1.2.3-2: Control Rod Waste per Year

Items Hafnium control rod Boron carbide control rod2

Control rod lifetime with continuous use 2 cycles 1 cycle

Exchange frequency 29 per 4 cycles 29 per 2 cycles

Control rod waste 5 per year 10 per year

Further to this, the generation of helium and tritium as a result of boron-10 neutron capture results in the stainless steel tubes containing boron carbide powder becoming gradually pressurised and swollen by an increase in the internal pressure and volumetric swelling. This swelling has the potential to result in very small scale, localised cracking of the tubes containing the boron carbide and the release of small quantities of tritium into the reactor water. Monitoring of tritium in the reactor water will detect any increase in the concentration of tritium and will enable Horizon to identify if a crack has occurred. Appropriate management of the operational lifetime of the boron carbide control rods will minimise the likelihood that cracking shall occur. However, in the event that a tube does crack discharges of tritium into the environment are not expected to increase by more than a few percent [Ref-35]. Following cracking there may also be a release of carbon-14. However this is considered to be small, 6 orders of magnitude less than the normal yearly carbon-14 discharge [Ref-36]. In hafnium type control rods, hafnium captures neutrons, and γ-rays are emitted. Thus, hafnium type control rods do not emit any gaseous products and therefore do not suffer from swelling related cracking.

3.1.2.4 Evidence: Reactivity Control - OPEX and Feedback

By using the recirculation flow control method, burnable gadolinia and control rods the UK ABWR does not require the injection of chemicals to control reactivity. Chemical controls in PWRs, through the introduction of boron and associated chemicals, produce quantities of tritium which is then disposed of to the environment. By eliminating the requirement for boron and associated chemicals the ABWR produces significantly less tritium. This is demonstrated in Table 3.1.2.4-1 which summarised discharge data from Japan [Ref-37] and the expected annual discharge from the UK European Pressurised Water Reactor (EPR) (UK EPR PCSR chapter 11.3) [Ref- 38].

Table 3.1.2.4-2: Comparison of Tritium Discharge between ABWR and PWR

ABWR annual UK EPR expected annual Radionuclide discharge (Bq/y) discharge (Bq/y)

H-3(gas) 1.2E+12 5.0 E+11

H-3(liquid) 2.0E+11 5.2 E+13

2 In the case where boron carbide control rods are used as the control cell control rods.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.1.3 Argument 1c: Efficiency of Fuel Use The efficiency with which the nuclear fuel is used in the Wylfa Newydd Power Station and the frequency with which it is changed will influence the amount of SF and HAW that is generated during operations. Reducing the generation of SF and HAW for a given energy output is an important part of optimising the nuclear fuel cycle from an environmental perspective. This applies to the selection of the nuclear fuel and the final choice for its management prior to disposal. The design of each UK ABWR prevents the unnecessary generation of waste and discharges associated with its management by ensuring that the minimum quantity of SF is produced per unit of electricity generated. The cores are arranged as an upright cylinder and contain a large number of fuel assemblies (3.1.3.1 Evidence: Configuration and Geometry of the Reactor Core). The design of the core has evolved over many years of BWR operation. Compared with a PWR, the BWR has a lower power density, larger fuel inventory, more fuel bundles, and smaller ratio of fuel assembly exchange. The flexibility of the fuel loading pattern allows fresh fuel to be placed at the core interior and old fuel to be placed at the core periphery. In addition, each fuel rod has axial enrichment distribution with lower enrichment at both top and bottom end. These BWR characteristics lead to increased fuel efficiency due to a decrease in neutron leakage. Additionally the BWR has “spectral shift operation” which also contributes to increased efficiency. In a BWR, the water in the core moderates (slows) neutrons created from fission which in turn go on to cause further fissions. Increased moderation (less boiling so fewer steam bubbles – also known as low void fraction) produces low energy spectrum neutrons which promotes fuel burn up. Decreased moderation (more boiling so more steam bubbles – also known as high void fraction) produces high energy spectrum neutrons so decreases fuel burn up but this also increases plutonium production. By taking advantage of these BWR characteristics, the core is operated at the high void fraction, from the start to the middle of the operation cycle, so that plutonium build-up is promoted. At the end of the operation cycle, the core is operated at the low void fraction (less reactor water boiling is allowed) so that plutonium burn-up is then promoted. The reactivity of the plutonium that is generated improves fuel efficiency. The first shutdown for refuelling will take place around 13 equivalent full power months after the start of initial power operations. Thereafter the operational length has been assumed to be 18 months although it can be varied between 12 and 24 months using GE14 fuel. The desired fuel cycle length is achieved by optimising the refuelling batch size and the average enrichment of the fuel bundles and is a balance between fuel efficiency and amount of radioactive waste produced (3.1.3.2 Evidence: Efficiency of Fuel Use - OPEX). The design arrangement of the core and the operating regime of the UK ABWR has been optimised to ensure that fuel is used as efficiently as possible and generation of thermal energy is maximised. This will reduce the amount of spent fuel that is generated at Wylfa Newydd per unit of electricity that is exported to the National Grid.

3.1.3.1 Evidence: Configuration and Geometry of the Reactor Core

The BWR has the ability to accept fuel design advancements owing to the core lattice configuration that separates the fuel assembly from the control rods. Since the initial deployment of the BWR, fuel designs have undergone numerous evolutions and reactors that were initially fuelled with 7x7 fuel bundles (i.e. bundles comprised of a 7x7 array of fuel pins) are now loaded with 10x10 bundles with substantial performance and reliability improvements. Relative to early designs, contemporary fuel bundles can generate more than twice the amount of energy with reduced enrichment requirements and demonstrate greatly increased reliability. Fuel technology is continuously

Wylfa Newydd Project – Best Available Techniques (BAT) Case

advancing and the ABWR is designed to utilise fuel technology for the entire BWR fleet - and the fleet fuel experience is applicable to the ABWR.

In addition, the ABWR has adopted an N-lattice geometry that provides a small increase in the intra-assembly bypass gap width relative to BWR/5-6 designs. The N-lattice benefits include an increased cold shutdown margin, moderate void reactivity coefficients, and increased margin to channel/control rod interference. This allows for greater core design and operational flexibility as well as simplicity. This is reflected in the equilibrium core design for the UK ABWR shown in Figure 3.1.3.1-1. In this core the fresh fuel assemblies are distributed throughout the centre region, alongside assemblies that have already been partially burned. The lower reactivity, higher burn-up assemblies are placed in the outer peripheral region to minimise neutron leakage. The core is operated at rated power and 90% core flow throughout most of the cycle utilising only deeply inserted control rods, thereby promoting spectral shift operation and achieving a high degree of fuel efficiency. This optimised core design satisfies all thermal margin design limits owing to the relatively low power density of the UK ABWR, in combination with a GE14 10x10 bundle design. The reload batch size and enrichment have been selected to achieve a batch average discharge exposure of 50 GWd/t.

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34

1 3 3 3 3 3 3 Burn-up 68

2 Bundle Type 3 3 3 3 3 3 3 3 3 3 3 3 F Cycle N 66

3 3 3 1 3 1 3 1 1 3 1 3 1 3 3 1 Cycle N-1 64

4 Control Cell 3 3 3 3 1 1 1 1 1 3 3 1 1 1 1 1 3 3 3 3 2 Cycle N-2 62

5 3 3 1 1 1 1 F 3 F 1 1 1 1 F 3 F 1 1 1 1 3 3 3 Cycle N-3 60

6 3 3 1 3 F 3 F 1 F 2 F 2 2 F 2 F 1 F 3 F 3 1 3 3 4 Cycle N-4 58

7 3 3 2 1 F 1 F 1 2 1 F 2 F F 2 F 1 2 1 F 1 F 1 2 3 3 56

8 3 3 1 1 F 1 F 2 F 1 F 2 F 2 2 F 2 F 1 F 2 F 1 F 1 1 3 3 54

9 3 1 3 F 1 1 2 F 2 1 2 F 2 1 1 2 F 2 1 2 F 2 1 1 F 3 1 3 52

10 3 1 F 1 F 2 2 2 F 2 1 2 F 2 2 F 2 1 2 F 2 2 2 F 1 F 1 3 50

11 3 3 1 3 F 2 F 2 F 2 F 2 F 2 F F 2 F 2 F 2 F 2 F 2 F 3 1 3 3 48

12 3 3 1 1 F 1 F 2 F 2 F 2 F 2 F 2 2 F 2 F 2 F 2 F 2 F 1 F 1 1 3 3 46

13 3 1 1 F 1 2 1 1 2 F 2 1 2 F 2 1 1 2 F 2 1 2 F 2 1 1 2 1 F 1 1 3 44

14 3 3 1 3 F 1 F 2 1 2 F 2 1 2 F 2 2 F 2 1 2 F 2 1 2 F 1 F 3 1 3 3 42

15 4 3 1 1 F 2 F 3 F 2 F 2 F 2 1 2 F F 2 1 2 F 2 F 2 F 3 F 2 F 1 1 3 4 40

16 3 3 3 1 1 F 2 F 2 F 2 F 2 F 2 F 2 2 F 2 F 2 F 2 F 2 F 2 F 1 1 3 3 3 38

17 3 3 1 3 1 2 F 2 1 2 F 2 1 2 F 2 2 2 2 F 2 1 2 F 2 1 2 F 2 1 3 1 3 3 36

18 3 3 1 3 1 2 F 2 1 2 F 2 1 2 F 2 2 2 2 F 2 1 2 F 2 1 2 F 2 1 3 1 3 3 34

19 3 3 3 1 1 F 2 F 2 F 2 F 2 F 2 F 2 2 F 2 F 2 F 2 F 2 F 2 F 1 1 3 3 3 32

20 4 3 1 1 F 2 F 3 F 2 F 2 F 2 1 2 F F 2 1 2 F 2 F 2 F 3 F 2 F 1 1 3 4 30

21 3 3 1 3 F 1 F 2 1 2 F 2 1 2 F 2 2 F 2 1 2 F 2 1 2 F 1 F 3 1 3 3 28

22 3 1 1 F 1 2 1 1 2 F 2 1 2 F 2 1 1 2 F 2 1 2 F 2 1 1 2 1 F 1 1 3 26

23 3 3 1 1 F 1 F 2 F 2 F 2 F 2 F 2 2 F 2 F 2 F 2 F 2 F 1 F 1 1 3 3 24

24 3 3 1 3 F 2 F 2 F 2 F 2 F 2 F F 2 F 2 F 2 F 2 F 2 F 3 1 3 3 22

25 3 1 F 1 1 2 2 2 F 2 1 2 F 2 2 F 2 1 2 F 2 2 2 F 1 F 1 3 20

26 3 1 3 F 1 1 2 F 2 1 2 F 2 1 1 2 F 2 1 2 F 2 1 1 F 3 1 3 18

27 3 3 1 1 F 1 F 2 F 1 F 2 F 2 2 F 2 F 1 F 2 F 1 F 1 1 3 3 16

28 3 3 2 1 F 1 F 1 2 1 F 2 F F 2 F 1 2 1 F 1 F 1 2 3 3 14

29 3 3 1 3 F 3 F 1 F 2 F 2 2 F 2 F 1 F 3 F 3 1 3 3 12

30 3 3 1 1 1 1 F 3 F 1 1 1 1 F 3 F 1 1 1 1 3 3 10

31 3 3 3 3 1 1 1 1 1 3 3 1 1 1 1 1 3 3 3 3 08

32 3 3 1 3 1 3 1 1 3 1 3 1 3 3 06

33 3 3 3 3 3 3 3 3 3 3 3 3 04

34 3 3 3 3 3 3 02 01 03 05 07 09 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 61 63 65 67 Figure 3.1.3.1-1 Reference Equilibrium Core Loading Map

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.1.3.2 Evidence: Efficiency of Fuel Use - OPEX

Fuel efficiency can be evaluated in several ways including:

• The reload batch size;

• The batch average discharge burn-up (or exposure);

• The enrichment required for a specified discharge exposure; and

• The amount of thermal energy produced per unit mass of fissile U235 loaded.

The UK ABWR with its moderate thermal power density, expanded core flow capability, large core size and favourable operating limits has the merit of maximising discharge burn-up with modest enrichment requirements, as demonstrated in the UK ABWR equilibrium core design. These features allow the ABWR to have fuel replacement requirements that are among the lowest in the BWR fleet.

A key feature of the ABWR is the large core flow range available at rated power conditions, from 90 to 111%, provided by ten RIPs. This core flow capability can be used to enhance a “spectral shift” operating strategy – spectral shift is an inherent BWR characteristic. In the BWR, the fuel operating in a high void fraction region (i.e., the upper core region) experiences a high energy neutron spectrum, resulting in the production of plutonium. The amount of plutonium increases in proportion to the burn-up period with high void fraction. A high void fraction can be achieved by operating at 100% of rated core power and 90% of rated core flow for an extended time. For the UK ABWR equilibrium core, operation at low flow is possible for the majority of the operating cycle. After approximately three-quarters of the operating cycle, core average axial power shifts toward the upper core region and the plutonium is utilised. Beginning and end of cycle void fraction and axial power profiles illustrating this spectral shift effect are shown in Figure 3.1.3.2-1 and 3.1.3.2-2 respectively. Contemporary fuel designs help to promote this spectral shift operating strategy, which significantly improves fuel utilisation. Fuel designs incorporate axial zoning of enrichment and gadolinia to further promote spectral shift operation by allowing a more bottom peaked axial power distribution while still satisfying thermal margins. Axial zoning also eliminates the need to insert shallow control rods for power shaping. In summary, each of the UK ABWRs can achieve a high level of fuel efficiency by adopting an enhanced spectral shift operating strategy made possible by a 90-111% flow range and contemporary fuel designs.

BWR fleet history and fuel evolution also have demonstrated dramatic improvements in fuel utilisation. BWRs that were loaded with 7x7 fuel bundles during the 1970’s typically achieved discharge exposures of approximately 20 GWd/t. BWRs loaded with 10x10 fuel bundles today can achieve discharge exposures of 50 GWd/t.

Finally, using the GE14 10x10 fuel bundles, fuel cycle lengths can be varied from 12 to 24 months. Shorter fuel cycles allow for better fuel utilisation and burnup but result in more outages over the plants lifetime. Due to the inherent generation of radioactive waste during an outage this also leads to greater radioactive waste generation.

Longer fuel cycles result in less outages so less radioactive waste is produced but fuel utilisation and burn up is affected. This is due to there being less certainty of the core’s starting conditions (because it is based on the end conditions of the previous cycle) so slightly more conservative estimates are applied when deciding how much fuel to replace. This ultimately leads to fuel being replaced prior to reaching its maximum burn-up.

An 18 month fuel cycle was assumed for GDA on the basis that it is a compromise between the two extremes. However, as stated, the actual cycle length to be implemented will be decided by Horizon and will be decided on by balancing factors such as fuel burn up with radioactive waste produced.

In summary, the UK ABWR coupled with modern fuel designs can support a wide range of operating plans while achieving a high level of fuel efficiency. Advances in fuel technology have consistently demonstrated reduced fuel requirements for the BWR fleet such that the GE14 (fuel type) equilibrium fuel cycle projection for Wylfa Newydd Power Station can be considered conservative when evaluating SF generation over the plant lifetime.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

1.8 100

Power Distribution 1.6

Void Fraction 1.4 75

1.2

1.0

50 Void FractionVoid (%) 0.8 Power distribution(relative power) 0.6

25 0.4

0.2

0.0 0 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Fuel Rod effective length Axial Position Fuel Rod effective length lower end (relative position from fuel rod effective length lower end) upper end

Figure 3.1.3.2-1 Axial Power and Void Fraction Distribution (Beginning of Cycle)

1.8 100

1.6

1.4 75

1.2 Power Distribution

1.0 Void Fraction 50 Void FractionVoid (%) 0.8 Power distribution(relative power) 0.6

25 0.4

0.2

Figure 3.1.3.2-2 Axial Power and Void Fraction Distribution (End of Cycle)

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.1.4 Argument 1d: Detection and Management of Failed Fuel The GE-14 nuclear fuel that will be used in each of the UK ABWRs is designed, manufactured and managed to minimise the potential for fuel failures to occur that could subsequently result in the release of FPs into the steam circuit (3.1.1: Argument 1a: Design, Manufacture and Management of Fuel). In the unlikely event that a fuel failure occurs and FPs enter the steam circuit, the UK ABWR has a range of features (3.2.2: Argument 2b: Charcoal Adsorbers for Noble Gases and Iodine) that allow for prompt detection and management of the failed fuel pin. The gamma radiation level of the gases entering the charcoal adsorber in the Off-Gas System (OG) is continuously measured. The measurements are indicated and recorded in the MCR. In addition, routine grab samples are spectroscopically analysed in the laboratory to obtain isotopic concentrations. Collectively this analysis is employed to indicate fuel failures (3.1.4.1 Evidence: Detection System in Gaseous Waste Treatment System). The design of each of the UK ABWR reactor cores allows the MCR operators to detect the location of the failed fuel bundle by the selective insertion of control rods around fuel assemblies (3.1.4.2 Evidence: Procedures for Locating Fuel Failure in Reactor Core). Insertion of the control rods reduces nuclear fission in immediately adjacent fuel assemblies and allows the operator monitoring the concentration of FPs in gases from the condensers to detect the source of the leak. The source of any leak is identified when changes in the concentration of FPs entering the OG correlate with the movement of control rods at specific locations within the reactor core. Following detection of the failed fuel the operator is then able to apply the failed fuel guidelines (3.1.4.2 Evidence: Procedures for Locating Fuel Failure in Reactor Core) (e.g. continue an operation if the activity is less than the prescribed limit). The design of the UK ABWR allows a graduated response to be taken to a fuel failure in the reactor and provides sufficient flexibility to allow Horizon to develop operating procedures to manage fuel failures associated with expected events (as defined in the Methodology for Expected Event Selection [Ref-3]) and accident conditions. At the following outage the failed fuel will be removed and transferred to the Spent Fuel Storage Pool (SFP) where it is cooled to minimise any release of FPs. In the SFP, a visual inspection will be carried out to identify the failed rod and to determine the cause of failure. Other techniques including UT and inspection using a fiberscope are available to the operator to support the inspection if required. Depending on the severity of the failure Horizon has a number of options that can be utilised to store the fuel and to ensure that the spread of contamination is minimised (3.1.4.3 Evidence: Management of Failed Fuel). Leaks of FPs from small cracks in the fuel cladding are infrequent but expected events. Horizon will be able to detect such leaks promptly and implement a response that will ensure that discharges of radioactivity to the environment remain very low.

3.1.4.1 Evidence: Detection System in Gaseous Waste Treatment System

The ‘Approach to Sampling and Monitoring’ report [Ref-39] describes the monitoring that is undertaken to measure the concentration of FPs entering the OG and to enable the detection of fuel failures and cladding defects. The monitoring arrangements include a gross gamma Continuous Emission Monitor (CEM) located at the inlet of the OG. The measurements are displayed and recorded in the MCR.

In the event of a fuel failure or cladding defect it is expected that radionuclides (such as Kr-83m, Kr-85, Kr-85m, Kr-87, Kr-88, Kr-89, Kr-90, Xe-131m, Xe-133, Xe-133m, Xe-135m, Xe-137, Xe-138, Xe-139) will be released directly from the failed fuel pin and ultimately enter the OG.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

In such an event, the CEM would alert operators to elevated gamma radiation levels in the OG and as such will act as the first indication of fuel failure.

During GDA Hitachi-GE demonstrated that analysing grab samples to identify specific radionuclides that are known to be released in the event of a failure will provide Horizon with the information required to manage such an event [Ref-39]. The off-gas isotopic grab samples would be taken periodically in accordance with Health Physics procedures and analysed in the laboratory. In the event that an increase in radioactivity is detected by the CEM, a System Abnormal Operating procedure would specify an increase in the frequency of collecting grab samples to support the diagnosis of the fuel failure.

3.1.4.2 Evidence: Procedures for Locating Fuel Failure in Reactor Core

The UK ABWR has been designed to enable the operator to identify and isolate a fuel bundle containing a fuel failure. When the abnormal increase of radioactivity is observed, MCR operating procedures are chosen based on the criterion of the activity level as follows:

1) Raise the level of monitoring. When the activity increases specifically and reaches a certain criterion (Criteria 1 in Figure 3.1.4.2-1), the operator increases the frequency of monitoring (e.g. increase the frequency of collecting a grab sample).

2) Confirm if a fuel failure event has occurred. When the activity increases to a certain criterion (Criteria 2 in Figure 3.1.4.2-1), the operator becomes aware that there is a fuel failure. The operator may conduct Power Suppression Testing (PST) to locate the failed fuel bundle as the activity level is sufficient to measure the response of the off-gas monitor during PST. A description of PST is provided in Figure 3.1.4.2-2.

3) Plant shutdown. When the activity increases up to a certain criterion (Criteria 3 in Figure 3.1.4.2-1) or continues to rise, the operator shuts down the plant.

Figure 3.1.4.2-1 also provides an overview of the procedures to be considered in the event of a fuel failure. Performance against the criterion is determined by analysis of grab samples (gas) and off-gas inlet radiation monitoring. The monitored activity depends on both the level of damage and the number of damaged fuel pins.

Figure 3.1.4.2-1 Operating Procedure against Fuel Failure

Wylfa Newydd Project – Best Available Techniques (BAT) Case

PST allows an operator to safely locate failed fuel in the core during operation and to subsequently enable the local management of the defect location to mitigate secondary damage and the associated fuel pellet material loss. This therefore helps to avoid mid-cycle outages following fuel failure.

The PST process consists of the following:

1) Transition to partial power with the insertion of control rods to prevent the damage of the failed fuel during PST.

2) Inserting and withdrawing one control rod or a certain group of control rods at a time. Failed fuel can be identified by the response of the OG monitor. Generally the fuel adjacent to the control rod with the largest activity change contains a failure.

3) Insertion of control rods around the suspected failed fuel channel to decrease its power and then keep the linear heat generation rate of its fuel rod below the limit. 4) Return to the rated power. Continue normal operations. Figure 3.1.4.2-2 Description of PST

After the plant has been shut down the concentration of FPs in the water inside each fuel assembly is measured to identify the failed fuel.

3.1.4.3 Evidence: Management of Failed Fuel

Once the plant has been shut down, the failed fuel is moved from the core to the SFP using the same process used for all other fuel. Once located in the SFP the operator will undertake visual inspection to identify the failed rod and to determine the cause of the failure. Typically visual inspection is carried out using an underwater camera. If the failed rod or the cause of the failure cannot be identified using visual inspection Horizon can use alternative techniques such as UT or using a fiberscope.

In accordance with relevant good practice [Ref-40] the option for storing failed fuel depends on the severity of the failure. For small pin hole failures such as those defined as an expected event (as defined in the Methodology for Expected Event Selection [Ref-3]) Horizon is likely to store the fuel directly in the SFP. The operator also has the option to isolate the fuel in a canister to prevent the spread of contamination if required. Horizon will develop and document the approach to managing failed fuel in the SFP [FAbc-1].

A number of different options for the containerised storage of failed fuel rods are available, ranging from removal and storage of a single pin to the storage of a whole assembly in a vented filtered container. A range of options are also available to provide additional containment if Horizon determines that it is appropriate to do so. These options are not foreclosed by the design of the UK ABWR and will be selected by Horizon based on a risk assessment performed on a case by case basis following identification of failed fuel [FAbc-1].

Whilst Horizon will select the optimal storage method, the storage of failed fuel is not expected to result in the spread of contamination or an increase in the radioactivity of the SFP water. This is because the fuel will start to be cooled once it has been placed into the SFP and the release of fission products from the fuel rod is predominantly dependent on the power increase that occurs in the core and not within the SFP.

3.1.5 Argument 1e: Commissioning, Start-up, Shutdown and Outage Procedures Processes that have the potential to occur during commissioning, start-up, shutdown and outage of the reactor could result in an increase in direct doses to operators and a small increase in the generation of radioactive wastes. These are: • The activation of unrecovered foreign material within systems from construction and commissioning activities;

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• An increase in the generation or mobilisation of Corrosion Products (CP) that are susceptible to activation as they pass through the reactor core; and • An increase in the incidence of SCC of key reactor components that will subsequently require replacement. There are a number of techniques that have been developed for use during commissioning, start-up, shutdown and outage of the plant that collectively reduce the generation of radioactive waste during each operating cycle. These techniques have been successfully deployed on the ABWR reactor fleet in Japan. Experience of operating BWRs in Japan has shown that as oxide films, which inevitably form on pipe and vessel internals, develop, they incorporate radionuclides, notably cobalt-60, which represent a significant contribution to worker doses during outages. As such, during commissioning some pipes and vessels may undergo (as necessary), processes to create oxide films on internal surfaces (3.1.5.1 Evidence: Alkali Pre-Filming Technique). This ensures that the oxide layers formed are done so in a radiologically clean environment which prevents the later incorporation of radionuclides. This contributes to a reduction in the radiological dose from pipes in the Reactor Water Clean-up System (CUW) of approximately 80% as demonstrated in Japanese BWRs. This approach also prevents the incorporation of non-radioactive cobalt-59 ions so a negative consequence is that cobalt-59 ions that are not accumulated on pipe and vessel internals have the potential to become activated as they pass through the reactor core. This will result in a very small increase in the concentration of radioactivity in spent demineralisers but it is deemed that the benefit from dose reduction outweighs the detriment associated with the small increase of radioactivity in the demineraliser resins. Deployment of Foreign Material Exclusion (FME) measures during construction and commissioning of systems can prevent debris remaining in water and steam systems during operation (3.1.5.2 Evidence: FME Strategy). Foreign material within systems after first criticality or when restarting after an outage are exposed to neutron radiation resulting in the generation of activation products. Foreign material can also degrade the performance of components causing premature failure and can cause fuel failures leading to FP releases into the coolant water. A FME Strategy sets out the overall approach and practicable steps that can be undertaken to avoid foreign material. Preventative actions e.g. housekeeping, substantiating FME risk areas and inspections are more effective than methods of recovering foreign material from the system after operation. If determined necessary at any point during the lifetime of the two UK ABWRs at the Wylfa Newydd Power Station, a prescribed injection of Hydrazine, Oxalic Acid and Potassium Permanganate (HOP), developed by Hitachi-GE, may be employed to decontaminate and remove radioactivity from the internal surfaces of pipes and vessels during outage. HOP has been demonstrated in service to deliver Decontamination Factors (DF) of between 10 and 80 with a commensurate reduction in average dose rate of approximately 95% (3.1.5.3 Evidence: HOP Decontamination Process). Additionally, the wastes that are generated from this process are likely to be suitable for processing through the Wylfa Newydd Power Station waste treatment facilities. Crud generated from ferrous materials that are present in the steam circuit after installation, maintenance or decontamination operations has the potential to pass through the reactor core and become activated. Plant start-up arrangements include recirculating condensate and feedwater through a condensate clean-up system prior to bringing the reactor back into service (3.1.5.4 Evidence: Water Conditioning). This allows any residual material in the condensate and feedwater systems to be captured in filters and minimises the potential to

© Horizon Nuclear Power Wylfa Limited 41 Wylfa Newydd Project – Best Available Techniques (BAT) Case generate manganese-54 and iron-59 from the activation of ferrous materials which would otherwise then require treatment as radioactive waste. Enhancements have been made to shutdown techniques through the development of Low Temperature Residual Heat Removal Shutdown Cooling (LT-SHC) (3.1.5.5 Evidence: LT- SHC). This reduces the temperature at which the Residual Heat Removal (RHR) is brought in to service for shutdown cooling which has the potential to reduce radioactivity deposition on the internal surfaces of pipes within the RHR system by approximately 80%. This in turn leads to a commensurate reduction in the volume of maintenance related radioactive waste and decontamination fluids. Horizon has also considered staggering the fuel cycles of the two reactors to mitigate the higher aqueous and gaseous discharges associated with having two reactors on an outage at the same time. However, it is recognised that this cannot always be guaranteed (3.1.5.6 Evidence: Fuel Cycle Staggering). The commissioning programme of the Power Station will structure the phases of construction testing, pre-operation testing and start-up testing. Identifying the suitable tests and acceptance criteria for SSC will ensure that they are not handed over to Horizon without demonstrating their functions and expected performance in environmental protection (3.1.5.7 Evidence: Commissioning Programme). The exact commissioning, Start-up, Shutdown and Outage Procedures have not yet been finalised but they will be based on a number of proven techniques and experience gained from operating previous generations of BWRs. The commissioning, Start-up, Shutdown and Outage Procedures will collectively aim to reduce the corrosion of reactor internals and to reduce the generation of activation products. This will aim to reduce to ALARP doses to workers and the activity of components that will eventually become radioactive waste at Wylfa Newydd.

3.1.5.1 Evidence: Alkali Pre-Filming Technique

Oxide films develop on the surface of carbon steel pipes and components during normal operation of the Power Station. If the oxide film is produced during reactor operation, cobalt-60 and other radioactive corrosion products entrained within the reactor coolant become deposited amongst the film and contribute towards worker dose during outages and increased ILW arisings. The alkali pre-filming technique internally produces an oxide film layer in a non-radiological environment prior to reactor operation which prevents the future formation of oxide films once reactor operation commences. This prevents these activation products being deposited in the oxide film so contributes to reducing operator dose and ILW arisings. Alkali pre-filming has the added benefit of being applicable to all carbon steel pipework and valves which have the potential for high radiological doses.

Horizon are currently reviewing pre-treatment/commissioning techniques which may be employed at the Wylfa Newydd Power Station. Alkali pre-treatment is a proven technique which Horizon may employ if deemed appropriate. However, the exact pre-treatment/commissioning techniques to be employed, if any, are currently under review [FAbc-5].

Figure 3.1.5.1-1 provides an illustration of how alkali pre-filming prevents the deposition of cobalt-60.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.1.5.1-1 Illustration of the Alkali Pre-Filming Technique

The alkali pre-filming technique achieves a reduction in the deposition coefficient as shown in Figure 3.1.5.1-2, which in turn leads to a reduced dose rate, as shown in Figure 3.1.5.1-3 [Ref-41].

Figure 3.1.5.1-2 Cobalt-60 Deposition Coefficient on CUW Piping (Carbon Steel)

A.P. = Power Station with alkali pre-filming

Figure 3.1.5.1-3 Dose Rate Comparison of Japanese Power Station’s with and without Alkali Pre-Filming (Carbon Steel)

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.1.5.2 Evidence: FME Strategy

Horizon will develop and implement a FME Strategy. This will provide guidance on how to develop process and supporting management arrangements for construction, commissioning and maintenance of the Power Station. The strategy will strongly focus on the prevention of foreign materials within systems before first operation, or before the return to service of a system following maintenance. Foreign material is any material not part of the defined system or component. Examples of foreign materials comprise debris, dirt, tools, rags, machining tailings, wiring grinding particles, paint chips, Personal Protective Equipment and any other item that could adversely affect the performance or chemistry of the system.

Failures in plant and equipment can occur due to unrecoverable debris remaining in the system from construction or commissioning activities. The malfunction, breakdown or failure of components within gaseous or liquid fluid containing systems that are active or potentially active will result in the generation of secondary radioactive waste in response to repair work. The shortened operating life of components or parts within radioactive waste, coolant or main steam systems results in the installation of new items that become radioactive waste in the future. Over the lifetime of the Power Station this leads to a higher volume of radioactive waste generated than predicted based on the design operating life of components. Systems containing foreign material that are susceptible to being exposed to neutron fields during operation, results in the generation of activation products. This has the consequence of increasing the production of radioactive waste and potentially introducing an additional operator dose penalty when maintenance is performed.

Horizon will incorporate relevant good practice presented in the Institute of Nuclear Power Operations’ (INPO) Guidelines for Achieving Excellence in FME report [Ref-47]. INPO’s guidelines are based on OPEX from Power Station operators that implement high performing FME programmes. To minimise the potential of component failures from foreign material: inspections of components during refuelling outages, direct unfiltered flow paths to the reactor are subject to inspection, cleaning and flushing and additional inspection using a borescope can be carried out. Design principles for incorporation in the Power Station include:

• Selection of components that are not susceptible to degradation or breakdown that results in the introduction of debris that can becomes a foreign material.

• Low point drains that are sized appropriately to enable system blowdown and flushing.

• Inspection openings and ports are available for pre-operation inspection and cleaning.

• Minimise open grating or decking over FME areas.

The FME Strategy for Horizon will adopt INPO [Ref-47] guidelines based on the following steps:

1. Preparation of work task

• Training of FME monitors that the role of controlling ingress and egress of tools and materials.

• Identification of systems and components at risk, based on proximity to task and potential foreign materials associated with the work task.

• Assessment of whether ventilation for the area could introduce contamination.

• Introduction of FME logs.

• Good housekeeping practices.

• Definition of FME controls on open systems and components.

2. Implementation of controls in work task

• Rationalisation of tools and materials.

• Inspection and cleaning of areas.

• Covers on breaches and opening of systems at risk.

• Consider physical barriers e.g. netting, sheeting and temporary partitions.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3. Closeout of work task

• Reconcile any items not logged as removed in the inventory.

• Verification of cleanliness.

• Before first operation or return to service of a system, consider flushing or borescope inspections.

• Monitoring the performance of FME controls.

4. Recovery of foreign materials

• Defined process for addressing unrecovered foreign material.

• Unrecovered foreign material list.

3.1.5.3 Evidence: HOP Decontamination Process

HOP was developed by Hitachi-GE and Kurita Engineering in order to clean pipe work by removing deposited radioactivity from internal surfaces. The chemicals of the HOP process are injected during outage resulting in a decontamination process that also removes any crud that is contained in the pipe work. If left in situ this crud could be transferred to the core, where it would become activated once the plant is operating and subsequently contribute towards the generation of radioactive waste.

The HOP process is undertaken at a temperature of 85-95°C and achieves a DF of between 10 and 80 [Ref-42].

The HOP process minimises the base metal corrosion of the pipes it comes into contact with, whilst the HOP chemicals are fully decomposed after use [Ref-42] and [Ref-43].

The HOP decontamination process can be used during outages as a measure to reduce dose rate. Further development on when to use the process will be explored by Horizon during operational and decommissioning phases of the Wylfa Newydd Power Station.

3.1.5.4 Evidence: Water Conditioning

Water conditioning during start-up

Before start-up, the concentration of dissolved oxygen in the coolant is saturated due to exposure to air. High oxygen concentrations are undesirable as they contribute to SCC. To prevent this form of SCC, de-aeration is conducted and the dissolved oxygen concentration is controlled by maintaining oxygen concentration below 200 ppb before control rod withdrawal. De-aeration of coolant to acceptance criteria will be carried out in accordance with Water Chemistry Control procedures. The presence of high concentrations of oxygen also accelerates the corrosion of fuel cladding and as such, the de-aeration of the coolant contributes to maintaining the fuel integrity.

Any crud generated in the condensate system and feedwater system during an outage has the potential to be transported into the reactor at start-up, consequently resulting in an increase in radioactive CPs. To prevent activation of the crud, commissioning procedures will condition that the feedwater and condensate system water is purified by re-circulation through the Condensate Filter (CF) to remove any crud prior to start-up. Further details of this process are provided in Section 5.1.8. Removing crud prior to it becoming activated reduces the radioactivity of the waste treatment facilities (e.g. radioactivity of the demineraliser resins).

Water conditioning during shutdown

The radioactive crud concentration in the reactor coolant spikes during the decreasing pressure and temperature phases of shutdown. Crud may be deposited in areas of reduced flow, which results in the formation of radioactive ‘hot spots’. The ABWR design minimises the formation of radioactive hot spots during shutdown by eliminating areas of stagnant flow or areas that may hold up material. Minimising the formation of radioactive hot spots during shutdown is effective at reducing operator radiological dose and also contributes to reducing the radioactivity of maintenance and decommissioning wastes [Ref-44].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Additionally, as stated in section 3.1.6 (Argument 1f: Water Chemistry), a number of elements of the chemistry regime are optimised to prevent crud formation such that the UK ABWR is envisaged to be a low crud plant.

Water conditioning during outage

During outage, the storage of the feedwater and condensate system will be optimised to minimise corrosion and the generation of crud. Storage options include:

• Draining the system so that it is dry,

• Filling the system with deoxygenated water, or

• Filling the system withed with oxygenated water and recirculating it.

The storage method will be determined by Horizon nearer to commissioning and will depend on a number of factors, including the outage duration and the work taking place. Before start-up, the feedwater and condensate system is purified by re-circulation through the CF to remove any crud generated during the outage [Ref-44].

Management of Condensate Demineraliser (CD) resins

During an outage the CD is isolated and is stored in deaerated water to prevent degradation of the resin. By mitigating the deterioration of the resin the frequency at which the resin requires replacement is reduced and thus lower quantities of resin wastes are generated.

The Maintenance Schedule will specify the frequency of draining liquid from the CD resins and their storage in demineralised water during an outage, the operation will be carried out in accordance with an Operating Instruction, both of which will be finalised in the Operation and Maintenance strategy prior to commissioning.

3.1.5.5 Evidence: LT-SHC

Reducing the temperature at which the RHR is brought in to service for shutdown cooling has the potential to reduce radioactivity deposition within the RHR system. Enhancements have been made to shutdown techniques by the adoption of LT-SHC. This reduces the temperature at which the RHR is brought into service, reducing the amount of radioactivity deposited on the internal surfaces of pipes and vessels by approximately 80%. The main benefit of this technique is that doses to workers are reduced during outage activities. However, the radioactivity of pipes and vessels will also be reduced which will lead to a commensurate reduction in maintenance related waste and decontamination fluids [Ref-45].

The LT-SHC method lowered the RHR system operation start temperature from 150°C to 110°C. In order to reduce radionuclide deposition on the RHR piping, the LT-SHC method will be applied in each of the UK ABWRs despite the commercial advantages to a faster cool-down (shorter cool down periods lead to shorter outages) [Ref-45].

Figure 3.1.5.5-1 outlines the process of the Low Temperature RHR Shutdown Cooling Method and Figure 3.1.5.5- 2 shows the reduction in radioactivity deposition compared to the Soft Shutdown Method [Ref-46].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.1.5.5-1 Low Temperature RHR Shutdown Cooling Method

Figure 3.1.5.5-2 Comparison of Radioactivity Deposition between the Soft Shutdown Method (without LT- SHC) and Low Temperature RHR Shutdown Cooling Method

3.1.5.6 Evidence: Fuel Cycle Staggering

Horizon has considered staggering the fuel cycles of the two reactors to mitigate the higher aqueous and gaseous discharges associated with having two reactors on an outage at the same time. However, it is recognised that this cannot always be guaranteed as in the initial start-up period both reactors will be running on irregular and shorter fuel cycles and will be put on outage more frequently due to initial commissioning testing of the reactors. Additionally, throughout normal operation issues such as an equipment failure coinciding with a planned outage could also result in both reactors being on outage simultaneously. It is therefore acknowledged that it is preferable not to have both reactors on outage at the same time and, where possible, fuel cycles will be staggered to aid this but it is also acknowledged that this cannot always be safely achieved and as such, Horizon will be prepared for dual outages over the lifetime of the Power Station. Finally, where a dual outage were to occur, Horizon would aim to stagger the start up to prevent a dual outage on the next cycle.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.1.5.7 Evidence: Commissioning Programme

Commissioning the Power Station must be undertaken to ensure effective testing of SSC to demonstrate design intent and correct environmental performance before operation. International Atomic Energy Agency (IAEA) guidance on Commissioning for Nuclear Power Plants [Ref-48] sets out that commissioning activities should also have the following purposes:

• Collection of onsite test data that provides baseline for future reference.

• Validation of plant procedures, checking accuracy and identifying elements that require update.

• Familiarisation and training of staff in operating and maintaining plant.

Horizon’s approach to commissioning will be a logical and progressive sequence of tests so that SSC is exposed to less onerous test conditions, before being exposed to more onerous conditions, and testing for individual components and systems to overall testing for integrated systems [Ref-49].

The IAEA’s guidance [Ref-48] on commissioning programmes, specifies they should provide the framework for scheduling tests, assigning competent persons and the timely production of documentation and records. The Horizon commissioning programme will cover all commissioning activities for SSC (including chemical pre- conditioning) and will be structured to ensure the following:

• That SSC performance will be sufficient to meet reliability requirements postulated.

• That components of the liquid effluent management system are suitable in meeting water re-use criteria and Discharge Criteria.

• That gaseous abatement systems are suitable to meet Discharge Criteria.

• That sub-components of the solid waste treatment systems are suitable in meeting WAC for treatment and disposal at an offsite premises.

• That SSC function will fulfil requirements in environmental protection.

The scope and method of tests in terms of functions, parameters and requirements of testing will be determined based on the SSC’s functions and the demands that are placed on its performance (Safety Functional Claims and specific environmental protection requirements). Acceptance criteria will be specified in test procedures and will be linked to the Wylfa Newydd BAT Case where appropriate. Test results will be reviewed by competent individuals to determine acceptability for further testing (where appropriate) or completion of commissioning. If test results indicate that SSC have failed to meet acceptance criteria, corrective action shall be carried out followed by re-testing. A graded approach in investigation and corrective action shall be taken that is commensurate with the SSC being tested, the issue identified and the stage of commissioning that the issue is revealed in.

The Power Station will have three commissioning phases: construction testing, pre-operational testing and start- up testing [Ref-49] (see below). Commissioning strategies for each phase will be used as a basis for developing the commissioning programme.

Construction testing strategy [Ref-50]

Construction testing commences during the latter part of construction. The scope of testing is at a component level; therefore, temporary supplies to suitably carryout testing may be relied upon, as it cannot be assumed that supporting systems (e.g. instrument air or demineralised water) will be available. The end of construction testing is when all components within a stated boundary of a pre-operational test have been tested correctly and back energisation of the 6.9kV switchboards has been achieved. Construction testing will be the responsibility of the Engineering, Procurement and Construction (EPC) contractor with appropriate witnessing by Horizon.

Horizon will be responsible for the production (or setting requirements and expectations for the EPC contractor or other contractors) of the following documentation: safe systems of work, procedures for verification and witnessing tests, competency arrangements, calibration programme, specification for defining content in

Wylfa Newydd Project – Best Available Techniques (BAT) Case

construction test documentation. These management arrangement will be developed at the appropriate time within the project. The EPC contractor will be responsible for the following documentation: list of construction tests required, construction test specifications, constructions test instructions and construction test reports. These will be developed by the EPC contractor after finalisation of EPC Contract with Horizon. The following activities for Power Station systems are expected to be undertaken by the EPC contractor:

• Mechanical systems

• Pipework flushing (comprises part of the FME Strategy (3.1.5.2 Evidence: FME Strategy).

• Pressure tests for design and welding / jointing tightness (supports the demonstration of leak tightness for tanks, containment vessels and waste systems (Evidence within 3.1.10 Argument 1j: Leak Tightness of Liquid, Gas and Mixed Phase Systems)).

• Valve actuation e.g. air, hydraulic and manual.

• Check the correct installation of filters and strainers, non-return valves, pipe lagging and component labels.

• Electrical systems

• Manual open and close of de-energised breakers.

• Resistance and earth testing of cables and components.

• Control & Instrumentation

• Calibration of instrumentation.

• Correct wiring connections between Control & Instrumentation systems and components.

• Civil

• Load tests of lifting points.

• Correct preservation regime e.g. paint, lagging or lining.

• Structural Integrity Test.

Turnover of SSC will occur between the EPC contractor and Horizon upon acceptance of results from construction testing.

Pre-operation testing (non-active commissioning) [Ref-51]

This second stage of commissioning comprises a series of tests conducted on systems of the Power Station. After pre-operational testing of particular SSC, there is a period of maintenance and operation until all systems have completed pre-operational testing. The following activities for Power Station systems are expected to be undertaken by the EPC contractor:

• Mechanical systems

• Pump flow rates and head.

• Trips and interlocks.

• Performance of filters and demineralisers.

• Electrical systems

• Switchboard interlocks (current checks / protection).

• Remote open and close of breakers.

• Control & Instrumentation

• Correct processing, transmission and display of positions, values and indications of equipment.

• Remote operation of equipment e.g. pumps, valves, breakers etc.

• Civil

• Pressure tests of rooms and buildings e.g. Integrated Leak Rate Test, HVAC differential pressure tests etc.

Horizon will be responsible for the production (or setting requirements and expectations for the EPC contractor or other contractors) of the following documentation: plant operating procedures, specifications for defining content in pre-operation test documentation, maintenance instructions, pre-operation test instructions and pre-operation test reports. These will be developed by the EPC contractor after finalisation of EPC Contract with Horizon.

Start-up testing (active commissioning) [Ref-52]

The first part of start-up testing is the acceptance and handover of SSC from the EPC contractor to Horizon (upon acceptance by Horizon). The handover comprises three batches:

• 1st batch – SSC that demonstrates security and emergency preparedness arrangements prior to receipt of nuclear material (fuel and neutron source) onsite.

• 2nd batch – SSC required for the safe (including environmental protection) receipt of nuclear material.

• 3rd batch – Remaining SSC.

Horizon will then undertake commissioning of SSC, tests include: system tests, fuel receipt and handling, first criticality and reactor operations over a series of power plateaus which demonstrate plant performance under static and transient conditions. The end of start-up testing occurs after successful completion of the warranty test and the COD.

3.1.5.8 Evidence: Shutdowns and crud

The start-up phase is found to present a similar operating trend to that observed during the shutdown phase, with a spike in activity (approximately 100 times that observed during power operation for Co-60) occurring due to deposited radionuclides on the fuel surface becoming disturbed as a consequence of the changing flow rate, reactor pressure, reactor water temperature and the change from reducing to oxidising conditions. Due to this being a physical phenomenon, radionuclides are assumed to be released in the same manner (albeit with different ratios), i.e. the activity associated with radionuclides is assumed to increase during the shutdown phase.

3.1.6 Argument 1f: Water Chemistry The fundamental function of the water coolant in each of the UK ABWR steam circuits is to transfer heat from the fuel and to generate the steam necessary to drive the turbines. The chemistry of the coolant is carefully managed to deliver the following: • Reducing chemistry related failures of the fuel cladding (including corrosion and deposit related failures); • Minimising the generation of CPs that could become activated in the reactor core; • Minimising the replacement of reactor and process components associated with material corrosion; • Minimising liquid discharges to the environment and the levels of radioactive waste generated; and • Reducing occupational exposure of workers to ionising radiation.

Early generations of BWRs employed a Normal Water Chemistry (NWC) regime which did not use any chemical dosing agents. The primary objective of this regime was to control the conductivity and chloride levels to prevent the occurrence of Trans granular Stress Corrosion Cracking (TGSCC) in austenitic stainless steels. However, the importance of Intergranular Stress Corrosion Cracking (IGSCC) was not recognised at the time. Subsequently, IGSCC resulted in repairs to reactor components and associated plant being undertaken which lead to the generation of significant quantities of radioactive waste thus increasing occupational exposures to workers and reduced the availability of the plant (3.1.6.1 Evidence: Hydrogen Water Chemistry with Noble Metal Chemical Addition). The NWC approach was effective at mitigating TGSCC but resulted in the presence of free oxygen and hydrogen peroxide (H2O2), both of which are strong oxidising agents. This created an oxidising environment within the reactor coolant water which led to the occurrence of IGSCC, although this was not recognised until the late 1970s (3.1.6.1 Evidence: Hydrogen Water Chemistry with Noble Metal Chemical Addition). Hydrogen Water Chemistry (HWC) was therefore employed in the 1980s to mitigate IGSCC. This involved direct injection of hydrogen into the reactor coolant water which combined with the free oxygen and H2O2 to form water. This created an overall reducing environment within the reactor coolant water and was successful in mitigating IGSCC. However, a disadvantage of the approach was that it also changed the chemical form of nitrogen-16 from soluble nitrate compounds to non-soluble/volatile forms such as nitrogen oxides and ammonia. This caused the nitrogen-16 formed in the reactor coolant water to partition into the steam phase which significantly increased operator doses in the turbine building (3.1.6.1 Evidence: Hydrogen Water Chemistry with Noble Metal Chemical Addition). To mitigate the undesirable volatilisation of nitrogen-16, Noble Metal Chemical Addition (NMCA) was therefore employed alongside HWC. This involved the addition of noble metals such as platinum and rhodium which catalyse the combination of injected hydrogen with free oxygen and H2O2. This reduced the amount of hydrogen which needed to be injected to achieve the desired combination which in turn reduced the amount of Nitrogen-16 which was volatised and transferred over to the steam phase. NMCA itself has also been optimised and the Wylfa Newydd Power Station will employ specifically the On-Line Noble ChemTM (OLNC) method (3.1.6.1 Evidence: Hydrogen Water Chemistry with Noble Metal Chemical Addition). In addition to preventing corrosion, the water chemistry is also optimised to mitigate the adverse effects of activation products, most notably cobalt-60. Three main elements of chemical control combine to achieve this: • During normal operations, corrosion product films (oxides) develop on material surfaces. Under NWC conditions, these are predominantly of ferrite structure (Fe2O3). Under the reducing conditions of HWC, an inverse spinel structured oxide (Fe3O4) forms which incorporates a greater amount of Co-60 compared to ferrite structured oxides. This is undesirable as it leads to increased operator doses. To mitigate this zinc, in the form of Depleted Zinc Oxide (DZO), is added to the reactor coolant water. The resultant zinc ions compete with cobalt ions for deposition sites and are favoured for incorporation into the corrosion film which reduces the amount of cobalt-60 that is deposited, especially under the highly reducing conditions of HWC+OLNC (3.1.6.2 Evidence: Zinc Injection). • The concentration of iron present in the feedwater is reduced and controlled to limit the amount of corrosion product containing iron oxide film that can form on the fuel and to ensure that what is formed does not readily dissolve/spallate as this results in the release of corrosion products, most notably cobalt-60. The UK ABWR is expected to be a low iron plant, with the option to implement CF bypass,

should iron control be required to promote formation of insoluble fuel deposits in order to mitigate release of Co-60. The CF bypass method is one of the standard methods used to control iron concentration in feedwater (3.1.6.3 Evidence: Iron Concentration Control). Additionally, oxygen injection and the selection of corrosion resistant materials also acts to reduce the concentration of iron cruds in the feedwater entering the reactor (3.1.6.4 Evidence: Oxygen Injection).

• Progressive improvements have been made to the water coolant clean-up system including the provision of a condensate clean-up system and a CUW. The condensate clean-up system contains a condensate filter/condensate demineraliser (CF/CD) which prevents corrosion products generated in the steam and condensate system from entering the reactor. This reduces the generation of activation products iron-59 and manganese-54 which require subsequent treatment as radioactive waste. The CUW then removes impurities and activation products from the reactor water. The two systems work together to minimise detrimental effects on the fuel performance, to reduce the generation of activation products and associated radioactive waste and to reduce occupational exposure to workers (3.1.6.5 Evidence: Condensate Clean-up System and Reactor Clean- up System to Remove Corrosion Products). The management of water chemistry has been a fundamental element undertaken by Hitachi-GE to reduce the failure of fuel, to reduce operator dose and to improve plant availability. Data available for BWRs show that by the early 1990s the number of fuel failures had reduced to approximately 10% of the peak experienced in 1974 and that availability of the plant continued to show an upward trend during this period (3.1.6.6 Evidence: Fuel integrity). The chemistry of the reactor coolant has evolved to include techniques that reduce the likelihood of small leaks from the fuel; corrosion of reactor components; and the creation and accumulation of activation products within the reactor. All of these contribute to minimising the activity of radioactive waste that will be disposed of from Wylfa Newydd. The regime being adopted is the result of decades of OPEX, is proven and is well understood.

3.1.6.1 Evidence: Hydrogen Water Chemistry with Noble Metal Chemical Addition

Early generations of BWRs used Normal Water Chemistry (NWC) whereby treatment was provided to control impurities within the reactor water but did not involve the injection of chemicals. Under these conditions the rates of SCC were high. As a result, HWC was introduced by injecting hydrogen into the reactor water as a way of reducing SCC [Ref-55]. The injection of hydrogen into coolant would be controlled by a Water Chemistry Control procedure. With HWC, the reactor water becomes a reducing environment where oxidising species such as oxygen and hydrogen peroxide are consumed by the recombination reaction with hydrogen. The effect of injecting hydrogen into the feedwater is to reduce the Electrochemical Corrosion Potential (ECP) and hence decrease crack growth rates. Whilst residual stress remains in the material surface, reduction of the corrosion potential has the effect of greatly reducing crack growth. Laboratory tests found that the addition of hydrogen to the feedwater to reduce the ECP to less than -230mV (relative to that of the standard hydrogen electrode) was found to be optimum in suppressing SCC.

A disadvantage of adding hydrogen to the reactor water was an increase in gamma radiation from the main steam line during transfer to the turbine. The primary source of the radiation increase resulted from increased volatilisation of nitrogen-16 under HWC conditions produced from neutron activation of oxygen-16. Although the half-life of nitrogen-16 is very short and does not impact on discharges to the environment it does result in an increase in gamma shine within the turbine building (T/B) during normal operations and therefore an increased dose to workers and the public.

To mitigate the undesirable volatilisation of nitrogen-16, Noble Metal Chemical Addition (NMCA) was therefore employed alongside HWC. This involved the addition of the noble metals such as platinum and rhodium which catalyse the combination of injected hydrogen with free oxygen and H2O2. This reduced the amount of hydrogen which needed to be injected (from about 2ppm to 0.15-0.35ppm [Ref-55]) to achieve the desired combination of hydrogen with free oxygen and H2O2 which in turn reduced the amount of Nitrogen-16 which was volatised and transferred over to the steam phase.

Classic NMCA involved the addition of platinum in the form of Na2Pt(OH)6 or rhodium in the form of Na2Rh(NO2)6 with the application being performed during outages at a frequency of approximately once every 5 years. However, Na2Rh(NO2)6 was found to contain impurities, specifically chlorine, which becomes activated and leads to the undesirable generation of radioactive chlorine-36 [Ref-56].

As such, OLNC was later developed and represents an optimised approach for achieving the desired catalyzation. OLNC allows the application to be performed at power at a frequency of once every year. OLNC employs only platinum (in the form of Na2Pt(OH)6) so eliminates the chlorine-36 production associated with the use of Na2Rh(NO2)6 [Ref-56]. This is additionally beneficial from a corrosion perspective, as chloride is also stated as a control parameter in the Water Quality Specification [Ref-65] (and EPRI water chemistry guidelines) due to its aggressive nature and influence on SCC.

Addition of Na2Pt(OH)6 does introduce sodium which becomes activated to form radioactive sodium-24 which, due to it being non-reactive in this environment, builds up in the reactor coolant water and the spent demineraliser resins [Ref-56]. However, this is largely non-volatile so does not readily partition to the steam phase and has a half-life of only 15 hours so is not persistent in solid radioactive wastes. As such, the benefit of adding the platinum catalyst is considered to outweigh the detriment of the introduction of sodium.

The benefit of NMCA is illustrated in Figure 3.1.6.1-1 which shows the main steam line dose rates on an ABWR as a function of feedwater hydrogen concentration highlighting the typical hydrogen concentrations and associated dose rates for the three regimes (NWC, HWC and HWC + NMCA) [Ref-56].

Figure 3.1.6.1-1 Relation between feedwater hydrogen concentration and main steam dose rate at Kashiwazaki-Kariwa-7

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.1.6.2 Evidence: Zinc Injection

During normal operation an oxide film is produced on the inside of stainless steel pipes which naturally incorporates cobalt-60 ions. An important adverse effect of HWC is that the reducing environment produced increases the rate of cobalt-60 uptake on this corrosion film which is undesirable as it leads to increased operator doses.

Zinc injection is currently used in BWRs for the reduction of pipe work dose rates. Zinc is injected continuously from the feedwater system [Ref-44]. Injecting zinc into the reactor circuit suppresses the build-up of cobalt-60 inside the oxide films that form on the surface of pipes and vessels which would result in increased direct radiological doses to operators [Ref-44]. It is reported that the impact of zinc was first recognised by the observation that nuclear power stations with high zinc concentrations in reactor water had lower radiation dose rates from pipe work. Through experimental and field studies, it was found that elevated concentrations of soluble zinc in reactor water reduced cobalt-60 build-up in the corrosion films on pipe work and components by promoting the formation of a protective oxide film. Radiation dose rates are lowered since zinc is favoured for incorporation into the oxide film relative to cobalt-60. While Natural Zinc Oxide (NZO) is effective at reducing operator doses, activation of the zinc-64 (48% of NZO isotopic composition) to form zinc-65 contributes to an increase in operator dose and the generation of radioactive waste. Consequently, Depleted Zinc Oxide (DZO), that is, zinc depleted in zinc-64, is commonly used for zinc injection [Ref-56]. When DZO is coupled with both iron control (see 3.1.6.3 Evidence: Iron Concentration Control) and cobalt source control (see 3.1.7 Argument 1g: Specification of Materials), low shutdown operator dose rates are achieved.

The reduction in cobalt-60 uptake on pipe and reactor internals through zinc addition is illustrated in Figure 3.1.6.2-1 below [Ref-58].

Figure 3.1.6.2-1 Laboratory Results Showing the Uptake of Co-60 by Type 304 Stainless Steel Under NWC, HWC, NWC + Zn, HWC + Zn and NMCA Treated Type 304 Under HWC +Zn Conditions

There have also been benefits observed from zinc addition with regards to stabilising the cobalt-60 containing deposits on the outer surface of fuel elements. The majority of cobalt-60 is produced by dissolved inactive

Wylfa Newydd Project – Best Available Techniques (BAT) Case

cobalt-59 from cobalt containing structural material depositing within the iron oxide layer formed on the outer surface of fuel elements as shown below [Ref-57].

59 59 Fe2O3 + Co -> CoFe2O4

Once deposited, the cobalt-59 is subjected to an intense neutron flux and becomes activated to form cobalt-60. 60 60 The now CoFe2O4 then dissolves and spallates giving rise to dissolved cobalt-60 ions and CoFe2O4 particulates being released into the reactor coolant water [Ref-57].

• The dissolved cobalt-60 ions then deposit on the oxide films produced on the internals of pipes (as discussed above, this is reduced by the presence of zinc) with any remaining residual cobalt-60 being removed from the reactor water by the demineralisers in the CUW.

60 • Additionally, a proportion of the released CoFe2O4 particulates accumulate in crud with the remainder being removed from the reactor coolant water by filters in the CUW.

Zinc is also favoured for incorporation in the iron oxide formed on the fuel elements such that ZnFe2O4 is more favourably formed over CoFe2O4. This therefore reduces the amount of cobalt-60 formed in the reactor [Ref-56].

Additionally, ZnFe2O4 is more tenacious than CoFe2O4 so when zinc is added it actually stabilises the oxide layer on the fuel and prevents the release of cobalt-60 ions and particulates allowing it to be removed with the fuel elements during refuelling. This is particularly important as it reduces the amount of cobalt-60 present (both dissolved and as particulates) in the reactor coolant water (as demonstrated in Figure 3.1.6.2-2 [Ref-56]) which in turn [Ref-56]:

• Reduces the amount of cobalt-60 available for incorporation into oxide films;

• Reduces the amount of cobalt-60 present in crud; and

• Reduces the amount of cobalt-60 present in spent demineralisers.

Figure 3.1.6.2-2 Effect of Zinc Addition on Cobalt-60 Concentrations in Reactor Coolant Water at a selection of BWRs employing HWC

A qualification programme for the use of zinc addition to reduce shutdown dose levels also indicated a further positive effect on the susceptibility of components and pipe work to IGSCC. Research concluded that low levels

Wylfa Newydd Project – Best Available Techniques (BAT) Case

of zinc reduced crack propagation rates under reducing conditions, and that high concentrations of zinc reduced crack growth rates even under oxidising conditions. Beneficial effects of zinc on crack initiation and growth of Alloy 600 were also observed in PWR environments. Therefore, work was initiated to find a specification to optimise the zinc concentration and hydrogen injection rate that would minimise operator dose rates during normal operations whilst providing IGSCC mitigation for the bottom region of the reactor vessel [Ref-44].

Finally, no gross adverse effects of hydrogen, zinc and NMCA, when added within prescribed amounts, have been observed. No gross hydriding has been identified either, although detection methods are not capable of detecting minor hydriding [Ref-59].

3.1.6.3 Evidence: Iron Concentration Control

During normal operation, corrosion products (such as iron, nickel, chromium, zinc and cobalt) are released from structural materials and are present in the reactor coolant water as soluble ions with iron representing the major component of these corrosion products [Ref-56].

The dissolved iron then deposits on the surface of the fuel elements as an oxide layer. As this oxide layer forms it also incorporates other dissolved corrosion products (such as nickel, chromium, zinc and cobalt) and these become part of the matrix of the oxide layer. Since this oxide layer is subjected to the intense neutron flux in the core, the incorporated corrosion products become activated giving rise to radioactive activation products. The oxide layer containing the activation products can then undergo spallation and, albeit limited, dissolution which then releases the activation products into the reactor coolant water as both particulates and soluble ions [Ref-56].

Limiting the amount of iron present in the feedwater of the reactor limits the amount of iron available to form the oxide layer. This in turn reduces the amount of dissolved corrosion products that become activated and released into the reactor coolant water [Ref-56].

Additionally, another benefit of limiting the feedwater iron concentration is that it ensures that any oxide layer produced (oxide layer formation can be reduced but cannot be eliminated) is less susceptible to dissolution and spallation. Oxide layer fuel deposits actually consist of two layers; a tenacious inner layer enriched in transition metals and an outer non-adherent layer. Reducing the feedwater iron concentration reduces the size of the outer layer, whilst maintaining the inner layer, thus preventing the release of activity [Ref-56]. This is important as it ensures that the activation product containing oxide layer stays attached to the fuel (and is removed with the fuel elements during refuelling) and does not readily become released into the reactor coolant water.

Conversely, there is OPEX to show that in some plants, if the iron concentration is reduced too low soluble monoxides such as NiO, CoO, ZnO may form on the fuel cladding surface instead. However, these species readily dissolve back into the reactor water resulting in increased reactor coolant water activation product concentrations [Ref-56].

The approach for the UK ABWR is therefore to control the iron concentration in the feedwater to <1.0 ppb and let it trend towards its natural level. If an increase in cobalt-60 is observed Horizon have the option to then operate under optimum iron control (iron concentration of 0.1-1.0ppb) using the condensate filter bypass [Ref-56]. Chemistry Quality Control procedures for sample collection, testing and analytical reporting will specify the utilisation of the CF bypass line from grab sampling and analysis of iron. Action levels will be set on iron concentration to determine when to actuate the bypass valve for control over the volume of water bypassing the CF.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Item ① ② ③ Method CF Bypass LPHD mixed in CD Downstream Fe injection into Feedwater Feature -Easy to operate after stabilizing -Effective for increasing sma ll -Flexible, Ea sy of CD iron removal efficiency amount of Fe -Short term, Auxiliary -No additional injection system -Rigid -Additional injection system is required -Increase in back-wash -Sufficient amount of Fe can not -Necessity of operation and frequency of CD be injected maintenance of injection system -Other corrosion products such as Co and Ni are fed into the reactor Range of Fe conc. 0.3-1 ppb 0.4-0.5 ppb 0.1-1 ppb control

Figure 3.1.6.3-1 Iron Concentration Control Methods

3.1.6.4 Evidence: Oxygen Injection

Since general corrosion correlates with Flow Accelerated Corrosion (FAC), the prevention of both crud formation and FAC of carbon steel in the feedwater and condensate system depends on the same countermeasures. It is known that a concentration of dissolved oxygen of more than 15 ppb can minimise the occurrence of both types of corrosion [Ref-44]. The dissolved oxygen concentration in the feedwater and condensate system is therefore maintained at or above 15 ppb to prevent corrosion [Ref-54] by application of a Water Chemistry Control procedure that instructs on the monitoring of oxygen and sets an action level for oxygen injection.

Figure 3.1.6.4-1 shows conductivity, iron concentration and dissolved oxygen concentration in the feedwater line when oxygen injection is implemented [Ref-53]. The oxygen concentration is measured at the inlet of the CF, outlet of the CD and outlet of the feedwater heater. The oxygen injection point is set between the CF and the CD and the concentration of dissolved oxygen ranges from 20 to 30 ppb in feedwater with oxygen injection [Ref-54]. The concentration of iron crud and conductivity decreases with oxygen injection, which indicates that the corrosion of carbon steel is reduced.

Figure 3.1.6.4-1 Conductivity, Iron Concentration and Dissolved Oxygen Concentration of Feedwater Line with Oxygen Injection

3.1.6.5 Evidence: Condensate Clean-up System and Reactor Clean-up System to Remove Corrosion Products

The amount of CPs in the feedwater and condensate system is controlled to minimise detrimental effects on the fuel performance, operator dose and the generation of radioactive waste. Increased concentrations of CPs within the core region has the potential to effect material integrity and fuel structural material integrity by accumulating on the fuel cladding and thereby decreasing thermal conductivity. By reducing the amount of CPs there are less activation products which results in a reduction in direct doses to operators and less radioactive waste being generated.

To prevent the incorporation of impurities, CFs and CDs are installed on the Condensate Clean-up System and a Filter Demineraliser (FD) is installed on the CUW. A description of these systems has been provided in 3.1.8 Argument 1h: Recycling of Water to Prevent Discharges.

FDs also help to reduce the occurrence of SCC as they reduce the concentration of chloride ions and sulphate ions which increase the SCC susceptibility of materials.

3.1.6.6 Evidence: Fuel integrity

The number of fuel failure events has declined since the late 1970s as illustrated in Figure 3.1.6.6-1 [Ref-53]. This is a result of fuel cladding material improvements (3.1.1 Argument 1a: Design, Manufacture and Management of Fuel) and water chemistry controls. Water chemistry measures introduced to maintain the integrity of the fuel include the purification of coolant and the introduction of measures that reduce the generation of CPs. The deposition of harmful impurities such as chloride and CPs on fuel surface (fuel deposits) strongly affects the integrity of fuel claddings. Fuel deposits have been reduced through methods such as oxygen injection (which is described in more detail in section 3.1.6.4 (Evidence: Oxygen Injection).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.1.6.6-1 Reduction in Number of Fuel Leaks Follow Material and Water Chemistry Improvements

3.1.7 Argument 1g: Specification of Materials Materials in the reactor are exposed to neutrons generated by nuclear fission. In some instances the materials will become radioactive by a process known as 'activation'. These 'activation products' are a significant source of direct doses to workers and are a source of radioactive waste. There are two main sources of activation products: • Structural materials within the reactor that are activated due to their proximity to nuclear fuel and the associated neutron flux. These materials become radioactive waste during maintenance and decommissioning tasks. • CPs that are suspended in the reactor water deposit on the surface of the fuel cladding and become activated. The activated elements can then re-dissolved into the reactor water, deposit on other reactor and pipe internals, and subsequently have the potential to contribute to an increase in dose to workers. The design of the UK ABWRs takes account of decades of experience in the design, operation and decommissioning of LWRs and, where practicable, uses materials that are less susceptible to corrosion, deposition and activation. Efforts to use alternative materials have sought to balance the benefits provided by the characteristics of the materials with their safety and environmental implications. Cobalt is present as an impurity in stainless steel and nickel based alloys that are used within each of the UK ABWR reactors and parts of the steam circuit. Cobalt is also a significant component of cobalt based alloys (e.g. Stellite®) that have been historically used in reactor systems due to their hard-wearing properties. The naturally occurring isotope cobalt-59 becomes activated by neutrons to create the radioactive isotope cobalt-60. Progressive evolutions of the BWR design have sought to reduce the amount of cobalt present in materials of construction. The design of each of the UK ABWRs now includes the specification of Low Cobalt Material (LCM) for a range of components that are particularly susceptible to direct activation or from which corrosion products may become activated (3.1.7.1 Evidence: Specification of Low Cobalt Materials (Reduction of Cobalt Based Alloys and Specification of Low Cobalt Materials)). The design also limits the use of Stellite® to those components where it is essential for material performance (3.1.7.2 Evidence: Substitution of Stellites®). CPs generated in the condensate system have been reduced by the progressive introduction of corrosion resistant steels over several evolutions of the ABWR design. The CPs contain

Wylfa Newydd Project – Best Available Techniques (BAT) Case the naturally occurring isotopes iron-56 and iron-58 which, when activated, become iron-59 or manganese-54 (3.1.7.3 Evidence: Introduction of Low Corrosion Materials). The CPs generated also contain nickel and chromium which, when activated, become cobalt-58 and chromium-51. In combination, these improvements have resulted in an expected approximate 50% reduction in the amount of cobalt-60 generated during the operation of the UK ABWR compared to if these improvements were not applied (3.1.7.4. Evidence: Specification of Materials – OPEX and Feedback). To ensure these benefits are realised, in the material selection process for SSC, material will be assessed according to factors that affect their environmental performance and availability (3.1.7.5 Evidence: Material Selection Factors). Materials of construction have been carefully selected to minimise the potential for activation from the nuclear reaction. Where practicable, materials that are particularly susceptible to activation have been replaced by other materials. This reduces the activity of solid radioactive waste that will be disposed of from Wylfa Newydd during operational and decommissioning phases. However, it has not been possible to replace all of these materials because their properties contribute to other important factors such as nuclear safety and operational efficiency.

3.1.7.1 Evidence: Specification of Low Cobalt Materials (Reduction of Cobalt Based Alloys and Specification of Low Cobalt Materials)

Improvements in material specifications have aimed to eliminate the use of cobalt based alloys wherever possible and reduce cobalt in stainless steels and Ni-based alloy that are used for large surface area or high flow rate components. LCM is used to reduce activated radioactivity and cobalt-60 concentrations in the reactor water. Reducing the activated radioactivity of components results in a reduction in activity of those components when they are disposed of as waste during maintenance and decommissioning [Ref-25] and [Ref-26]. Preventing the generation and release of cobalt-60 into the reactor water minimises the generation of radioactive waste when the reactor water is treated by the CUW.

Cobalt based alloys are almost completely eliminated inside the RPV of the ABWR. This has been achieved by:

• Eliminating jet pumps found in conventional BWRs, which had Stellite® present in the slip joints.

• The use of iron and nickel based alloys for control rod rollers and pins rather than cobalt based alloys.

There is also the potential to replace valve seats manufactured from cobalt based alloys that are located external to the reactor vessel with alternative materials.

A materials specification will be developed to limit the cobalt content of stainless steel, Ni-based alloys and welding materials. This specification will apply to those materials which are particularly susceptible to activation or dissolution with subsequent activation. A design review and subsequent ALARP assessment was performed to assess the use of LCM [Ref-60]. The design review explored opportunities to increase the use of LCM in the design of the UK ABWR. The review evaluated the cost of applying LCM, the potential reduction in direct dose to workers and the generation of radioactive waste for a range of reactor components. It was concluded that it was both ALARP and BAT to manufacture the components in the reactor from LCM ( 0.05% cobalt) with some exceptions which only make a small contribution to activation and cobalt release. ≦ The materials used in the reactor and steam system consist mainly of austenitic stainless steel, nickel based alloys, carbon steel and low alloy steel components. The use of cobalt based alloys and the cobalt content of the alloys used in the core region (control rods, fuel bundle hardware, steam separator and dryer and some internal reactor components), as well as final feedwater heater tubes, is minimised to reduce the creation of activation products and associated potential for gamma radiation. To ensure that the concentration of cobalt is

Wylfa Newydd Project – Best Available Techniques (BAT) Case

minimised a material specification has been developed for the components of these systems that require the cobalt content to be less than 0.05% [Ref-62].

It is recognised that the use of LCM in specific components reduces both direct doses to operators and the generation of radioactive waste. Hitachi-GE therefore undertook an assessment of the availability of material with a cobalt content that is less than the current limit of 0.05% proposed to be used in the UK ABWR during GDA. The assessment, based on Japanese suppliers concluded that the limited availability and additional cost of reducing the cobalt content limit to lower than 0.05% was grossly disproportionate compared to the benefits [Ref-60]. A preliminary review of European suppliers confirmed the findings of the assessment.

LCM will therefore be specified where appropriate in each of the UK ABWRs [Ref-62]. Procurement of SSC will be subject to review by environmentally competent persons and material specifications will be defined by the codes and standards assigned to that SSC.

3.1.7.2 Evidence: Substitution of Stellites®

Stellites® are cobalt based alloys with hard-facing characteristics. Stellite® surfaces exposed to reactor coolant will contribute to cobalt in feedwater which will become activated in the reactor core producing cobalt-60. In BWR designs, Co-based alloys such as Stellites® are used in taps, valves and some parts of the internal core support structure and primary pumps [Ref-62].

In the UK ABWR design, where suitable justification can be made, cobalt based alloys will be replaced by cobalt-free materials with confirmed sufficient material characteristics, or, in some cases, by an adequate nickel and iron base [Ref-62]. Examples of material substitution are:

• The use of Wear Proof Material (WPM), a nickel based alloy, for rollers of the control rods; and

• Nitronic®-60 (an iron based alloy) for control rod pins.

The cobalt content of both alternate materials is 0.25% or less.

This will contribute to reducing the activity of radioactive decommissioning waste at the end of the plant lifetime [Ref-25] and [Ref-26].

3.1.7.3 Evidence: Introduction of Low Corrosion Materials

To reduce iron input into the reactor water, extensive use of low alloy steel and stainless steel in pipe work systems and other components will be implemented. The effect of adopting corrosion resistant steel, along with the introduction of a dual condensate polishing system and oxygen injection, has reduced the iron feedwater concentration from between 5 to 10 ppb to between 0.1 to 1 ppb. This has reduced the generation of iron-59 and manganese-54 crud on the fuel cladding and thus reduced the generation of radioactive waste and dose rates.

Other significant corrosion products are nickel and chromium which become activated to form cobalt-58 and chromium-51. Cobalt-58 radiologically exhibits almost the same behaviour as cobalt-60. Chromium-51 has a short half-life and low gamma energy which means it is less significant in terms of radioactive waste generation and radiation exposure.

This will contribute to reducing the activity of radioactive decommissioning waste at the end of the plant lifetime [Ref-25] and [Ref-26].

3.1.7.4 Evidence: Specification of Materials – OPEX and Feedback

The improvements of using stainless steels and nickel based alloys (Inconel®) with reduced concentrations of cobalt have led to a reduction in the amount of cobalt-60 activated in the core. This is illustrated in Figure 3.1.7.4-1 [Ref-62].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.1.7.4-1 Reduction of the Generation of Co-60 as a Result of Using LCM and Reducing Stellite®

3.1.7.5 Evidence: Material Selection Factors

Hitachi-GE have applied factors that affect the integrity of SSC and dose to inform the selection of materials [Ref- 63]. Horizon will also apply these factors in the assessment of materials for purchasing LLI and in the supply of SSC that can affect environmental radiation protection. The factors support fulfilment of four overarching goals (listed below) that ensure the selection of an optimised material.

• Material supports SSC in the delivery of environmental protection requirements;

• Material supports SSC in meeting its design life and minimises the generation of radioactive waste;

• Material supports SSC in optimising the radiological impact to workers and members of the public; and

• Materials are available for the manufacturing of SSC.

Material Specifications, Equipment Design Reports and other applicable design information for SSC will be assessed by Horizon SMEs to determine whether the factors have been adequately met and demonstrate BAT, consideration will be given to UK RGP and ALARP principles to ensure radiological impacts an SSC’s material through the lifecycle of the WN Power Station is ALARA.

3.1.8 Argument 1h: Recycling of Water to Prevent Discharges Water is used as the coolant within the UK ABWR and to raise the steam that is used in the turbines to generate electricity. The coolant becomes contaminated with radioactive material as it passes through the reactor and around the steam circuit. A high concentration of radioactivity in the steam circuit is undesirable as it can result in increased exposure to workers during operational and maintenance activities. The disposal of the contaminated water is also undesirable as it could degrade the environment and expose members of the public to potentially harmful ionising radiation.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

The design of the UK ABWRs at the Wylfa Newydd Power Station includes the following systems for recycling water that is used during operations and maintenance: • The Condensate Water Clean-up System treats water that has passed through the turbines and has been condensed back to water. Filters remove any solid matter that could either damage the fuel or become activated in the reactor core. Demineralisers are used to remove ions that have become activated or have the potential to become activated in the reactor core or deposit on internal surfaces of pipes and vessels (3.1.8.1. Evidence: Condensate Water Clean-up System). • The CUW continuously draws water from the reactor and passes it through a FD system to control conductivity and remove impurities. This has the combined effect of reducing the potential for corrosion of fuel cladding, minimising the generation of CPs on internal surfaces and reducing the potential for CPs to become activated in the reactor core. Once treated, the water is returned to the reactor in the main feed line. The quality of the water passing through the system is continuously monitored to ensure that the characteristics are within defined parameters and that the system is performing as expected. In the unlikely event that the characteristics of the liquid fall outside of the defined parameters, the liquid is passed through the FD again so that the characteristics of the liquid meet the defined parameters. During start-up and shutdown operations, any excess reactor water can be transferred to the Low Chemical Impurities Waste (LCW) to enable an operator to manage the reactor water level of the system. Following treatment, this water is then returned to the Condensate Storage Tank (CST) and is then available to be recycled back into the reactor water circuit (3.1.8.2.Evidence: CUW). • The Fuel Pool Cooling and Clean-up System (FPC) and the Suppression Pool Clean-up System (SPCU) use shared demineralisers to maintain the water quality. The characteristics of the water in these two areas is broadly similar as water from the Suppression Pool (S/P) is pumped to the equipment lay-down pool during refuelling operations. At the end of refuelling, water from the reactor well is returned to the S/P where its quality continues to be managed by the SPCU. This represents an improvement on previous generations of the BWR which used water from the CST during refuelling operations which was subsequently transferred to the liquid effluent management system and then back to the CST. The quantity of water previously extracted from the CST per refuelling operation was approximately 2500 m3 in the BWR-5. Eliminating this process has resulted in reducing the size of the CST and the inventory of water in the reactor circuit (3.1.8.3Evidence: FPC and SPCU). • The LCW treats liquid effluent from the SFP, FPC and CUW, along with effluent waste from equipment drains, etc. The LCW consists of a filter and a demineraliser which ensures that the water quality meets the criteria for the CST and subsequent reuse in the plant (3.1.8.4. Evidence: LCW Treatment System). The Turbine Gland Steam System (TGS) uses water extracted from the CST to produce steam that is used in the turbine gland seal. Following use in the turbine gland, 98% of the steam is condensed and is subsequently returned to the main condenser and is made available for reuse (3.1.8.7 Evidence: Recycling of Water within Steam Circuit - OPEX and Feedback). The design of the water treatment systems include demineralisers to remove soluble material including radionuclides. Demineraliser (also referred to as ion-exchange) systems are

Wylfa Newydd Project – Best Available Techniques (BAT) Case utilised throughout the nuclear industry to remove soluble material including radionuclides from liquid processes to maintain the water quality within target values. They are recognised by the EA to represent BAT for this purpose and are widely considered RGP (3.1.8.5 Evidence: Nuclear Industry Application – Demineralisers). As such, their use in the UK ABWR was considered to represent BAT. At GDA the resin selection was determined for the UK ABWR and Horizon have adopted the same resin selection for the Wylfa Newydd Power Station. There is flexibility provided to change the resin selection following commencement of operations if deemed appropriate. This flexibility was considered to represent BAT at GDA and this position has not changed for the Wylfa Newydd Power Station as it prevents the foreclosure of options for Horizon to change the resin selection should superior resins be developed as technology evolves (3.1.8.6 Evidence: Demineraliser Media). Recycling water in the UK ABWR mitigates the requirement to make routine liquid discharges from the steam circuit during the operational life of the facility (3.1.8.7 Evidence: Recycling of Water within Steam Circuit - OPEX and Feedback). Water is also reused during decommissioning activities to prevent the generation of new aqueous waste (3.1.8.8 Evidence: Reuse of Water during Decommissioning Activities). There are two possible treatment routes for final discharge of aqueous waste (3.1.8.9 Evidence: Treatment routes for final steam circuit liquid disposal): • Treatment through the existing liquid effluent management system following retrofitting of upgrades; or • Treatment through a bespoke liquid treatment system brought in specifically for decommissioning. The option employed will not be decided until closer to site decommissioning, as technologies and BAT itself will evolve between now and at the end of commercial operation of the Power Station. The decision will take account of the assessed characteristics of the aqueous waste and the Decommissioning Strategy that will be cognisant of technological advances during the lifetime of the Power Station. Design elements are employed on the UK ABWR to prevent contamination of reactor coolant but it will still inevitably become contaminated with impurities, some of which will be radioactive, during routine operation of the Power Station. Processes will be employed to remove these impurities and allow this water to be continuously recycled within the plant. This in turn will reduce the amount of radioactive aqueous waste discharged to the environment and thus contribute to the environmental performance of the Power Station.

3.1.8.1 Evidence: Condensate Water Clean-up System

The configuration of the Condensate Water Clean-up System is illustrated in Figure 3.1.8.1-1. CFs remove any crud in the system which could either damage the fuel or become activated in the reactor core. Demineralisers are used to remove ions that have become activated or have the potential to become activated in the reactor core or become deposited on internal surfaces of pipes and vessels.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.1.8.1-1: Condensate Water Clean-up System

Design improvements led to the CD being used in combination with a pre-filter known as the CF [Ref-62]. Prior to the introduction of the CF any crud in the water remained in the system and contributed to increased operator dose and an increase in the radioactivity of maintenance wastes. The pre-filter removes insoluble ions and the crud content resulting in the following improvements:

• Operator dose reduction;

• Reduction in radioactivity of maintenance wastes; and

• Prevention of the demineraliser blocking as a result of solids build up.

The performance of the CF is summarised in Table 3.1.8.1-1.

Table 3.1.8.1-1: Impact of Introducing Demineraliser Pre-filters

Before (system before After (Improved & standardised improvement) system)

Ion removal & crud removal for Requirement for Condensate Ion removal reduced radiation exposure around Polishing System reactor

Outlet metallic impurities Approx. 10 ppb

Occupational Exposure Approx. 10person·Sv

The CD is an external non-regenerative type (demineralising by mixed bed ion exchange resins) and processes approximately 900 m³/h of water from the condenser [Ref-61]. The demineraliser removes soluble salts (substances present in ionic form) from the condensate allowing the radioactivity to be disposed of as solid waste once the resins are spent. The CDs will be non-regenerative, but will require frequent “washing”, to remove deposited Crud.

3 Condensate Pre-Filter is one of the improvements of plant dose rate reduction.

3.1.8.2 Evidence: CUW

To maintain the parameters of the reactor water and to minimise the transfer of impurities into the steam circuit a CUW is provided. The CUW system conditions the coolant using filter demineralisers to remove impurities in order to reduce the potential for corrosion of fuel cladding, minimise the generation of CPs on internal surfaces and reduce the potential for CPs to become activated in the reactor core.

In earlier BWR systems the typical flowrate through the CUW was in the region of 1%. However, in the Wylfa Newydd Power Station this has been increased to 2% [Ref-64] which results in a superior clean-up capability.

The configuration of the CUW is illustrated in Figure 3.1.8.2-1.

Feedwater Line

From RHR

Figure 3.1.8.2-1 CUW

The CUW removes impurities contained in the reactor water and maintains the quality of the reactor water within the determined range in order to prevent:

• Corrosion of the equipment and pipe work of the reactor cooling system;

• Decreasing the heat transfer efficiency by avoiding the adhesion of impurities on the fuel surface; and

• Radioactive contamination of the reactor cooling system and the related equipment.

The CUW system, through the use of filter demineralisers, seeks to maintain the coolant within the design specifications presented in Table 5.1.8.2-1 [Ref-65].

Table 3.1.8.2-1: Design Specification Coolant Conditions Following Treatment in the CUW

Conductivity (25°C) 100 μS/m or less

Clˉ 100 ppb or less

pH (25°C) 5.6 to 8.6

In practice the system performs much better than the conditions in Table 3.1.8.2-1 and can achieve 10 μS/m at 25°C. In the unlikely event that the parameters of the coolant fall outside of the design specification prescribed in Table 3.1.8.2-1, the operator can review the efficacy of the treatment system. If required the filter demineralisers can be backwashed and the coolant can be recirculated through the filters and demineralisers so that the characteristics of the liquid meet the defined parameters. During start-up and shutdown operations, any excess reactor water can be transferred to the LCW to enable an operator to manage the reactor water level of the system. Details of the treatment in the LCW are described in Section 3.2.6 (Argument 2f: Configuration of Liquid Management Systems). Following treatment in the LCW the water is returned to the CST Tank where it is reused.

3.1.8.3 Evidence: FPC and SPCU

The FPC provides two functions:

• Removal of decay heat from the SF Storage Facility in the SFP; and

• Removal of impurities from the water of the fuel pool.

Figure 3.1.8.3-1 shows the configuration of the FPC which includes heat exchangers for the removal of decay heat and filter demineralisers for the removal of impurities. The filter demineralisers are designed to process approximately 250m³/h of water [Ref-61].

The purpose of water chemistry management of the SFP is primarily to maintain the fuel integrity in the storage pool, including the integrity of the storage rack and pool itself and to minimise the radioactivity level increase in the pool. For this purpose, coolant is cleaned up and its temperature is controlled during normal operations.

Coolant purification has two purposes:

• To protect the fuel, fuel rack and pool structure from corrosion; and

• To maintain the clarity of coolant to allow visual inspection of the fuel assembly located under water.

In the initial stages of BWR development, a large amount of activated crud was dispersed inside the pool during refuelling resulting in higher operator doses. However, in recent years, because of the reduction of CPs from the feedwater line, the problem of elevated radiation levels during refuelling has been dramatically improved. The water chemistry of the SFP in operation is sufficiently controlled such that no failure caused by water chemistry related damage has been experienced in ABWR designs.

Further detail on the FPC and the role that it plays within the whole SFP is provided in the Pre-Construction Safety Report (PCSR) Chapter 19: Fuel Storage and Handling [Ref-66].

Skimmer Surge Tanks

D/S Pit Reactor well SFP

FPC Hx

FPC-P

FPC F/D

Figure 3.1.8.3-1 FPC

Figure 3.1.8.3-2 shows the SPCU system configuration. This system shares the same filter demineraliser as the FPC which removes impurities from the pool water in the Suppression Chamber (S/C).

As well as maintaining the water quality of the S/P, the SPCU also has the function of transferring water to the upper pools (Equipment Lay Down Pool and Reactor Well) prior to refuelling. See Figure 3.1.8.3-3. This represents an improvement to the previous BWR-5 design which requires the upper pools to be filled with water from the CST. By adopting the SPCU and utilising the SP water, a decrease in demand on the radioactive waste treatment facility and the CST has been achieved, allowing for its capacity to be reduced by 40% [Ref-61]. Figure 3.1.8.3-4 summarises this design improvement.

DS Pit Reactor well SFP

FPC F/D S/P

SPCU-P

Figure 3.1.8.3-2 SPCU (S/P Water Clean-up Mode)

DS Pit Reactor well SFP

FPC F/D S/P

SPCU-P

Figure 3.1.8.3-3 SPCU (Water Filling Operating Mode)

Figure 3.1.8.3-4 Design Improvement of Adopting the SPCU

3.1.8.4 Evidence: LCW Treatment System

LCW is liquid effluent from the SFP, FPC and CUW, along with liquid effluent from equipment drains, etc. LCW from the SFP, FPC and CUW is excess liquid that is removed from these systems during reactor start-up and shutdown operations to help maintain the water balance of the plant. The LCW treats the liquids which are then returned to the CST for reuse.

The LCW consists of filters, for the removal of insolubles, demineralisers, for the removal of solubles, and sampling pools, as shown by Figure 3.1.8.4-1. The LCW therefore allows LCW to be treated and reused rather than being discharged during normal operation. Further detail on the LCW is provided in PCSR Chapter 18.2: Radioactive Waste Management [Ref-67].

Figure 3.1.8.4-1: Schematic of the Configuration of the LCW

At GDA Hitachi-GE undertook an assessment [Ref-68] to compare the different treatment technologies available for the LCW. The assessment compared demineralisers (ion exchange), reverse osmosis membrane and cross- flow filtration against a range of criteria such as OPEX, reliability, maintainability, solid waste generation, DF and cost. Demineraliser technology scored significantly higher than the other two techniques, outperforming them in the following areas: • Lowest impact on the generation of solid waste; • Lowest capital cost; • Highest reliability safety; • Highest maintainability; • Highest resistance to radiation; and • Low/no requirement for chemicals.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

The assessment therefore identified the use of demineralisers, as well as a filter, as the preferred option for treatment of LCW. The use of demineralisers is identified as relevant good practice (RGP) as discussed in section 3.1.8.5.

The performance of the LCW that is achieved in standard ABWRs has been demonstrated through OPEX and feedback. Analysis of samples taken from the inlet and outlet of the LCW demonstrates a DF of approximately 1,000. This high level of performance is required to ensure that liquids are suitable to be reused within the Power Station.

3.1.8.5 Evidence: Nuclear Industry Application – Demineralisers

Demineralisation (also known as ion exchange) removes soluble materials from aqueous streams. The process of demineralisation is the removal of soluble salts (substances present in ionic form) from aqueous waste using ion exchange resins which retain certain substances which are then converted into solid waste when the demineralisation medium is spent. Ion exchange resins are recognised as RGP in the nuclear industry for the removal of soluble radionuclides and are used extensively on nuclear power stations in the UK and internationally. In its GDA Public Consultation Document on the UK EPR [Ref-69] the Environment Agency stated that “at this time, filtration by cartridge filter, ion exchange and, for aqueous waste incompatible with ion exchange, evaporation are BAT for use in the UK EPR”.

3.1.8.6 Evidence: Demineraliser Media

Ion exchange media selection depends on the properties of the target ion, the presence of other competing ions in the feed stream, availability and cost. The capacity of the media relates to how much target ion a particular type of media can hold.

Ion exchange media typically used in demineralisers in UK nuclear installations are either made of:

• Organic resins, which can carry various functional groups that provide a cation or anion exchange effect; or

• Inorganic ion exchangers, some of which act as adsorbers rather than ion exchangers and, to make them more efficient, are fabricated into beads or microporous gels with a high surface area.

The media to be initially used in the Wylfa Newydd Power Station UK ABWRs will be the same as was specified at GDA. This is the same as is used in in the Japanese ABWR and is consistent with the types listed above. It was decided BAT to initially employ these media due to them being proven and backed with considerable OPEX.

The choice of demineralisation media represents a balance between three factors:

• The overall type of wastes to be treated. Some waste streams have a specific and consistent composition which allows a particular demineralisation media to be used at maximum efficiency. However, in other cases the composition of the waste stream and its variability require a more general-purpose demineralisation media to be used.

• The nature of the nuclide to be removed. If a specific nuclide is to be removed from the waste stream, a specialised demineralisation media can be chosen. However, if a more general clean-up is required a media that offers an overall good performance needs to be used (e.g. incorporating anion and cation beds).

• The quantities of secondary waste arisings. A demineralisation media might offer good selectivity for a particular radionuclide or waste stream but may have a short lifetime, so resulting in higher secondary solid waste arisings.

The CUW, the FPC, SPCU, CF/CD and LCW all use a combination of filters and demineralisers to treat the water to allow reuse.

However, the ion exchange media is a consumable and the design of the demineralisers provides flexibility for Horizon to change the type of media used in the demineralisers throughout the operational life of the plant. This is

Wylfa Newydd Project – Best Available Techniques (BAT) Case

important as it does not foreclose options to employ different media in the future which will facilitate BAT being continually employed throughout the lifetime of the Power Station. This flexibility is deemed to also represent BAT.

The flexibility provided by the demineralisers is limited to the modification of the ion-exchange resin used based on:

• Cation and anion ratio changes which is linked to the water quality to be treated; and

• The selection of low- Total Organic Carbon (TOC)-elution ion-exchange resin, high-spec-ion- exchange resin, etc.

The design of the demineralisers on these systems is sufficiently flexible to allow Horizon to select the most appropriate ion exchange media.

3.1.8.7 Evidence: Recycling of Water within Steam Circuit - OPEX and Feedback

The treatment systems described in the evidence above allow for the water of the steam circuit, the SFP and the S/C to be recycled and thus alleviates the need to make liquid discharges from these systems under normal operating conditions.

The TGS design uses water extracted from the CST to produce steam that is subsequently used in the turbine gland seal. The use of CST water as the supply for the gland steam evaporator rather than purified water allows Horizon to manage the water balance of the plant without having to make additional discharges of aqueous radioactive waste.

Figure 3.1.8.7-1 shows the water balance of the plant. The majority of the water is recycled within each of the treatment systems described above however, in some cases, water may be sent to the LCW for further treatment. This is then recycled back to the CST for reuse. There are therefore no routine discharges from these systems. The only liquid discharges from each of the UK ABWRs are from the Controlled Area Drain System (CAD) and occasionally the High Chemical Impurity Waste (HCW) System, as described in 3.2.6 Argument 2f: Configuration of Liquid Management Systems.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.1.8.7-1 Water Balance of the Plant Showing the Recycling of Water

3.1.8.8 Evidence: Reuse of Water during Decommissioning Activities

Water is required during decommissioning of the Wylfa Newydd Power Station for processes such as chemical decontamination and underwater cutting of reactor internals. Rather than introducing clean water from outside of the system to carry out these processes, which would then become contaminated and lead to increased volumes of radioactive discharges, a proportion of the plant water that has been retained within the Power Station during the 60 years of operation could be used. This reused plant water has relatively low levels of radioactivity due to decay over the plant lifetime whilst its retention and reuse during decommissioning activities reduces the activity further. Reusing the plant water saves the generation of 5260m3 of aqueous waste that would have been required for decommissioning activities. Of the 11710m3 of water retained after shut down of the plant, around 45% could be reused as shown in Table 3.1.8.8-1. Ultimately Horizon will determine the extent that plant water will be reused during decommissioning activities [Ref-27].

Table 3.1.8.8-1 Reuse of plant water for decommissioning activities

Required water volume during Retained Reused No. Decommissioning activity decommissioning (m3) water (m3) rate (%)

System chemical 1 decontamination before 600 dismantling

Dismantling of reactor - - 2 3310 internals and RPV

Dismantling of reactor 3 1350 shielding wall

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Required water volume during Retained Reused No. Decommissioning activity decommissioning (m3) water (m3) rate (%)

Total 5260 11710 45

Prior to eventual discharge to the environment the aqueous waste is treated by the existing Radioactive Waste Building (Rw/B) equipment or another temporary liquid waste processing facility. The discharges will not exceed the operational discharge limits proposed by Horizon [Ref-37].

3.1.8.9 Evidence: Treatment routes for final steam circuit liquid disposal

Decommissioning operations will require continued operation of the HCW and CAD systems. The systems associated with the reactor, SFP and associated systems (LCW, FPC, CUW, and SPCU) will be operated as appropriate to the requirements for processing and discharge of the water contained within them. These will be gradually phased out and replaced with bespoke equipment relevant to the reducing volume and levels of contamination of the wastes being generated.

Aqueous waste arising from the primary circuit decontamination process will either be treated by modifying the installed liquid effluent management system or installing a bespoke system if this proves unfeasible [Ref-70].

3.1.9 Argument 1i: Secondary Neutron Sources Secondary neutron sources provide additional neutrons, at a controlled rate, to assist with reactor start-up. PWRs typically use antimony-beryllium neutron sources which generate tritium. The cladding for antimony-beryllium neutron sources is typically manufactured from stainless steel which is porous to tritium. Any tritium that is generated can therefore diffuse through the cladding and into the reactor coolant. Both UK ABWRs will use californium-252 as the start-up neutron source, instead of antimony-beryllium. An advantage of using californium-252 over antimony-beryllium neutron sources is that they do not generate tritium as a by-product (3.1.9.1 Evidence: Selection of Neutron Source Materials). The UK ABWR design also means that secondary neutron sources are only required during the start-up phase of the first fuel cycle. Due to the flexibility provided by the reactor design (as described in 3.1.3 Argument 1c: Efficiency of Fuel Use), whereby fuel with a range of burn- ups can be used, secondary neutron sources are unlikely to be required after the first cycle and can therefore be removed. The selection of materials for secondary neutron sources is limited. There are very few materials that have the combined features of neutron generation and a sufficiently long half-life to make them suitable for operations in a nuclear reactor. Neutron sources typically fall into one of three types which are spontaneous fission sources, alpha-neutron sources and photoneutron sources. The selection of the source material will have the largest impact on the amount of tritium that enters the coolant. The reference case for the UK ABWR is that californium-252 sources in stainless-steel cladding (3.1.9.2 Evidence: Selection of Neutron Source Cladding) will continue to be used as there is considerable OPEX available (3.1.9.3 Evidence: Secondary Neutron Sources - OPEX and Feedback) and that the quantity of tritium generated is relatively low. Secondary neutron sources are essential for reactor start up. Some secondary neutron sources produce tritium as a by-product. Californium-252 secondary neutron sources are used in the UK ABWR because they produce very little tritium compared to the most common alternatives. This will therefore reduce the amount of tritium discharged to the environment from Wylfa Newydd.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.1.9.1 Evidence: Selection of Neutron Source Materials

The neutron source assemblies consist of two main components – a long cylindrical tube assembly, referred to as the source holder, which is supported by the core plate and the top guide. The assembly is installed and removed in the same manner as a local power range monitor (LPRM) instrument tube, with a spring-loaded upper tip that fits under the top guide. Each source holder houses a neutron source pin assembly containing the californium-252 neutron source material.

96.9% of californium-252 (2.6 years half-life) undergoes alpha decay to generate curium-248 (3.4x105 years half- life) while the remaining 3.1% of decays are spontaneous fission. Use of californium-252 eliminates the production of tritium. Tritium is generated from antimony beryllium as a result of the following reaction chain:

9 1 4 6 4 Be+0n→2 α +2He

6 t1/ 2 =807ms 0 − 6 2 He→−1 β +3Li 6 1 4 3 3 Li+0n→2 α +1H

3.1.9.2 Evidence: Selection of Neutron Source Cladding

The californium-252 is in the form of palladium-californium composite wire. The palladium-californium wires are contained in welded stainless steel capsules (inner capsule) of Type 316L stainless steel. The inner capsule in turn is inside a 316L outer capsule, which provides a second encapsulation and forms the neutron source pin assembly, exposed to the reactor coolant. The wire consists of a uniform distribution of californium in a palladium alloy matrix. The palladium matrix itself constitutes a form of encapsulation.

The compatibility of the palladium-californium wire with 316L stainless steel capsule is established. Palladium has a relatively high melting point (1552°C) and is resistant to degradation by oxidation and corrosion. Regarding chemical compatibility, no reactions have been found between the palladium-californium wire and stainless steel up to 1000°C.

3.1.9.3 Evidence: Secondary Neutron Sources - OPEX and Feedback

The californium-252 neutron source was commercially available from the 1970s and it was first used for the start- up of a commercial reactor in 1973. Today, the use of californium-252 neutron sources for start-up is applied for all new reactors in Japan and the U.S.

From the 1980s Hitachi-GE has applied the californium-252 neutron sources to the start-up of eight reactors. Californium-252 neutron sources are unloaded 1-2 cycles after start-up.

These sources have also been used for reactors that have experienced an extended shutdown period of 7 to 10 years. In this case, the irradiated fuel alone may not have provided an adequate source of neutrons.

3.1.10 Argument 1j: Leak Tightness of Liquid, Gas and Mixed Phase Systems The design of each UK ABWR includes the provision of containment and ventilation systems that are intended to ensure that radioactive substances are retained within designated facilities during normal and fault conditions, and that they only enter the environment via appropriately permitted routes. Containment systems are also provided to confine radioactive substances. This is to prevent the unnecessary spread of radioactive contamination within facilities that would contaminate plant and equipment and result in the generation of avoidable radioactive waste. Containment systems also isolate workers from potential exposure to radioactive substances. Collectively, these systems contribute to minimising the amount of radioactive waste produced. Containment systems have common objectives related to worker safety and environmental protection which are delivered by effective design, manufacture, installation and operation. Hitachi-GE and Horizon have developed a series of design policies and

Wylfa Newydd Project – Best Available Techniques (BAT) Case principles that aim to ensure that both safety and environmental protection are an inherent part of the design (3.1.10.1. Evidence: Application of Design Features for Leak Tightness). The design of the UK ABWR has evolved to enhance the leak tightness of the reactor coolant circuit by rationalising the amount of pipework associated with plant operations and by improving the performance of welds, seals and connections (3.1.10.1. Evidence: Application of Design Features for Leak Tightness). Where contained materials have the potential to be released, systems will be provided for detection and containment in segregated ventilation systems and drains that are specific to the characteristics of the waste. Design policies have been established for the following waste facilities and systems: • Liquid Effluent System(s) (3.1.10.2 Evidence: Design Policies and Principles for Leak Tightness in Liquid Effluent System(s); • OG System(s) (3.1.10.3 Evidence: Design Policies and Principles for Leak Tightness in OG System(s); • Containment Vessel (3.1.10.4 Evidence: Design Policies and Principles for Leak Tightness in the Containment Vessel); • HVAC (3.1.10.5 Evidence: Design Policies and Principles for Leak Tightness in HVAC System); and • SFPs (3.1.10.6 Evidence: Design Policies and Principles for Leak Tightness in the SFPs). Additionally, design elements have been incorporated to prevent atmospheric argon leaking into the system which would otherwise have the potential to become activated (3.1.10.7 Evidence: Design Policies to Prevent Atmospheric Argon Leaking into the Coolant Systems and 3.1.10.11) (Evidence: Improvements in Turbine gland seal design). The design also includes a system for leak detection and isolation (3.1.10.9 Evidence: Leak Detection and Isolation System). Design improvements have led to improvements in leak tightness (3.1.10.10 Evidence: Improvements to Leak Tightness). Systems are used in the UK ABWR to ensure that all process fluids and wastes are contained within vessels, pipes, tanks, ducts etc. These systems primarily limit the spread of radioactivity and the unnecessary creation of additional radioactive waste which, in turn, reduces the activity of radioactive waste that will be disposed of from Wylfa Newydd.

3.1.10.1 Evidence: Application of Design Features for Leak Tightness

The following design features included in the UK ABWR design have been developed by Hitachi-GE and Horizon to reduce leakage and releases of radioactive material [Ref-71]:

• Extremely stringent leak rate requirements specified for all equipment, piping and instruments and will be confirmed using as-installed helium leak tests of the entire process system in accordance with Testing and Commissioning procedures.

• Use of welded joints wherever practicable.

• Specification of valve types with extremely low leak rate characteristics (i.e., bellow seal, double stem seal, or equivalent).

• Routing of drains through steam traps to the main condenser.

• Specification of stringent seat-leak characteristics for valves and lines discharging to the environment via other systems.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Confirmation that these design features can be found in the penetration, piping and joint design specifications is provided in [Ref-72], [Ref-73] and [Ref-74].

3.1.10.2 Evidence: Design Policies and Principles for Leak Tightness in Liquid Effluent System(s)

The items summarised below have been taken into consideration when designing the system for treating liquid effluents and the facilities related to them in order to prevent leakage of liquid radioactive substances from these facilities and to prevent their uncontrolled discharge to the environment [Ref-71]:

• To prevent the occurrence of leaks, the liquid effluent system shall be made of suitable materials and shall be provided with tank water level detectors;

• As a general rule, the drain pipes and vent pipes which empty outside of the system shall be provided with lockable valves or the equivalent for closing. Work control for isolation and de- isolation of valves would be in accordance with Operating Instructions. However, those which are used with a high frequency shall be provided with a drain leading to tanks, sump pits or the equivalent;

• Should radioactive liquids leak out, there shall be provisions making it possible to detect the leaks promptly and to remove and decontaminate the leaked liquids easily;

• The components of the liquid effluent management system shall be provided with secondary containment, such as bunds to prevent spreading of any leaks inside the Power Station. Bunds shall be provided on inlets and outlets connected to points outside the Power Station to prevent leakage outside of the Power Station. Outdoor devices and outdoor pipes shall be designed so that any leaked liquids will be collected inside facilities such as shielding walls or pipe ducts;

• Alarms provided for the tank water level or for leak detection shall be designed to alert either the MCR or the radioactive waste system control room, so that it will be possible to inform the operators reliably of abnormalities;

• The facilities shall be designed to ensure that facility drainage systems are segregated and do not connect with drainage channels discharging liquid wastes outside of the site in an uncontrolled manner;

• Welded joints are adopted for connections between equipment and pipe work for radioactive waste unless disconnections are required for maintenance; and

• Waterproof/impermeable coating is applied to the floor and wall where radioactive waste may leak.

The application of some of these design principles is described in Table 3.1.10.2-1. Further details can be found in PCSR Chapter 18.2 [Ref-67] and in Chapter 31 [Ref-25] where they contribute to reducing radioactive decommissioning waste arisings.

Table 3.1.10.2-1: Materials of the Radioactive Liquid Effluent Disposal and Treatment Facility

Property of retained water Materials Typical example

High corrosive potential due to the Selection of stainless steel or Concentrated waste storage ion concentration of chloride, carbon steel & lining. tank and pipe work. sodium, and sulphate being high.

Corrosive environment as a result Selection of stainless steel. HCW collection tank and pipe of acid or alkali. work.

Low corrosive potential combined Selection of stainless steel. LCW collection tank, sludge with a high crud density. storage tank and pipe work.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Property of retained water Materials Typical example

Low corrosive potential combined Selection of stainless steel. Sample tank and pipe work. with requirement to maintain water quality.

Others. Carbon steel. Other pipe work.

3.1.10.3 Evidence: Design Policies and Principles for Leak Tightness in OG System(s)

Leakage of off-gas into other parts of the Power Station is prevented by operating the off-gas charcoal adsorber at negative pressure. As a general rule, the valves in contact with radioactive gases are bellow seal types [Ref- 75]. Further details can be found in PCSR Chapter 18.3 [Ref-76].

3.1.10.4 Evidence: Design Policies and Principles for Leak Tightness in the Containment Vessel

The inner surface of the UK ABWR containment design is lined with a steel plate which acts as a leak tight membrane [[Ref-77], Page 18]. All normally wetted surfaces of the liner in the S/P are made of stainless steel. Penetrations through the liner for the drywell head, equipment hatches, personnel locks, piping and electrical and instrumentation lines are provided with seals and leak tight connections.

3.1.10.5 Evidence: Design Policies and Principles for Leak Tightness in HVAC System

The air flow inside buildings will be such that the air flows from areas of low radioactive contamination to areas of higher radioactive contamination, and the reverse flow will not be allowed. Contaminated air generated locally will be exhausted through a hood where practicable. To prevent the spread of contaminated air inside a room, the contaminated area will be at a local negative pressure.

For penetrations of piping or trays between rooms where the radioactive contamination classification differs, an air-tight sealing will be provided to reduce air leakage from the room with higher radioactive contamination to the room with lower radioactive contamination. No openings except hatches or doors will be provided between the rooms and an air-tight seal will be provided for the opening provided between the rooms [Ref-78].

3.1.10.6 Evidence: Design Policies and Principles for Leak Tightness in the SFPs

In order to prevent leakage of water from the fuel pool, the fuel pool has been designed without any exhaust ports, and an emergency make-up water system will be provided. Water leakage detectors and water-level alarm devices will be provided in order to monitor any possible leakage of the fuel pool water.

The fuel pool and the cask pit do not have drain outlets to prevent the leakage of fuel pool water. The FPC is designed to recirculate water that is released into the skimmer surge tank beyond the skimmer weir. In addition, check valves are placed on the pipes that lead to the fuel pool to avoid the release of fuel pool water by a siphon action [Ref-61].

3.1.10.7 Evidence: Design Policies to Prevent Atmospheric Argon Leaking into the Coolant System

Argon-40, a constituent of air, can leak into the main condenser where it is entrained within the reactor coolant and passed through the reactor core where it becomes activated forming argon-41. This is then transferred to the steam phase and, being non-condensable, is ultimately transferred to the OG via the SJAE where it is then discharged to the environment contributing to radioactive discharges. Measures are taken during the design and manufacture of the main condenser in order to minimise the amount of leakage that occurs, therefore minimising the amount of argon-41 that can be produced.

3.1.10.8 Evidence: Design Policies on Isolation

Horizon must define schemes, operating regimes and maintenance instructions to ensure adequate isolation is achieved [FAbc-2]. These should include:

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• Mechanical systems containing radiological hazards must be isolated, vented, drained and purged as required by operating instructions and maintenance procedures;

• The highest standard of isolation that is reasonably practicable must be used i.e. method of isolation should be proportionate to the radiological hazards involved and the risk of failure of the isolating equipment;

• The consequences of a release in exposing workers, members of the public and the generation of avoidable radioactive waste must be assessed; and

• Use of standard mechanical isolation schemes must be used that provide guidance and rank the effectiveness of isolation methods e.g. single valve, double valve, double valve and bleed etc.

3.1.10.9 Evidence: Leak Detection and Isolation System

The leak detection and isolation system consists of temperature, pressure and/or flow sensors with associated instrumentation, alarm, and/or isolation functions [Ref-79]. This system detects and indicates and/or alarms following leakage and provides signals to close containment isolation valves, as required, in the following systems: • Main steam lines;

• CUW;

• RHR System;

• Reactor Core Isolation Cooling System;

• Feedwater System;

• Emergency Core Cooling Systems; and

• Other miscellaneous systems.

An additional radiation sensor with an alarm function is provided for detecting a main steam or pre-filtered coolant leakage. Small leaks are generally detected by monitoring the air cooler condensate flow, radiation levels, equipment space temperature, and drain sump fill-up and pump-out rates. Large leaks are also detected by changes in reactor water level, drywell pressure, and changes in flow rates in process lines.

Manual isolation control switches are provided to permit the operator to manually initiate isolation of the main steam line from the control room. In addition, each Main Steam Isolation Valve (MSIV) is provided with a separate manual control switch in the control room which is independent of the automatic and manual leak detection isolation logic.

The Leak Detection System (LDS) provided to detect leakage of coolant from the reactor coolant pressure boundary during normal operation ensures that a leakage of about 3.8l/min [Ref-80] can be detected within one hour. This is achieved by monitoring the volume of condensed water in the gas coolers located in the containment vessel and the water level in the sumps also located in the containment vessel.

3.1.10.10 Evidence: Improvements to Leak Tightness

A number of design improvements have been incorporated into the design of the UK ABWR to prevent leaks which contribute to the generation of radioactive waste [Ref-81].

The MSIVs are designed to isolate the steam line primary in containment preventing the loss of coolant and any release of radioactive materials. Previous BWR designs had three MSIVs on the main steam line. The third MSIV controlled and prevented leakage from the first two MSIVs in the case of an accident. However improvements to the valves of the first two MSIVs resulted in a low leak rate. The third MSIV was therefore no longer required along with all the leakage control system associated with it. Figure 3.1.10.10-1 highlights the third MSIV and associated leakage control system, as in previous BWR designs, which was removed from the ABWR design as a result of the valve improvements.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Division A (Same as Div.B,C,D)

AO AO MO Inside MSIV-LCS Main Steam Lines To Turbine R P MSIVs MSIVs Third MSIVs V Outside MSIV-LCS

MO MO

B C D B C D Drywell

AO MO Suppression Pool

AO MO

Figure 3.1.10.10-1: Eliminated Equipment from the MSIV Leakage Control System Following Valve Improvements

The improvements to the design and orientation of the first two, MSIVs is detailed in Table 3.1.10.10-1 which led to improved performance of the MSIV negating the requirement for the third MSIV.

Table 3.1.10.10-1: Improvements to MSIVs

Phase Items Description

Improvement of disc form Tighter contact between the two sides of the valves discs.

Design 2 MSIVs were initially installed inclining at an angle of 15 Improvement of instalment degrees due to space limitations, but installing upright improved angle seating on the valve.

The disc was able to contact firmly to the seat by slitting along Flexible structure of disc the outer periphery disc.

Improvement of disc lapping An automatic disc lapping process rather than a manual process. process

Stellite weld overlay at valve The seal properties were improved by Stellite weld overlay at Maintenance seat seat side. Whilst this introduces small amounts dissolved Stellite in the system which can lead to increased cobalt-60 levels in the

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Phase Items Description

reactor coolant, the benefits to the seal properties were deemed to outweigh this small detriment.

Elimination of difference in The seal properties were improved by eliminating the difference level on the seat side generated by weld overlay at seat side.

OPEX and feedback from the existing fleet of BWRs in Japan has demonstrated that leakage rates have reduced to below 5%/day. The elimination of the third MSIV and the MSIV leakage control system reduced the worker dose, the costs for maintenance and the radioactive waste during the decommissioning phase.

3.1.10.11 Evidence: Improvements in Turbine gland seal design

A design change was introduced in 1976 to improve the safety and environmental performance of the turbine gland seal. Figure 3.1.10.11-1 shows the comparison of the existing TGS design with the original BWR design. In the original BWR design, the steam seal regulator (SSR) used reactor steam as the gland sealing steam. The reactor sealing steam is collected to the gland steam condenser (GSC), and any steam that is not condensed is then released to the environment via the main stack. This resulted in reactor steam containing elevated levels of radioactive contamination, including noble gases, being released directly to air. The TGS design was replaced by the Separate Steam Seal System (SSSS) and multi-phase gland seal as shown in the same figure which resulted in the following modifications:

• Introduction of a multistage turbine gland (TG) to prevent the loss of containment of reactor steam into the turbine building even under TGS steam failure, and

• Replacement of the reactor water/steam with CST water as the source for the TG seal steam to reduce the source term in the seal steam, including the elimination of noble gases, in order to reduce the radioactivity of gaseous radioactive waste discharges.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.1.10.11-1: Evolution of TGS design

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.2 Claim 2 - Minimise the Radioactivity in Radioactive Waste Disposed to the Environment The Wylfa Newydd Power Station employs a range of features to reduce the discharge or disposal of radioactivity from those radioactive wastes that are unavoidably created during operations. The Arguments presented in support of this Claim are considered to demonstrate compliance with the standard BAT conditions [Ref-5]: • Condition 2.3.2(a) ‘The operator shall use the best available techniques in respect of the disposal of radioactive waste pursuant to the permit to minimise the activity of gaseous and aqueous radioactive waste disposed of by discharge to the environment.’ • Condition 2.3.3(a) ‘The operator shall use the best available techniques to exclude all entrained solids, gases and non-aqueous liquids from radioactive aqueous waste prior to discharge to the environment.’ The Wylfa Newydd Power Station design contains a range of features that contribute to the substantiation of this Claim including: • Provision of an OG which includes processes to reduce radioactivity in the gaseous phase prior to discharge to the environment. • Provision of off-gas charcoal adsorber within the OG to abate short lived noble gasses. • A Heating Ventilating and Air Conditioning (HVAC) system that prevents the uncontrolled discharge of radioactive substances. • Treatment techniques for aqueous wastes that minimise the discharge of radioactivity to the environment. • Charcoal adsorbers to minimise the radioactivity associated with wastes that require disposal. In developing the Arguments presented to demonstrate the validity of Claim 2, the following REPs [Ref-7] are considered to be relevant and have been taken into account: • Principle ENDP14 ‘Best available techniques should be used for the control and measurement of plant parameters and releases to the environment, and for assessing the effects of such releases in the environment.’ • Principle ENDP15 ‘Best available techniques should be used to prevent and/or minimise releases of radioactive substances to the environment, either under routine or accident conditions.’ • Principle ENDP16 ‘Best available techniques should be used in the design of ventilation systems.’

3.2.1 Argument 2a: Off-Gas Waste Treatment Facility Gaseous radioactive wastes will be generated during the operation of the reactor. Significant efforts are expended to eliminate as much of the radioactivity in wastes as is practicably possible, but the disposal of gaseous waste to the environment is required to ensure the safe and efficient operation of the Power Station. Some of the radionuclides that are carried in the steam do not condense in the condenser. These radionuclides are carried by the Steam Jet Air Ejector (SJAE), which is used to maintain

Wylfa Newydd Project – Best Available Techniques (BAT) Case the vacuum in the condenser, and require treatment and disposal as gaseous radioactive waste. The design of the Wylfa Newydd Power Station includes two dedicated OG, one servicing each reactor, which collect, convey, treat and discharge gaseous radioactive waste from the condenser (3.2.1.1 Evidence: Configuration of the Off-Gas Waste Treatment Facility). The OG operates continuously during operation (3.2.1.2 Evidence: Off-Gas Waste Treatment Facility - Operational Philosophy) and includes processes to reduce radioactivity in the gaseous phase prior to discharge to the environment. A separate Argument is presented for the most significant treatment processes; the decay of noble gases and iodine through the use of charcoal adsorbers (3.2.2 Argument 2b: Charcoal Adsorbers for Noble Gases and Iodine) and filtration of airborne particulate matter (3.2.4 Argument 2d: Filtration of Airborne Particulate Matter). Some of the radionuclides in the off-gas such as tritium and carbon-14 do not undergo treatment in the OG and are discharged directly to the environment via the main stacks. This is because the assessment of treatment techniques for these radionuclides (3.2.1.3: Evidence: Assessment of Gaseous Treatment Techniques for Tritium and Carbon-14) has shown that the costs of installation and operation are very high and the reduction in impacts on members of the public and the environment is low. Additionally, the majority of the tritium in the gaseous phase will be removed from the off-gas by the OG recombiner and OG condenser. In-process monitoring is carried out in order to ensure that the OG is performing as expected (3.2.1.4 Evidence: In-process Monitoring to Support Demonstrating the Application of BAT). The exact monitoring equipment to be employed has not yet been decided, and will not be until nearer operation so the Wylfa Newydd Power Station design provides sufficient space (3.6 Claim 6 – Horizon shall Apply BAT When Characterising and Quantifying Gaseous and Aqueous Radioactive Waste Discharges) to allow Horizon to carry out in-process monitoring to confirm that what has been identified as BAT is performing as expected. A system is provided to treat radioactive gases that are present in the steam from the reactor. This system is effective at treating most of the radioactive gases and it will significantly reduce the amount of radioactivity that will be discharged to the environment from Wylfa Newydd. Some radioactive gases are not treated because the cost of treatment is considered grossly disproportionate to the benefits that will be received by members of the public.

3.2.1.1 Evidence: Configuration of the Off-Gas Waste Treatment Facility

The Wylfa Newydd Power Station will have two OG, one servicing each reactor, and these OG reflect RGP from across the nuclear industry and the experience gained from operating previous generations of the BWR and Japanese ABWR. The configuration of the OG reflects this experience ensuring that gaseous radioactive waste discharges into the environment are minimised.

The design of the OG has been developed to address two primary functions:

• The safe recombination of flammable gases (hydrogen and oxygen), which are generated by radiolytic decomposition of reactor water, to reduce the possibility of a hydrogen explosion. Whilst this was not primarily designed to contribute to the environmental performance of the Power Station, it does serve as an effective method for removing tritium from the gaseous effluent. This is because tritium is a hydrogen isotope so the performance of the hydrogen recombiner also removes tritium from the off-gas. This is achieved as tritiated hydrogen gas is converted to tritiated water which is then condensed by the OG condenser and returned to the Condensate Storage tank where it is

Wylfa Newydd Project – Best Available Techniques (BAT) Case

reused within the plant [Ref-82]. The specification and performance of this system is addressed in detail within the PCSR Chapter 18.3 [Ref-76].

• To minimise and control the release of radioactive gases and particulates into the atmosphere by delaying and filtering the off-gas waste process stream to adequately decay short lived radioactive isotopes and filter out particulate matter.

A number of different techniques are used for the treatment of off-gas from nuclear installations in order to reduce the radioactivity of discharges. The techniques are primarily focussed on holding up short lived radionuclides to allow them to decay and thereby reduce the overall radioactivity discharged to the environment. A summary of the different gaseous short lived radionuclide treatment techniques is provided in Table 3.2.1.1-1.

Table 3.2.1.1-1: Treatment Techniques for Short Lived Radionuclides in the OG

Treatment technique Description

Compressed gas Following passage through a recombiner the gas is compressed and directed to one storage of several storage tanks. Compressed gas storage systems achieve a reduction in activity release by storing the FP gases under pressure in large tanks to allow them time to decay. Although systems of this type have been used in previous generations of BWR, compressed gas systems are no longer being proposed for new nuclear power stations because of the large storage volumes required and the operating complexities introduced by the use of a compressor downstream of the SJAE.

Cryogenic distillation The krypton and xenon are condensed out of the gas stream as it passes through a distillation column operated at very low temperatures. As the process temperature is

lowered from ambient conditions, the gases begin to liquefy, with the amount of liquefaction of each gas depending upon its boiling point and vapour pressure. The condensed krypton and xenon are collected in a sump located in the distillation column. When the sump becomes full, the liquefied krypton and xenon are transferred to a gas cylinder for storage and decay prior to eventual release. Distillation systems are capable of very low release rates, but when compared to a refrigerated charcoal system, the gained incremental reduction in release is generally an insignificant part of the total Power Station release. The use of cryogenic distillation also adds considerably to construction and operating costs.

Charcoal Adsorbers The off-gas is passed through a charcoal adsorber which will hold the FP gases long enough to allow those having short half-lives to decay. To increase the adsorption efficiency of the charcoal, any water vapour remaining with the gas is extracted by a moisture removal subsystem. The charcoal is contained in several tanks operated in series downstream of the moisture removal equipment.

Cryogenic charcoal Following the recombiner the gas is cooled and passed through a moisture separator and desiccant dryer which serves to keep ice crystals from plugging the cryogenic

charcoal adsorber. The charcoal adsorber is maintained at a temperature of approximately -170oC by using the nitrogen gas that boils off a liquid nitrogen bath to remove the heat of adsorption. The -185oC liquid nitrogen bath is used to cool the decontaminated effluent gas from the cryogenic charcoal adsorber prior to passage through the regenerative heat exchanger which returns it to ambient temperature. The saturated bed is regenerated by passing heated nitrogen gas through the bed and causing krypton and xenon to desorb. The gases are then compressed and stored in gas cylinders prior to release. This technique has similar limitations to cryogenic distillation.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

The Wylfa Newydd Power Station uses activated charcoal adsorbers which are widely used in the nuclear industry and considered BAT by the Organisation for Economic Co-Operation and Development (OECD) report [Ref-9]. ABWRs also have years of operating experience of using charcoal adsorbers. The review of other techniques in Table 3.2.1.1-1 identified significant challenges associated with the other techniques.

Figure 3.2.1.1-1 provides an illustration of the OG for the Wylfa Newydd Power Station [Ref-75].

Figure 3.2.1.1-1 Schematic Drawing of the Off-Gas Waste Treatment Facility [Ref-75]

The OG comprises the following unit operations [Ref-75]:

• OG preheater: Prevents formation of water in the recombiner;

• OG Recombiner: Combines hydrogen and oxygen to remove the potential to generate an explosive atmosphere;

• OG condenser: Condenses steam to reduce its volume and cool the off-gas;

• OG cooler condenser: Cools the off-gas to reduce the moisture content prior to treatment within the charcoal adsorbers;

• OG refrigeration facility: Supplies a cooling source for the OG cooler condenser;

• OG charcoal adsorbers: Designed to hold-up short lived radionuclides (3.2.2 Argument 2b: Charcoal Adsorbers for Noble Gases and Iodine);

• OG filter: Removes particulate matter from the off-gas including any particulate from the charcoal adsorber (3.2.4 Argument 2d: Filtration of Airborne Particulate Matter);

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• OG ejector: Maintains negative pressure within the charcoal adsorbers;

• OG blower: Maintains negative pressure within the charcoal adsorbers during reactor start-up;

• OG blower after cooler: Cools off-gas that has been heated by the OG blower during plant start- up; and

• OG charcoal adsorber heating ventilation handling unit: Controls the temperature of the charcoal adsorbers.

• TGS filter: Removes particulate matter from the TGS and MVP discharge.

A more detailed description of the OG is provided in PCSR Chapter 18.3 [Ref-76].

3.2.1.2 Evidence: Off-Gas Waste Treatment Facility - Operational Philosophy

The OG is designed to operate continuously in a steady state whilst the Wylfa Newydd Power Station is operating. In the event that operational parameters do not remain within Discharge Criteria and specified limits it will be practice to shut-down the reactor until the performance of the OG can be demonstrated to have returned to normal. The operational philosophy for the OG is described in detailed in the OG operational regime document [Ref-83].

The exact operational parameters which, if breached will lead to a reactor shut down, are still being determined but it is envisaged they will be determined by operations and reactor chemistry SMEs based on OPEX and EPRI data.

3.2.1.3 Evidence: Assessment of Gaseous Treatment Techniques for Tritium and Carbon-14

Tritium and carbon-14 will be present within the gaseous waste that passes through the OG. Charcoal adsorbers such as those provided in the Wylfa Newydd Power Station design are not an appropriate abatement technique for these longer lived radionuclides. An assessment has therefore been carried out by Hitachi-GE to support the application of BAT for the abatement of tritium and carbon-14 [Ref-82].

The discharge rates of tritium and carbon-14 in the gaseous discharge streams during normal operation have been calculated based on the source term (See Section 5 of the EP-RSR Application), a summary is presented in Table 3.2.1.3-1.

Table 3.2.1.3-1: Summary Table of Total Gaseous Discharges

Annual Discharge form the Radionuclide power station (2 units)(Bq/y)

H-3 5.5E+12

C-14 1.8E+12

The OECD issued a report in 2003 (Effluent release options from nuclear installations technical background and regulatory aspects [Ref-10]) that stated that “in the particular cases of tritium and carbon-14, there are no abatement techniques in place to reduce discharges”. The report identified that carbon-14 and tritium from nuclear power stations is discharged without treatment in liquid or gaseous form to the environment and that common practice in many instances is to allow dilution to take place within the plant processes, followed by further dilution and subsequent dispersion upon release to the environment. The report went on to say that tritium removal from gas streams and liquid streams can be uneconomic, especially when concentrations are low.

At GDA Hitachi-GE undertook an assessment of the treatment options for carbon-14 and tritium [Ref-82]. The techniques assessed are detailed in Table 3.2.1.3-2.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Table 3.2.1.3-2: Options Assessed for the Treatment of Carbon-14 and Tritium

Carbon-14 Treatment Techniques Tritium Treatment Techniques

Alkaline scrubbing Molecular sieving

Dehumidification using packed bed or plate Gas adsorbtion by wet scrubbing columns

Fluorocarbon or ethanolamine absorption

Molecular sieve adsorption

Cryogenic distillation

The assessment of techniques for the removal of carbon-14 and tritium from gaseous wastes indicate that a number exist, although none are used on operational reactors and are thus not considered to be RGP. In addition to this, IAEA Technical report No.421 [Ref-84] concluded that methods for the separation of carbon-14 and tritium from gaseous wastes are costly and require high energy consumption and that application of these separation technologies may therefore be limited by their high cost.

The Hitachi-GE study reached the following conclusions:

Tritium

The assessment of tritium abatement techniques identified that none of the treatment processes are currently used on operational LWRs and the techniques would require significant development work to enable the techniques to be suitable for use on an operational reactor. Further to this, very little tritium is actually discharged in gaseous form from the OG as the OG recombiner, which recombines hydrogen and oxygen and the OG condenser, which cools and condenses the hydrogen depleted off-gas to separate any moisture and return it to the main condenser, transfers the majority of the tritium into the liquid phase. As tritium is a hydrogen compound; the performance of the recombiner and OG condenser therefore also removes the tritium from the off- gas. The hydrogen and therefore any tritium is converted to water and is returned to the CST where it is reused within the plant. Recirculating tritium within the process water for the lifetime of the Power Station delivers additional benefits from decay.

Carbon-14

The assessment of techniques that have the potential to abate carbon-14 gaseous discharges also identified that none of the treatment processes are currently used on operational LWRs. Based on the evidence available, alkaline scrubbing was identified as a potentially viable technique although a programme of research and development would be required to explore the costs and benefits in sufficient detail to determine its actual viability. At GDA it was considered to be grossly disproportionate to undertake this research and development work on the basis that:

• Although carbon-14 makes a significant contribution to offsite dose, the offsite dose remains very low.

• Development work on alkaline scrubbing treatment techniques has identified significant costs including those associated with the disposal of secondary wastes.

Hitachi-GE therefore concluded that not abating carbon-14 in the off-gas represented BAT.

This was later revisited by Horizon in a further options assessment [Ref-85] to establish whether this initial approach still represented BAT in the site specific design. This assessment built on the initial GDA assessment but further investigated:

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• Whether there had been, since the GDA assessment, any significant technological advances in the technologies assessed which would make them more feasible;

• Whether there were any new technologies that were not available at the time of the GDA assessment which could be used to abate H-3 or C-14; and

• Whether any of the site specific factors (mainly there being two UK ABWRs rather than one) challenged the initial GDA approach.

This assessment found that there had not been any significant technological advances since the GDA assessment and that none of the site specific factors challenged the GDA approach. It therefore concluded that the initial GDA approach of not abating the two isotope still represented BAT in the site specific design [Ref-85].

However, as per the EP-RSR permit conditions, Horizon will continually review developments in C-14 abatement technology to establish the suitability of any technological advances in this area for use at the Wylfa Newydd Power Station.

Other measures can be used to minimise tritium production at source rather than implementing treatment techniques [Ref-82] (as summarised in Table 4-1). The impacts of the disposal into the environment of both carbon-14 and tritium are minimised by the use of discharge stacks to discharge gaseous effluents at height (3.5.1 Argument 5a: Gaseous Discharge System - Main Stack). This ensures that the gaseous effluents emitted are dispersed into the environment in a manner which minimises impacts on members of the public and the environment.

3.2.1.4 Evidence: In-process Monitoring to Support Demonstrating the Application of BAT

Monitoring is carried out in order to ensure that the OG is performing as expected. The performance of the charcoal adsorbers are influenced by temperature. As a result, temperature in the charcoal adsorber room is continuously measured to ensure it is within set parameters. Manual sampling and analysis is also carried out at the charcoal adsorber outlet to ensure the radioactivity levels are maintained at a normal level.

3.2.2 Argument 2b: Charcoal Adsorbers for Noble Gases and Iodine Low concentrations of FPs such as noble gases and iodine will be present in the off-gas from the reactor coolant circuit due to: • Tramp uranium; • Fission of fuel material; • Fission of structural uranium; and • Ternary fission in fuel. The concentration of these FPs from within the fuel will increase in the unlikely event of a failure in the fuel cladding. The majority of these FPs have relatively short half-lives and undergo rapid decay. Retention of the FPs in the gaseous waste treatment facility for a period prior to discharge reduces the amount of radioactivity that will enter the environment. The design of the OG system includes a series of four charcoal adsorbers that are filled with charcoal (3.2.2.1: Evidence: Configuration of Charcoal Adsorbers). The use of charcoal adsorbers is common practice in the UK nuclear industry (3.2.2.2: Evidence: Use of Charcoal Adsorbers). The purpose of these charcoal adsorbers is to retain the FPs for a defined period during which they undergo radioactive decay. The rate at which the FPs are adsorbed onto and then de-adsorbed from the surface of the charcoal is dictated by the chemical and physical properties of the charcoal, the FPs and other species passing through the charcoal adsorbers. The charcoal adsorber system has been designed to retain isotopes of xenon for a period of approximately 30 days and isotopes of krypton for approximately 40

Wylfa Newydd Project – Best Available Techniques (BAT) Case hours. Calculations have been undertaken (3.2.2.3: Evidence: Calculations to Support Charcoal Adsorber Size for Xenon and Krypton) that show that the use of the charcoal adsorbers contributes to a reduction in the amount of radioactive krypton and xenon gas discharged to the environment to 1/30,000. The calculations used to design the charcoal adsorbers have not specifically taken account of the presence of isotopes of iodine. However, assessment has demonstrated that, as a result of a combination of the high boiling point of iodine and the properties of the charcoal in the charcoal adsorbers, the charcoal adsorbers are effective at reducing the amount of iodine that is discharged to the environment (3.2.2.4: Evidence: Calculations to Support Abatement of Iodine). Evolution of the ABWR design has introduced a number of improvements (3.2.2.5: Evidence: Charcoal Adsorbers for Krypton, Xenon and Iodine - Design Improvements) to the system for retaining noble gases and iodines. These improvements have increased the length of time that krypton and xenon gases are retained within the gaseous waste treatment system, from one day for all gaseous wastes to the current 30 days for isotopes of xenon and 40 hours for isotopes of krypton, whilst improving the overall reliability of the system. They have also contributed to reducing the amount of iodine discharged. Radioactive argon is also present within the off-gas and is held up by the charcoal adsorbers for 7 hours which achieves a reduction in the amount of argon gas discharged to the environment to 1/14 (3.2.2.6 Evidence: Charcoal Adsorbers for Radioactive Decay of Argon). The Wylfa Newydd Power Station design will include a dedicated temperature controlled room for the charcoal adsorbers which allows the performance of the system to be maintained whilst reducing the quantity of solid waste generated from drying the off-gases. The system provided to treat radioactive gases contains charcoal adsorbers which retain some of the radioactive gases for a defined period of time. During this retention period the radioactive gases undergo radioactive decay. This reduces the radioactivity of these gases to less than 1/500 of that in the gases entering the charcoal adsorbers and will substantially reduce the amount of radioactivity discharged to the environment from Wylfa Newydd.

3.2.2.1 Evidence: Configuration of Charcoal Adsorbers

The OG charcoal adsorbers comprise four vertical towers filled with charcoal. Each charcoal adsorber contains 18 tonnes of charcoal and are connected in series. The OG charcoal adsorbers receive a flow rate of off-gas of 40m³/h during normal operations and 80m³/h during start-up. The flow rate is expected to normally be below the design flow rate of 40m³/h and, as a result, the off-gas benefits from more hold-up than the design basis calculations shown below [Ref-86]. The number of charcoal adsorbers and the quantity of charcoal within them is determined by the required hold-up time and the quantity of gas they receive. This is consistent with the Japanese ABWR design.

Evidence supporting the selected hold-up time is provided in Section 3.2.2.3 (Evidence: Calculations to Support Charcoal Adsorber Size for Xenon and Krypton). The specification of the charcoal used in the OG charcoal adsorbers is provided in the Wylfa Newydd Power Station design. Confirmation that the charcoal will deliver the required performance will be demonstrated during tests carried out prior to active commissioning. In-process monitoring is provided that will allow Horizon to demonstrate that the activity of gaseous radioactive waste discharges have been minimised. The charcoal adsorbers are designed for 60 years of operation without a requirement to replace the charcoal. Calculations supported with OPEX from 20 operational nuclear power stations demonstrate that the charcoal will not deteriorate over the 60 years of operation proposed for the Wylfa Newydd Power Station [Ref-87].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.2.2.2 Evidence: Use of Charcoal Adsorbers

The use of charcoal adsorbers for the abatement of short lived radionuclides is common practice in the nuclear industry and subject to being appropriately optimised is considered to be BAT, as confirmed in the OECD report [Ref-10]. The OECD report identifies that charcoal adsorbers are appropriate for the abatement of noble gases in gaseous discharges and states that they achieve an economically beneficial retention of radioactive noble gases.

Charcoal adsorbers are used extensively in the nuclear industry and are installed at various UK sites including Sizewell B. They are also provided for the treatment of similar wastes within the design of the PWR proposed for installation at Hinkley Point C.

The Environment Agency [Ref-88] guidance states that “for nuclides with short half-lives that decay to stable (or less hazardous) nuclides, storage prior to discharge represents an option for abatement”. This process of decay reduces the activity in the gas and therefore minimises the amount of radioactivity discharged to the environment. Charcoal adsorbers offer a more passive system (fewer moving parts), overall better safety performance and lower operator dose (less pump maintenance) than alternative techniques. Although delay tanks are a viable option, they offer no gaseous treatment benefits over charcoal adsorbers, as discussed in Section 3.2.1.1.

Delay tanks with a hold-up time of 24 hours were used in previous generations of BWR. The OG including the delay tanks provided a decrease in gaseous radioactive waste of 1/300 (3.2.2.3 Evidence: Calculations to Support Charcoal Adsorber Size for Xenon and Krypton). Delay tanks have subsequently been replaced with charcoal adsorbers in the Wylfa Newydd Power Station design. This change is considered to be a design improvement that will decrease gaseous radioactive waste discharges to approximately 1/30,000 of the radioactivity at the inlet of the Charcoal Adsorber as demonstrated in 3.2.2.3 (Evidence: Calculations to Support Charcoal Adsorber Size for Xenon and Krypton).

3.2.2.3 Evidence: Calculations to Support Charcoal Adsorber Size for Xenon and Krypton

The size of the charcoal adsorbers have been derived using design basis calculations. The calculations used are provided as Calculation 3.2.2.3-1, 2 and 3 [Ref-89].

Calculation 3.2.2.3-1 is used to calculate the quantity of radioactivity for each individual radionuclide at a period in time by inputting the initial quantity of radioactivity and the delay constant.

−λt A(t) = A 0 ∙ e where,

A(t): Quantity of radioactivity at time t [Bq/s]

A 0 : Initial quantity of radioactivity [Bq/s]

λ : Decay constant [1/s]

Calculation 3.2.2.3-1 Radioactive Decay Calculation

After calculating the quantity of radioactivity, Calculation 3.2.2.3-2 calculates the DF and then the decay rate. The DF is the ratio of initial specific radioactivity to final specific radioactivity.

( ) Decay rate = = 1 𝐴𝐴 𝑡𝑡 𝐷𝐷𝐷𝐷 𝐴𝐴0 Calculation 3.2.2.3-2 Decay Rate Calculation

Calculation 3.2.2.3-3 allows for the comparison of the relationship between hold-up time and decay performance.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

. [ / ] ( ) = = Decay rate = ( ) . [ / ] 𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅𝑅 𝑎𝑎𝑎𝑎 𝑡𝑡 𝑋𝑋𝑋𝑋30𝑑𝑑+𝐾𝐾𝐾𝐾40ℎ 3 25𝐸𝐸+02 𝐵𝐵𝐵𝐵 𝑠𝑠 1 𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼 𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐𝑐 𝑜𝑜𝑜𝑜 𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟𝑟 𝑋𝑋𝑋𝑋+𝐾𝐾𝐾𝐾 9 68𝐸𝐸+06 𝐵𝐵𝐵𝐵 𝑠𝑠 30000

Calculation 3.2.2.3-3 Decay Rate for Xenon after 30 days and Krypton after 40 hours

The Japanese ABWR provides a hold-up period of 30 days for xenon and 40 hours for krypton [Ref-89]. Using Calculation 3.2.2.3-1, 2 and 3, an assessment has been undertaken to determine if this hold-up period represents BAT for the Wylfa Newydd Power Station. The outcome of the assessment is presented within Table 3.2.2.3-1. Table 3.2.2.3-1 demonstrates that radioactive xenon and krypton are reduced to approximately 1/30,000 of the radioactivity at the inlet of the charcoal adsorber using the Japanese ABWR OG (30 days for xenon and 40 hours for krypton). Based on the design basis calculations most benefit is derived during the early part of the retention period and significantly reduces prior to attaining the end of the retention period of 30 days for xenon and 40 hours for krypton. Increasing the retention period would require the inclusion of additional charcoal adsorbers which is considered to be grossly disproportionate based on the design basis calculations presented in Figure 3.2.2.3-1.

Table 3.2.2.3-1: Radioactive Decay of Xenon and Krypton

Hold-up time for 0 (At the xenon(day) / krypton Charcoal 1/24 20/26.6 30/40 40/53.3 50/66.7 (hour) Adsorber inlet)

Quantity of radioactive 6.81E+06 3.27E+04 9.13E+02 2.45E+02 6.66E+01 1.84E+01 xenon (Bq/s)

Quantity of radioactive 2.87E+06 1.16E+03 7.28E+02 8.09E+01 1.10E+01 2.54E+00 krypton (Bq/s)

Quantity of radioactive 9.68E+06 3.39E+04 1.64E+03 3.26E+02 7.76E+01 2.09E+01 xenon + krypton (Bq/s)

DF for xenon + krypton - 2.86E+02 5.90E+03 2.97E+04 1.25E+05 4.62E+05

The quantity of charcoal in the OG charcoal adsorber that is required to hold up xenon and krypton contained in the off-gas for the established periods (xenon: 30 days, krypton: 40 hours) is calculated by Calculation 3.2.2.3-4 and Calculation 3.2.2.3-5.

× = ×(1 + ) 𝑊𝑊 𝑡𝑡𝑡𝑡 𝑀𝑀 𝑎𝑎 Where: 𝐾𝐾 M: Quantity of charcoal (t)

W: off-gas flow rate (m3/h) tH: Hold-up time (h)

K: Dynamic adsorption coefficient (m3/t)

: Margin ratio (-)

𝑎𝑎 Calculation 3.2.2.3-4 Required Charcoal Quantity Calculation

Wylfa Newydd Project – Best Available Techniques (BAT) Case

56.4×40 = = ×1.15 = 72.02 = 72 (t) 36

𝐾𝐾𝐾𝐾 𝑀𝑀 56.4×30×24 = = ×1.15 = 71.84 = 72 (t) 650

Where: 𝑋𝑋𝑋𝑋 𝑀𝑀

OG flow rate: 56.4 m3/h calculated using the normal flow rate (40m3/h) but with the actual temperature (25°C) and pressure (78.45kPa)

Calculation 3.2.2.3-5 Required Charcoal Quantity Calculation

The results of calculations 5.2.2.3-4 & 5 demonstrate that 72 tonnes of charcoal adsorber is required to achieve the required hold-up time of 30 days for xenon and 40 hours for krypton.

Figure 3.2.2.3-1 Hold-up Time versus Release Rate for Krypton and Xenon

3.2.2.4 Evidence: Calculations to Support Abatement of Iodine

Radioactive iodine is mostly retained within the condensate of the main condenser as a result of its high boiling point. As such, little radioactive iodine is carried over into the OG. Whilst the OG charcoal adsorbers are designed to hold-up radioactive xenon and krypton it has been calculated that it also facilitates the decay of any radioactive iodine that is carried over from the main condenser. This is because radioactive iodine has favourable characteristics that promote its retention within the OG charcoal adsorbers.

The capacity of the OG charcoal adsorbers to adsorb gaseous substances is effected by dynamic adsorption equilibrium constants of the charcoal. Through experimentation the charcoal adsorption capacity for iodine and methyl iodine has been determined and the specifications of the charcoal adsorbers are presented in Table 3.2.2.4-1 [Ref-90].

A separate study was conducted to review abatement techniques used in the nuclear industry for iodine and to support the demonstration that the charcoal adsorbers provided in the Wylfa Newydd Power Station design are

Wylfa Newydd Project – Best Available Techniques (BAT) Case

BAT. The review drew upon conclusions from the OECD report [Ref-10] that “volatile iodine FPs from nuclear reactors are abated using carbon filter beds. The filter beds are porous which provide a very high effective surface area for adsorbing the iodine gas. It compared this to a number of techniques such as: • Silver based solid sorbents;

• Cadmium based solid sorbents;

• Lead based solid sorbents;

• Alkali liquids;

• Iodox liquid;

• Mercurex liquid;

• Electrolyte scrubbing;

• Fluorocarbon Absorption; and

• Organic solvents.

It concluded that, although there are other techniques available which can abate iodine, no other techniques have been used on nuclear power stations other than charcoal adsorbers. Other techniques are used on fuel reprocessing facilities on the basis that carbon is not feasible for use. This is due to issues with other techniques such as:

• Costs;

• Inferior abatement when compared with charcoal adsorbers;

• Large volumes of secondary wastes produced; and

• Additional safety concerns associated with handling corrosive and/or chemically toxic agents.

Wylfa Newydd Power Station charcoal adsorbers are also required to abate a number of short lived radionuclides in addition to iodine and charcoal adsorbers are effective at abating these radionuclides. Additionally, charcoal adsorbers are considered technically mature. To this effect, the cost of further enhancement of abatement techniques, or the development of new techniques for the Wylfa Newydd Power Station would not be justified or cost effective. As such, it was considered BAT to employ charcoal adsorbers for the abatement of iodine.

Further to this, the IAEA publication [Ref-91] also states that for nuclear power stations “elemental iodine is generally removed by a physical adsorption process and nuclear power stations almost exclusively use activated carbon for the removal of radioiodine”.

Table 3.2.2.4-1: Specification of Standard Wylfa Newydd Power Station Charcoal Adsorbers

Radioactive Gas Iodine-131 Iodine-133

Iodine Methyl Iodine Iodine Methyl Iodine

Activated charcoal quantity (M) [t] 72

Off-gas flow rate (W) [m3/h] 56.39

Dynamic adsorption equilibrium 5.16E+04 9.22E+03 5.16E+04 9.22E+03 constant (K) [m3/t]

Hold-up time (h) 6.58E+04 1.18E+04 6.58E+04 1.18E+04

half-life (h) 1.93E+02 1.93E+02 20.80 20.80

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Radioactive Gas Iodine-131 Iodine-133

Iodine Methyl Iodine Iodine Methyl Iodine

Decay constant (λ) [/h] 3.59E-03 3.59E-03 3.33E-02 3.33E-02

DF 5.20E+102 2.36E+18 infinity 2.80E+170

The DF for iodine and methyl iodine is greater than that of xenon and krypton whilst the quantities within the off- gas are lower. This means that the OG charcoal adsorber has a large capacity to decay iodine and methyl iodine. Thus it is reasonable to conclude that discharges of iodine and methyl iodine to the environment from the OG will be very low.

3.2.2.5 Evidence: Charcoal Adsorbers for Krypton, Xenon and Iodine - Design Improvements

Figure 3.2.2.5-1 provides an illustration of the changes introduced in the 1970’s and 1980’s to improve the performance of the OG [Ref-62]. This change resulted in the replacement of the delay tanks with charcoal adsorbers. The delay tanks provided a minimum hold up period of 1 day and operated as a batch process. The charcoal adsorbers deliver a minimum hold-up of 30 days for xenon and 40 hrs for krypton and operate continuously.

Figure 3.2.2.5-1 Design Improvements to the Off-Gas Charcoal Adsorber System

The result of increasing the hold-up time by providing charcoal adsorbers resulted in a significant decrease in the concentration of radioactive krypton and xenon gases released to the environment. A comparison of discharges from each system is provided in Figure 3.2.2.5-2.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

1.0E+08 Storage tank system Charcoaladsorber system 1.0E+07

1.0E+06 3.39E+04

1.0E+05

1.0E+04 3.26E+02

1.0E+03

1.0E+02

1.0E+01 Xe+Kr

RadioactiveNoble Gas ReleaseRate (Bq/s) 1.0E+00 0 １5 10２ 15３ 20 25 30 35 40

30 35 40 45 50 Hold Time(day) Kr hold-up time [hour]

Figure 3.2.2.5-2 Release Rate of Radioactive Krypton and Xenon Over Krypton and Xenon Gas Hold Time Following Design Improvements

3.2.2.6 Evidence: Charcoal Adsorbers for Radioactive Decay of Argon

As detailed in section 3.2.2.3 the primary focus of charcoal adsorbers is for the radioactive decay of krypton and xenon. Section 3.2.2.4 describes how there is additional, greater decay benefits for radioactive iodine. Argon-41 is also present within the OG as a result of activation of argon-40, a constituent of air, which leaks into the main condenser where it is entrained within the coolant and passed through the reactor (3.1.10.7 Evidence: Design Policies to Prevent Atmospheric Argon Leaking into the Coolant System).

The charcoal adsorbers offer 7 hours of hold up time for argon-41 which achieves a reduction in radioactivity of 1/14 of its activity prior to abatement as calculated in Table 3.2.2.6-1 and shown in Figure 3.2.2.6-1.

Table 3.2.2.6-1 Radioactive Decay of Argon

Hold-up time for 0 (At the Charcoal 4.7 7 9.3 11.7 24 argon (hour) Adsorber inlet)

Quantity of radioactive 8.14E+05 1.39E+05 5.72E+04 2.36E+04 9.73E+03 9.03E+01 argon (Bq/s)

DF for argon - 5.86E+00 1.42E+01 3.45E+01 8.37E+01 9.01E+03

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.2.2.6-1 Hold-up Time versus Release Rate for Argon, Krypton and Xenon

As displayed in Table 3.2.2.6-1, the DF achieved for argon is much greater following the 24 hours of decay that would have been achieved in the old delay tank design than under the 7 hours of decay in the more recent charcoal adsorber design. Whilst the charcoal adsorber does not perform as well for argon as the delay tank, the design change represented an improvement in a number of other aspects:

• Hazard reduction – delay tanks are high pressure systems which increase the likelihood of a leak whereas the charcoal adsorber operates under negative pressure so, even in the event of a crack or rupture, the holdup system does not leak.

• Reliability – the delay tank operates in batches with repeated start and stop operations and include a number of valves, whilst the charcoal adsorbers operate in a steady state mode.

• The charcoal adsorbers offer much greater abatement of krypton and xenon using more reliable and passive technology, particularly following a fuel failure where levels of radioactivity increase;

• Less risk of operator error due to a passive, continually operating system. This complements reliability but is a safety improvement as well.

Options to further optimise the abatement of argon-41 have been identified within Table 3.2.2.6-2.

Table 3.2.2.6-2 Options to Further Optimise Ar-41 Abatement

Advantages Disadvantages

Zeolite Adsorber Zeolite has a uniform fine pore Installation of an additional zeolite filter would be diameter which can selectively costly and would require additional space in the adsorb particular gases, for T/B. In addition, the zeolite filter would require example argon. periodic replacement which would result in an increase in secondary waste arisings.

Introduction of a A storage tank could provide Additional cost of installation and high space storage tank in further decay time for argon-41. requirements in the T/B. Reintroduction of the addition to the disadvantages associated with delay tanks. existing charcoal adsorbers

Increase number/ More charcoal adsorbers would Additional cost of installation and high space size of charcoal provide an increased delay time requirements in the T/B resulting in the adsorbers to increase and therefore improve the requirement to provide approximately 247 tonnes residence time to 24 abatement of argon-41. (24/7×72) of additional charcoal in order to hrs achieve the equivalent of a 24hr hold up period.

It has been demonstrated that the existing charcoal adsorbers demonstrate an improvement for the abatement of the off-gas beyond that achieved by the delay tanks provided on earlier BWR designs. However, it is recognised that the delay tanks have the potential to delay argon-41 for a longer period than that offered by the charcoal adsorbers provided in the ABWR design. The transition to charcoal adsorbers overcame a number of disadvantages associated with the operation and safety of delay tanks and improved the reliability and performance of abatement for short lived radionuclides contained within the off-gas.

The review of options that could potentially increase the abatement of argon-41 identified that a dedicated abatement system would need to be provided in the form of alternative technology such as a zeolite based adsorber, reintroduction of delay tanks or increasing the number of charcoal adsorbers.

Qualitative assessment of the additional techniques for further abatement of argon concludes that all three options would be prohibitively expensive compared to the potential benefit and would require an extension to the T/B. Owing to the fact that argon-41 only contributes a small amount of the dose [Ref-37], it is considered grossly disproportionate to abate argon further than the abatement achieved in the charcoal adsorbers.

3.2.3 Argument 2c: Heating, Ventilation and Air Conditioning System Gaseous radioactive waste will be discharged to the environment via appropriately permitted outlets. The air pressure in facilities handling radioactive substances are typically maintained at a lower level than atmospheric pressure to ensure that air flows into the facility from the external environment. This prevents the uncontrolled discharge of any radioactive substances through doors, windows and gaps in the building fabric. The negative pressure within the facility is maintained by a ventilation system which continuously extracts air from within the Power Station facility buildings and discharges it to the environment during normal operations. The design of the Wylfa Newydd Power Station includes a HVAC system which delivers the combined function of: • Preventing the uncontrolled discharge of radioactive substances; • Providing a pleasant working environment for workers;

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• Ensuring optimal working conditions for plant and equipment; and • Delivering safety related functions to protect workers in the event of a release of radioactivity. The configuration of the HVAC system ensures independent operation of sub-systems in principal areas of the plant (3.2.3.1: Evidence: Configuration of HVAC System). It also makes efficient use of the air that is drawn in to the system by allowing it to flow from areas of lower contamination risk to areas of higher contamination risk. HVAC sub-systems that serve areas of the plant where radioactive substances are present have filters to remove any particulate matter prior to discharge to the environment via appropriately permitted outlets. The approach to selecting a filter type that is appropriate for the characteristics of the gaseous waste is described in a separate Argument (3.2.4 Argument 2d: Filtration of Airborne Particulate Matter). HVAC sub-systems that serve areas of the plant where radioactive substances are handled do not provide any abatement other than filters. Additional abatement systems are not considered necessary because, under normal operations, the amount of radioactivity that is expected in the large volumes of waste air drawn through the HVAC system will be very low (3.2.3.2 Evidence: HVAC Discharges). Horizon considers that sufficient evidence has been provided to demonstrate that the HVAC system philosophy represents BAT. To support the demonstration that those parts of the design that have been identified as BAT are operating as expected, the Wylfa Newydd Power Station design provides sufficient space and access to allow Horizon to carry out in-process monitoring (3.2.3.3 Evidence: In-process Monitoring to Support Demonstrating the Application of BAT in the HVAC System). HVAC systems are used for a variety of functions including the treatment and discharge of air from areas where radioactive substances are present. Filtration is used to control hazards associated with radioactive substances and, where present, this will contribute to reducing discharges of radioactivity from Wylfa Newydd. Filtration is not provided in areas where the hazards associated with radioactive substances are considered to be very low.

3.2.3.1 Evidence: Configuration of HVAC System

The HVAC System will be provided to maintain environmental conditions within the Power Station and provide a cascade air flow from areas of low contamination to areas of higher contamination.

The HVAC system will be segregated into sub-systems according to the principal areas and functions of the Power Station. The key HVAC sub-systems include those provided for the: • Reactor area;

• Emergency diesel generator electrical equipment area;

• Turbine area;

• Heat exchanger area;

• Instrument and control power supply panel area inside control building;

• MCR;

• Backup Building ECR;

• Backup Building diesel generator electrical equipment area;

• Controlled area inside the Rw/B; and

• Service Building (S/B).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

The HVAC system has been designed to ensure that appropriate abatement is provided on those systems that serve areas that have the potential to generate gaseous radioactive waste, these are summarised below and further detail is provided in PCSR Chapter 16.3 [Ref-78]. Ultimately all HVAC systems discharge via appropriately permitted outlets.

HVAC System for reactor area

Each reactor unit will have a dedicated reactor area HVAC system, the configuration of which is shown in Figure 3.2.3.1-1. This system provides ventilation and air-conditioning to the reactor area within the Reactor Building (R/B) during normal operations, and this system consists of a supply and exhaust air treatment facility. Air is supplied to the reactor areas and the exhaust air collected from each area is discharged from the corresponding unit’s main stack via the exhaust air treatment facility by the exhaust fans. Filtration will be provided on this system to minimise discharges of radioactive particulate matter. High Efficiency Particulate Air (HEPA) filters that are compliant with UK standards such as NVF/DG001 (An Aid to the Design of Ventilation of Radioactive Areas) [Ref-92] will be installed in Wylfa Newydd Power Station. Two stages of multiple safe change HEPA filter units will be provided. The requirement for two stages of filtration was primarily driven by the area classification; during operations the reactor area is a C2 area however this rises to C3 during maintenance.

A small amount of water vapour is expected to be released via the HVAC system as a result of evaporation from the SFP. This water vapour may contain some radioactivity, the majority of which is made up of tritium. The FPC maintains the water quality of the SFP and heat exchangers remove the decay heat and reduce the temperature of the water. The water temperature within the SFP is maintained below 52°C for operability reasons and the safety of operators. Maintaining the water temperature below this limit also reduces the evaporation rate and therefore discharges via the HVAC. Temperature monitoring is provided in the SFP to ensure that temperatures remain below this limit.

HVAC System for T/B

Each reactor unit will have a dedicated T/B HVAC system, the configuration of which is shown in Figure 3.2.3.1-2.This system provides ventilation and air-conditioning to the turbine area within the T/B during normal operations, and this system consists of a supply and exhaust air treatment facility. Air is supplied to each T/B, and the exhaust air is discharged to the environment via the corresponding unit’s main stack. Exhaust air from potentially contaminated areas of the T/B or component vents is collected, filtered and discharged to the atmosphere through the T/B high radiation area exhaust system. HEPA filters that are compliant with UK standards such as NVF/DG001 (An Aid to the Design of Ventilation of Radioactive Areas) [Ref-92] will be installed in Wylfa Newydd Power Station. Two stages of multiple safe change HEPA filter units are provided in the C2 area of the T/B which becomes a C3 area during maintenance and a single stage of HEPA filtration is provided in the area of the T/B which remains C2 during maintenance.

The Radiological Protection management arrangement for the Designation and Classification of Areas will support the RPA and Health Physics group in implementing changes to the designation of contamination areas.

Figures 3.2.3.1-1 to 3.2.3.1-2 show the configuration of the HVAC systems for the reactor area and T/B.

HVAC System for Controlled Area inside the Rw/B

The HVAC system for the controlled area inside the combined Rw/B provides ventilation and air-conditioning within the Rw/B during normal operations, and this system consists of a supply and exhaust air treatment facility. Air is supplied to the Rw/B, and the exhaust air collected from each area is discharged from the unit 1 main stack via the exhaust air treatment facility by the exhaust fans. Filtration will be provided on this system to minimise discharges of radioactive particulate matter. HEPA filters that are compliant with UK standards such as NVF/DG001 (An Aid to the Design of Ventilation of Radioactive Areas) [Ref-92] will be installed in Wylfa Newydd Power Station. Multiple safe change HEPA filter units are provided, and one stand-by housing is installed.

HVAC System for S/B

The S/B HVAC system provides ventilation and air conditioning within the combined S/B during normal operations, and this system consists of a supply and exhaust air treatment facility. Air is supplied to the S/B, and

Wylfa Newydd Project – Best Available Techniques (BAT) Case

the exhaust air collected from each area of the S/B is discharged from a local exhaust opening via the exhaust air treatment facility by the exhaust fans. HEPA filters that are compliant with UK standards such as NVF/DG001 (An Aid to the Design of Ventilation of Radioactive Areas) [Ref-92] will be installed in Wylfa Newydd Power Station. Two stages of multiple safe change HEPA filter units are provided in the C2 area of the S/B which becomes a C3 area during maintenance and a single stage of HEPA filtration is provided in the area of the S/B which remains C2 during maintenance.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.2.3.1-1 Outline of the Reactor Area HVAC

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.2.3.1-2 Outline of the T/B HVAC

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.2.3.2 Evidence: HVAC Discharges

The realistic radioactive discharge to atmosphere from the HVAC systems serving each building has been estimated by identifying the main radioactive discharge sources (such as SFP) in the source term topic report [Ref-91]. Dose rates from the HVAC system were compared with and without the use of HEPA filters [Ref-93]. Dose rates from HVAC discharges when HEPA filters are not installed are approximately 2.2μSv/y. In comparison, dose rates from HVAC discharges with HEPA filters are approximately 2.1μSv/y. Thus it can be seen that the use of HEPA filters provides a very small benefit in terms of dose reduction during normal operation, however HEPA filters are installed in Wylfa Newydd Power Station that are compliant with UK standards such as NVF/DG001 [Ref-92].

Hitachi-GE also conducted a study at GDA into possible abatement techniques within the HVAC system for iodine. However, it was concluded that the limited source term, high dilution, secondary waste generation and high capital costs for abatement techniques were grossly disproportionate [Ref-94].

It must be noted however that the reasoning for employing HEPA filtration is driven by nuclear safety, specifically in reducing radioactive discharges in the event of an accident. However, since they do offer an albeit small reduction in particulate discharges during normal operation, they do contribute to the environmental performance of the plant and are therefore included in this BAT case.

3.2.3.3 Evidence: In-process Monitoring to Support Demonstrating the Application of BAT in the HVAC System

Monitoring is carried out in order to ensure that the HEPA filters are performing as expected during normal operations and will perform as expected in accident conditions. In situ challenge testing will be carried out by Horizon to ensure filters are performing within defined parameters. Differential Pressure monitoring will be used to indicate when filters are blocking and need changing [Ref-95].

3.2.4 Argument 2d: Filtration of Airborne Particulate Matter The Wylfa Newydd Power Station will employ appropriate filtration techniques to ensure that the concentration of particulate matter within the gaseous radioactive waste stream is minimised during normal conditions and fault events. HEPA filtration is considered to be RGP in the UK nuclear industry for the abatement of particulate matter and it is the design intent to provide filtration on the Wylfa Newydd Power Station HVAC system as appropriate (3.2.4.1: Evidence: Application of Filtration in the Nuclear Industry). The Wylfa Newydd Power Station has been subject to considerable optimisation that has minimised the amount of particulate matter that has the potential to become mobilised within the building areas serviced by the HVAC systems. As a result, during normal operations, concentrations of particulate matter are not expected to be significant. The number and specification of the filters to be used to abate airborne particulate matter has been determined through the application of BAT; however, filter selection will be influenced by the functional requirements to mitigate activity releases in fault events (3.2.4.2: Evidence: The Basis for Filter Selection). Therefore the DF performance of filtration systems will exceed the challenge that is expected to be experienced during normal operations. Attempts were made to optimise the extract air flow rates within the HVAC systems in order to reduce the number of filter assembly changes. However, extract air flow rates are primarily determined based on reducing risk and controlling potential airborne contamination (3.2.4.3 Evidence: Filtration of Airborne Particulate Matter – Air Flow Rate). Temporary methods for abating gaseous wastes generated from decommissioning activities will be deployed and it is expected that connections will be made to existing HVAC systems (3.2.4.4 Evidence: Use of Local, Temporary Filtration During Decommissioning).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Filtration, where used in the HVAC systems, removes particulate matter (including radioactive particulates) from the air stream. The rate of the air flow through the HVAC systems influences the number of filters that will become radioactive waste and which will require disposal from Wylfa Newydd. It has not been possible to reduce the rate of air flow (and the associated number of waste filters) because of the safety functions that the HVAC system is required to provide.

3.2.4.1 Evidence: Application of Filtration in the Nuclear Industry

The standard technique for the removal of particulate matter from gaseous effluents collected by HVAC systems in nuclear installations is the use of HEPA filters (filters that have a capture efficiency of 99.99% as stated in Technical Guidance from the Environment Agency [Ref-97] and the Nuclear Industry Safety Directors Forum [Ref- 92]). During normal operations at the Wylfa Newydd Power Station, HVAC serving radiologically controlled areas will be abated by HEPA Filtration [Ref-78].

The HEPA filtration systems for Wylfa Newydd Power Station meets the industry codes of practice for testing, design and operation with particular reference to UK standards such as NVF/DG001 [Ref-92]. It can be considered that the systems used for the Wylfa Newydd Power Station are consistent with those used for existing installations in the UK and employed on most other nuclear installations internationally.

3.2.4.2 Evidence: The Basis for Filter Selection

HEPA filters will be used on the exhaust air treatment facility of the reactor area, T/B, Rw/B and S/B HVAC systems. They are circular safe change HEPA filters as displayed in Figure 3.2.4.2-1.

Figure 3.2.4.2-1 Safe Change HEPA filter unit Outline Drawing

Banks of HEPA filters are commonly used to remove radioactive solid particulate matter in dry atmosphere environments. HEPA filters achieve DFs of greater than 1,000 when several filters are arranged in series and parallel combinations. These technologies are relatively mature and it is unlikely that new techniques will significantly enhance their effectiveness [Ref-8].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Exhaust air from the potentially contaminated areas is discharged from appropriately permitted outlets. The extraction system for C2 areas (low contamination areas) includes a single stage HEPA filter whilst the extraction system for C3 areas (high contamination areas) includes two stages of HEPA filters [Ref-78].

3.2.4.3 Evidence: Filtration of Airborne Particulate Matter – Air Flow Rate

Previously, attempts were made to reduce the number of air changes in the UK ABWR HVAC system compared to previous BWR and ABWR designs [Ref-98]. The design modifications that were made to achieve this were:

• The introduction of a refrigerator for cooling supply air;

• Reduction of design air change rate; and

• Reduction of building capacity.

These design modifications resulted in a reduction in the air flow rate which had the potential to reduce the number of filter changes required and thus would reduce the volumes of solid waste generated. Another benefit would be a reduction in energy consumption as a result of reduced demand on the fans.

Following further review and ALARP assessment [Ref-95], as well as a review of compliance with NVF/DG001 [Ref-92], it was concluded that air flow rates should be increased for the Wylfa Newydd Power Station. The primary driver for this increase was contamination control achieved by controlling the number of air changes and the velocity requirement for the flow from C2 areas to C3 areas. The flow rates are shown in Table 3.2.4.3-1.

Table 3.2.4.3-1: Reduction in Air Flow Rate from the BWR Design, ABWR Design and Wylfa Newydd Power Station Design

Wylfa Newydd Power Station BWR generation 5 ABWR (KK-6/7) (flowrate per reactor Power Station Area (flowrate per reactor (flowrate per reactor unit) (See Section 6 of unit) unit) the EP-RSR Application)

Reactor Area 225,000 m³/h 170,000 m3/h 228,285 m3/h

T/B 400,000 m³/h 328,000 m3/h 353,143 m3/h

Rw/B 185,000 m³/h 80,000 m3/h 175,547 m3/h

Whilst attempts were made to optimise air changes in order to reduce the generation of solid waste, the air flow rates are driven by the safety function of the HVAC system.

A benefit of increasing the air flow rate to similar rates found in generation 5 BWRs is an increase in the degree of dilution that will be achieved when the HVAC extract is combined with the treated off-gas prior to being discharged from the main stack. As shown by Section 3.5.1.3 Evidence: Gaseous Discharge – Dilution Factor, the dilution achieved is significant and supports the demonstration that the impact associated with discharges to air have been optimised.

3.2.4.4 Evidence: Use of Local, Temporary Filtration During Decommissioning

Radioactive gaseous waste generated during decommissioning will be discharged via the HVAC system, using local, temporary filtration and extract where appropriate.

3.2.5 Argument 2e: Optimisation of the Turbine Gland Seal The TGS gland steam exhauster discharges gaseous radioactive waste into the main stack where it combines with the OG and HVAC extract before being discharged into the

Wylfa Newydd Project – Best Available Techniques (BAT) Case

environment. In 1976 a design change was introduced to prevent the leakage of reactor steam from the turbine into the turbine building and to avoid discharging unabated reactor steam directly into the environment. A diagram of the original design and the new design comprising a multi stage seal and separate steam seal supply system is show in Figure 3.1.10.11-1. The new design is more robust at preventing reactor steam leakage and is designed to return reactor steam to the main condenser rather than discharging part of it unabated to the environment. The SSSS prevents the leakage of air into the seal and subsequently into the main condenser. Air entering the main condenser would reduce the condenser vacuum pressure leading to a turbine trip (the current design would not function without the design modification that introduced the SSSS). Air that leaks into the main condenser also has the potential to become activated in the reactor increasing the generation of radioactive waste that would require management in the OG. The SSSS uses steam created by the TGS. The TGS extracts water from the CST and converts it to steam using reactor steam in the turbine gland evaporator. Part of the turbine gland steam used in the turbine gland is then extracted with reactor steam into the main condenser. The remainder is extracted with any air in leakage to the gland steam condenser. 98% of the turbine gland steam is condensed in the main condenser and gland steam condenser and is subsequently reused in the plant. The remaining residual steam (steam not condensed in either the main condenser or GSC) is discharged via the gland steam exhauster to the main stack. The discharge contains tritium and very small quantities of iodine and particulates and has been conservatively calculated to contribute approximately 2% of the dose from the Wylfa Newydd Power Station. A HEPA filter is provided on the extract from the gland steam exhauster to remove the particulate matter. Condensing 98% of the steam in the main condenser and GSC also prevents the discharge of 98% of the tritium and iodine as it is entrained in the TGS steam. The new design minimises the generation of radioactive waste compared to the original BWR design. It is recognised that the design does result in the discharge of gaseous radioactive waste. The abatement of the CST water prior to use and minimising the amount of residual steam that is discharged by condensing it results in the contribution to dose being very small (3.2.5.1 Evidence: TGS Impact on Gaseous Discharges). As a result of this contribution to the gaseous radioactive waste discharges from the Wylfa Newydd Power Station an assessment of options was undertaken to determine if the TGS could be further optimised. A robust process was implemented to both identify and assess the options (3.2.5.2 Evidence: TGS Options Assessment). During the assessment it was identified that to either reduce the source term of the steam or to reduce the radioactivity of the gaseous radioactive waste would result in significant cost (in terms of time, trouble and effort). A number of options also resulted in the generation of a large volume of aqueous radioactive waste that would require discharge (weakening Argument 1h: Recycling of Water to Prevent Discharges). Improvements in the TGS design since the 1970’s have improved the leak tightness of the plant. Reactor steam is prevented from leaking into the turbine building where it has the potential to result in a discharge to the environment. The leakage of air into the main condenser is also prevented (3.1.10.11 Evidence: Improvements in Turbine gland seal design). The combined performance improvements delivered by the TGS and TG in terms of minimising the radioactivity of gaseous radioactive waste, preventing the leakage of reactor water/ stream into the turbine building, minimising the in-leakage of air into the main

Wylfa Newydd Project – Best Available Techniques (BAT) Case

condenser whilst preventing the generation of additional volumes of aqueous radioactive waste is considered to represent a significant benefit beyond that achieved by the original design. The assessment that explored opportunities to further optimise the TGS concluded that the costs (in terms of time, trouble and cost) are grossly disproportionate compared to the benefits in terms of dose reduction. The benefits of the existing design in addition to the other improvements detailed within this demonstration of BAT report that reduce the radioactivity of the CST water contribute to the application of BAT. The TGS has evolved to improve the efficiency of the turbine and to reduce the creation of gaseous activation products in the reactor. The current design minimises the generation of aqueous radioactive waste but will result in a small increase in the activity of gaseous radioactive waste that will be discharged from Wylfa Newydd. The cost of treating the gaseous waste from the TGS to remove the radioactive gases is considered to be grossly disproportionate to the benefits that will be received by members of the public.

3.2.5.1 Evidence: TGS Impact on Gaseous Discharges

The radioactivity within the residual steam has been calculated based on the source term within the CST. The extract from the gland steam exhauster is provided with a HEPA filter to remove any particulate matter before it being combined with the HVAC system in the main stack. Although a HEPA filter is provided this will only remove particulate matter and as such it has been assumed that it will have less impact on gaseous radioactive waste discharges which will mainly consist of tritium and iodine.

The changes of the design of the TGS will have a positive impact on the OG system as combined TGS and TG prevent the in-leakage of air into the main condenser. Air that enters the main condenser can become entrained within the condensate and subsequently activated within the reactor resulting in greater challenge to the OG system, therefore minimising the generation of activation products and reducing the demand on the OG system. The baseline TGS design is not connected to the HVAC or OG system. These discharges are combined in the main stack.

Condensate from the main condenser is returned to the CST and as such is subject to the same treatment as other water that is reused within the plant.

It is acknowledged that the TGS would result in a separate waste stream for gaseous tritium and iodine which will be discharged via the main stack independently of the OG and HVAC systems. Neither iodine nor tritium is abated via the TGS route. The additional discharge and the impact on dose is given in Table 3.2.5.1-1.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Table 3.2.5.1-1: Quantification of TGS discharge

Discharge Discharge Dose from Dose from Discharge from Dose from from OG from HVAC HVAC TGS TGS system OG system system system system system (Bq/y)6 (μSv/y) (Bq/y)4 (Bq/y)5 (μSv/y) (μSv/y)

Noble gases 1.8E+12 0.0E+00 0.0E+00 5.8E+00 0.0E+00 0.0E+00 (Kr, Xe, Ar)

Carbon-14 9.1E+11 0.0E+00 0.0E+00 6.2E+01 0.0E+00 0.0E+00

H-3 0.0E+00 1.3E+12 1.6E+12 0.0E+00 1.2E+00 1.5E+00

Iodine 0.0E+00 2.7E+08 4.8E+07 0.0E+00 8.6E-01 2.3E-02

Particulate 0.0E+00 4.0E+04 2.1E+05 0.0E+00 1.2E-04 4.7E-04

Total 6.8E+01 2.1E+00 1.5E+00

Contribution(％) 95.0 2.9 2.1

Note: The annual discharges from each discharge route (OG, HVAC, and TGS) are set by considering expected maximum discharge of each discharge route. (i.e. OG, TGS: 12 month operation, HVAC: 11month operation +1 month outage).

Note: Table 3.2.5.1-1 reflects discharge data which was available at the time of the options assessment. The source term has since been updated to reflect the TGS discharges and as such, these figures vary slightly to those presented in Section 5 of the EP-RSR application. These figures are not intended to replace those presented in Section 5, nor aligned with those presented in Section 5. They were calculated for the options assessment to illustrate an estimate of the contribution that the TGS has to total discharges and dose and should therefore be interpreted by the reader as simply the information present at the time of the options assessment.

The preliminary dose calculations are performed using DPUR (Dose per Unit Release) described in the Initial Radiological Assessment Tool (IRAT) methodology, which was developed by the Environment Agency. This DPUR is the same as the coefficient used in Stage 1 of E8 Prospective Dose Modelling [Ref-99]. The coefficient in the dose evaluation is very conservative. For that reason, the full prospective dose results shown in E8 will be lower than that calculated here using the Stage 1 methodology.

The decontamination factor of 99.9% has been used as a removal factor of HEPA for particulate in the discharge assessment.

3.2.5.2 Evidence: TGS Options Assessment

An assessment was carried out by Hitachi-GE to systematically review the TGS to identify those options that have the potential to impact on the generation of radioactive waste or on-site and off-site dose resulting from the TGS.

4 Calculated from discharge data taken from [Ref-99].

5 Calculated from discharge data taken from [Ref-99].

6 Calculated from discharge data taken from [Ref-99].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

The discussions and outcomes of the workshop are documented in detail in the TGS: Demonstration of BAT document [Ref-100] and are summarised below.

The options considered in the assessment were: • Revert to original BWR design (i.e. go backwards because the new design was the wrong choice).

• Replace CST with HS,

• Replace CST with purified water,

• Further treatment of CST water,

• Replace sealing steam with air/ nitrogen,

• Use less steam for seal,

• Connect TGS discharge to main condenser or replace GSC with main condenser,

• Install drain separator (demister plate) at TGS discharge,

• Larger GSC,

• Connect TGS discharge to OG condenser inlet,

• Install cooler condenser at TGS discharge,

• Add abatement at TGS discharge, and

• Existing baseline TGS system design (the do nothing option).

The options assessment identified that the performance of the options was a balance between reducing gaseous radioactive waste discharges and increasing costs (in terms of time trouble and effort). It is also recognised that the increase in worker dose, industrial safety and non-radioactive environmental impact effectively result in an increase or reduction in cost (in terms of time trouble and effort) as the expectation is that a reduction in performance would result in a requirement for additional costs to apply mitigation when implementing the option. It was also identified that operability and aqueous radioactive waste were key attributes in differentiating the performance of the different options.

The outcome of this assessment identified that the baseline design represents an evolution of the design utilised in the early 1970s and continues to provide benefits in terms of safety and environmental performance. The baseline design supports the arguments presented within the Demonstration of BAT report in terms of minimising liquid waste discharges, whilst preventing leakage of reactor water/ steam from the TG and air into the TG. The use of CST water to produce the TGS seal steam does result in a gaseous radioactive waste discharge although this is considered to be optimised compared to the design used before 1976. The CST water is also subject to abatement to remove particulates and soluble radionuclides prior to use.

A robust process has been implemented to both identify and assess options that have the potential to further optimise the performance of the TGS system. This assessment has concluded that the contribution to dose from this system is already very small reflecting the fact that the design is already optimised and that the introduction of further modifications would result in very high costs (in terms of time, trouble and effort). On this basis it is concluded that the baseline TGS system is BAT and that further optimisation of the design is grossly disproportionate compared to the benefit in terms of both onsite and offsite dose reduction [Ref-100].

3.2.6 Argument 2f: Configuration of Liquid Management Systems Significant effort has been made to minimise the generation of aqueous radioactive waste. However, the safe and efficient operation of the Wylfa Newydd Power Station requires that small volumes of radioactive liquids are disposed of to the environment, after appropriate treatment.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

The design of the Wylfa Newydd Power Station includes a centralised liquid effluent management system (3.2.6.1 Evidence: Centralised liquid effluent management system) that collects, conveys and, when essential discharges aqueous radioactive wastes (3.2.6.2 Evidence: Configuration of Liquid Effluent Management System) from both reactor units. The option of employing two dedicated liquid effluent management systems, one servicing each reactor unit, was considered, but it was decided that BAT was represented by a single, centralised liquid effluent management system (3.2.6.1 Evidence: Centralised liquid effluent management system). There is a system of segregated drains (3.2.6.2: Evidence: Configuration of Liquid Effluent Management System) that allow wastes with broadly similar characteristics to be collected separately prior to the application of any treatment. Liquid effluent is then routed, depending on its characteristics, to one of three treatment systems (collectively known as the liquid effluent management system); the LCW, the HCW and the CAD. The segregated drainage system ensures that treatment techniques can be targeted on specific characteristics of the waste stream and enhances the overall performance of the liquid management system. Treated water that meets the criteria for re-use within the reactor circuit is sent to the CST from where it can be pumped back into the reactor cooling circuit (see 3.1.8 Argument 1h: Recycling of Water to Prevent Discharges). Treatment systems include filtration, demineralisation and evaporation processes to remove certain materials and radionuclides which are considered in separate Arguments (3.2.8 Argument 2h: Demineralisers for Distillates from the HCW Evaporator) and (3.2.9 Argument 2i: Evaporation of HCW). The design has been developed by adopting a series of design policies (3.2.6.3: Evidence: Design Policies for the Liquid Effluent Management System) relating to the minimisation, segregation, containment, treatment and discharge of radioactive liquid effluent. The LCW and CAD are designed to treat liquid effluent to a condition that is suitable for reuse within the facility. Only in the event that the volume of liquids within the cooling circuit and pools exceeds the maximum working capacity or in the few expected instances where treated HCW does not meet reuse criteria will treated/conditioned effluent be discharged from the facility rather than being returned for re-use. Design data demonstrates that the quantity of aqueous radioactive waste disposed of to the environment via this route each year will be extremely low. The water within the cooling circuit will only be discharged to the environment at the end of the operational life of the facility following appropriate treatment and assurance monitoring (3.2.6.4 Evidence: Key Parameters and Water Balance). CAD will be routinely discharged to the environment following treatment and assurance monitoring but the volume and radioactivity of this is expected to be very low. Some of the radionuclides in the aqueous effluent such as tritium and carbon-14 do not undergo treatment in the liquid effluent management system. The majority of techniques (3.2.6.5 Evidence: Assessment of Liquid Treatment Techniques for Tritium) that could be used to treat these radionuclides have been shown to have installation and operating costs that are very high, and are considered grossly disproportionate compared with the low impacts associated with discharges. Therefore tritium and carbon-14 is typically returned to the reactor system where it will reach an equilibrium concentration. Only a small proportion of liquid effluent is disposed of to the environment during the operational phase of the Power Station and this is only in the event that liquids within the cooling circuit and pools exceeds

the maximum working capacity. Additionally, concentrations are low and significant challenge is associated with treating low concentrations of tritium and carbon-14 in liquid effluents. As such, disposal to the environment subject to reassurance monitoring is therefore considered to be the BAT. The Wylfa Newydd Power Station design provides sufficient space to allow Horizon to carry out in-process monitoring to confirm that what has been identified as BAT is performing as expected (3.2.6.6 Evidence: In-process Monitoring to Support Demonstrating the Application of BAT). Significant effort has been made to minimise the generation of aqueous radioactive waste but some will need to be discharged to the environment. The liquid effluent management system has been developed to target specific characteristics of aqueous waste and to reduce the amount of radioactivity that will be discharged from Wylfa Newydd.

3.2.6.1 Evidence: Centralised liquid effluent management system

Early in the design of the Wylfa Newydd Power Station a liquid effluent management system options assessment was undertaken [Ref-101] which explored the merits and detriments of both a single, centralised liquid effluent management system and of two dedicated liquid effluent management systems, one servicing each reactor unit. This initial options assessment concluded that employing two dedicated liquid effluent management systems, one servicing each reactor unit, represented BAT.

This options assessment was based on the draft design at the time whereby the reactor units and supporting buildings were situated such that if a single Rw/B were to be employed, it would have been situated close to one unit but some distance from the other. This was undesirable as it would have resulted in the requirement to move unconditioned radioactive waste, spent ILW resins being of most concern, over a large distance. It was therefore decided that employing two dedicated Rw/B, one servicing each reactor unit, represented BAT.

Because liquid effluent would also require movement over this distance, and influenced by the Rw/B decision as the liquid effluent management system is housed within this, it was decided that having two dedicated liquid effluent management systems also represented BAT [Ref-101].

However, as the design of the Wylfa Newydd Power Station evolved, the layout as a whole was reviewed. It was noted that if the position of the two units was revised, many duplicated facilities across the site, the Rw/B being one of them, could be combined into a single facility servicing both units. Furthermore, these combined facilities could be placed centrally and equidistance from the two units. This design represented BAT so was employed [Ref-102].

This decision then prompted a revisit of whether to employ one or two liquid effluent management systems based on the new design. Since the initial issue with regards to the movement of unconditioned radioactive waste and effluent associated with having one liquid effluent management system was mitigated by the revised design, and because the new design resulted in a single Rw/B, it was decided BAT to employ only one liquid effluent management system within this Rw/B to service both reactor units. This was decided because one liquid effluent management system servicing both units would result in reduced operational radioactive waste arisings, reduced decommissioning radioactive waste arisings, reduced maintenance burden associated with the facility and a reduced overall footprint of the building [Ref-102].

3.2.6.2 Evidence: Configuration of Liquid Effluent Management System

The liquid effluent management system segregates and collects liquid effluent generated which has the potential to become contaminated during normal operation of the Power Station. Treatment of the segregated liquid effluent is determined based on the properties of the effluent. Typically treated liquids are reused after being

treated by the liquid effluent management system. However, in some cases they may be discharged after reassurance monitoring has demonstrated that concentrations of radioactive substances are very low.

The treatment capacity and the configuration of the systems provided in the liquid effluent management system are designed to manage the maximum design basis volume of liquid effluent.

The liquid effluent management system comprises of the following three systems [Ref-103]:

LCW

LCW originates from:

• Equipment drains (which includes T/B LCW sump, CUW blow down, reactor well drain, R/B LCW sump etc.);

• Equipment blow down water (waste generated only in the case of periodic inspection); and

• SFP, FPC and CUW.

The LCW consists of filters, for the removal of insolubles, demineralisers, for the removal of solubles, and sampling pools. Treated liquids are returned to the CST. The selection of these treatment techniques was based on a Hitachi-GE assessment undertaken during GDA (as discussed in Section 3.1.8.4).

Since the nature of the waste has not changed since GDA this decision remains valid.

HCW

HCW originates from:

• Rw/B HCW sump;

• Chemical drains; and

• Equipment blow down water (waste generated only in the case of periodic inspection).

The HCW comprises an evaporator for removal of impurities and a demineraliser for removal of residual solubles. The selection of these treatment techniques was based on a Hitachi-GE assessment undertaken during GDA (described in sections 3.2.8 and 3.2.9). Treated liquids are either transferred to the CST for reuse or in limited circumstances disposed of to the environment following reassurance monitoring.

Since the nature of the waste has not changed since GDA this decision remains valid.

CAD

This stream comprises various liquid effluents generated by plant and equipment in the Wylfa Newydd Power Station facility’s controlled areas, which are not otherwise captured by the HCW and LCW. For example, liquid derived from local air-conditioning units located in the controlled area. The quantity of CAD generation depends on the temperature and the humidity in the building. The CAD is discharged to the environment following reassurance monitoring.

The liquid effluent management system will be installed in the combined Rw/B and will service both reactor units [Ref-102]. Figure 3.2.6.2-1 shows the liquid effluent management system and further details are provided in PCSR Chapter 18.2 [Ref-67].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.2.6.2-1 Liquid Effluent Management System Flow Sheet

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.2.6.3 Evidence: Design Policies for the Liquid Effluent Management System

The design of the liquid effluent management system has been developed based on the following design policies [Ref-103]:

• The liquid effluent management system shall separate, collect and treat liquid effluents. As a general rule, the treated liquids shall be reused, and discharge of radioactive substances shall be minimised as far as practical.

• The treatment capacity of the liquid effluent management system and the in-line configuration of the system shall be designed so that it will be able to deal adequately with anticipated cases in which the maximum amounts of waste liquids are generated. The components of the liquid effluent management system shall be made of suitable materials, taking into consideration the properties of the liquid effluents.

• The following items shall be taken into consideration in designing the facilities for treating liquid effluents and the facilities related to them in order to prevent leakage of liquid radioactive substances from these facilities and to prevent their uncontrolled discharge outside of the site:

• In order to prevent the occurrence of leaks, liquid effluent management system unit operations shall be made of suitable materials to prevent leaks and where appropriate will be provided with level alarms to aid detection of leaks and to prevent overfilling.

• As a general rule, the drain pipes and vent pipes which empty outside of the system shall be provided with caps or similar closing devices. However, those which are used with a high frequency shall have drain vents leading to tanks or sump pits.

• Should radioactive liquids leak out, there shall be provisions making it possible to detect the leaks promptly and to remove and decontaminate the leaked liquids easily.

• The SSC of the liquid effluent management system will either be partitioned in an independent section, or bunded in order to prevent spreading of any leaks inside the facilities. Weirs shall be provided on inlets and outlets connected to points outside the facilities in order to prevent leakage outside of the facilities. Outdoor devices and outdoor pipes shall be designed so that any leaked liquids will be collected inside facilities such as shielding walls or pipe ducts. Floor and wall surfaces of facilities in which there is a possibility that aqueous radioactive wastes might collect will be constructed from materials that minimise the potential for leakage.

• Alarms for the tank water level or leakage detection shall be designed so that the output can be displayed either in the MCR or in the Rw/B control room, so that it will be possible to inform the operators reliably of abnormalities, and so that they will be able to take suitable measures.

• The facilities shall be designed so that no floor surfaces within the facilities shall be located on drainage channels discharging liquid wastes outside of the site in an uncontrolled manner. They shall also be designed so that no apertures connected to uncontrolled drainage channels shall be located inside the related facilities.

• The liquid effluent management system shall be designed so that it can be centrally monitored and controlled in the Rw/B control room. It shall also be possible to monitor this in the MCR.

A design review has been undertaken which has provided confirmation that the design policies have been appropriately addressed and implemented during the development of the Wylfa Newydd Power Station design [Ref-105].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.2.6.4 Evidence: Key Parameters and Water Balance

Table 3.2.6.4-1 provides a summary of the yearly estimated quantity of aqueous waste generation from each treatment system based on the cycle average (i.e. average of 17 month operation and 1 month outage) and the ratio that each treatment system contributes to the total volume of aqueous waste.

Table 3.2.6.4-1 : Estimated Quantities of aqueous waste being treated

Volume (m³/y) (cycle System Ratio to total average)

LCW 10,000 83%

HCW 400 3%

CAD 1,600 13%

Total 12,000 100%

Common practice for operators of ABWRs is to re-use treated LCW and HCW within the primary circuit. The liquid effluent management system is designed to process the liquid effluent to meet the design criteria of treated water that can be returned to the CST for reuse [Ref-103], as shown in Table 3.2.6.4-2. If, after treatment the LCW or HCW water quality does not meet the criteria for reuse (or discharge in certain circumstances for HCW waste), the treated water would be returned to the Collection Pool or Tank from where it originated via the reprocessing line for re-treatment [Ref-106].

Table 3.2.6.4-2: Minimum Criteria to be achieved by Treatment to Allow Reuse

Item Criteria

Conductivity 100 μS/m

Clˉ 20 ppb

pH 5.6 - 8.0

2ˉ SO4 20 ppb

TOC 400 ppb

In some cases Horizon may be required to discharge a small proportion of the treated HCW to the environment, following reassurance monitoring, in order to maintain the water balance of the Power Station. Detail on the operations that may result in the requirement to discharge surplus water is included in [Ref-107]. Examples include:

• Water drained from the system during inspection of equipment; and

• Activities during start up and shut down that result in the use of pure water such as using house steam as the driving force for the SJAE until switch over to the main steam.

Although only a proportion of the treated HCW will be discharged, if it is assumed that all treated HCW is discharged to the environment, the maximum volume of water from the HCW is 400m3/year. This is approximately 4% of the total quantity of liquid effluent (10,000m3/year + 400m3/year) generated from LCW and

Wylfa Newydd Project – Best Available Techniques (BAT) Case

HCW. However, OPEX demonstrates that only a small number and volume of discharges from the HCW are made [Ref-108]. The approaches taken to minimise discharges from the HCW are [Ref-107]:

• Retaining and recycling water within the Power Station. During operation, water is returned to the CST for reuse. During outages, several water tanks, including the CST are used to manage the storage of water during the plant outage. Upper pools can also be utilised as water storage during outage.

• Minimising the addition of treated fresh water to the system.

The Quantification of Discharges and Limits report [Ref-37] calculates the annual discharges from the HCW based on the assumption that all processed HCW is discharged to the environment.

All treated water from the CAD system is discharged to the environment after reassurance monitoring to confirm that the water quality satisfies the discharge limits. Non-radioactive components including environmentally sensitive organic species are described in Section 8 of “Other Environmental Regulations” [Ref-109]. CAD aqueous waste is collected in the CAD collection tanks, and then sampled and analysed before being discharged to the environment. Usually CAD aqueous waste is discharged to the environment without processing as it is not radiologically contaminated waste. If sampling and analysis identifies that the CAD water quality does not satisfy Discharge Criteria, the CAD aqueous waste is transferred to the HCW collection tanks for processing using the HCW [Ref-106]. The outline water balance for liquid effluents in the Power Station is shown in Figure 3.2.6.4-1.

The limits that discharges from the CAD and HCW must meet when discharged to the environment will be defined in Section 5 of the EP-RSR Application.

Figure 3.2.6.4-1: The Outline Water Balance of Liquid Effluents in the Plant

3.2.6.5 Evidence: Assessment of Liquid Treatment Techniques for Tritium

Hitachi-GE undertook an assessment of techniques for the abatement of tritium in liquid wastes [Ref-82]. The assessment considered the following techniques:

• De-tritiation and tritium removal/separation;

• Decay of liquids;

• Evaporation; and

• Conversion of tritiated water to solid waste.

Although the assessment identified that there are treatment processes which could, in theory, be used to reduce liquid tritium discharges from a Wylfa Newydd Power Station, the cost and resources of their implementation is considered grossly disproportionate to the benefits in reduced public and environmental dose (which is very low) from the use of such processes. This position is strengthened by the fact that none of the treatment processes are currently used on operational ABWRs or other LWRs and would require significant development work to make them viable for operational use. There are currently no processes implemented on an industrial scale for the treatment of tritium in liquid phase for an ABWR.

By recycling and not making liquid discharges of reactor water; the majority of the tritiated water remains within the Power Station for its lifetime. Some tritiated water will be lost via evaporation and when discharges are made from the HCW to maintain the water balance of the plant. The residence time of the reactor water within the plant for the lifetime of the Power Station will contribute to minimising the radioactivity that becomes waste as a result of radioactive decay. The half-life of tritium is 12.3 years whilst discharges will not occur until after 60-80 years allowing the radioactive decay of proportions of the tritiated water.

3.2.6.6 Evidence: In-process Monitoring to Support Demonstrating the Application of BAT

The components of the liquid effluent management system that require in-process monitoring are listed in Table 3.2.6.6-1 along with detail on the monitoring methods applied [Ref-39].

Table 3.2.6.6-1 Components of the Liquid Effluent Management System Requiring In-process Monitoring and a Description of the Monitoring Techniques

Treatment Monitor Location of monitoring Objective Type Facility

LCW Manual Manual sampling & Analysing suspended solid and Manual sampling analysis at LCW conductivity to confirm that the sampling & collection tank. properties of collected liquid analysis. effluent are suitable for the LCW.

Conductivity Inlet and outlet of LCW Monitoring conductivity to Continuous meter demineraliser. confirm the processing measurement. performance of the

demineraliser.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Treatment Monitor Location of monitoring Objective Type Facility

Manual Manual sampling & Analysing to confirm that the Manual sampling analysis at LCW sample properties of processed water sampling & tank. satisfy the reuse criteria analysis. (Conductivity, pH, chloride ion - 2- (Cl ), sulphate ion (SO4 ) and TOC).

HCW Conductivity Inlet and outlet of HCW Monitoring conductivity to Continuous meter demineraliser. confirm the processing measurement. performance of the

demineraliser.

Manual Manual sampling & Analysing to confirm that the Manual sampling analysis at HCW sample properties of processed water sampling & tank. satisfy the reuse criteria analysis. (Conductivity, pH, chloride ion - 2- (Cl ), sulphate ion (SO4 )) and TOC (or Discharge Criteria).

CAD Manual Manual sampling & Analysing to confirm that the Manual sampling analysis at CAD properties of collected liquid sampling & collection tank. effluent satisfy the Discharge analysis. Criteria.

3.2.7 Argument 2g: Sizing of Tanks, Vessels and Liquid Containment Systems The design of the Wylfa Newydd Power Station includes a series of tanks to manage the liquid effluents from the various drains on systems provided across the plant. These tanks have been designed (3.2.7.1: Evidence: Capacity of Tanks and Vessels) to provide the capacity necessary to store aqueous waste for treatment and prior to discharge. For each unit, there are a separate series of tanks for each of the four dedicated treatment systems described in Section 3.2.6.2. The tanks provide sufficient capacity to accumulate aqueous waste from operational activities and expected events (as defined in the Methodology for Expected Event Selection [Ref-3]). The size of the tanks ensure that operators have enough time to undertake sampling and analysis of wastes prior to making any decisions to discharge aqueous waste to the environment or to subject it to additional treatment. All tanks are fitted with a series of alarms that indicate when the tanks contain a pre-defined volume of liquid. In addition all tanks are contained in bunds to capture any spills from over-filling (3.2.7.2 Evidence: Secondary Containment of Tanks and Vessels). The tanks sizing, based on the calculations in 3.2.7.1, are therefore considered to be appropriate for normal operations and as such are considered to be BAT. The aqueous radioactive waste tanks, vessels and containment have sufficient capacity to manage, treat and sample wastes that will be generated during normal operations and any associated expected events.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.2.7.1 Evidence: Capacity of Tanks and Vessels

The tanks and vessels provided as part of the liquid effluent management system have been sized [Ref-110] using design basis calculations to allow the collection, storage and recycling or discharge (depending on the particular system) of aqueous waste from the Wylfa Newydd Power Station under normal operating conditions. The sizing of these tanks and vessels ensures that:

• Capacity is sufficient to ensure that Horizon is not constrained by capacity when determining how to treat, re-use or dispose of liquid effluents; and

• The total site storage capacity of tanks and vessels is sufficient to retain estimated annual liquid volume production during normal operations (i.e. including volumes produced during shutdowns and other operational phases that generate aqueous waste).

In early design phase, typically the capacity of the tanks and vessels is derived from which ever calculation [(1) or (2)] derives the greater capacity, where:

(1) Is the normal quantity of waste generated over 2 days which can be stored in 1 tank; or

(2) Is the maximum quantity of waste generated in 1 day which can be stored in 2 tanks.

³ × × 1.2 × 1.1 = Tank volume m³/tank (1) Normal quantity of waste a day m 2 days 1 tank

³ × × 1.2 × 1.1 = Tank volume m³/tank (2) Maximum quantity of waste a day m 1 day 2 tank

The calculated capacity of each tank and vessel is subsequently provided with a process margin (1.2) and free board margin (1.1) to ensure a conservative tank volume and to ensure that sufficient space is available within the tanks and vessels to allow for alarms and level detection as required. The capacity of the tanks is derived from the maximum volume that can be generated by the feed process plus a process margin and a free board margin. The process margin ensures that the capacity of the tank is conservative and allows some flexibility to Horizon. The free board margin ensures detection systems and alarms can be appropriately installed and that they will allow sufficient response times for operators without a potential loss of containment. Further detail on the process margin and free board margin is provided in [Ref-111]. The liquid effluent management system for each of the individual waste streams each have two sample tanks which allows capacity for one to be on duty whilst the other is being sentenced [Ref-67].

3.2.7.2 Evidence: Secondary Containment of Tanks and Vessels

Secondary containment will be provided for the tanks and vessels that form part of the liquid effluent management system. The design of the secondary containment will ensure that 110% of the contents of the largest tank within the secondary containment can be retained. This is based on the Pollution Prevention Guidance for above ground oil storage [Ref-112]. In addition, all containment surfaces will be constructed from materials that are impermeable to the wastes they are designed to contain.

3.2.8 Argument 2h: Demineralisers for Distillates from the HCW Evaporator As described in Section 3.2.6.4 (Evidence: Key Parameters and Water Balance) the normal operating mode of the Wylfa Newydd Power Station is to reuse treated HCW. However, when managing the water balance of the Power Station Horizon operators may be required to dispose of treated HCW to the environment. The evaporator in the HCW is effective at concentrating and containing the majority of the radioactivity from the HCW liquid into a solid

Wylfa Newydd Project – Best Available Techniques (BAT) Case

form that can be disposed of, therefore minimising the radioactivity of liquid effluent that may potentially be discharged to the environment. However, some volatile radionuclides are carried over from the evaporator to the distillate and require additional treatment (polishing) to further minimise radioactivity before the waste is potentially discharged to the environment. The HCW includes a demineraliser, similar to those employed throughout the nuclear industry (3.1.8.5 Evidence: Nuclear Industry Application – Demineralisers), to remove soluble radionuclides from liquid processes and to allow the radioactivity to be managed as a solid waste (3.2.8.1 Evidence: Configuration of the Demineraliser Provided in the HCW). Filtration is unnecessary as the evaporator retains solid matter in the concentrate. The demineraliser is capable of using a variety of resins (3.1.8.6 Evidence: Demineraliser Media) which allows Horizon to make its selection based on operating requirements, compatibility with subsequent disposability requirements and any prevailing regulatory requirements. This flexibility is considered to represent BAT. The demineraliser will ensure that the concentration of FPs and activation products will be low in the event that the HCW is required to be disposed of to the environment. In-process monitoring will be undertaken to demonstrate that the demineralisers are working as expected and will alert operators if they are not. Treated HCW will, from time to time, be disposed of to the environment. The treatment involves evaporation to remove solids and demineralisation to remove volatile radionuclides. Collectively, these treatments reduce the amount of radioactivity discharged from Wylfa Newydd to the environment.

3.2.8.1 Evidence: Configuration of the Demineraliser Provided in the HCW

The demineraliser on the HCW is used to remove soluble radionuclides from liquids and to allow the radioactivity to be managed as a solid waste. Details of the demineraliser are presented in Table 3.2.8.1-1.

Table 3.2.8.1-1: Description of the HCW Demineralisers

Number of Type of System Demineralisers Demineraliser

Mixed bed bead HCW 1 type ion exchanger

Following treatment by the evaporator, the radioactivity of the distillate is low and as a result the radioactivity of the spent demineraliser resin is characterised as Low Level Waste (LLW). The radioactivity of the spent resin is shown in Table 3.2.8.1-2 below. The resin of the HCW demineraliser is exchanged when the conductivity at the outlet of the HCW demineraliser exceeds the CST collection criteria of 1μS/cm.

Table 3.2.8.1-2 Radioactivity of Spent Resins from the HCW

Radioactivity (Bq/g) Source of spent resin (see Section 5 of EP- RSR application)

HCW spent resin 6.83E+00

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.2.9 Argument 2i: Evaporation of HCW Liquid effluent collected in the chemical drain contains substances with properties that interfere with waste treatment facilities and can cause corrosion of process equipment. This waste is referred to as HCW and is segregated from the remainder of the process wastes which are referred to as LCW. The impurities in HCW must be removed before the water can be treated by demineralisation and returned to the process or disposed of to the environment. The design of the Wylfa Newydd Power Station includes an evaporator to treat the HCW to remove the impurities. The evaporator has been designed to ensure discharges are minimised as far as reasonably practicable and are standard components in ABWRs operating in Japan (3.2.9.1: Evidence: Nuclear Industry Application - Evaporators) and (3.2.9.2: Evidence: Configuration of Evaporation System). Sampling and monitoring is carried out within the HCW in order to demonstrate the performance of the evaporator and demineralisers and to ensure treatment meets the reuse or Discharge Criteria (3.2.9.3 Evidence: HCW Sampling and Monitoring). The design of the evaporator has undergone a number of improvements to account for the increase in demand resulting from inclusion of floor drains in the HCW stream (3.2.9.4 Evidence: Evaporation System - Design Improvements). These improvements have further reduced the corrosion potential of the re-circulated HCW and reduced the accumulation of scale that could impact on the performance of the evaporator. Use of the evaporator allows a high proportion of the HCW to be returned to the process. All liquid effluent from the evaporator is passed through demineralisers that remove any soluble/volatile radionuclides that are carried over (3.2.8 Argument 2h: Demineralisers for Distillates from the HCW Evaporator). The residues from the evaporator which contain the majority of the radioactivity are converted to and disposed of as solid radioactive waste. HCW contains chemical impurities that could damage the reactor and associated process equipment. The evaporator allows these chemical impurities to be removed so that the HCW can be reused in the process. This means that HCW will only be disposed of to maintain the water balance of the plant and that discharges of aqueous radioactive waste to the environment from Wylfa Newydd will be minimised.

3.2.9.1 Evidence: Nuclear Industry Application - Evaporators

Evaporators are used for the treatment of radioactive liquid effluents; they are widely used throughout the nuclear industry and are standard practice for BWRs, ABWRs and PWRs. The Wylfa Newydd Power Station proposes to make use of evaporation to minimise the discharge of radioactive liquids from the HCW. Evaporation of liquid effluents results in the production of a sludge-like concentrate that will contain the bulk of the radioactivity from the HCW.

Evaporation leads to significant volume reductions compared with other techniques. Depending on the chemical composition of the liquid effluent and the evaporator type, a DF of between 1,000 and 10,000 is typically achieved. The decontaminated water is collected in the vapour phase whilst most of the radioactivity is retained as a solid (non-volatile components such as salts containing most radionuclides) at the bottom of the condenser.

The IAEA report on the Handling and Processing of Radioactive Waste from Nuclear Applications [Ref-113] describes how evaporation is widely used in the nuclear industry as an effective method for chemical and radiological purification of liquid effluent. An Oslo and Paris Convention on Protection of the Marine Environment

Wylfa Newydd Project – Best Available Techniques (BAT) Case

of the North East Atlantic (OSPAR) report [Ref-114] found that amongst the Contracting Parties of the OSPAR agreement, evaporation is a widely used technique in the treatment of liquid discharges arising from nuclear power generation.

During GDA Hitachi-GE undertook a study to determine which treatment technology to use on the HCW [Ref-68]. The study assessed a number of different treatment techniques; evaporation, reverse osmosis membrane, ion exchange, ultra-filtration and micro filtration and gave each technique a score for a range of criteria including OPEX, reliability, maintainability, solid waste generation, DF and cost. Overall, evaporation scored highest in the study. The evaporation technique outperformed the other techniques in important areas including:

• The highest DF of 103 to 104;

• The lowest impact on solid waste generation due to the evaporator’s high volume reduction performance; and

• Reliability and safety due to the evaporator’s extensive operating performance.

The evaporation technique performed less well in terms of capital cost and layout impact (size of installation) however these were deemed less significant compared to the main benefits identified above.

3.2.9.2 Evidence: Configuration of Evaporation System

The evaporator is part of the HCW and receives waste liquid from the HCW collection tank which has undergone pH neutralisation. The forced circulation evaporator provided within the Wylfa Newydd Power Station design has appropriate capacity. A diagram of the HCW showing the evaporator is provided in Figure 3.2.9.2-1. Following treatment in the evaporator the treated liquid is collected in the HCW distillate tank before further treatment in the HCW demineraliser.

Concentrated HCW from the bottom drain of the evaporator will be transferred into the concentrated waste tank for a defined period of time as set out in the controlling Operating Instructions to allow decay of short half-life radionuclides.

Concentrated HCW is then transferred for treatment as identified within the waste processing BAT options assessment study [Ref-115]. The properties of the evaporator concentrate residue are described in the Hitachi- GE Radioactive Waste Management Arrangements document [Ref-116]. A study was conducted to identify the optimal conditioning and disposal route [Ref-115] for this waste stream. Further detail on this is provided in Section 3.3.4.1 (Evidence: Solidification of Concentrated HCW).

Figure 3.2.9.2-1 The HCW Sampling Points

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Table 3.2.9.2-1 below shows the level of decontamination achieved across the HCW [Ref-117].

Table 3.2.9.2-1 Activity across the HCW (Source Term Topic Report [Ref-117]

HCW Collection HCW Distilled Water HCW Sample Tank Nuclides Tank (Bq/cm3) Tank (Bq/cm3) (Bq/cm3)

FPs and CPs (excluding H-3) 1.9E+01 1.9E-01 1.9E-02

3.2.9.3 Evidence: HCW Sampling and Monitoring

The HCW includes a number of sampling and monitoring points to confirm the performance of the system and to ensure the criteria for reuse or discharge is met [Ref-119]. The locations of these monitoring points are shown in Figure 3.2.9.3-1. The purpose of the sampling points is described in Table 3.2.9.3-1; further detail of sampling can be found in the approach to sampling and monitoring report [Ref-39]. CE : Conductivity F : Flow rate Evaporator SP : Sampling (manual) Demister (5) PSP : Proportion Sampler RE : Radiation Monitor Sump Demineraliser SP pump Condenser CST F Distillate Sample (reuse) Collection CE (6) Tank Tank Tank (3) Release SP SP Concentrated CE CE F RE to canal (1) (2) Waste Tank (4) (4) Bead Resin Storage Tank PSP

Figure 3.2.9.3-1 Outline of HCW and Monitoring/Sampling Points

Table 3.2.9.3-1 Items and Purpose for Each Monitoring and Sampling Point

Monitoring / Measured Items Purpose of Measurement Sampling No

(1) Conductivity (Manual sampling & analysis) Chloride ion To confirm the properties of liquid effluent supplied to the evaporator. Suspended solid

Radioactivity pH, etc.

(2) Suspended solid (Manual sampling & analysis) Chloride ion, etc. To manage the concentration of concentrated liquid effluent in the evaporator, these items are measured periodically.

(3) Conductivity Monitoring of the carryover from the evaporator to confirm whether resultant liquid effluent (distilled water) is suitable for demineralisation process.

(4) Conductivity (Inlet/Outlet) To confirm the processing performance of demineraliser.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Monitoring / Measured Items Purpose of Measurement Sampling No

(5) Reuse criteria or (Manual sampling & analysis) Discharge Criteria To confirm that the properties of processed water satisfy reuse criteria (or Discharge Criteria).

(6) Radiation During the discharge, radiation levels are continuously monitored and if the monitor signal exceeds the pre-set value, discharge automatically stops. (*1)

(*1) Before the discharge, manual sampling and activity check is performed at the sampling tank in accordance with the Radiation Protection procedure for Collection and Monitoring of Samples including Discharge Collection Tanks.

3.2.9.4 Evidence: Evaporation System - Design Improvements

A number of design improvements have been introduced into the design of the ABWR that have resulted in a reduction in the amount of radioactivity that is discharged into the environment. The improvements include:

Volume reduction of floor drain waste (reuse of distillate water) by evaporation [Ref-103]

In early generations of BWRs floor drain waste was disposed of to the environment following filtration. This resulted in the dominant discharge of radioactivity to the environment originating from the floor drains (approximately 90%). Design improvements have resulted in the floor drain being treated by the HCW where evaporation and demineralisation processes are applied to treat it. This has resulted in a number of benefits which include a significant reduction in the radioactivity discharged to the environment and an increase in the volume of water that is available for reuse within the Power Station. Treatment and reuse of the floor drain resulted in an approximate reduction in radioactive discharges to 1/10 [Ref-62].

Figure 3.2.9.4-1 Released Radioactivity (except tritium) from Liquid Effluent Management System to Environment (A Plant and B plant in Japan)

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Adoption of a forced circulation evaporator

As a result of combining the floor drain waste and HCW, a more durable and reliable evaporator was required to meet the requirements of the combined waste stream. The original evaporation system was a single barrel or natural circulation multi-barrel type. This was not sufficiently robust due to the:

• Occurrence of pitting corrosion and crevice corrosion; and

• Occurrence of blocking scale in the heat exchanger tube.

To prevent blocking by scale, a forced circulation (by pump) evaporation system was developed by Hitachi-GE. This type of evaporator increases the velocity in the heat exchanger which reduces the occurrence of blocking scale. This improved evaporation system has been adopted in the Standard ABWR design and will be provided within the Wylfa Newydd Power Station. The design changes that lead to the improvements stated above are illustrated in Figure 3.2.9.4-1.

Figure 3.2.9.4-1 Illustration of Evaporator Design Improvements pH Adjustment [Ref-82]

The capability to adjust the pH of waste entering the evaporator has been provided within the ABWR design; this system enables the corrosion potential of the waste to be reduced.

3.2.10 Argument 2j: Radioactive Decay of Solid and Liquid Wastes Over time the amount of radioactivity associated with all radioactive wastes will reduce as a result of radioactive decay. The rate at which the radioactivity reduces depends on the 'half- life' which is different for each of the radionuclides present within the waste streams. Allowing solid and aqueous radioactive waste to undergo radioactive decay before disposing of it to the environment or another premises will reduce the amount of radioactivity that is disposed

Wylfa Newydd Project – Best Available Techniques (BAT) Case

of in the waste. The reduction of radioactivity will be a function of the half-life of the radionuclides and the length of time over which the waste is stored (3.2.10.1 Evidence: Nuclear Industry Application - Decay Storage). The design of the Wylfa Newydd Power Station includes a number of features that allow solid and aqueous radioactive wastes to undergo decay prior to disposal (3.2.10.2 Evidence: Storage of Solid HLW Activated Metal), (3.2.10.3 Evidence: Option of Radioactive Decay of Sludge and Spent Resin Waste) and (3.2.10.4 Evidence: Decay Storage of Concentrated Liquid Waste). However, it was also recognised that in some instances, this is not feasible (3.2.10.3 Evidence: Option of Radioactive Decay of Sludge and Spent Resin Waste). During 60 years of operation the RPV and a number of reactor internals become activated and are categorised as Intermediate Level Waste (ILW) during decommissioning. Utilisation of radioactive decay is applied to ILW during decommissioning in order to reduce its radioactivity (3.2.10.5 Evidence: Decay storage of HAW during decommissioning). The period of time that an operator stores a waste and the decision to apply pre-treatment techniques prior to storage will need to be balanced against the key attributes such as the requirement to reduce the hazard presented by the form of the waste. Any decay that can be achieved prior to disposal will contribute to reducing the radioactivity that requires disposal and as such contributes to the application of BAT. The radioactivity associated with any solid or aqueous radioactive wastes that are stored at Wylfa Newydd will reduce over time as a result of ‘radioactive decay’. Where practicable, Horizon will take advantage of this natural process particularly for wastes that contain radionuclides with short half-lives. Horizon recognises that its plans will have to balance nuclear safety requirements with any environmental benefits associated with decay storage.

3.2.10.1 Evidence: Nuclear Industry Application - Decay Storage

Decay storage is a recognised practice in the nuclear industry. The OECD report on Effluent Release Options from Nuclear Installations [Ref-10] states that wastes should be capable of safe storage prior to final disposal in a repository in support of optimisation of waste disposals.

The Department for Environment, Food and Rural Affairs’ (DEFRA’s) Radioactive Waste Management Advisory Committee [Ref-121] also stated that safe storage means the safe containment and storage of a package of waste, possibly for several decades, before its final disposal in a repository.

The Nuclear Industry Safety Directors Forum, in its report on BAT for the Management of the Generation and Disposal of Radioactive Wastes [Ref-8], detailed how the adoption of the waste hierarchy is embedded in UK policy for the management of solid, liquid or gaseous radioactive wastes. One of the principles of the waste hierarchy, in relation to decay storage, is minimising quantities of waste requiring disposal by adoption of decay storage.

Environment Agency guidance [Ref-88] states that for “nuclides with short half-lives that decay to stable (or less hazardous) nuclides, storage prior to discharge represents an option for abatement”. The process of decay storage reduces the activity and therefore minimises the amount of radioactivity discharged to the environment. The guidance also noted that “For radioactive decay to reduce the inventory of any nuclide to 10% of its initial value requires a storage period of between three and four half-lives. This delay option is therefore viable only for shorter lived nuclides”.

3.2.10.2 Evidence: Storage of Solid HLW Activated Metal

The solid HLW activated metal wastes comprise [Page 20 of [Ref-122]]:

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• Hafnium Control rods,

• Boron Carbide Control Rods,

• Local Power Range Neutron Monitor,

• Start-up Range Neutron Monitor,

• Traversing in-core probe; and

• Neutron Source Unit.

It has been assumed that this waste will be stored initially in the SFP for a minimum of 5 years following removal from the reactor due to the thermal loading of the waste and the campaign cost benefit requirements [Page 20 of [Ref-122]].

Following this initial storage period, the waste will then be removed from the SFP and placed into Non-fuel Waste Containers (NFWC). The NFWCs will then be placed into overpacks and stored in the sites combined SF Storage Facility pending final disposal in a GDF [Ref-123].

3.2.10.3 Evidence: Option of Radioactive Decay of Sludge and Spent Resin Waste

LLW Resins

It was recognised that whilst there was an opportunity to dry LLW resins and subject them to decay storage to permit a change in classification, decay storage timescales would be in the order of 15-20 years and would therefore require the development of a suitably sized buffer store. This would increase the potential doses to operators and is inconsistent with Licence Condition 32 (Accumulation of Radioactive Waste), as Horizon would be instigating the long-term storage of disposable radioactive waste onsite. Decay storage to aid reclassification of the waste was therefore dismissed [Ref-122].

LLW bead resins will however be subject to decay storage within the Rw/B for a shorter, 8 year period. This is because when produced, these resins will represent the upper limits of LLW with regards to radioactivity [Ref-102] and review of the source term and supporting dose modelling has indicated that the contact dose rates of packages containing these wastes will exceed transport limits. It was therefore decided that these resins will be subject to this decay storage period, not to reduce their classification, but to reduce the activity levels such that the contact dose rates from the resultant waste packages are sufficiently low to allow transportation [Ref-102]. Although this decision is the result of an ALARP argument, there is also an environmental benefit to this as this will reduce the amount of radioactivity disposed of.

ILW Resins

It was also recognised that there was an opportunity to subject ILW resins to decay storage but this was also dismissed after an options assessment identified that decay storage would not result in a reclassification to LLW [Ref-122].

However, due to the decision to increase the time between batch processing of WSILW (see 3.4.1.3 Evidence: Waste processing and Packaging Facilities), and the subsequent decision to defer the construction of the ILW Storage Facility (See 3.4.1.4 Evidence: Waste Storage Capacity) the raw waste storage capacity within the Rw/B has been increased to allow up to 10 years accumulation. As such, whilst not strictly decay storage, these resins will be subjected to up to 10 years decay which will allow the decay of short lived radionuclides so will contribute to reducing the radioactivity of the waste.

3.2.10.4 Evidence: Decay Storage of Concentrated Liquid Waste

The concentrated liquid waste from the evaporator in the HCW is collected in concentrated waste tanks and can be stored for one year to reduce the radioactivity by decay. After this period of decay, the concentrated waste is mixed with a solidifying agent and placed into drums.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Halogenous nuclides are expected to reduce significantly in radioactivity over the decay period of several months. FPs, other than halogenous nuclides, have much longer half-lives and as such will not significantly decay over short time periods (e.g. months or even years).

Taking into consideration the above, the storage period in the tank of the concentrated waste has been determined as one year to allow halogenous nuclides to decay. The relationship between the decay period and the radioactivity concentration in the concentrated waste tank based on the Japanese BWR plant is shown in Figure 3.2.10.4-1. Actual precise storage capacity will be decided during detailed design phase.

Figure 3.2.10.4-1 Relationship between Decay Period and Radioactivity

3.2.10.5 Evidence: Decay storage of HAW during decommissioning

As the Wylfa Newydd Power Station enters the decommissioning phase, SF will be stored in the SFP within the R/B for a period of up to 10 years before it is transferred to the SF storage facility. Following the removal of SF from the R/B, reactor internals will be dismantled and treated as solid ILW which will then be stored in the ILW storage facility for a maximum period of 100 years before being transferred to the GDF for final disposal. If the waste is stored for the maximum period of 100 years a reduction in radioactivity of 82% would be achieved as shown in Table 3.2.10.5-1.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Table 3.2.10.5-1 Decay in storage of ILW solid waste

Activation Radioactivity (Bq/tonne) Weight Products (tonnes) 6 years 100 years

Reactor Core shroud, Top guide, 220 5.9E+15 1.1E+15 internals Core plate, etc.

RPV RPV, RPV cladding 580 5.1E+10 1.1E+9

Total 800 1.6E+15 3.1E+14

Reduction rate (100%) 82%

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.3 Claim 3 - Minimise the Volume of Radioactive Waste Disposed of to Other Premises The Arguments presented in support of this Claim are considered to demonstrate compliance with the standard BAT conditions [Ref-5]: • Condition 2.3.2(b) ‘The operator shall use the best available techniques in respect of the disposal of radioactive waste pursuant to the permit to minimise the volume of radioactive waste disposed of by transfer to other premises.’ The Wylfa Newydd Power Station design contains a range of features that contribute to the substantiation of this Claim including: • Design changes that will minimise the volumes of operational and decommissioning waste arisings. • Provision of a number of features that will allow Horizon to adopt an operating philosophy that will minimise the quantity of solid radioactive waste associated with routine operations and maintenance. • Provision of dedicated facilities for the management, treatment and storage of solid radioactive waste. • Reducing the quantity of solidified HCW that are generated. • Availability of a range of decontamination techniques during decommissioning. • The use of the clearance and exemption process to remove wastes from radioactive waste controls. In developing the Arguments presented to demonstrate the validity of Claim 3 the REPs [Ref- 7] have been taken into account. The following REP is considered to be specifically relevant to this Claim: • Principle RSMDP3 ‘the best available techniques should be used to ensure that production of radioactive waste is prevented and where that is not practicable minimised with regard to activity and quantity.’

3.3.1 Argument 3a: Design to Minimise the Volumes of Operational and Decommissioning Waste Arisings The operation, maintenance and subsequent decommissioning of the Wylfa Newydd Power Station will generate solid radioactive waste that will require management and treatment before it is consigned for either disposal to other premises or storage on-site pending future off-site disposal. The design of the Wylfa Newydd Power Station has evolved to reduce the quantities of solid radioactive waste that will be generated during its life-cycle and to ensure that those wastes that are unavoidably created are compatible with waste management techniques typically used in the UK. The design changes that have had the greatest impact on the volume of solid waste generated are: • RIPs are used to circulate the coolant around the reactor in the Wylfa Newydd Power Station. These internal pumps have removed the pipe work and associated plant used in previous evolutions to pump the coolant and will contribute to a reduction in the quantity of waste generated from maintenance and decommissioning operations. RIPs also have the added benefits of reducing occupational exposure to workers, decreasing the amount of power required to

Wylfa Newydd Project – Best Available Techniques (BAT) Case

recirculate the coolant and reducing the cost of vessels and pipe work (3.3.1.1. Evidence: Introduction of Reactor Internal Pumps) compared to the previous design. • Evolution of the BWR design replacing the PCV comprising a free standing Steel Containment Vessel (SCV) and separate Biological Shield Wall (BSW) with the Reinforced Concrete Containment Vessel (RCCV) with integrated steel liner of the ABWRs. This design improvement delivers the same functionality whilst reducing the volume of metal and concrete that will require disposal at decommissioning (3.3.1.2 Evidence: Evolution of the Primary Containment Vessel). • The Wylfa Newydd Power Station’s PCV is smaller than those used in previous evolutions of the BWR. This has the effect of reducing the size of the RCA by approximately 15% which will reduce the quantity of waste generated from general occupancy, housekeeping and decommissioning (3.3.1.2 Evidence: Evolution of the Primary Containment Vessel). • Introduction of techniques that reduce the amount of SCC experienced on reactor components. These techniques contribute to a reduction in the frequency that components require replacement as a result of corrosion. Consequently, this will result in a reduction in the quantity of waste associated with the replacement of damaged components and any related maintenance activities (3.3.1.4 Evidence: Techniques to Reduce Stress Corrosion Cracking). • The introduction of Hollow Fibre Filter (HFF) or pleated filters that have eliminated the powder resins that were generated on previous evolutions of the BWR from the use of pre-coat filters. The use of HFF’s or pleated filters is expected to reduce the quantity of solid radioactive waste by 27 tonnes per year (3.3.1.5 Evidence: Replacement of Pre-coated Filters). • The introduction of a range of design changes that contribute to a reduction in the volume of operational and decommissioning waste (3.3.1.6 Evidence: Review of Further Design Changes). • Where possible, embedded pipework has been minimised. This will contribute to reducing the potential volume of radioactive concrete waste arisings during decommissioning (3.3.1.7 Evidence: Reduction in the use of Embedded Pipework). • As discussed in Section 3.4.1 (Argument 4a: Provision of Solid Waste Management Facilities) the Rw/B and ILW Storage Facility, facilities that are common to both units, have been combined into single centralised facilities servicing both units which will reduce radioactive decommissioning arisings. The above improvements have been developed and implemented as a result of Hitachi-GE's on-going commitment to improving performance. Hitachi-GE has a comprehensive research and development programme that explores opportunities to reduce the materials used during construction and operations. In contrast to this the baseline design of the TGS system comprises of a number of unit operations that will require maintenance, periodic replacement and end of life disposal during decommissioning. The baseline TGS system design therefore contributes to the generation of solid radioactive waste.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Evolution of the UK ABWR has resulted in rationalisation/improvement of process equipment and reduction of the size (and associated materials of construction) of key structures. These features that will significantly reduce the quantity of solid radioactive waste that will require disposal from Wylfa Newydd during operations and, in particular, during decommissioning.

3.3.1.1 Evidence: Introduction of Reactor Internal Pumps

The method of re-circulating water in the reactor has undergone a number of evolutionary changes since the BWR concept was originally conceived. Figure 3.3.1.1-1 provides a summary of the evolutionary steps [Ref-124], which have progressively reduced the amount of process equipment, pipework and materials associated with the reactor.

Figure 3.3.1.1-1 Evolution of Water Coolant Re-circulating Designs

The most recent evolutionary change, from the latest BWR to the standard ABWR, is the introduction of RIPs. The RIPs are located within the reactor and remove the requirement for re-circulation pipe work, valves and pumps located external to the reactor that are used in previous generations of BWR. The use of RIPs provides the following benefits:

Reduction in decommissioning waste

The elimination of external pipes and valves contributes to a reduction in the volume of radioactive waste generated at decommissioning. Table 3.3.1.1-1 shows the weight of Reactor Recirculation System (RRS) main components in the BWR and ABWR. In the ABWR, most of the major pipe work with respect to RRS is installed in the PCV. As a result, the weight of RRS piping and valves has been reduced by approximately 40 tonnes and 20 tonnes, respectively. The weight of the RRS pumps has increased slightly in the ABWR compared to the BWR as the ABWR has ten RIPs compared to the BWR which has two Primary Loop Recirculation System (PLR) Pumps. However, overall, rationalising RRS from PLR Pumps in the BWR to RIPs in the ABWR has reduced the weight of pumps, piping and valves by 50 tonnes.

Table 3.3.1.1-1 Weight of RRS Main Components in BWR and ABWR Based on Hitachi-GE Experience

Piping Valve Pump Total [tonnes] [tonnes] [tonnes] [tonnes]

Power Station 40 20 40 100 A(BWR)

Power Station 0 0 50 50 B(ABWR)

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Piping Valve Pump Total [tonnes] [tonnes] [tonnes] [tonnes]

Difference -40 -20 10 -50

Reduction in the size of the PCV

Removing components external to the reactor has reduced the space required and subsequently contributed to the size of the PCV being reduced (Figure 3.3.1.3-1).

Decrease in Operational Exposure to Workers

The elimination of the external recirculation piping and valves means that there is no requirement for inspections during operation and lower radiation in work areas around the reactor vessel. The need to maintain an external recirculation system has also been removed which further reduces occupational exposures.

Improvement in safety

The internal recirculation system means that an external pipe break with an associated loss of coolant is no longer possible. In the event of an accident the reactor core will remain covered in water. This extends the time available to initiate a response from other safety related systems installed to control abnormal reactor conditions.

Reduction in power consumption

Rationalisation of piping, valves and jet pumps has reduced the amount of power required to drive the recirculation system.

3.3.1.2 Evidence: Evolution of the Primary Containment Vessel

The ABWR PCV incorporates a number of improvements over the BWR-5 design that will result in a reduction in the volume of radioactive waste requiring disposal. Table 3.3.1.2-1 provides details on how the functionality of the PCV has changed.

Table 3.3.1.2-1 Optimisation of PCV Primary Functions

BWR-5 ABWR

Function PCV BSW RCCV Liner RCCV

Withstand pressure x x

Leak prevention x x

Shielding x x

The PCV of the BWR-5 utilises a free-standing SCV that delivers both the withstand pressure and leak prevention function. A separate BSW is provided within the R/B to meet the shielding requirements. The ABWR design utilises a RCCV with integrated internal RCCV liner; this configuration delivers all three functions [Ref-125].

The evolution of the PCV has resulted in the following benefits:

• Reduction in the mass of steel that will require disposal at decommissioning by approximately 40% (Figure 3.3.1.2-1);

• Reduction in the R/B volume (Figure 3.3.1.2-1); and

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• Reduction in the volume of concrete that will require disposal at decommissioning by integrating the RCCV and RCCV liner.

Figure 3.3.1.2-1 Comparison of BWR-5 PCV and ABWR RCCV

PCV (metal component): Approx. 3850 tonnes RCCV (metal component): Approx. 2290 tonnes

Description: PCV main body, hatch, penetration Description: RCCV liner, hatch, penetration portion, portion, diaphragm floor, vent piping, γ-shield, diaphragm floor, vent piping, γ-shield, RPV pedestal etc. RPV pedestal etc.

3.3.1.3 Evidence: Design of the Primary Containment Vessel

Figure 3.3.1.3-1 and Figure 3.3.1.3-2 illustrate the volume of the PCV in the BWR and the ABWR respectively. Compared with the volume of the BWR-5, the ABWR reduced the size of RCAs by approximately 15% as shown by Calculation 3.3.1.3-1 [Ref-126].

Percentage volume reduction = (1- ) ∙ 100 Volume of PCV in ABWR , ³ Volume of PCV in BWR Percentage volume reduction = (1- , ³) ∙ 100 18 460m Percentage volume reduction = 14.5%21 600m

Calculation 3.3.1.3-1: Percentage Volume Reduction of PCV

The reduction in the size of the PCV has been achieved as a result of two improvements:

• Introduction of RIPs. As discussed in Section 3.3.1.1 (Evidence: Introduction of Reactor Internal Pumps) eliminating the large bore pipe work associated with the coolant recirculation system allows the RPV to be located at a lower position within the PCV, and thereby enable the height of the PCV to be reduced.

• Reinforced concrete containment vessel. Replacement of the SCV used in the BWR with a RCCV in the standard ABWR. As shown in Figure 3.3.1.3-1 and Figure 3.3.1.3-2, the RCCV is cylindrical in shape, whilst the SCV has a circular cone structure. By adopting a cylindrical structure, the volume of the PCV has been reduced whilst the layout of equipment has been improved.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.3.1.3-1 The Configuration of the PCV in BWR-5

Figure 3.3.1.3-2: Configuration of the PCV in ABWR

3.3.1.4 Evidence: Techniques to Reduce Stress Corrosion Cracking

SCC results in the formation of cracks that if unidentified have the potential to result in sudden failure of an affected part. Mitigation of SCC focuses on material selection, improvements to the fabrication process (such as welding techniques) and the use of techniques to improve the operating environment, which is discussed in this section.

During normal operation of the reactor, water radiolysis results in the formation of oxygen and hydrogen peroxide. The presence of oxygen and hydrogen peroxide increases the ECP of the water in the reactor which contributes

Wylfa Newydd Project – Best Available Techniques (BAT) Case

to increased occurrences of SCC. Hitachi-GE has adopted HWC and NMCA (See Argument 1f) which results in a reduction in the ECP as a result of the recombination reaction shown below [Ref-127].

2 + 2

𝐻𝐻 2+ 𝑂𝑂2 → 2𝐻𝐻2𝑂𝑂

A consequence of injecting hydrogen into the feedwater𝐻𝐻2 𝐻𝐻2𝑂𝑂 is2 an→ increase𝐻𝐻2𝑂𝑂 in the amount of nitrogen-16 that is produced. Nitrogen-16 has a very short half-life and as such does not contribute to the gaseous radioactive discharges from the Power Station into the environment. However, nitrogen-16 does contribute to worker dose and as such, Hitachi-GE has introduced NMCA which allows less hydrogen to be injected resulting in a reduction in concentrations of nitrogen-16 to levels equivalent to those achieved prior to the introduction of hydrogen injection. Further detail is provided in Section 3.1.6.1 (Evidence: Hydrogen Water Chemistry with Noble Metal Chemical Addition).

Monitoring of the ECP enables the performance of the combined hydrogen injection and NMCA system to be optimised based on operational performance data.

Avoiding SCC reduces the requirement for maintenance and the replacement of reactor component parts. These activities would result in the generation of significant volumes of primary and secondary radioactive waste.

3.3.1.5 Evidence: Replacement of Pre-coated Filters

Filters are provided within the treatment systems for both condensate and LCW to remove iron crud (Claim 1). Early generations of BWR used pre-coat filters to perform this function. Pre-coat filters are coated in a resin that requires disposal as ILW once it is spent. A design change introduced the improvement of replacing pre-coat filters in the condensate and LCW with HFF or pleated filters which are not coated with a resin [Ref-62]. The introduction of alternative filter types has resulted in the elimination of the associated ILW spent powder resin waste stream. HFF and pleated filters also provide an increase in performance, improving the water quality. Design basis calculations have been undertaken to determine the volume of solid waste that has been avoided by the replacement of pre-coated filters (Table 3.3.1.5-1).

Table 3.3.1.5-1: Impact of Replacing Pre-coat Filters with HFF’s for the CF

Criteria Pre-coated filter HFF

Inlet App. 8 to 10 ppb Fe crud Water quality concentration Outlet Approx. 0.5 to 1.5 ppb Approx. 0.1 ppb

≦ Spent resin volume 0kg/Year Waste Approx. 30,000kg/Year reduction (including crud) Crud volume Approx. 100kg/Year

3.3.1.6 Evidence: Review of Further Design Changes

Hitachi-GE has identified a number of design changes that also contribute to a reduction in the generation of radioactive waste [Ref-124]. A summary of these changes has been provided in Table 3.3.1.6-1.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Table 3.3.1.6-1: Summary of Design Changes that Contribute to Minimising Volume of Radioactive Waste

Items Description

Introduction of larger capacity Safety Relief Valve The number of SRVs was decreased from 18 to 16 (SRV) valves by increasing the capacity of each SRV thus reducing the amount of maintenance and decommissioning waste.

Elimination of CUW motor operated valves Removal of two motor-operated valves connected between the CUW and the feedwater lines.

Elimination of the third Main Steam Isolation Valve The third MSIV was eliminated from the design by reducing the leak rate of MSIVs.

Elimination of pumps in the RHR system The method for sealing RHR injection pipes with water was rationalised by using Make-up Water Condensate System (MUWC) pumps.

Elimination of S/P drain line for warm-up water The main drain line for warm-up water is rationalised. This allowed for the elimination of the drain pipes connected to S/P.

Reduction of the number of valves monitored by the The valves to be monitored were reduced based on LDS previous designs so that fewer valves are monitored.

Installation of pumps on points with lower temperature Warm up pipes are eliminated by changing the composition of the CUW.

Adoption of boron racks in the fuel pool Adoption of boron racks allowed for more efficient storage of SF which increased the capacity of the SF in the SFP. More efficient storage allowed for a reduction in the volume of the SFP area and thus the associated waste.

Elimination of interim loops in the R/B Cooling Water Elimination of pipework that could become plated with System (RCW) activated products and subsequently require disposal as HAW.

Reduction in the capacity of Standby Gas Treatment Reduction in the capacity of SGTS Fan due to the low System (SGTS) Fans Secondary Containment leak rate and thus less plant requiring disposal at end of life.

It is also recognised that some of the design developments have resulted in the addition of plant items which may ultimately lead to greater volumes of waste being generated during maintenance and decommissioning. For example, moisture separator re-heaters were introduced to improve the thermal efficiency of the plant but may become a waste during maintenance and decommissioning.

3.3.1.7 Evidence: Reduction in the use of Embedded Pipework

Reducing embedded pipework decreases the potential for infiltration of radioactive wastes and substances to inaccessible areas [Ref-128], which can lead to the spread of contamination and the generation of radioactive

Wylfa Newydd Project – Best Available Techniques (BAT) Case

concrete waste arisings at decommissioning. Leaks from embedded pipework can also be difficult to detect and result in doses to workers during subsequent decommissioning [Ref-26].

During GDA, Hitachi-GE received a Regulatory Observation on the use of embedded pipework which prompted them to perform an optioneering study [Ref-129] to reduce its use. The study explored alternative options to embedded pipework and Hitachi-GE are re-designing several systems in response, the result will be a reduced amount of embedded pipework used in the Wylfa Newydd Power Station [Ref-26]. This will therefore contribute to reducing the potential volume of radioactive concrete waste arisings during decommissioning.

3.3.2 Argument 3b: Selection of Methods to Minimise Solid Waste Generation The methods adopted by Horizon for operations and maintenance will influence the quantity of solid radioactive wastes requiring treatment, storage and disposal. The design of the Wylfa Newydd Power Station includes a number of features that will allow Horizon to adopt an operating philosophy that will minimise the quantity of solid radioactive waste associated with routine operations and maintenance: • Space is provided at key work locations within the designated areas to allow operators and maintainers to segregate wastes depending on their physical, chemical, radiological and biological properties. This will ensure that wastes do not become contaminated with substances that require more robust treatment, storage or disposal options (3.3.2.1 Evidence: Segregation of Waste). • The provision of office accommodation outside of controlled areas which reduces occupancy of controlled areas and the associated generation of waste from office equipment and consumables (3.3.2.2 Evidence: Locate Offices Outside of Controlled Areas). • The provision of space to store tools, scaffolding and maintenance equipment within designated areas to minimise the amount of equipment that is routinely taken in/out of designated areas for maintenance activities and subsequently disposed of (3.3.2.3 Evidence: Storage Facilities for Tools and Other Maintenance Equipment). • Deployment of reusable protective clothing for workers and visitors in the RCA to minimise the volume of soft waste disposed of in the RCA (3.3.2.4 Evidence: Re- usable protective clothing in the RCA). The clothing will be cleaned at an offsite active laundry facility. • The adoption of a maintenance philosophy that selects and applies the optimum maintenance regime for SSC preventing the potential of generation of avoidable radioactive waste (3.3.2.5 Evidence: Maintenance Philosophy). • Specification of performance parameters for surveying filters in operation to identify when they require replacement to avoid their further performance degradation affecting their environmental protection functions (3.3.2.6 Evidence: Performance Monitoring Maintenance). The flexibility afforded by these features enables Horizon to develop an approach that is appropriate to their operational needs and the regulatory requirements in force at the time. It is considered that these features support the demonstration of BAT for minimising solid radioactive waste generation at this time. Offices, stores and maintenance areas at Wylfa Newydd are designed to prevent the unnecessary introduction of equipment, tools and consumables to areas where they

Wylfa Newydd Project – Best Available Techniques (BAT) Case

might become contaminated with radioactivity. Areas for the segregation and storage of solid radioactive waste are also provided. Horizon will develop an operating philosophy that, when combined with these design features, will minimise the quantity of radioactive waste that is created during operations.

3.3.2.1 Evidence: Segregation of Waste

The Best Available Techniques for the Management of the Generation and Disposal of Radioactive Wastes report [Ref-8] recognises that effective waste collection, segregation, processing and storage is an effective means of reducing the activity, mass or volume of waste arisings.

The design of the Wylfa Newydd Power Station enables Horizon sufficient flexibility to segregate, collect, store and process waste in a manner that allows BAT to be applied to the management and disposal of the waste. Methods for the collection, segregation, processing and storage of waste have been subject to detailed assessment which is reported in the BAT Optioneering report [Ref-115] and Radioactive Solid Waste Monitoring Requirements report [Ref-130].

The segregation of radioactive wastes is dependent upon their accurate characterisation as defined in the Radioactive Waste Characterisation Procedure. The radioactive waste streams and SF stream are described in Table 3.3.2.1-1 [Ref-116]:

Table 3.3.2.1-1: Summary of Solid Radioactive Waste and Spent Fuel Streams

No. Title Description Category Form Arising during

Miscellaneous dry, low activity Very Low Dry active wastes in various forms; Operations & 1 Level Waste Solid waste including metals, concrete Decommissioning (VLLW) cloths, paper, etc.

Arising from filter changing in air treatment facilities from exhausts from R/B, T/B Operations & 2 HVAC filters LLW Solid including high radiation), Rw/B Decommissioning S/B and radioactive waste treatment Buildings.

Arising from the HCW demineralisers; styrene Operations & 3 Bead resin LLW Wet divinylbenzene copolymer Decommissioning matrix.

Arising from the HCW/ Concentrated Waste System 4 Concentrates (CONW) evaporator, LLW Wet Operations comprises particulate and previously dissolved species.

Includes spent HFF Miscellaneous membrane, plastic sheets, Operations & 5 LLW Solid combustible paper, wood, cloth, oil, Decommissioning activated carbon from the OG.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

No. Title Description Category Form Arising during

Miscellaneous Spent HFFs , metal, pipes, Operations & 6 non- cables, lagging, gas filters, LLW Solid Decommissioning combustible concrete and glass

Arising from backwashing of 7 Sludge (crud) various filters from the CF ILW Wet Operations system and the LCW HFFs.

Arising from the CUW and FPC/FPCU FD; cross linked Operations & Powder resin ILW Wet polystyrene matrix. Contains Decommissioning particulate CP. 8 Secondary waste arising from Ion exchange system decontamination of ILW Wet Decommissioning resin RPV, RIN and closed loop systems.

Cruciform shape metallic construction containing HAW stainless steel tubes in each Higher activity wing of the cruciform filled with High Level 9 metals – Solid Operations boron carbide powder. Waste (HLW) control rods Hafnium is also employed to at arising, ILW perform the same function of at disposal) reactivity control.

Zircaloy box which surrounds the fuel bundle. Approx. 4.3m long and 15 × 15cm Higher activity HAW square. Channel boxes will 10 metals – Solid Operations stay with spent fuel so (Remain with Channel boxes disposal “item” is a spent fuel fuel elements) assembly (fuel bundle in a channel box).

HAW Various reactor core Higher activity 11 components from SRNM and (HLW at Solid Operations metals - others LPRM systems. arising, ILW at disposal)

Non-combustible, largely Contaminated metal and concrete items. and irradiated Some very large and requiring 12 LLW Solid Decommissioning metal and size reduction. concrete Including SFP furniture (e.g. fuel racks).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

No. Title Description Category Form Arising during

Non-combustible, largely Contaminated metal items. Some very large 13 and irradiated and requiring size reduction. ILW Solid Decommissioning metal Including SFP furniture (e.g. fuel racks).

14 SF Used fuel elements. HLW Solid Operations

Wastes that have been segregated based on the waste forms and properties displayed in Table 3.3.2.1-1, are stored, transferred and processed independently of each other, to prevent mixing and cross contamination which allows them to be effectively treated based on their physical and chemical properties in accordance with the waste hierarchy. Figure 3.3.2.1-1 shows the waste flows of wastes and indicates how each waste stream is segregated based on waste form and properties and how this then allows for each waste stream to undergo the appropriate process method before storage or disposal.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[This page is intentionally blank]

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Main Cartridge Demineraliser Condenser Filters Spent Fuel Multi Purpose Canister (MPC) in Transfer Cask CUW CF CD Non-Fuel Waste Canister (NFWC) in Transfer Cask Filter CAD Sump tank Reactor Demineraliser Packaged Spent Backwash Sludge Filter Cartridges receiver tank Control Rods Dispatch for Direct Spent Filter Media Direct Disposal Packaging for Characterisation Disposal LAW LLW & VLLW Transport (LLW/VLLW) Spent Resin/Crud Floor Drains Equipment Drains reciever tank Equipment Floor Drains Drains Reactor Components Combustible Packaging for Dispatch for Incineration Waste Transport

LCW Sump Tank SPCU Fuel Pool Spent HVAC Filters Compactable Packaging for Dispatch for LCW Sump tank CAD Sump tank Supression Pool Powder Characterisation Resin LAW Waste Transport Supercompaction Drywell LCW and Crud Sump Filter Turbine Building - 108 Demineraliser Segregated Metal Waste Heterogeneous Packaging for Dispatch for Metal Fuel Pool Characterisation (suitable for LAW Transport Recycling recycling) Drywell HCW (from all Buildings) Sump FPC

Reactor Building - 101 Potential Decontam LAW Decontam Secondary No current treatment route (oils/contaminated Spent Bead Resin Wastes land) Spent HVAC Filters If Req’d Filter Backwash Sludge

Outage Only Spent Cartridge Filter Media Out of Scope Conventional Waste Management Lower Activity Waste Management Facility - 246

Spent HVAC Filters

Supernate recycling Radioactive Effluent Management Process Wet Solid ILW Process Condensate Equipment drains Floor Drains HLW Spent Fuel Storage Tank Powder Resin Cement Powders THISO Storage Cask Storage Cask Storage Tank

LCW Spent Fuel Storage Facility - 201

LCW Collection Hollow fibre LCW Demineraliser In drum tank filters Sample tank Filter Crud ILW 3m3 Drum Final solidification Storage Tank Process Tank Waste Package in 3m3 Drum Filter Backwash Sludge

Drummed Spent 3m3 Drum Filter Media 3 (LAW) 3m Drum Final Waste Spent Bead Resin Supernate recycling Package HCW Pre-mixed Bead Resin Cement Grout & Cement Powders ILW Storage Facility - 202 HCW Collection HCW Storage Tank Evaporator Demineraliser Waste tank Sample Tank

Fill and Grouted Third CONW Storage LLW Evaporator Concentrate Sludge Solidify in Height ISO Tank Process Tank CAD Container Container

CAD Collection LCW Sump tank HCW Sump tank Sample tank Wet Solid LLW Process Tank

Radioactive Waste Building - 104

Laboratory Decontamination Discharge to Process Drains Showers Environment Key

Spent Fuel Wet Solid ILW Dry Solid LLW - HVAC Contaminated Effluent

CAD Sump tank HCW Sump tank Dry Solid ILW Wet Solid LLW Dry Solid LLW Clean or very low contamination effluent

Outage only Off-site Disposition Service Building - 109

Figure 3.3.2.1-1: Integrated Waste and Effluent Treatment System

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[This page is intentionally blank]

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.3.2.2 Evidence: Locate Offices Outside of Controlled Areas

Where possible, the Wylfa Newydd Power Station design locates office accommodation outside the RCA [Ref- 131]. This minimises the number of operators routinely accessing RCAs and the volume of office related consumables routinely taken into RCAs.

Minimising the number of operator visits into controlled areas and reducing the volume of consumables that are taken into RCAs reduces the potential to generate secondary wastes. Shadow Site Local Rules and Local Rule Notices that control RCA personnel access and egress will be enforced before active commissioning. This will provide time for the operatives to practice and for behaviours to become normalised prior to active commissioning.

3.3.2.3 Evidence: Storage Facilities for Tools and Other Maintenance Equipment

Minimising the amount of maintenance equipment, tools, spares, consumables, protective clothing and packaging that are taken into RCA is recognised as RGP within the nuclear industry. The Wylfa Newydd Power Station design will provide dedicated areas for the storage of maintenance equipment and tools. This will allow Horizon to minimise the amount of equipment and tools that require removal from RCAs following use and that have the potential to become radioactive waste. Retention of maintenance equipment and tools within RCAs after use will also alleviate the requirement to bring new/replacement equipment and tools into the RCAs.

The Radiation Protection procedure on the Control of Items Leaving a Controlled Area will be a management control for preventing unnecessary items entering the RCA. Items stored and used within the RCA could become contaminated. The Radioactive Clearance and Exemption process sets the requirements for collecting characterisation information, the acceptance criteria for clearance and the decision points for determining if an item is compliant for clearing from the RCA. The Radiation Protection procedure on the Monitoring of Items Leaving a Control Area supports the execution of the Clearance and Exemption process. The Radiation Procedure prescribes how items will be monitored by radiological protection surveyors for the purpose of collecting data on an item’s surface radioactivity.

3.3.2.4 Evidence: Re-usable protective clothing in the RCA

Horizon’s Local Rules for RCAs within the Wylfa Newydd Power Station will stipulate the use of protective outer clothing to avoid personal clothing becoming contaminated and preventing the spread of contamination beyond the RCA. The use of protective clothing in the RCA aligns with the Ionising Radiations Regulations 1999 Approved Code of Practice [Ref-132]. Personnel will wear protective clothing (hereafter to be referred to as coveralls) prior to entering the RCA and remove them upon egress from the RCA. Horizon can select to use coveralls designed to be used once or small number of times in the RCA, they are then disposed of as soft combustible LAW suitable for offsite incineration (3.3.3.2 Evidence: Incineration). An alternative option is for Horizon to deploy robust, washable coveralls for use in the RCA. At appropriate intervals the coveralls would be packed and transferred to an active laundry facility for washing and decontamination, the cleaned coveralls are then returned to Horizon. Over the lifetime of the Wylfa Newydd Power Station, the deployment of reusable coveralls will result in a lower total volume of operational soft waste generated than compared with using disposable coveralls. The practice of using disposable coveralls also represents a less preferred waste management level of the waste hierarchy.

An optioneering study was carried out to determine whether Horizon should utilise an offsite operated laundry service or design, build and operate a facility within the Wylfa Newydd Power Station. The study concluded that [Ref-101]:

• The environmental performance (based on environmental risk, benefit and impact) of an off-site active laundry facility was determined to be more favourable than an onsite active laundry facility operated by Horizon in the Power Station;

• Whilst the capital costs associated with designing, building and commissioning a dedicated on-site facility were unknown, they were thought to be considerable. No capital cost would be associated with employing an off-site facility; and

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• Operational costs associated with employing an off-site facility would be less than for employing a dedicated on-site facility.

Based on these factors, Horizon will establish a contract with an offsite active laundry service provider for its reusable coveralls [Ref-133] [FAbc-3].

3.3.2.5 Evidence: Maintenance Philosophy

The maintenance philosophy adopted for the Wylfa Newydd Power Station will result in the suitable maintenance of components that could potentially contribute to the generation of radioactive waste. Components will be maintained to prevent their unnecessary, premature replacement or to prevent the degradation of their reliability causing an increase in failure rate, resulting in full or partial replacement.

For the development phase of the project, Horizon has reported on its maintenance philosophy [Ref-134] to describe the maintenance strategies that will need to be developed at appropriate timescales in the project and how the Maintenance Schedule will evolve. The Maintenance Schedule will be a repository of maintenance requirements set on SSC and will be utilised upon handover from the EPC contractor at Hold Point Stage 5 of the project (Section 2.6 Management Arrangements) before the first delivery of fuel on site.

HMS procedures on the management of BAT SSC [Ref-20], [Ref-21] and [Ref-22] will prescribe the assessment of SSC to identify any special requirements associated with maintenance. The outputs will feed into the Maintenance Schedule and the optimum maintenance regime for SSC will be selected for maintaining the effectiveness of SSC. The different types of maintenance regime are outlined below:

• Regulated maintenance requirements – Examination, Inspection, Maintenance and Testing (EMIT) requirements for SSC fulfilling nuclear safety functions derived from the Safety Case.

• Preventive maintenance – Activities carried out to maintain a component within design operating conditions and to extend its life, including predictive (condition-based) and periodic/planned (time- based) actions.

• Life cycle management – The planned replacement or refurbishment of major assets based on service life, such as for turbines, generators, reactor internal pump motors, the emergency diesel generator system. The design of major components will factor in their life cycle and replacement prior to the end of commercial operations. The design of SSC that is redundant (on standby or backup) should facilitate on-line replacement through the provision of sufficient space to allow the replacement to be installed next to the operating one and a simple changeover mechanism to place the new one in-service.

• Emergent corrective or elective maintenance – Activities involving work on any SSC that has failed or is significantly degraded (corrective), or which has identified potential or actual degradation as minor (elective).

The maintenance philosophy for the Wylfa Newydd Power Station ensures that non-critical items that have the potential to contribute to the generation of radioactive waste will only be replaced based on performance subject to recommendations from the manufacturer.

3.3.2.6 Evidence: Performance Monitoring Maintenance

The operating performance of SSC that supports the demonstration of BAT will be monitored and if its performance is assessed as inadequate to sufficiently fulfil its environmental protection function, corrective action will be undertaken. The performance parameters of HEPA filters and FD media are based on recommendations from the manufacturer and they will be specified in the Surveillance Test Maintenance procedures that prescribe the routine inspections of plant and equipment.

Replacement of HEPA filters

The Wylfa Newydd Power Station design will allow Horizon to replace filters based on performance. Monitoring of the differential pressure across the filters will be provided to allow replacement of the filters at a pre-defined set point. This is expected to reduce the frequency of filter changes. In areas where high particulate matter © Horizon Nuclear Power Wylfa Limited 146

Wylfa Newydd Project – Best Available Techniques (BAT) Case loading is expected pre-filters will be used to minimise the potential to blind HEPA filters which will extend the operational life of the HEPA filters.

Replacement of FD media

Similar to HEPA filter replacement, the replacement of a FD element is determined by either a pressure drop higher than the specified value across the unit or by a conductivity rate at the FD outlet higher than the specified value during normal operation.

The acceptable performance parameters of HEPA Filters and FDs will be specified in the Surveillance Test Maintenance procedures that prescribe the routine inspections of plant.

3.3.3 Argument 3c: Application of Volume Reduction Processes for Solid Waste All solid radioactive waste is stored, transported and disposed of in containers that have been designed to meet the requirements of relevant legislation and regulatory guidance. In the majority of cases the waste is disposed of in the container in which it is transported to the waste management facility. Making the most efficient use of space in containers has the combined effect of reducing the size of storage facilities, decreasing the number of vehicle movements during transportation and minimising the demand on disposal capacity at appropriately permitted disposal sites (3.4 Claim 4 – Selecting the Optimal Disposal Routes for Wastes Transferred to Other Premises). The design of the Wylfa Newydd Power Station includes dedicated facilities for the management, treatment and storage of solid radioactive waste. These facilities will include a provision for managing control rods that have reached the end of their useful life. The decision on how the control rods will be treated and subsequently stored will impact on the efficiency of storage and therefore the amount of space required. A study was conducted into the management of control rods which concluded that they would be subjected to decay storage followed by size reduction prior to final GDF disposal (3.3.3.1 Evidence: Size Reduction of Control Rods). The Wylfa Newydd Power Station will not include an on-site incinerator or compactor as the current best practice for volume reduction by these means is to perform this task at an off- site industrial facility (3.3.3.2 Evidence: Incineration, 3.3.3.3 Evidence: Solid Waste Compaction). Facilities will be provided for the sorting, segregation, size reduction and packaging of wastes for onward disposal/further treatment. These techniques will increase the amount of waste that can be put into each waste container thus reducing the overall volume of solid radioactive waste that is disposed of from Wylfa Newydd. Where applicable, additional off-site treatment will further reduce this volume.

3.3.3.1 Evidence: Size Reduction of Control Rods

Control rods are a cruciform shape in cross-section which, if disposed of without any size reduction, have the potential to create a large amount of voidage in the selected disposal package, potentially resulting in a greater number of waste packages or larger sized waste packages. Use of size reduction techniques will ensure that the control rods are able to be packaged in a more efficient manner which ultimately reduces the volume of waste that will require disposal.

It has been assumed that spent control rods will be stored initially in the SFP for a minimum of 5 years following removal from the reactor due to the thermal loading of the waste and the campaign cost benefit requirements [Ref-122].

Following this initial storage period, spent control rods will then be placed into a buffer store in a yet to be determined location to allow some radioactive decay prior to size reduction (where necessary) and packaging in

Wylfa Newydd Project – Best Available Techniques (BAT) Case an appropriate container, pending final GDF disposal [Page 58 of [Ref-122]]. Additionally, during size reduction operations, characterisation will be undertaken to determine whether the cold end of the control rods can be managed as LLW.

3.3.3.2 Evidence: Incineration

Incineration is a widely used and effective method of reducing the volume of combustible wastes. The majority of the radioactivity will either be captured in the ash which will be disposed of as solid waste or in abatement systems. The solid waste produced (ash and filters) will be significantly less than the original volume of waste consigned to the incinerator. Thermal treatments are reviewed in the Environment Agencies Requirements Working Group (EARWG) best practice guidance which concludes that thermal treatments are an effective volume reduction technique for combustible materials [Ref-135]. The EARWG best practice guidance states that volume reduction factors of up to 100:1 can be achieved.

In its Integrated Waste Management report, the Nuclear Decommissioning Authority (NDA) [Ref-136] has identified the opportunity to make more use of waste incineration which is currently used in the UK by existing nuclear facilities as a method of volume reduction. The main advantages of using incineration as a technique for treating combustible waste (once the waste is determined to be suitable for incineration) includes the significant volumetric reduction in the waste form. Volume reduction of waste provides the opportunity to improve the efficiency with which the Low Level Waste Repository (LLWR) is used, which is a national resource. Additionally, the option of incineration reduces the use of uncontaminated material for disposal (e.g. grout) and other resources that would have been used in the management, handling and long-term storage or disposal of the waste LLWR.

As with compaction, incineration which takes place off-site does not contribute to the minimisation of waste transferred to other premises but it does provide a means to significantly reduce the volume of waste ultimately requiring disposal. Incineration also provides a very effective route for the disposal of specific waste streams, such as oils, where the alternatives (e.g. encapsulation) result in the generation of large volumes of solid waste.

The Hitachi-GE Waste processing Study - BAT Optioneering report [Ref-115] assessed waste management routes including incineration for the management of waste streams. The assessment identified that incineration is the preferred management technique for the combustible waste stream. LLW filters were assessed according to the disposal costs per m3 for three identified options:

• Compaction;

• Direct disposal to LLWR; and

• Incineration.

Off-site incineration was identified as the preferred option for LLW filters as it has the cheapest disposal costs per m3.

Incineration was also assessed as an option for LLW resins. However, following the assessment, it was determined that bead resin would not meet the WAC requirement on total gamma/beta dose for incineration. For this reason, it was estimated that the incineration route for the LLW bead resin is not a viable option for this waste stream.

Should any aspect of the waste stream not meet the WAC for incineration, the contingency option is off-site compaction, as the volume of waste being disposed of would still be reduced. Direct disposal to the LLWR is the least favoured option for combustible wastes.

3.3.3.3 Evidence: Solid Waste Compaction

The EARWG identifies compaction as a widely used method to reduce the volume of dry radioactive waste. Compaction reduces the waste volume by reducing the amount of voidage in the waste [Ref-135].

Low force compaction can be carried out both on and off-site and is a straightforward and effective volume reduction technology which is commonly used in the UK. Compaction does not produce any secondary solid wastes and any dust created can be contained by an appropriately designed secondary containment system. © Horizon Nuclear Power Wylfa Limited 148

Wylfa Newydd Project – Best Available Techniques (BAT) Case

The typical volume reduction factor achieved using low force compaction is between 3 and 10 as stated in the EARWG best practice guidance.

High force compaction is typically carried out off-site at centralised facilities. High force compaction reduces the volume of waste to a higher degree compared to low force compaction. The EARWG Best Practice in Waste Minimisation report [Ref-135] states that it can achieve a volume reduction factor of up to 25. Although high force compaction will take place off-site so does not contribute to the Claim that waste transferred to other premises will be minimised, it is recognised as the best practice for specified wastes and will reduce the volume that is ultimately disposed of.

As discussed in Section 3.4.2, Horizon will employ the waste processing capabilities available to them under the LLWR Ltd Waste Services Agreement and through the wider supply chain. To this effect, compaction will not be employed on-site so does not contribute to the minimisation of waste transferred to other premises but it does provide a means to significantly reduce the volume of waste ultimately requiring disposal.

3.3.4 Argument 3d: Solid Waste, Minimising the Quantity of Solidified HCW HCW contains impurities that increase the risk of corrosion and the associated generation of CPs. HCW must be treated to make it suitable for either reintroduction to the process water system or for disposal to the environment. An evaporator is used to separate water from the impurities contained in the HCW. The evaporated water is collected, sampled and analysed prior to a decision on re-use or discharge to the environment. The concentrated liquor that remains in the evaporator contains all of the chemical impurities and the majority of the radioactivity associated with the HCW. This concentrated waste is solidified and subsequently disposed of as solid waste (3.3.4.1 Evidence: Solidification of Concentrated HCW). The concentrated liquid waste from the evaporator is solidified in cement. Modifications to the drainage systems within the R/B has substantially reduced the quantity of HCW that is generated which has had a commensurate reduction in the quantity of concentrated liquor that is produced in the evaporator. A series of modifications have been made to the UK ABWR’s drainage system which will reduce the amount of HCW that is generated at Wylfa Newydd. This in turn will reduce the volume of solid radioactive waste recovered from the of the HCW in the evaporator.

3.3.4.1 Evidence: Solidification of Concentrated HCW

Hitachi-GE has undertaken an assessment to demonstrate the application of BAT for the management of concentrated HCW liquors as part of its waste processing study [Ref-115]. The assessment was based on the principle that concentrated liquid wastes will be co-managed with activated carbon and resin waste streams. This assessment has concluded that the preferred technique for the management of concentrated liquid wastes is solidification using an in-line cement immobilisation process.

The assessment considered the following options:

• Solidification using cement immobilisation;

• Polymer immobilisation; and

• Off-site incineration.

Cement immobilisation was selected as the preferred option because:

• Drying of the resin followed by off-site incineration was the top ranking option at the workshop, subject to confirmation that the waste complied with the LLWR WAC. It was subsequently estimated that the bead resin would fail to meet the LLWR WAC requirement for limits on contact dose level, and for this reason it was concluded that this route is not a valid option for this waste

Wylfa Newydd Project – Best Available Techniques (BAT) Case

stream. However, if alternative thermal treatment opportunities become available in the future then these should be considered.

• Polymer encapsulation was ranked second at the workshop by a small margin ahead of cement encapsulation. However, it was determined that polymer encapsulation has limitations around the maturity of the technique for LLW. There is also some uncertainty over the long-term stability of the waste form, and its acceptability against the LLWR WAC.

Following these GDA assessment findings the matter was revisited by Horizon [Ref-122] in an optioneering study. This study recognised the benefits of polymer encapsulation, but also highlighted the challenges on polymer encapsulation formulation. It also highlighted the tendency for the solvent soluble organic content to seep out of cementitious matrices and that this organic content would probably not meet the LLWR WAC.

It was therefore decided that it would be best managed by employing appropriate pre-treatment of the liquor to allow it to meet the LLWR WAC and to then include this liquor in the LLW resin stream which will be subject to cement encapsulation. The study concluded that this approach represented BAT [Ref-115] and the IWS [Ref-70] confirms that this approach will be adopted.

3.3.5 Argument 3e: Application of Decommissioning Techniques to Reduce the Activity and Volume of Decommissioning Waste It is recognised that decommissioning operations will generate significant quantities of waste. A number of decommissioning techniques are available in order to reduce the activity and volumes of this waste. System decontamination using chemical decontamination techniques is carried out in order to remove radioactive contamination attached to inner surfaces of pipes (3.3.5.1 Evidence: System decontamination during decommissioning). Decontamination after dismantling can lead to a lowering of the radioactivity category of the wastes (3.3.5.2 Evidence: Decontamination after dismantling), potentially reducing it to out of scope waste. It is envisaged that the early stages of decommissioning will involve an extensive programme of site characterisation, followed by defueling of the reactors and decontamination of the primary circuit to remove radioactivity from the site. Following this site characterisation, all radioactive decommissioning wastes identified will be considered for suitability against a range of decontamination techniques to lower the total volume of radioactive waste disposed of. The details of the types of decontamination techniques to be employed have not yet been decided on as decontamination technology is constantly evolving. As such, Horizon will decide on the precise techniques to be employed closer to decommissioning to ensure that BAT at the time of decommissioning is employed (3.3.5.3 Evidence: Horizon Decommissioning Strategy). Design elements of the Power Station have been aligned with the philosophy of ‘design for decommissioning’. Applying ‘design for decommissioning’ will be demonstrated through the selection of materials with properties resistant to becoming contaminated and the rationalisation of substances that would be hazardous during decommissioning (3.3.5.3 Evidence: Horizon Decommissioning Strategy). Significant quantities of solid radioactive will be generated during the decommissioning of Wylfa Newydd. The decommissioning techniques have yet to be selected but the volume of solid radioactive waste that is anticipated will be reduced by ‘designing for decommissioning’.

3.3.5.1 Evidence: System decontamination during decommissioning

System decontamination involves the removal of radioactivity, which is attached on the inner surfaces of pipes, with the use of chemicals. The chemical decontamination process is executed prior to dismantling of the plant

Wylfa Newydd Project – Best Available Techniques (BAT) Case and uses an oxidation-reduction reaction. Carrying out chemical decontamination prior to dismantling reduces the risk of the spread of contamination that could occur during cutting and disassembly of pipework. There are a number of techniques as displayed in Table 3.3.5.1-1.

Table 3.3.5.1-1 Chemical Decontamination Techniques Secondary waste waste Secondary Decontamination decontamination decontamination Classification of Classification Temperature of Removal target chemical method method volume liquid Soft Hard crud crud Base DF

(chromium (iron metal oxide)

oxide)

CORD 10- Very X X 90°C method* 100 small

10- Very HOP method† X X 90°C Oxidation- reduction 100 small dissolution decontamination T-OZON 10- Very X X 90°C method method‡ 100 small

DfD method§ X X X 100 Small 90°C

3.3.5.2 Evidence: Decontamination after dismantling

After pipes and plant items are dismantled, decontamination using physical, mechanical and chemical methods can be applied in order to reduce the levels of contamination. Removing contamination can lower the waste category which aids with waste handling and disposal. A range of techniques are available including the following: • Dry ice blasting;

• Ultrasonic cleaning;

• Water flushing;

• Scabbling;

• Shaving;

• Chemical fog;

• Chemical gel (TechXtract);

• Chemical foam;

• Strippable coatings; and

• Electropolishing.

Ultimately the selection of the decontamination techniques will be made by Horizon but should take the following into account:

* Chemical Oxidation Reduction Decontamination (CORD) method (AREVA tech.)

† Hydrazine, Oxalic Acid, Potassium Permanganate (HOP) method (Hitachi-GE Nuclear Energy tech.)

‡ Toshiba Ozone Oxidizing Decontamination for Nuclear Power Plants (T-OZON) method (Toshiba tech.)

§ Decontamination for decommissioning (DfD) method (EPRI tech.) © Horizon Nuclear Power Wylfa Limited 151

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• The material requiring treatment;

• Safety, ensuring there is no spread of contamination;

• Efficiency;

• Minimising worker dose;

• Cost Effectiveness;

• Secondary waste generation; and

• Feasibility of Application.

3.3.5.3 Evidence: Horizon Decommissioning Strategy

The consideration of decommissioning requirements is an integral part of new nuclear build requirements. Prospective operators of new Power Station are required under the Energy Act 2008 [Ref-137] to develop a detailed FDP, which encompasses a Decommissioning Waste Management Plan and a Funded Arrangements Plan, as part of the planning process prior to obtaining permission to operate the plant. In addition to these requirements Horizon is also giving appropriate consideration to ‘design for decommissioning’ in the development of the Wylfa Newydd Power Station design including the following:

• The application of advanced technologies and materials in the Wylfa Newydd Power Station is intended to ensure that waste volumes are minimised and the dose uptake of workers during decommissioning is ALARP.

• The Wylfa Newydd Power Station employs design features that eliminate the generation of activated or contaminated oil, a potentially difficult and hazardous waste stream.

• More generally, in the construction of the Wylfa Newydd Power Station, porous, easily contaminable materials will be avoided or coated where practicable. Surface finishes of plant and equipment are chosen with due attention to their potential operational environment and designed to be readily decontaminated as necessary [Ref-25].

In identifying waste routes for decommissioning wastes the following considerations will be made:

• Selection of decommissioning techniques will take into account the generation of secondary wastes and ensure that waste routes are available for them.

• Wherever practicable wastes will be pre-characterised to enable segregation as close to source as possible, to maximise downstream waste route optimisation and avoid unnecessary cross- contamination of wastes. This is particularly important where systems may include radioactive and non-radioactive elements.

• Decontamination techniques will be selected with due regard to the disposability of any decontamination agents and any resulting secondary wastes from their usage. This is of particular significance where aggressive chemical decontamination methods are to be employed.

In order to improve the predictability of the nature and quantity of decommissioning waste arisings throughout the operational period, the following practices and approaches will be utilised:

• Gathering of OPEX data for other nuclear power plants;

• Preparation of volume and activity calculations based on plant history and usage;

• Collection and retention of relevant characterisation data during outages and other maintenance activities.

• Logging accident events involving leakages and release of radioactivity or hazardous substances.

• Retaining the investigation reports of such events to provide information on the type and location of potential contamination.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• A system of recording data from routine and special Health Physics surface surveys.

• Implementation of management processes to avoid the unnecessary creation of wastes through contamination and cross-contamination.

Effective early characterisation will assist in reducing uncertainty, improving hazard identification and refining the scope, schedule and cost basis for the decommissioning phase. It is envisaged that the early stages of decommissioning will involve an extensive programme of site characterisation, followed by defueling of the reactors and decontamination of the primary circuit.

The solid radioactive waste processing and storage facilities will be amongst the last buildings to be removed from site [Page 58 of [Ref-70]].

3.3.6 Argument 3f: Application of Waste Characterisation to Minimise the Volumes of Waste Sent for Disposal Characterisation of radioactive waste involves determining its physical, chemical, biological and radiological properties. It may be carried out as a standalone process or during other processing such as during segregation. It may be required for record keeping, moving waste between steps and also to determine the best method for managing the waste. The joint regulatory guidance on management of higher activity wastes [Ref-138] provides a clear explanation of regulatory expectations on the requirement to characterise legacy radioactive wastes. The UK has a history of failing to adequately characterise radioactive wastes at source and of failing to retain adequate waste characterisation records. These failings have often resulted in the need to apply significant conservatisms in calculations, which in turn leads to over categorisation of waste which increases the volume of higher activity wastes requiring disposal. It has also led to extreme difficulty/expense in attempting to gain characterisation information retrospectively. By implementing characterisation requirements from the outset, Horizon will be able to optimise the segregation of wastes and then apply the waste hierarchy to the separate waste streams. The requirement placed on Horizon is therefore to put in place suitable arrangements for waste characterisation such that wastes are suitably and sufficiently characterised at the point of arising and at appropriate points in their processing, on a case by case basis and as defined by the specific application. Horizon will define internal WACs that set the requirements of the characteristics (physical, chemical, biological etc.) of waste (3.3.6.1 Evidence: WAC). The characterisation information for wastes will be assessed against the applicable internal WACs to determine whether it adheres to the disposal route that has been identified as representing BAT. At this stage of the project waste management arrangements are being planned for future development so it is too early to definitively state what characterisation techniques will be employed. Instead, an initial assessment of characterisation requirements, on an individual waste stream basis, has been carried out. This took account of requirements stated within external WACs owned by waste service providers (3.3.6.1 Evidence: WAC). This assessment defined broad characterisation requirements of the various radioactive waste categories relevant to the Power Station (3.3.6.2 Evidence: Characterisation Requirements). It has also been used to inform Horizon in the provision of sufficient space in the design of the various waste handling facilities (3.4.1 Argument 4a: Provision of Solid Waste Management Facilities), to allow appropriate equipment to be employed during commissioning and operation. Characterisation records shall be retained within a quality assured records management system with defined retention periods and will be used to support decision making steps (i.e. retrieval, buffer storage, processing, packaging and transport) to consign radioactive waste offsite for treatment or disposal (3.4.2 Argument 4b: Optimal Disposal Route Selection). © Horizon Nuclear Power Wylfa Limited 153

Wylfa Newydd Project – Best Available Techniques (BAT) Case

This approach prevents the foreclosure of options and allows Horizon to defer the decision of BAT with regards to specific equipment used to such a time when sufficient information on waste streams is available to support decision making. The expectation is that equipment will be selected and deployed in time for commissioning. Therefore at the appropriate time to ensure delivery of equipment before commissioning, an assessment will be undertaken that will consider the information on waste streams and the latest technology. Horizon’s operating philosophy at Wylfa Newydd will be to characterise the physical, chemical, radiological and biological properties of waste as early as possible. Whilst the characterisation techniques have yet to be selected, the design of the UK ABWR provides sufficient space to allow Horizon to deploy a wide range of such techniques.

3.3.6.1 Evidence: WAC

The requirements for characterisation of individual waste streams will be defined in WAC. At this stage it is considered that WACs will fall into two categories [Ref-139]:

• Internal WACs – These WACs define the criteria within Horizon for the monitoring, characterisation, sorting and segregation of discharges and different radioactive wastes streams to ensure that they comply with downstream processes. This will provide the means to control wastes from point of production, thereby ensuring effective application of the waste hierarchy and optimisation of waste routing – this is key to the demonstration of BAT on a case by case basis. The Internal WACs will be produced at a future phase within the Power Station Project as part of the Operational Strategy. The Internal WACs are reliant upon accurate waste characterisation data and sufficient detailed performance information of waste treatment processes and technologies.

• External WACs – These WACs will be defined by external organisations operating treatment processes and permitted disposal facilities. They require operators to apply BAT in determining options to treat or dispose of waste. Horizon’s internal WACs will be written to inform waste producers of the physical characteristics expected after waste management processes that demonstrates BAT has been applied and will be compatible with an external WAC. This enables Horizon to send waste off-site for onward treatment and disposal in a compliant and optimised manner. External WACs considered include:

• Radioactive Waste Management Ltd (RWM) Waste Packaging Specifications (WPS) [Ref- 140];

• Low Level Waste Repository Ltd (LLWR) LLW Disposal WAC [Ref-141];

• LLWR WAC - Combustible Waste Treatment [Ref-142];

• LLWR WAC – Metallic Metal Treatment [Ref-143];

• LLWR WAC – Super-compactable Waste Treatment [Ref-144]; and

• LLWR WAC – Very Low Level Waste Disposal [Ref-145].

3.3.6.2 Evidence: Characterisation Requirements

An assessment of characterisation requirements [Ref-139] identified those appropriate to managing waste generated in the Power Station. These requirements are summarised in Table 3.3.6.2-1.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Table 3.3.6.2-1: Summary of Characterisation Requirements

Waste Category Characterisation Requirements

VLLW Considered a sub-section of LLW. Requirements identical to LLW.

Waste Form Requirements Physical, chemical and biological properties. Evidence that the presence of ‘restricted’ materials are within defined constraints (e.g. reactive metals, putrescible wastes, organic materials, soluble solids, oxidising agents, pressurised containers, complexing agents, ion exchange resins, hazardous and polluting materials etc.) and packaged in accordance with any prescribed requirements of the waste service provider’s WAC. The Radiological Characteristics of waste need to be determined. Records of the total and specific radioactivity content, radionuclide data, fissile content, radiation dose rate information, presence of sealed sources, package external contamination levels etc. to LLW demonstrate that all criteria are within defined acceptable limits derived from the disposal facility Environmental and Safety Cases. Waste Package Requirements Evidence of compliance with packaging requirements including maximum gross weight, labelling, external dose rates, venting arrangements, compliance with transport regulations etc. In most cases the waste container will conform to a prescribed design and may be provided by the disposal facility service provider. In any case where a non-standard container is utilised further requirements will be defined to demonstrate that the container conforms fully to the disposal facility’s WAC.

Waste Form Requirements

Radioactivity: total activity, radionuclide composition, radionuclide fingerprint, criticality safety, radiogenic heat output, radiation stability, homogeneity, surface dose rate, dose rate and identification of volatile/non-volatile radionuclides.

Chemical properties: chemical stability, composition, reactivity, corrosiveness, chemical compatibility, gas generation, toxicity, organic content, material specifications for resins.

Physical properties: permeability, porosity, homogeneity, density, voidage, load resistance, dimensional stability, impact resistance, weight distribution, particle size ILW distribution, rheology, settling behaviour, flow characteristics, fire resistance, thermal conductivity and freeze/thaw stability.

Biological properties: biological degradation, potential for organic decomposition. Waste Package Requirements

Handling: shape, dimensions, lifting arrangements, gross package weight, ease of stacking, impact resistance, gas build-up/venting systems and package identification.

Radiological protection: surface contamination levels and surface dose rates. Container durability: container material, container closure, and container degradation.

HLW Same as for ILW.

SF Same as for ILW.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.4 Claim 4 – Selecting the Optimal Disposal Routes for Wastes Transferred to Other Premises The Rw/B in the Wylfa Newydd Power Station will be designed with sufficient space and services for the installation and operation of SSC necessary for the characterisation, treatment and storage of radioactive wastes. Horizon will undertake activities to manage radioactive waste to enable the selection of an optimal waste disposal route for each waste stream. This BAT Case presents the disposal routes which have been selected, following site specific optioneering. However, it is also recognised that some disposal route selections are dependent on factors outside of Horizon’s control and in these cases the decision on the disposal route of the waste will not be made until sometime in the future. These such instances have been flagged as forward actions. The Arguments presented in support of this Claim demonstrate compliance with the standard BAT conditions [Ref-5]: • Condition 2.3.3 (b) ‘Characterise, sort, segregate solid and non-aqueous liquid wastes, to facilitate the disposal by optimised disposal routes.’ • Condition 3.1.3. ‘The operator shall dispose of each form of solid and non- aqueous liquid radioactive waste by an optimised disposal route for that waste form.’ The Wylfa Newydd Power Station design contains a range of features that contribute to the substantiation of this Claim including: • Areas of space for the segregation and sorting of radioactive wastes. • Facilities to treat and process radioactive wastes. • Areas of space for temporary and long-term storage radioactive waste packages. The Wylfa Newydd Power Station will be operated to manage radioactive wastes to meet expectations in the following: • Agreements in principle obtained for the disposal of LAW that will be generated during the lifetime of the Wylfa Newydd Power Station. • Disposability assessments prepared for HAW and SF. In developing the Arguments presented to demonstrate the validity of Claim 4 the REPs [Ref-7] have been taken into account. The following REP is considered to be specifically relevant to this Claim: • Principle RSMDP7 ‘When making decisions about the management of radioactive substances, the best available techniques should be used to ensure that the resulting environmental risk and impact are minimised.’

3.4.1 Argument 4a: Provision of Solid Waste Management Facilities A range of facilities and equipment are required to ensure the effective and efficient use of available waste management routes. The design of the Wylfa Newydd Power Station Rw/B includes the space and services to install the equipment necessary to undertake the characterisation (3.4.1.1 Evidence: Waste Characterisation and Assessment Capabilities), sorting (3.4.1.2 Evidence: Segregation and Sorting Capabilities), treatment and packaging (3.4.1.3 Evidence: Waste processing and Packaging Facilities) and storage (3.4.1.4 Evidence: Waste Storage Capacity) of waste prior to consignment to an appropriately permitted waste management service supplier. These facilities reflect the outputs of the © Horizon Nuclear Power Wylfa Limited 156

Wylfa Newydd Project – Best Available Techniques (BAT) Case waste strategy (3.4.1.5 Evidence: Waste Strategy) that has been developed for the wastes that will be produced by the Wylfa Newydd Power Station. The current size and configuration of the Rw/B is considered to offer Horizon the flexibility to select and implement techniques that will reflect its operational needs and the regulatory requirements in force at the time (3.4.1.6 Evidence: Size and Configuration of Rw/B provides flexibility). The provision of such a flexible facility is considered to represent BAT at this stage. Horizon’s operating philosophy at Wylfa Newydd will be to segregate, sort, treat and package waste on the basis of their physical, chemical, radiological and biological properties. Horizon have undertaken strategic options assessments to determine the process options to be employed and whilst the exact process equipment to be employed has yet to be selected, the design of the UK ABWR provides sufficient space to allow Horizon the flexibility to deploy a wide range of such equipment.

3.4.1.1 Evidence: Waste Characterisation and Assessment Capabilities

The design of the Wylfa Newydd Power Station will include sufficient space to undertake sampling and non- destructive assay of waste prior to transfer off site for disposal [Ref-115]. As well as waste assay to determine the radiological properties of waste items, the characterisation and assessment facilities will enable Horizon to determine the waste item’s physical and chemical properties. This allows for the effective segregation of waste and the identification of suitable waste routes based on the characteristics of each distinct waste stream and ensured that wastes are compliant with internal and external WACs and CfAs. Waste management practices proposed for the Wylfa Newydd Power Station will produce packages consistent with those already produced in the UK nuclear industry.

Appropriate characterisation techniques will be employed to adequately assay all wastes although the exact equipment and techniques employed will be decided on at a later date to prevent foreclosure of options. The provision of sufficient space for waste characterisation within the design of the Wylfa Newydd Power Station provides flexibility for Horizon to select the equipment that best represent BAT prior to commissioning. Sampling and non-destructive assay techniques currently available in the UK have been assessed as suitable for use in characterising solid wastes from Wylfa Newydd Power Station [Ref-130] giving confidence that the deferral of the decision will not prevent wastes being suitably characterised. This approach is therefore deemed to represent BAT.

3.4.1.2 Evidence: Segregation and Sorting Capabilities

The design of the Wylfa Newydd Power Station provides sufficient space to allow Horizon to carry out those techniques detailed in the Radioactive Waste Management Arrangements [Ref-116] [Ref-115].

Horizon has undertaken a number of optioneering assessments to inform BAT with regards to sorting and segregation facilities which will inform future design decisions on the facilities needed in sorting and segregating waste. These decisions are detailed below:

• Following a review of the radioactive waste management philosophy [Ref-118], LAW (with the exception of wet solid LLW) will be sorted, segregated, packaged and initially monitored at source to support onsite movement. It will then be consigned to the Lower Activity Waste Management Facility (LAWMF) for comprehensive characterisation and further packaging to support onward management decisions and to ensure and demonstrate compliance with WACs and CfAs. Compliant packages will then be accumulated within the facility to allow batch wise shipment to onward treatment/disposal facilities. The LAWMF will be a centralised facility servicing both reactor units [Ref-102].

• There will be two dedicated SFPs (one servicing each reactor unit) located within the corresponding R/B. Each of these SFPs will have a dedicated sorting area which is provided to

Wylfa Newydd Project – Best Available Techniques (BAT) Case

allow sorting and size reduction of HLW (primarily control rods and neutron monitoring equipment) prior to onsite storage [page 24 of [Ref-146]].

This approach allows for the segregation and sorting of waste into diversified waste streams prior to the appropriate treatment and/or disposal option for each waste stream (3.4.1.3 Evidence: Waste processing and Packaging Facilities and 3.4.2 Argument 4b: Optimal Disposal Route Selection). Details relating to the selection of techniques that will be adopted for the segregation of wastes are provided in Section 3.3.2 (Argument 3b: Selection of Methods to Minimise Solid Waste Generation).

An overview of the two segregation and sorting facilities is provided below.

LAWMF

In early iterations of the Wylfa Newydd design, raw LLW and VLLW was to be consigned to the LLW treatment facility where it would be subject to sorting, segregation, shredding and low force compaction. This treatment was to ensure it met the appropriate WAC to allow its onward treatment/disposal (as appropriate) in line with the selected disposal routes detailed in Section 3.4.2. (Argument 4b: Optimal Disposal Route Selection). Additionally, there was to be a decay storage area for the decay of spent bead resins and a buffer storage area to accumulate packaged waste to allow batch consignment to onward treatment/disposal facilities.

Through engagements with LLWR and related nuclear industry waste service providers, Horizon has become better informed on emerging national strategy and practice in this area [Ref-102]. This prompted and facilitated the development of a LAW management philosophy [Ref-118] which resulted in the following design changes [Ref-102]:

• Waste will be sorted, segregated, packaged and initially monitored at source rather than in the LAWMF thus simplifying the processing undertaken in the LAWMF. To support this, each building anticipated to generate radioactive waste will then have a dedicated packaged radioactive waste collection point where waste packages will be routinely and regularly collected and taken to the LAWMF;

• The requirement for a raw radioactive waste buffer store within the LAWMF has been removed. Instead, there will be a small packaged waste buffer store to allow the accumulation of waste consignments; and

• Decay storage of bead resins will be undertaken in raw form in shielded tanks within the Rw/B prior to processing allowing the removal of the packaged waste buffer storage area in the LAWMF.

Due to this, the LAWMF will be more of a management facility than a processing facility, possessing only the capability to:

• Undertake detailed characterisation to:

• Support onward management decisions and facilitate wastes being managed as appropriately as possible in line with the NWP and the principles of BAT; and

• To ensure and demonstrate compliance with relevant WACs and CfAs.

• Further package waste to support transport and consignment to onward management/disposal facilities;

• Accumulate waste packages to allow batch-wise shipment to onward treatment/disposal facilities; and

• Repackage waste if required.

This approach will still ensure that the waste will be in a form that will allow its onward treatment/disposal (as appropriate) in line with the selected disposal routes detailed in Section 3.4.2. (Argument 4b: Optimal Disposal Route Selection), but is deemed to represent BAT for the following reasons [Ref-102]:

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• With regards to LAW, it aligns Horizon’s radioactive waste strategy with that of UK national strategy and of the National Waste Programme allowing efficient and optimised onward management and disposal of LAW.

• The reduction in the amount of raw, unconditioned radioactive waste within the LAWMF reduces worker dose and represents ALARP;

• It drastically simplifies the processing undertaken in the LAWMF mitigating the requirement for much of the previously assumed processing equipment (compactor, sorting tables, gloveboxes etc.) which leads to:

• Reduced operational secondary radioactive waste arisings as, linked to the reduction in unconditioned radioactive waste in the facility, less equipment and disposables will become contaminated over the lifetime of the facility;

• Reduced decommissioning waste arisings as, linked to the reduction in unconditioned radioactive waste in the facility, there will be less equipment and plant to become contaminated over the lifetime of the facility;

• A reduced maintenance burden of the facility.

• A reduced footprint of the facility on the Development Consent Order (DCO) plot plan [RD-1];

• A reduced capital cost.

• Decaying bead resins prior to processing as opposed to following reduces occupational exposure associated with processing and is consistent with ALARP.

SFP – Sorting Area

Whilst principally the SFPs are storage facilities used to safely store SF and HLW prior to onsite dry storage, within each SFP there will be a sorting and size reduction area. This area provides sufficient space and shielding to allow the safe sorting of HLW and SF items and, if appropriate, the size reduction of HLW (principally control rods and neutron monitors) prior to placement into suitable storage canisters [Ref-146].

3.4.1.3 Evidence: Waste processing and Packaging Facilities

Techniques to treat waste were selected for GDA (3.4.2.1 Evidence: Waste processing Techniques and Disposal Routes) to ensure effective and efficient treatment, to enable wastes to be disposed of in accordance with the Radioactive Waste Management Arrangements [Ref-116].

Facilities are required within the Power Station to receive, process, package/condition and export the following waste types [Ref-70]:

• Dry solid HLW (control rods, blade guides, neutron flux monitors, probes and neutron sources);

• Wet solid ILW (WSILW) (sludge from backwashing of HFFs in the LCW and powder resins and crud from filter demineralisers in water clean-up (CUW and FPC) systems);

• Wet solid LLW (WSLLW) (bead resins from the LCW, HCW and Condensate Demineraliser (CD) systems and concentrates from the evaporator in the HCW system);

• Dry solid LAW (contaminated PPE, monitoring swabs and items generated from routine operations and maintenance (for example, wood, cloth, metal, pipes lagging, gas filters, concrete, glass and spent filtering materials));

Additionally, whilst not strictly waste, facilities must also be provided to receive and package SF to facilitate onsite storage.

Optioneering for the Wylfa Newydd Power Station has concluded that [Ref-102]:

• LLW (with the exception of WSLLW) will be sorted, segregated, packaged and initially monitored at source [Ref-118]. It will then be consigned to the LAWMF for onward management, further

Wylfa Newydd Project – Best Available Techniques (BAT) Case

packaging and consignment to an onward offsite treatment/disposal site (as appropriate). The LAWMF will be a centralised facility servicing both reactor units [Ref-102].

• There will be a single, centralised Rw/B servicing both reactor units which will include processing facilities for WSILW and WSLLW [Ref-102].

• There will be two dedicated SFPs (one servicing each reactor unit) located within the corresponding R/B. Each of these SFPs will have a dedicated sorting area which is provided to allow sorting and size reduction of HAW prior to onsite storage [page 24 of [Ref-146]].

A summary of how the waste processing and packaging facilities will interface with other parts of the Wylfa Newydd Power Station is provided in Figure 3.4.1.3-1.

Figure 3.4.1.3-1 Summary of the flow of waste into and out of the waste processing and packaging facilities

An overview of the processing facilities is provided below.

LAWMF

As stated in Section 3.4.1.2 (Evidence: Segregation and Sorting Capabilities), the LAWMF will be more of a management facility than a processing facility and further details can be found in Section 3.4.1.2.

Rw/B

The Rw/B is a radioactive waste management building which will house the WSILW and WSLLW treatment facilities. It will also house the liquid effluent management system but this is discussed separately in Section 3.2.6 (Argument 2f: Configuration of Liquid Management Systems).

In early iterations of the Wylfa Newydd design the Rw/B was to only house the liquid effluent management system and the WSLLW and WSILW systems were to be housed in attached but separate buildings. This was later revised following optioneering which concluded incorporating the WSLLW and WSILW process activities within an extended Rw/B represented BAT and resulted in three process buildings per reactor unit being consolidated into one [Ref-101].

This gave rise to two dedicated Rw/B, one servicing each reactor. This was then further optimised following rationalisation of the plot plan, as stated in Section 3.2.6.1 (Evidence: Centralised liquid effluent management system), by combining these two dedicated Rw/B into a single, centralised facility servicing both units. This approach realised the following benefits [Ref-102]:

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• Only one set of each processing equipment (liquid effluent management system, WSLLW and WSILW) will be required:

• Operational radioactive waste arisings will be reduced;

• Radioactive decommissioning waste arisings will be reduced;

• The maintenance burden associated with the facility will be reduced;

• The overall footprint of the site will be reduced; and

• Packaging campaigns will potentially be more efficient.

As such, it was therefore decided that the two UK ABWRs at the Wylfa Newydd Power Station will be serviced by a single, centralised Rw/B [Ref-102].

The WSILW and WSLLW treatment facilities will treat both WSILW and WSLLW separately dependent on their treatment technique.

• The WSLLW treatment part of the facility will process spent resins and concentrated liquid waste (sludge) from the evaporator to allow disposal. Space requirements are based on the requirement to immobilise WSLLW using a cementation process.

• The WSILW treatment part of the facility will process the WSILW into an immobilised form compliant with NDA RWM requirements using a cementation technique. For the purposes of the GDA, space requirements were based on the requirement to immobilise WSILW using a cementation process. However, Horizon is considering different options based on industry programmes and it is recognised that Horizon may adopt an alternative treatment process. The space provided within the design of the Wylfa Newydd Power Station is considered to be sufficient to allow Horizon the flexibility to adopt an alternative treatment technique if this is demonstrated to be BAT during subsequent assessments.

Further detailed design considerations are included in the System Design Descriptions for WSILW [Ref-149] and WSLLW [Ref-150] and in PCSR Chapter 18.4 [Ref-147].

The main plant items in the WSLLW treatment part of the facility are anticipated to include:

• A LLW process tank;

• Cement grout preparation equipment (powder feeder, grout feed hopper and grout pump);

• In-Line Mixer; and

• A THISO filling position.

As stated these are anticipated plant items based on the System Design Descriptions for WSLLW [Ref-150]. It must be noted that a decision has been made to defer the design and installation of the WSLLW processing element of the Rw/B [Ref-102] so plant items listed above might change. This decision was made because the processing capability will not be required until 13 years after Unit 1 COD and deferring the design and installation of this capability offered the following benefits which contributed to this decision representing BAT [Ref-102]:

• It presents Horizon the opportunity to gather OPEX and real time characterisation data for the waste prior to design and installation of any treatment process. This will allow the design of the process to be tailored to the actual waste’s characteristics rather than an estimation of these based on the current source term.

• It avoids a situation where process and handling plant will be dormant for a significant period of time prior to use.

• It presents the opportunity to reduce the capital cost of the project as deferral of the design and installation would place the associate costs on the operational phase of the Power Station.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• WSLLW processing is a well understood process with the supply chain (such as Tradebe Inutec) offering suitable mobile plant services. These could be employed on a campaign basis, which could lead to additional cost savings.

• It does not foreclose options allowing BAT at the time of first use (taking into account technological advances between now and then) to be implemented.

To support this deferral decision, sufficient space and appropriate interfaces (process tanks, services and pipework routes) must be provided in the design of the Rw/B to allow incorporation of future processing equipment. Dimensional data for a Tradebe Inutec mobile solidification system, a proven design which at present could process this waste, was therefore obtained to bound the Rw/B envelope [Ref-102]. This ensures that, should there not be any future technological advances in this area, Horizon can still treat the waste using a currently available and proven technique giving confidence that no orphan wastes will result from the deferral decision.

The main plant items in the WSILW treatment part of the facility include:

• ILW resin storage tanks;

• ILW filter backwash sludge storage tanks;

• Process Tank;

• The ability to send waste to the WSLLW system if waste turns out to be below the ILW threshold. This will reduce ILW arisings;

• Cement grout preparation equipment (powder feeder, mixer, grout feed hopper and grout pump);

• Cement powder feed system; and

• Modular ILW solidification plant.

The baseline option for the processing of WSILW was for it to be a batch wise process being undertaken every 2-5 years which would have required the availability of WSILW processing from the onset of operations. However, an options assessment process based upon Principle RSMDP11 of the NRW REPs [Ref-7], assessed this option against two other options [Ref-102]:

• Store the waste in “raw” form in tanks in the Rw/B throughout the operational phase – This option inferred having sufficient powder resin and sludge storage capacity in the Rw/B for the operational life of the plant and that the total plants arisings would be treated in a single batch process at the end of the plant’s life.

• Design and implement WSILW processing, transfer and storage to align with the SF Storage Facility construction timescales - This option requires interim tank storage for powder resin arisings of up to 10 years in the Rw/B. WSILW processing capability would be installed in the Rw/B but would not be required for up to 10 years. The ILW Storage Facility would be constructed in the same area and at the same time as the SF Storage Facility.

The study concluded that to design and implement WSILW processing, transfer and storage to align with SF Storage Facility construction timescales was the favoured approach for the following reasons [Ref-102]:

• Since this approach would produce no ILW packages until 10 years after COD, it removes the initial requirement for an ILW Storage Facility. As such the construction of the ILW Storage Facility can be delayed by 10 years which presents a number of benefits:

• It will bring the construction of the facility in line with that of the SF Storage Facility. This presents the opportunity for a combined construction programme for the SF and ILW storage Facilities. Since these facilities share a High Security Area (HSA) there will be economies and efficiencies associated with co-construction – this may include a smaller initial site perimeter fence;

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• With regards to construction of the SF Storage Facility, and due to the two facilities being situated so close to each other, there is an ALARP argument for undertaking the construction of the storage facilities at the same time. This is because construction of the SF Storage Facility next to an active ILW Storage Facility would increase the dose and therefore risk posed to construction workers;

• The decision would lead to a shorter design life requirement for the ILW Storage Facility; and

• The estimated £80-100m cost of the ILW Storage Facility would be placed on the operational phase of the Power Station leading to a significant reduction in capital costs.

• It presents potential to defer procurement and installation of some of the WSILW processing equipment which will prevent the foreclosure of options ensuring that BAT at the time of implementation (taking into advances in technology) is employed. This is particularly important as Horizon have become aware of ongoing developments in thermal treatment technology through ongoing interactions with the NDA, which would likely be suitable to treat this waste;

• The decision opens the opportunity for homogenisation and sampling of the waste before finalising the process design. This will allow the design to be based on actual waste data rather than predicted characteristics such that the process can be better tailored to the actual wastes produced. This will reduce operational uncertainty and related risks during processing;

• Since the step posing the most risk of dose to workers is the processing step which follows the raw storage period, increasing the raw storage period will reduce the risk posed during processing as it will allow the activity to further decay thus resulting in reduced dose to workers. There is therefore an ALARP argument for the decision; and

• Deferring installation of some of the WSILW processing equipment would reduce the associated maintenance burden (if the equipment is not installed it does not require maintenance) and avoids a situation where process and handling plant will be dormant for a significant period of time prior to use.

As such, and in line with this conclusion, it was decided that the raw WSILW storage capacity would be increased to allow 10 years’ worth of accumulation and the batch processing would be undertaken every 10 years.

SFP

As described in Section 3.4.1.2, the SFP will have a sorting area in which control rods will be size reduced. It will also have another area where both SF and HLW containers are loaded. Although not technically treatment, placing the SF and HLW in these containers (following sufficient cooling time on the SFP) it a large step in its journey towards becoming passively safe. This then allows it’s consignment to the SF Storage Facility for on-site storage pending the availability of a GDF [Ref-70].

3.4.1.4 Evidence: Waste Storage Capacity

The design of the Wylfa Newydd Power Station includes the capacity to temporarily store ILW, HLW and SF as detailed in the Radioactive Waste Management Arrangements [Ref-116]. As stated in Section 3.4.1.2 (Evidence: Segregation and Sorting Capabilities), LAW will be promptly processed and consigned to onward management/disposal facilities so will not be stored on-site [Ref-102].

The capacity of the storage facilities is sufficient to allow Horizon to optimise storage times through the application of BAT.

ILW Storage Facility

The ILW storage facility will store solidified WSILW in 3m3 drums for an extended period prior to disposal to the GDF in accordance with RWM requirements. Further work has been identified to consider optimisation of the use of this facility for decommissioning wastes and it is currently envisaged that a separate ILW Storage Facility for decommissioning waste arisings will be built at the end of the plant’s life [FAbc-7]. Further detailed design considerations are included in the System Design Description [Ref-148].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

At GDA, Hitachi-GE explored ILW storage options and assumed that ILW would be stored in on-site facilities, the design of which would be suited to the specific packages/containers which would be stored in them. The stores will be built in accordance with industry guidance on the storage of HAW packages [Ref-120].

As discussed in Section 3.2.10, the option to decay store ILW was identified as a potential option, but this was discounted during an optioneering study as it was concluded that the ILW resin would not decay to LLW [Ref- 122].

Horizon then undertook a workshop study on the storage of ILW and concluded that it was the preferred option to employ a single centralised facility servicing both reactors for the reasons outlined below [page 41 of [Ref-101]]:

• The ILW storage facility will be a large heavily shielded building and as such, having a shared facility is beneficial with regards to minimising the capital cost of constructing and commissioning such a facility.

• Operator dose is likely to be lower for a shared facility, as only one set of equipment would require operation, maintenance and decommissioning.

• The environmental impact of constructing and operating a shared facility would be less than constructing and operating two facilities, requiring less energy and resource use.

• There are a number of constraints with the existing plot plan and as such, a shared facility will ensure land use is minimised.

• Having one facility in which all ILW packages are stored will minimise the complexity of managing the packages and associated records during storage. A shared facility shall ensure greater consistency in how ILW packages are managed at Wylfa Newydd, aiding the eventual disposal of ILW to the GDF.

• A shared facility would be beneficial from a security perspective, minimising the number of radwaste storage facilities requiring surveillance after site closure. Fewer facilities, resulting in a smaller area of land to patrol, would also have a longer term benefit in terms of reducing security costs.

SF Storage Facility

The baseline Wylfa Newydd design reflected a single HLW storage facility servicing both reactor units and a single SF storage facility also servicing both reactor units. Early optioneering [Ref-101] suggested that the two facilities could be combined which initiated further work in this area. Subsequent work then identified the following benefits of combining the two facilities:

• SF and HLW share a common storage cask arrangement so the storage requirements for HLW and SF are very similar. As such, combining the two stores eliminates the duplication of a number of key elements in the design, notably [Ref-102]:

• The seismically qualified floor slabs;

• The airlock entry systems;

• The shielded cask transfer and inspection bays and cask pits;

• The aisle restack positions; and

• The Vertical Cask Transfer (VCT) machines.

• These facilities are large so having single centralised facilities is beneficial with regards to minimising the capital cost of constructing and commissioning [Ref-101].

• Operator dose is likely to be lower for single centralised facilities as, for each facility, only one set of equipment would require operation, maintenance and decommissioning [Ref-101].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• The environmental impact (energy consumption, carbon emissions and resource use) of constructing and operating shared facilities would be less than for individual dedicated facilities [Ref-101].

• Shared facilities would have a smaller footprint than individual dedicated facilities so would minimise the use of land [Ref-101].

• Having shared facilities minimises the complexity of managing packages and associated records during storage. Shared facilities ensure greater consistency in how packages are managed during storage, aiding the eventual disposal of the waste to the GDF [Ref-101].

• Both HAW and SF is likely to require repackaging prior to disposal to the GDF. Having shared facilities would therefore minimise the complexity of managing the repackaging of the waste prior to disposal [Ref-101].

• Shared facilities benefit nuclear security arrangements, as it minimises the number of radioactive waste storage facilities requiring surveillance and physical security after site closure. Fewer facilities results in a smaller area of land to patrol, would also have a longer term benefit in terms of reducing security costs [Ref-101].

It was therefore decided during the site wide optimisation that the two would be combined to make a single HLW/SF storage facility (known as the SF Storage Facility) servicing both reactor units. As with the baseline HLW and SF storage facilities, this combined facility is designed to safely and securely store packaged HLW and SF for a period, currently envisaged to be approximately 100 years, pending the availability of a GDF. The combined facility was deemed to represent BAT over the baseline design for the following reasons [Ref-102]:

• It will lead to a reduction in the overall footprint when compared to the two individual storage facilities;

• It will reduce decommissioning waste arisings at the end of the facility’s life;

• It will reduce in the overall capital costs of the storage capability; and

• Since both waste types require increased security arrangements, the combined facility results in a decreased security burden.

An overview of the current design of the combined SF Storage Facility is provided in Figure 3.4.1.4-1 [Ref-102]. It must be noted that this is the current design but this isn’t due to be implemented until 10 years after COD. As such Horizon may change this design if there are advances in cask/cask storage system technology. Horizon are still deciding the exact casks and cask storage systems to be employed. Typically, the choice of cask will dictate the choice of cask storage system to be implemented.

However, Holtec are currently implementing their newly developed UMAX below grade dry storage technology at a site in New Mexico, USA, close to the already established Waste Isolation Pilot Plant (WIPP) facility [Ref-104]. This technology is capable of accepting all fuel cask designs such that matching specific cask/cask storage systems are not required. Since this technology is currently available, this gives confidence in Horizon’s decision to defer the selection of spent fuel cask as this should not impact their capability to safely store the selected fuel casks. This gives Horizon the opportunity not to foreclose options in regards to fuel cask / fuel cask storage system selection allowing them to take advantage of potential technological advances in this area between now and time of implementation.

Horizon will be required to undertake revised direct shine dose assessments from the final combined facility to ensure that the revised, combined store does not challenge, from a dose perspective, the concept of ALARA [FAbc-6]. However, it is not envisaged that it will.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.4.1.4-1 Combined SF Storage Facility

3.4.1.5 Evidence: Waste Strategy

The current waste strategy is presented in the IWS [Ref-70] and is briefly summarised below.

Dry Solid LAW

Horizon strategy for the management of dry solid LLW will be characterisation, sorting and segregation at source, followed by confirmatory checking in the LAWMF, in order that the waste can be disposed of in line with the Waste Hierarchy. Where practicable wastes will be reclassified, recycled or volume reduced with disposal to LLWR or similar as a last resort.

WSLLW

Horizon’s current strategy for the management of WSLLW will be direct cementation in a THISO and disposal to LLWR. This is to ensure that the packaged waste volumes are minimised as far as is practicable. It is important to note that this strategy may change because, as stated in 3.4.1.3 (Evidence: Waste processing and Packaging Facilities), a decision has been made to defer the design and installation of the WSLLW processing element of the Rw/B. One of the reasons for this decision was to prevent the foreclosure of options. As such, it must be noted that this strategy may change in the future if technological advances render this current strategy no-longer BAT.

WSILW

Horizon strategy for the management of WSILW will be cementation in thin walled stainless steel containers; this will result in a passively safe wasteform for storage, and designed to be suitable for disposal in the GDF. Formulation development work will be undertaken in the future to ensure that the packaged waste volumes are minimised as far as is practicable, and to support Letter of Compliance (LoC) Submissions.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

HLW

Horizon strategy for the management of dry solid HLW is decay storage in suitable containers and shielded concrete overpacks followed by retrieval, size reduction, packaging and conditioning in thin walled stainless steel containers compatible with GDF disposal requirements. HLW final waste packages will be stored in a shielded store on site awaiting the availability of the GDF.

Horizon strategy for the management of SF is storage in suitable canisters and shielded concrete overpacks within a dedicated on-site store. Following the availability of a GDF, spent fuel will be retrieved and repackaged prior to consignment to the GDF for final disposal.

3.4.1.6 Evidence: Size and Configuration of Rw/B provides flexibility

The Rw/B shall have sufficient space and services for the installation of required waste handling facilities. For example, suitably sized tanks for the accumulation of 10 years’ worth of raw WSILW will be installed in the Rw/B and current compatible and proven technology for the treatment of WSLLW has been used to bound the dimensions of the WSLLW process area within the Rw/B [Ref-102].

3.4.2 Argument 4b: Optimal Disposal Route Selection The UK Government and Devolved Administrations published a revised policy for the long term management of solid LLW in 2007. This policy recognised that the previous preference for disposing of LLW from nuclear sites to the national LLWR was no longer sustainable and that alternatives for the management of these wastes were required. The revised policy requires nuclear operators to consider a range of options when developing plans for the management of solid LLW. These options are to be based on the waste hierarchy and are to take into account a broad range of environmental and sustainability principles in addition to those related to the risk of exposure to potentially harmful ionising radiation. Since 2007 the nuclear industry and its suppliers have made significant progress in developing alternatives to the disposal of LLW to the national LLWR. A range of techniques have been implemented that allow LLW to be: • Minimised at source; • Re-used/recycled; • Volume reduced prior to disposal; and • Disposed of at alternative sites to the national LLWR. The design of the Wylfa Newydd Power Station, and the Radioactive Waste Management Arrangements [Ref-116] developed to manage the waste, recognise that a range of waste management options are available for the management of LLW that will arise during the operation of the Power Station. Management of waste items as LAW is as a consequence of them not being appropriate to the clearance and exemption process or not meeting criteria for clearance (3.3.2 Argument 3b: Selection of Methods to Minimise Solid Waste Generation). Strategic consideration of options related to the provision of on-site waste processing facilities has concluded that incinerators, metal treatment and disposal facilities will not form part of the generic design for the Wylfa Newydd Power Station. This decision is consistent with the findings of a number of strategic studies that are discussed further in Section 3.4.5 (Argument 4e: Compatibility of Existing UK Waste BAT Studies). Evidence is provided to support the use of selected waste management techniques (including those provided by LLWR) for the disposal of LLW (3.4.2.1 Evidence: Waste Processing Techniques).

Agreements in principle from the LLWR (3.4.3 Argument 4c: Agreement in Principle for Waste Routes - LAW) indicate that Horizon has viable waste routes for its solid LAW, as it will be acceptable for treatment or disposal via waste services offered by LLW (3.4.2.2 Evidence: Waste Routes for LAW). The avoidance of oils used in active processes during normal operations, means there is no oily LAW route (3.4.2.2 Evidence: Waste Routes for LAW). The design of the Wylfa Newydd Power Station solid waste management facilities allows Horizon a high degree of flexibility in the selection and deployment of LLW treatment techniques as discussed in Section 3.4.1 (Argument 4a: Provision of Solid Waste Management Facilities). Implementing a Waste Management System (WMS) to cover all aspects of managing radioactive waste in the Power Station is the tool for generating accurate and comprehensive records that support consignment of radioactive waste to an optimum offsite treatment facility or disposal site (3.4.2.3 Evidence: Waste Management System). Horizon’s operating philosophy at Wylfa Newydd will be to make the best use of the wide range of waste management routes for solid LLW and to minimise the amount of solid LLW disposed of to LLWR. Horizon will invest in the segregation, sorting, treatment and packaging techniques that will be necessary to implement this approach.

3.4.2.1 Evidence: Waste Processing Techniques

The techniques required to manage waste from the Wylfa Newydd Power Station were assessed by Horizon. Waste processing equipment identified for use in the Wylfa Newydd Power Station had initially been considered to be fixed type plant but the opportunity to use mobile equipment for a number of processes has not been foreclosed [Ref-102]. A summary of the techniques chosen to be employed following assessment has been provided below:

Solid LLW

The preferred waste processing techniques for solid LLW are [Page 58 of [Ref-122]]:

• Recyclable metals – Transferal to an off-site facility for metal melting. In the UK, the LLWR offers a metal melting and recycling service via its waste services agreement. Alternatively, the metals can be sent directly to a metal melting facility;

• Non-compactible wastes - Direct disposal to the LLWR;

• Filters - Off-site high force compaction in a suitable facility (such as Winfrith); and

• Combustibles - Transferal to an approved and permitted off-site incineration facility.

The National Strategy for the Management of LAW in the UK has undergone a fundamental and wide reaching transformation in recent years. This has been driven by the development of a National Waste Programme under the direction of the Nuclear Decommissioning Authority (NDA) which has been implemented through their subsidiary, Low Level Waste Repository Ltd (LLWR Ltd). This has resulted in significant progress in optimising the use of national assets, especially in preserving the storage capacity at the Low Level Waste Repository and in providing new, innovative and optimised waste routes [Ref-102].

Horizon is therefore involved in ongoing stakeholder engagement with LLWR and other supply chain waste service providers to ensure that Horizon is fully informed on the development of the NWP, and to ensure that Horizon are able to act effectively in the role of intelligent customer [Ref-102].

The above techniques are therefore recognised as currently preferred methods but it is recognised that these may change as the NWP develops and new techniques evolve. Horizon will continually review this in line with its permit conditions [Ref-102].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

WSLLW

It has been concluded by options assessment that the preferred technique for processing WSLLW (resins, concentrated liquid wastes, and activated carbon) is cement immobilisation, using in-line cementation process [Ref-122].

This option was selected in preference to the other identified options (polymer immobilisation, dewatering and compaction, and off-site incineration) for the following reasons [Page 59 of [Ref-122]]:

• Polymer encapsulation has limitations around the maturity of the technique for LLW in the option considered of a polymer injection system. There is uncertainty over the long-term stability of the waste form and its ability to meet the LLWR WAC. Further study of developing work would be required before this could be identified as a preferred solution.

• Further development work would be needed before it can be proven that the WSLLW can meet the WAC for off-site compaction. A drying plant may be required in additional to the dewatering process, to enable the waste to achieve the acceptable moisture content levels for compaction.

• It is estimated for the Hitachi-GE BAT study that the bead resin would not meet the conditions for acceptance for the LLWR incinerator. However, there is the opportunity to dry the material to make it passively safe, and decay store on-site until it is suitable for transfer off-site for incineration.

WSILW

Horizon options assessment concluded that the preferred technique for managing WSILW (resins and crud) is cement immobilisation followed by storage on-site and final GDF disposal [Ref-122]. This option was selected in preference to the highest ranking options of in-container vacuum drying with storage in robust containers and polymer immobilisation which are described below. Due to gaps in information for these two techniques, research and development would be required and the options assessment study revisited to establish greater certainty on which technique represents BAT.

• In-container vacuum drying of WSILW with storage in robust containers – This option was the top ranked option at the study, and is considered the best of the options scored. The total volume of WSILW to be generated for the Power Station will be low, making storage in robust containers a cost effective solution. However, it was recognised that additional development work would be required to substantiate design elements of the packages, such as a lid ventilation and seal system. This option also has a potentially high worker dose penalty associated with maintenance activities; therefore, the option may not be ALARP.

• Polymer (in-drum mixing) immobilisation storage and disposal – This option was ranked third in the study. Polymer immobilisation of ion exchange resins has been successfully implemented at Trawsfynydd. However, polymer immobilisation has limitations relating to the maturity of the technique in processing ILW. There is uncertainty on the longevity of the immobilised waste form, in particular as a result of the potential radiolytic degradation of the matrix over an extended time period.

3.4.2.2 Evidence: Waste Routes for LAW

The waste management techniques for solid dry LLW and WSLLW can produce waste forms viable for consigning to disposal routes. The selected waste routes for VLLW and LLW has been demonstrated using the services provided by LLWR through an agreement-in-principle (3.4.3 Argument 4c: Agreement in Principle for Waste Routes). The LLWR framework comprises four main waste services see Section 3.3.6.1 (Evidence: WAC), as detailed in the Radioactive Waste Management Arrangements document [Ref-116]. The waste routes identified allow for alternative options other than direct disposal to the LLWR, which is consistent with the 2007 UK policy for the long term management of solid LLW [Ref-151].

• Metallic waste processing service;

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• Combustible waste processing service, managed via the WSC (alternatively, combustible waste is not sent via the LLWR, instead it is sent directly to an incineration service provider).

• Super compaction waste processing service; and

• LLW disposal service.

The current assumption is that contaminated oil is not expected to arise from the normal operation of general plant and equipment associated with the Wylfa Newydd Power Station. However, in the unlikely event that a small amount of lightly contaminated oil is generated during non-routine operations, the main credible option is incineration. This could be facilitated by the LLWR via its WSC with Tradebe. The oil would first be separated from any water present to reduce the volume of waste sent for treatment.

3.4.2.3 Evidence: Waste Management System

A WMS can provide structure to activities in packing, transporting and disposing of radioactive waste. The IWS [Ref-70] sets out that Horizon’s WMS will be used to compile and record quality data associated with generation, characterisation, categorisation, processing, storage, transport and disposal of radioactive waste. Information managed by the WMS will be used to inform decision making on which of waste route is suitable for a radioactive waste item. The WMS will be developed to align with the IAEA Safety Guide on The Management System for the Processing, Handling and Storage of Radioactive Waste [Ref-152].

3.4.3 Argument 4c: Agreement in Principle for Waste Routes - LAW Solid and non-aqueous LAW are subject to requirements that the consignor of the waste must fulfil before it can be accepted. Compliance with the WAC is a requirement of the terms and conditions agreed between contracting parties. Compliance with WAC is also a requirement of the NRW's standard environmental permit template for disposals of radioactive waste from nuclear licensed sites as they are 'instructions' given by the person to whom the waste is consigned to. Hitachi-GE engaged with the LLWR on their waste services provided through the LLWR’s common contractual arrangement and obtained an 'agreement-in-principle' for LAW disposal routes (3.4.3.1 Evidence: Agreement in Principle). This process provided assurance that the wastes generated from operating the Wylfa Newydd Power Station (will comply with the LLWR’s WAC for treatment and disposal routes. The agreement in principle covers the following waste routes: • Metallic waste for physical decontamination and recycling; • Combustible waste for volume reduction by incineration; • VLLW for disposal at appropriately permitted commercial landfills; • Compaction of compressible LAW followed by disposal in the national LLWR; and • Disposal of non-compressible LAW in the national LLWR. Since the waste streams to be produced have not changed in characteristics since GDA, the agreements in principle demonstrate that LAW that will be produced by the Wylfa Newydd Power Station are compatible with the range of waste routes and services available in the UK for such wastes. This allows wastes to be disposed of using waste routes (3.4.2 Argument 4b: Optimal Disposal Route Selection) that are identified as BAT. Horizon will have the ability to setup waste contracts for waste management services with the LLWR or with other permitted waste service providers (3.4.3.2 Evidence: Waste Contracts). LLWR Ltd. confirmed during GDA that all of the solid LLW anticipated from an operating UK ABWR was compatible with one or more of the waste management services offered under its WSC. The solid LLW expected at Wylfa Newydd is the same as that for the UK ABWR. Whilst recognising that alternative waste © Horizon Nuclear Power Wylfa Limited 170

Wylfa Newydd Project – Best Available Techniques (BAT) Case management options exist, Horizon has concluded that it will be able to dispose of all of its solid LLW under the WSC offered by LLWR Ltd.

3.4.3.1 Evidence: Agreement in Principle

The LLWR provides a range of waste service treatment options for LAW and operates the LLW disposal site. At GDA Hitachi-GE submitted a waste enquiry form to LLWR to seek agreement in principle from the LLWR for the waste disposal routes identified in Section 3.4.2 (Argument 4b: Optimal Disposal Route Selection). A letter from the LLWR giving ‘Agreement in Principle’ for VLLWs and LLWs was received by Hitachi-GE [Ref-153] and [Ref- 154]. Hitachi-GE re-assessed waste generating activities as presented in this document, following an update of the generic site source term. It was determined that radioactive wastes remain complaint with the LLWR’s WACs [Ref-155]. Horizon’s decision to move from a one UK ABWR unit site to a two unit site will not change the characteristics of radioactive wastes; therefore, this agreement in principle is still considered valid.

3.4.3.2 Evidence: Waste Contracts

The LLWR range of waste services has been developed by the National Waste Programme team to support implementation of the 2007 Strategy [Ref-151]. Horizon could use the services offered by LLWR, although, it is recognised that in the future new arrangements could be selected the optimal disposal route at the time. When operating the Power Station, Horizon will need to make business decisions based on commercial and logistical requirements which may indicate that there are other ways of dealing with specific wastes. For example, Horizon may decide to make arrangements directly with a waste service provider, rather than utilising the LLWR framework contract. Before Horizon can transfer radioactive waste offsite to waste service providers, commercial service level agreements will be needed to setup and approved between Horizon and each permitted waste service provider that Horizon has identified as representing an optimum waste route [FAbc-4].

3.4.4 Argument 4d: Disposability Assessments for Higher Activity Wastes HAW and SF generated in the Wylfa Newydd Power Station will have a level of radioactivity that exceeds acceptance criteria for existing waste disposal routes. Instead, HAW and SF will be stored until the UK GDF is available. Any treatment and storage arrangements must be in accordance with the current management and/or disposal concepts for these wastes. RWM is the source of authoritative guidance regarding management and disposal concepts. RWM undertakes assessments to determine the degree to which proposals for the management of HAW and SF accords with management and disposal concepts developed by them. During GDA Hitachi-GE engaged extensively with RWM during the development of the GDA submission. Disposability assessments were prepared for the following: • Disposal of SF in the proposed national GDF (3.4.4.1 Evidence: Disposability Assessment – Spent Fuel); and • Disposal of HAW in the proposed national GDF (3.4.4.2 Evidence: Disposability Assessment – Intermediate Level Waste). Methods for processing HAW at the Wylfa Newydd Power Station will meet RWM disposability criteria. This allows wastes to be disposed of using waste routes that are identified as BAT. It is recognised that minor alterations to arrangements may be made to ensure BAT is still applied as RWM packaging plans mature. RWM confirmed during GDA that all of the HAW and SF anticipated from an operating UK ABWR was compatible with current management and/or disposal concepts. The HAW and SF expected at Wylfa Newydd is the same as that for the UK ABWR. Horizon has concluded that it will be able to manage and/or dispose of all of its HAW and SF in accordance with such concepts.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.4.4.1 Evidence: Disposability Assessment – Spent Fuel

Hitachi-GE consulted RWM on whether HAW generated by the UK ABWR would be acceptance for disposal at the future UK GDF. Hitachi-GE submitted supporting information on ILW [Ref-34] and SF [Ref-156] to RWM, to allow them to undertake Disposability Assessments. It was reviewed by RWM and issues were identified with regards to the proposed packaging plans. These issues are proposed for addressing as part of future site specific disposability assessment work as they are related to maturing packaging plans, rather than fundamental issues with disposability [Ref-157]. The exception to this is the quantification of Hf-178n which has now been quantified.

The Radioactive Waste Management Arrangements document [Ref-116] describes the characteristics of SF from a Wylfa Newydd Power Station and the options for packaging, storage and disposal. These options are based upon good practice as currently applied in other countries (including one option planned for implementation at Sizewell B in the UK) and what was considered BAT at this time.

High level optioneering on SF storage [Ref-158] concluded that the dry cask storage system should be adopted for GDA. However, it noted that the exact selection of the appropriate option would be decided by Horizon and would be subject to the application of BAT within the decision making process. Horizon undertook a detailed optioneering study which concluded that both dry storage in canisters in vertical concrete over-packs, and dry storage in canisters in horizontal concrete over-packs would be suitable for the storage of spent fuel arising from the Wylfa Newydd Power Station. The SF cask system is projected to be required 11.5 years after the start of commercial operation [Ref-159]. Due to the substantial lead time for when SF will need to be packaged in a cask, the optimum time to schedule a decision to a select a SF cask system has been the subject of a study [Ref-159]. Based on this, suitable casks will be selected through options assessment at a later stage.

3.4.4.2 Evidence: Disposability Assessment – Intermediate Level Waste

At GDA the option of cement encapsulation (for solid items) and solidification (for wet/slurry wastes) into unshielded stainless steel containers was selected as the packaging method to be adopted in the submission to RWM [Ref-34] to produce a Disposability Assessment. It was reviewed by RWM and issues were identified with regards to the proposed packaging plans. These issues are proposed for addressing as part of future site specific disposability assessment work as they are related to maturing packaging plans, rather than fundamental issues with disposability [Ref-157].

The IWS [Ref-70] establishes that that these methods will be employed so there is no requirement for RWM to undertake further assessment. Horizon is now required to engage in the Letter of Compliance (LoC) process to obtain endorsement of the proposed methods.

3.4.5 Argument 4e: Compatibility of Existing UK Waste BAT Studies Selection of appropriate waste routes is an important element of demonstrating that waste management practices form part of an integrated strategy that is focussed on waste minimisation, the application of the waste hierarchy and demonstrating the application of BAT. A series of strategic BAT assessments have been prepared by LLWR at the behest of the NDA to examine the degree to which certain waste routes underpin the development of IWS’ for producers of radioactive waste and support delivery of the waste hierarchy. These assessments adopted a systematic, robust and transparent approach to determining the strategic BAT option for groups of radioactive waste with broadly similar characteristics. During GDA Hitachi-GE undertook a series of assessments to determine the degree to which the findings of these studies are applicable to the following types of waste that will be generated by the Wylfa Newydd Power Station: • Metal waste that has radioactive contamination on its surfaces (3.4.5.1: Evidence: Review of LLWR Metallic Waste Strategic BAT Assessment). • Combustible wastes that are lightly contaminated with beta and gamma emitting radionuclides and/or very lightly contaminated with alpha emitting radionuclides

Wylfa Newydd Project – Best Available Techniques (BAT) Case

(3.4.5.2: Evidence: Review of LLWR Combustible Waste Strategic BAT Assessment). • Waste with very low levels of radioactivity (3.4.5.3: Evidence: Review of LLWR VLLW Strategic BAT Assessment). Horizon considers that the LLWR assessments are applicable to the UK ABWR and that BAT was demonstrated at a strategic level for metallic wastes, combustible wastes and wastes with very low levels of radioactivity. Since these assessments remain unchanged, and because the waste streams anticipated to be generated have not changed, this conclusion can also be assumed for the Wylfa Newydd Power Station. Future waste optimisation and BAT studies on LAW routes will be assessed against waste routes defined in the Wylfa Newydd BAT Case and IWS [Ref-70]. Selected waste routes for solid radioactive wastes anticipated at Wylfa Newydd were assessed against a series of strategic BAT assessments prepared by LLWR Ltd. The treatment methods proposed by Horizon for metallic wastes, combustible wastes and wastes with very low levels of radioactivity are consistent with the findings and the conclusions of the LLWR Ltd. assessments.

3.4.5.1 Evidence: Review of LLWR Metallic Waste Strategic BAT Assessment

The LLWR strategic BAT assessment for the management of metal waste [Ref-160] evaluated options for the management of metallic LLW against a number of safety, environmental, economic, and technical attributes. The assessment concluded that in the short term the best option is waste processing at an overseas metal recycling facility and in the longer term the development and use of UK metal recycling facilities. The report identified a number of advantages associated with this option including:

• Secondary solid waste generation is significantly reduced compared to the on-site or national disposal options.

• The option allows a significant quantity of material to be recycled thereby reducing the overall demand on raw materials and energy requirement of processing raw materials.

Hitachi-GE identified in its waste processing assessment [Ref-115] that recyclable metals will be consigned under the LLWR WSC. Horizon’s IWS [Ref-70] has identified this waste processing route for metallic LLW.

3.4.5.2 Evidence: Review of LLWR Combustible Waste Strategic BAT Assessment

The LLWR strategic BAT assessment for the management of combustible LLW [Ref-161] evaluated options against a number of safety, environmental, economic, and technical attributes. The overall recommendation was to identify a thermal waste treatment service provider with a view to establishing contractual and commercial arrangements for the disposal of this waste form at the earliest opportunity. Additionally it was recommended that the practice of compaction be continued for those waste streams that could not be readily routed for thermal treatment.

The conclusions of the strategic BAT assessment are deemed to be consistent with Horizon’s IWS for managing combustible wastes [Ref-70].

3.4.5.3 Evidence: Review of LLWR VLLW Strategic BAT Assessment

The LLWR strategic BAT assessment evaluated the relative merits of options for the long-term management of VLLW [Ref-162]. The main conclusions of this report were:

• The application of the waste hierarchy to VLLW is supported and reinforced through the study;

• Whilst the performance of the disposal options are broadly comparable;

• Options that represent existing co-disposal facilities generally scored well together with new disposal facilities located on-site or adjacent to nuclear licensed sites;

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• On-site disposal scored better than other new disposal facility options;

• There is little differentiation between new local, regional and national facilities; and

• Use of the LLWR at Drigg scores consistently low.

• Depending on the waste stream, site specific issues and the WAC of any of the options, VLLW may require disposal via an option that does not score highly in this assessment. This may still be consistent with the strategic BAT assessment where specific arguments and justifications can be made.

The conclusions made by the LLWR report are deemed to be consistent with Horizon’s IWS [Ref-70].

Dry active waste is the only VLLW stream identified for the Wylfa Newydd Power Station. It is a mixed waste that will arise during reactor operations and decommissioning. The waste consists of contaminated PPE, monitoring swabs, plastic, equipment, structures and contaminated plant. These wastes will require specific removal, handling, sorting and size reduction techniques depending on their physical form and characteristics prior to treatment. The preferred strategy for this mixed dry active VLLW is to recycle the metallic portion where practicable and to dispose of the remainder, to permitted disposal sites within the UK.

3.5 Claim 5 - Minimise the Impacts on the Environment and Members of the Public from Radioactive Waste that is Disposed of to the Environment The design of the Wylfa Newydd Power Station has focused on reducing the amount of radioactivity in gaseous and liquid wastes that are discharged from the facility. However, where discharges of radioactivity to air and water are unavoidable, techniques have been adopted to ensure that the subsequent impacts to members of the public are ALARA. The Arguments presented in support of this Claim are considered to demonstrate compliance with the standard BAT conditions [Ref-5]: • Condition 2.3.2(c) The operator shall use the best available techniques in respect of the disposal of radioactive waste pursuant to the permit to dispose of radioactive waste at times, in a form, and in a manner so as to minimise the radiological effects on the environment and members of the public. The Wylfa Newydd Power Station design contains a range of features that contribute to the substantiation of this Claim including: • Minimising the impact of discharges to the environment by means of optimising the design and operation of any discharge outlets. In developing the Arguments presented to demonstrate the validity of Claim 5, REPs [Ref-7] have been taken into account. The following REPs are considered to be specifically relevant to this Claim: • Principle RSMDP7 ‘When making decisions about the management of radioactive substances, the best available techniques should be used to ensure that the resulting environmental risk and impact are minimised. • Principle RPDP1 ‘All exposures to ionising radiation of any member of the public and of the population as a whole shall be kept ALARA, economic and social factors being taken into account.’ • Principle ENDP2 ‘Radiological impacts to people and the environment should be avoided and where that is not practicable minimised commensurate with the operations being carried out.’

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• Principle ENDP16 ‘Best available techniques should be used in the design of ventilation systems.’ • Principle DEDP3 ‘Facilities should be designed, built and operated using the best available techniques to minimise the impacts on people and the environment of decommissioning operations and the management of decommissioning wastes.’

3.5.1 Argument 5a: Gaseous Discharge System - Main Stacks Significant efforts have been expended to remove radioactivity from gaseous wastes generated in the Wylfa Newydd Power Station but some radioactivity will be discharged to the environment. The location of the discharge points will have a bearing on the impact to members of the public and the environment. As the gaseous discharges will be continuous, as a result of the OG and maintaining negative pressure in buildings, timing of discharges is not a consideration for minimising the impact. The Wylfa Newydd Power Station will have two main stacks, one servicing each reactor, which will be located on the roof of the R/Bs. These stacks will receive gaseous wastes from the OG and HVAC systems and will be the discharge route for the majority of the gaseous radioactive waste produced at the Power Station. At GDA the location of the main stack was selected because of its proximity to the systems that feed into it, the height provided by the R/B, of which it is part of, and the structural strength of the R/B (3.5.1.1 Evidence: Main Stack - Location). Preliminary modelling of the impacts associated with discharges was undertaken at GDA to demonstrate the relationship between the height of the stack and the impact to members of the public from the radioactivity of the waste that is discharged (3.5.1.2 Evidence: Stack Height Determination – Main stack) and (3.5.1.3 Evidence: Gaseous Discharge – Dilution Factor). The assessment also considered the costs (costs are impacted by relevant factors including seismic requirements, civil engineering and hazards) of the engineering associated with increasing stack-height and explored at what point further increases in the height of the stack becomes grossly disproportionate to the benefits realised from reductions in impacts to members of the public and the environment. Following these initial assessments, Horizon undertook a more detailed assessment which considered site specific factors, included more comprehensive dose and dispersion modelling and considered financial and engineering implications. This assessment concluded that main stack heights of 75m would represent BAT as it provided optimal dispersion (3.5.1.2 Evidence: Stack Height Determination - Main Stack). The design of the Wylfa Newydd Power Station’s main stacks include the provision of equipment that allows for representative sampling of gaseous radioactive waste that is discharged to the environment [Ref-39]. Horizon will be responsible for defining the environmental monitoring programme which will allow the actual impacts on members of the public and the environment from discharges to be retrospectively determined. The concept design of waste facilities (Spent Fuel Storage Facility, ILW Storage Facility and LAWMF) shared by both UK ABWRs in the Power Station site will be developed further and the arrangements for discharges of any gaseous waste will be confirmed. Gaseous radioactive waste which is unavoidably created will be discharged from each reactor unit at Wylfa Newydd to the environment via a main stack. The height of the stacks ensures that the gaseous radioactive waste is diluted and dispersed so that radiological impacts (dose) to members of the pubic and the environment are very low. The cost of increasing the stack heights is grossly disproportionate to any benefit that would be received by members of the public.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.5.1.1 Evidence: Main Stack - Location

The main stacks are located on the roof of the R/Bs [Ref-163]. This location provides the following benefits:

• Minimises the length of duct work required to transfer gaseous radioactive waste between the source and the final discharge point; and

• It benefits from the height and structural strength provided by the R/B. The additional height contributes towards effective dispersion.

3.5.1.2 Evidence: Stack Height Determination - Main Stacks

The main stack heights of the four existing Japanese ABWRs are 57m, 75m, 98m and 100m. For the purposes of GDA, a stack height of 57m was used for preliminary dose modelling [Ref-99].

The height of both Wylfa Newydd Power Station main discharge stacks was determined using appropriate tools and techniques in accordance with the ‘Approach to Optimisation’ [Ref-12]. The optioneering report [Ref-164] concluded that a height of 75m was BAT and would ensure that effective dilution and dispersion of gaseous radioactive waste is achieved, in order to minimise the dose to members of the public and the environment. This was determined by detailed assessment that took account of the following:

• Local site topography

• Wind patterns

• Dispersion modelling taking into account site characteristics (topography, wind patterns etc.)

• Dose to the public and the environment

• Off-site doses due to on-site releases in normal and accidental operation

• On-site doses to workers

• Industrial safety

• Cost aspects

• Design of HVAC systems

• Technical limitations

• Planning requirements, including visual impact

The stack height assessment was governed by the proportionality principle to ensure that optimisation is achieved (the point at which any additional increase of the stack height would incur costs that are grossly disproportionate to the benefits provided).

The stack height optioneering report [Ref-99] demonstrated that:

• The benefits of increasing the main stack heights, which increases radionuclide dispersion and decreases dose to receptors, is matched by the detriments of increasing industrial safety hazards, visual impact and cost.

• Increasing the main stack heights up to 70m improved dispersion with a proportionate increase in cost.

• Between 70m and 80m, the benefit realised starts to diminish with regards to reducing ground level radionuclide concentration against the increasing cost of the main stacks.

• Above 80m it becomes grossly disproportionate to increase the height of the main stacks with regards to benefit gained in terms of ground level radionuclide concentration and dose reduction against the cost of the main stacks.

• Analysis of the dispersion modelling and off-site impacts (Section 4.1.1) demonstrated that optimal dispersion was achieved at 75m.

• Stack heights above 75m would require further structural substantiation which would cause delays of at least 3 months, although it is recognised delays could be considerably longer. Additionally, these programme delays will have associated costs which are likely to be considerable as the delays would likely impact other areas of the design.

3.5.1.3 Evidence: Gaseous Discharge – Dilution Factor

Gaseous radioactive waste with the highest concentration of radioactivity is subject to treatment within the OG. This waste stream has a relatively small volumetric flow compared to the combined discharge from the HVAC systems which have very low concentrations of radioactivity. Following appropriate abatement the gaseous radioactive waste from the OG is therefore subject to considerable dilution prior to being discharged to the environment via the main stacks.

Additionally, the TGS exhaust is combined with the OG prior to being discharged so is subject to the same dilution and subsequent dispersion that is achieved for the OG.

The volumetric flow data is based on the following:

Unit 1

• Reactor area HVAC: 228,285 m3/h [Ref-96];

• T/B HVAC: 353,143 m3/h [Ref-96];

• Rw/B HVAC: 175,547 m3/h [Ref-96];

• Turbine gland air and steam : 3,160 m3/h;

• Driving air for OG ejector: 146 m3/h; and

• OG: 40 m3/h.

Unit 2

• Reactor area HVAC: 228,285 m3/h [Ref-96];

• T/B HVAC: 353,143 m3/h [Ref-96];

• Turbine gland air and steam : 3,160 m3/h;

• Driving air for OG ejector: 146 m3/h; and

• OG: 40 m3/h.

The total volumetric flow from the HVAC, turbine gland steam and OG ejector driving air volumetric flow under normal operations is therefore:

• 760,281 m3/h for Unit 1;

• 584,734 m3/h for Unit 2.

Along with the HVAC systems, each main stack also serves the corresponding OG. The volumetric flow rate from each OG via its corresponding main stack during normal operations is 40 m3/h [Ref-75]. Table 3.5.1.3-1 presents the dilution factor that each unit’s OG is subjected to. This has been calculated using the method in Calculation 3.5.1.3-1.

Table 3.5.1.3-2: Dilution Factor for Gaseous Radioactive Waste from the OG

Volumetric flow from HVAC, Volumetric Dilution turbine gland steam and OG flow from OG factor ejector driving air (m3/h) (m3/h)

Unit 1 760,281 40 approx. 19,000

Unit 2 584,734 40 approx. 14,500

Calculation 3.5.1.3-1 Dilution Factor Calculation

Dilution factor = Vf / Vi Where; Vf: The volume of exhaust air from the HVAC systems, turbine gland steam and OG ejector driving air (m3/h); and Vi: The volume of exhaust air from the OG (m3/h).

The stack flow rate is a major determining factor in the dispersion of radioactive gaseous waste, as such the acceptable parameters for stack flow will be specified in Discharge Criteria. The stack flow will be continuously monitored at the MCR. Minor alert thresholds will be defined in the Discharge Criteria for tolerable variations to the stack flow (e.g. +/- 5%) of the set normal figure. Major alert thresholds will also be defined for significant variations in stack flow (e.g. +/-10%) of the normal figure that require immediate response and resolution. The meeting or exceeding minor and major alert thresholds will initiate alarms at the alarms panels in the MCR and at control cabinet in the stack equipment room. MCR operatives will respond to minor and major levels alarms using Operating Procedures.

3.5.2 Argument 5b: Aqueous Waste Management Significant efforts have been expended to remove radioactivity in the liquid effluent management system and recycle water in the Wylfa Newydd Power Station, but aqueous waste containing low radioactivity will need to be discharged to the environment to manage water balancing in the plant (3.1.8 Argument 1h: Recycling of Water to Prevent Discharges). The design of the Wylfa Newydd Power Station liquid effluent management system includes sampling equipment that allows aqueous waste to be sampled prior to discharge to the environment. This will allow the Horizon to undertake analysis and to confirm that the characteristics of the waste conforms to Discharge Criteria that are based on specific limitations and conditions within the site's EP-RSR permit. The Approach to Sampling and Monitoring report assesses the equipment required to take samples of the discharge and allow Horizon to provide a true and accurate record of the radioactivity discharged to the environment [Ref-39]. This is discussed in more detail in Claim 6. The location and timing of these discharges will have a direct bearing on the impact to members of the public and the environment from operations of the Wylfa Newydd Power Station (3.5.2.1 Evidence: Evidence: Cooling Water Discharge Location). Prior to discharge radioactive aqueous waste will be mixed with the Circulating Water (CW) from the Circulating Water System (CWS) and with water from the Reactor Service Water (RSW) and Turbine Service Water (TSW). The flow rate of the CWS (the largest of the three systems) from each UK ABWR will depend on the operational status of the reactor unit and will be between approximately 8,000m3/hr and approximately 200,000m3/h (3.5.2.2 Evidence: Aqueous Waste Discharges - Dilution).

The design of the liquid effluent management system allows the timing of aqueous waste discharges to be controlled, to take account of any prevailing environmental conditions and regulatory requirements (3.5.2.3 Evidence: Control and Management of Aqueous Waste Discharges). Dose models used to determine the impacts to members of the public and the environment from aqueous waste discharges reflect the generic site description and are insensitive to the position of the discharge provided it is within 10km of the site and within 5km of the shore (3.5.2.4 Evidence: Dose modelling). Following GDA dose modelling, it was decided that the discharge outfall would be located next to the current Magnox station outfall to the west of Wylfa Head approximately 50m from the coastline. Further modelling of this specific discharge location has confirmed that adequate mixing and dilution will be achieved (3.5.2.4 Evidence: Dose modelling) and demonstrates that the impacts of the discharges will be very low. Additionally, radioactive liquid waste generated during decommissioning will be treated, either by using the installed equipment from the operating phase, or new temporary waste treatment facilities (3.5.2.5 Evidence: Liquid Waste Treatment during Decommissioning). This decision will be subject to assessment within the decommissioning BAT Case. Horizon shall be responsible for defining the environmental monitoring programme which will allow the actual impacts on members of the public and the environment from discharges to be retrospectively determined. Aqueous radioactive waste which is unavoidably created will be discharged from Wylfa Newydd to the environment with the CW, RSW and TSW. The design and operation of the liquid effluent management system and CWS ensures that the aqueous radioactive waste is diluted and dispersed so that radiological impacts (dose) to members of the pubic and the environment are very low.

3.5.2.1 Evidence: Cooling Water Discharge Location

The CW, TSW and RSW, along with any liquid waste from the HCW and CAD, will be discharged into the Irish Sea adjacent to the Wylfa Newydd Power Station.

Aqueous waste from the HCW and CAD, are mixed with the CW, TSW and RSW discharge in the Seal Pit prior to being released into the discharge pipe. The outfall of the discharge structure will be located adjacent to the existing outfall of the Magnox station west of Wylfa Head approximately 50m from the coastline [Page 7 of [Ref- 165]]. This location was chosen to ensure that low activity liquid wastes, already diluted by the CW, TSW and RSW, would be further diluted and dispersed in the marine environment.

3.5.2.2 Evidence: Aqueous Waste Discharges - Dilution

The discharge rate of CW, TSW and RSW from the Wylfa Newydd Power Station will depend on the operational status of the reactor units and the abstraction rate of water, which is dependent upon the Wylfa tidal range. Table 3.5.2.2-1 details the CWS, TSW and RSW abstraction rates for the Power Station [Ref-166].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Table 3.5.2.2-1: Cooling Water Abstraction Rates

At Low Astronomical Tide At High Astronomical Tide System on single UK ABWR unit (LAT) (m3/hr) (HAT) (m3/hr) Circulating Water System (CWS) [Ref- 184,800 206,976 166] Turbine Building Service Water (TSW) 7,400 8,288 [Ref-166] Reactor Building Service Water (RSW) 9,000 10,080 [Ref-167]

Subtotal for single UK ABWR unit 201,200 225,344 Total for Power Station (twin UK ABWR 402,400 450,688 units)

CW from the CWS merges with water from the TSW and RSW in the Seal Pit [Ref-168]. Low activity aqueous waste from the HCW system (which also treats CAD which does not meet discharge criteria) will also be discharged into the Seal Pit. The discharge rate of radioactive aqueous waste is calculated to be 575m3/year, per reactor [Page 8 of [Ref-165]]. The aqueous waste is treated by the systems described in Section 3.2.6 (Argument 2f: Configuration of Liquid Management Systems) and HCW should not be discharged except as needed to maintain water balance. The HCW and CAD aqueous wastes will be subject to a very large amount of dilution within the Seal Pit prior to being discharged into the environment.

Additionally, the high flow rate of the discharge plume from the outfall will disperse and dilute aqueous waste with the surrounding sea water [Page 10 of [Ref-165]].

3.5.2.3 Evidence: Control and Management of Aqueous Waste Discharges

Management of discharges will take into account tidal, hydrological and geomorphological features and other factors that could affect the dilution and dispersion of radioactive aqueous waste.

The design of the Wylfa Newydd Power Station liquid effluent management system includes grab sample equipment on sample tanks. This allows the operator to sample and then analyse the characteristics of aqueous waste to confirm it meets Discharge Criteria, before transfer to the Seal Pit. If aqueous waste does not meet its Discharge Criteria it will be recirculated through the HCW treatment system. Low activity aqueous waste will not be discharged from the HCW to the environment except when needed to maintain water balance in the UK ABWR. The controlled release of aqueous waste will occur if the following conditions are met:

• Analytical results of collected samples prove that the aqueous waste meets Discharge Criteria.

• Tidal conditions promote sufficient dilution (low tides, especially Neap tides are sub-optimal).

Timing the discharge of aqueous waste from the outfall will be decided by a SQEP and only when aqueous waste is compliant with Discharge Criteria. The sampling tanks will be sized appropriately to provide sufficient capacity to store aqueous waste during outage periods [Page 8 of [Ref-165]].

3.5.2.4 Evidence: Dose modelling

The EA has developed a simple and cautious assessment methodology of the critical group dose for the prospective assessment of public doses [Ref-169] and [Ref-170], known as IRAT methodology.

Part of this approach involves using site specific data and a detailed modelling tool such as PC-CREAM. This model predicts the transfer of radionuclides in the environment and is the accepted tool for modelling radionuclide dispersion and associated dose uptake when assessing the radiological impact of routine continuous discharges. The PC-CREAM model is based on the assumption that discharges are made into a single, well mixed, local marine compartment represented in the model by a 10km x 10km box extending seaward from the coastline, i.e. the resolution of the model output would not distinguish between a discharge of radioactivity from an outfall located a few metres or a few kilometres from the shore. © Horizon Nuclear Power Wylfa Limited 180

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Modelling of thermal plumes suggests that mixing increases with distance from the shore. The thermal modelling showed that this location offers reduced environmental impacts and provides optimal mixing of thermal energy from the discharge plume [Page 7 of [Ref-165]]. Since thermal energy will disperse into the sea water in the same way as radioactivity, this thermal modelling can be assumed to be indicative of the radioactivity dispersion.

The results of the PC-CREAM modelling used for GDA have demonstrated that the discharges from the Wylfa Newydd Power Station are well below the public dose limit of 1mSv and 0.3mSv from any source [Ref-99] and these discharges have been assumed for the Wylfa Newydd Power Station. It can be therefore assumed that even if there was an approved methodology to achieve the required modelling resolution, such modelling would not show any significant difference in doses between nearshore and offshore shore locations. This is despite the differences in the level of mixing and dilution that occurs at different distances from the shoreline.

To ensure that expected dilution rates are achieved and there is no accumulation of radioactivity near shore, an environmental monitoring programme will be undertaken by Horizon and independently by the Centre for Environment, Fisheries and Aquaculture Science (CEFAS) on behalf of NRW. Such programmes will measure sediments, and seaweeds amongst other environmental indicators around the Power Station. The information gathered during such programmes will determine the level of accumulation of radioactive material directly around the outfall and dose rates to the local environment. If unsuitable conditions are determined during environmental monitoring, investigation will be undertaken to determine the cause. Adequate mitigation measures will be identified and proposed to the NRW.

3.5.2.5 Evidence: Liquid Waste Treatment during Decommissioning

As stated in Section 2.5, at the end of the plant’s lifetime a separate Wylfa Newydd Decommissioning BAT Case will be produced. This is because decommissioning is not anticipated to be performed for the UK ABWR until at least 60 years after COD and it is expected that significant advancements in technology will occur in that time. It is therefore likely that today’s decommissioning technologies, currently considered to represent BAT, will likely not be considered BAT at the time of decommissioning. Additionally, the site end state has not yet been established [Ref-25] and this will influence the decommissioning approach. As such, it is not possible, nor is it appropriate, to definitively state how decommissioning will be managed. However, broad assumptions, based on current best practice have been assumed.

It is assumed that radioactive liquid waste generated during decommissioning will be treated, either by using installed equipment from the operating phase, or temporary waste treatment facilities, depending on which is most appropriate at the time [Ref-25]. The types of radioactive liquid waste anticipated include [Ref-25]:

• Liquid waste resulting from equipment flushing and draining.

• Liquid waste from floor drains.

• Liquid waste from system decontamination.

• Decontamination waste solution after dismantling and decontamination of equipment.

3.6 Claim 6 – Horizon shall Apply BAT When Characterising and Quantifying Gaseous and Aqueous Radioactive Waste Discharges This claim addresses the application of BAT to those systems and processes that are designed to provide a true and accurate record of gaseous and aqueous radioactive waste discharges. Further information and detail on these systems is discussed in Section 6 of the EP-RSR application (Approach to Sampling and Monitoring). The demonstration that BAT is being applied to the selection of techniques that will be used to characterise solid radioactive waste is addressed in Sections 3.3.6 (Argument 3f: Application of Waste Characterisation to Minimise the Volumes of Waste Sent for Disposal) and 3.4.1 (Argument 4a: Provision of Solid Waste Management Facilities). The systems

Wylfa Newydd Project – Best Available Techniques (BAT) Case used to sample and monitor gaseous and aqueous discharges support the fulfilment of the following objectives: • Quantifying gaseous and aqueous radioactive waste discharges to the environment, • Demonstrating performance against notification levels, and • Providing plant performance data. This will allow Horizon to demonstrate compliance with the limits and conditions of the EP- RSR Permit and the associated operational arrangements and criteria. Gaseous and aqueous radioactive waste discharge monitoring will also be used for performance monitoring and to determine if the plant is operating as expected. The discharge monitoring requirements for nuclear installations are defined in the EP-RSR permit conditions [Ref-5]. The most applicable permit condition to this Claim is Condition 3.2.1, which states that an operator shall: • “(a) take samples and conduct measurements, tests, surveys, analyses and calculations to determine compliance with the conditions of this permit; and • (c) Use the best available techniques when taking such samples, conducting such measurements, tests, surveys, analyses and calculations, and carrying out such an environmental monitoring programme and retrospective dose assessment.” The design of the sampling and monitoring systems employed at the Wylfa Newydd Power Station include features that enable representative samples to be collected and measured at the point of final discharge. This is achieved by collecting and analysing representative samples using BAT in order to provide robust data for assessment of the radiological impacts to the public and the environment [Section 6.1.1.2 of [Ref-171]]. In developing the Arguments presented to demonstrate the validity of Claim 6, REPs have been taken into account [Ref-7]. The following REPs are considered to be specifically relevant to this Claim: • Radioactive Substance Management Developed Principle 9 (RSMDP9): Radioactive substances should be characterised using the best available techniques so as to facilitate their subsequent management, including waste disposal. • Radioactive Substance Management Developed Principle 13 (RSMDP13): BAT consistent with relevant guidance and standards, should be used to monitor and assess radioactive substances, disposals of radioactive wastes and the environment into which they are disposed. • Engineering Developed Principle 4 (ENDP4): Environment protection functions under normal and fault conditions should be identified, and it should be demonstrated that adequate environment protection measures are in place to deliver these functions. • Engineering Developed Principle 10 (ENDP10): Facilities should be designed and equipped so that BAT is used to quantify the gaseous and liquid radioactive discharges produced by each major source on a site. • Engineering Developed Principle 14 (ENDP14): BAT should be used for the control and measurement of plant parameters and releases to the environment, and for assessing the effects of such releases in the environment.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.6.1 Argument 6a: Identification of those Parameters in the Gaseous and Aqueous Discharges to be Measured Gaseous and aqueous radioactive waste discharges from the Wylfa Newydd Power Station will be quantified to support the demonstration of compliance with the limits and conditions of the EP-RSR Permit and to determine plant performance. The radionuclides that will be subject to measurement have been selected based on the following: • A review of EC radionuclides of importance to ensure that all radionuclides of international importance were captured in the site limits and monitoring programme (3.6.1.1 Evidence: European Commission recommendation 2004/2/Euratom). • The identification of significant radionuclides based on UK NRW guidance (3.6.1.2 Evidence: Significant radionuclide selection process). • A review of the pollution inventory reporting requirements (3.6.1.3 Evidence: Pollution Inventory). Following this assessment the following radionuclides have been identified as requiring limits and will need measuring either through direct measurement or sampling and analysis: • Gaseous discharges: noble gases, argon-41, iodine, carbon-14 and tritium (3.6.1.2 Evidence: Significant radionuclide selection process). • Aqueous discharges: tritium (3.6.1.2 Evidence: Significant radionuclide selection process). Sampling methods and analysis techniques (Arguments 3.6.3 and 3.6.4) will be used for radionuclides requiring monitoring. The design of the Power Station will provide sufficient space for installing instruments to sample identified radionuclides in discharges, as well as instruments to sample radionuclides not identified as requiring monitoring at this stage of the project (3.6.5 Argument 6e: Space, capability and flexibility for maintenance activities and servicing). A robust and systematic process was applied to identify the radionuclides and groups of radionuclides that should be monitored to demonstrate compliance with the limits and conditions of the expected permit.

3.6.1.1 Evidence: European Commission recommendation 2004/2/Euratom

The EC recommendation 2004/2/Euratom [Ref-173] contains a list of radionuclides of interest arising from gaseous and aqueous discharges released from nuclear power stations in the EC. The groupings given in this document were also used to produce grouped limits. These recommendations were produced to achieve standardised reporting across Europe. The EC recommendation was reviewed at GDA to identify and document the significant radionuclides based upon source term data for normal operations [Ref-174]. The radionuclides included are given in Tables 3.6.1.1-1 and 3.6.1.1-2.

Table 3.6.1.1-1. Radionuclides and groupings from 2004/2/EURATOM for gaseous discharges

Noble gases

Ar-41 Kr-85m Kr-85 Kr-87

Kr-88 Kr-89 Xe-131m Xe-133m

Xe-133 Xe-135m Xe-135 Xe-137

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Xe-138

Particulates (excluding iodines)

Cr-51 Mn-54 Fe-59 Co-58

Co-60 Zn-65 Sr-89 Sr-90

Zr-95 Nb-95 Ag-110m Sb-122

Sb-124 Sb-125 Cs-134 Cs-137

Ba-140 La-140 Ce-141 Ce-144

Pu-238 Pu-239+240 Am-241 Cm-242

Cm-243 Cm-244

Iodines

I-131 I-132 I-133 I-135

Non-particulates

C-14 H-3

Table 3.6.1.1-2. Radionuclides and groupings from 2004/2/EURATOM for liquid discharges

Tritium

H-3

Other radionuclides

Cr-51 Mn-54 Fe-55 Fe-59

Co-58 Co-60 Ni-63 Zn-65

Sr-89 Sr-90 Zr-95 Nb-95

Ru-103 Ru-106 Ag-110m Sb-122

Sb-124 Sb-125 Te-123m I-131

Cs-134 Cs-137 Ba-140 La-140

Ce-141 Ce-144 Pu-238 Pu-239+Pu-240

Am-241 Cm-242 Cm-243 Cm-244

3.6.1.2 Evidence: Significant radionuclide selection process

Regulatory guidance on limits setting [Ref-175] specifies criteria for when NRW would normally set limits to radionuclides and groups of radionuclides. Table 3.6.1.2-1 presents the criteria and subsequent decisions for

Wylfa Newydd Project – Best Available Techniques (BAT) Case determining the significant radionuclides and radionuclide groups that will be subject to limits being set and, therefore, will need to be monitored (See Section 9 of the EP-RSR Application). The applicable gaseous and liquid radionuclides and radionuclide groups in Table 3.6.1.2-1 are based on radionuclides and groupings stated in EC recommendation 2004/2/Euratom [Ref-173] (3.6.1.1 Evidence: European Commission recommendation 2004/2/Euratom).

Table 3.6.1.2-1: Significant radionuclides for gaseous and aqueous discharges

Applicable gaseous Applicable liquid Criterion specified in guidance [Ref-175] radionuclides radionuclides

Radiological impact on people: exceeds 1 μSv None, all radionuclide doses Carbon-14, Tritium per year below 1 μSv per year

Radiological impact on non-human species: None, total dose below None, total dose below 40 exceeds 40 μGy/hour 40 μGy/hour μGy/hour

Quantity of radioactivity discharged: exceeds Argon-41, Carbon-14, Tritium 1TBq per year Tritium

Collective dose: 500-year collective dose None, total collective dose Carbon-14 exceeds 1 man-sievert per year below 1 man-sievert per year

No additional radionuclides No additional radionuclides Constrained under national or international identified as a result of this identified as a result of this agreements or is of concern internationally requirement requirement

No additional radionuclides have been identified that Indicators of plant performance Noble gases, iodine would act as indicators of plant performance

Beta/gamma radionuclides were initially identified as applicable for both gaseous and aqueous discharges. However, discharges from the Power Station are likely to be at levels that will not be detectable within samples taken for measurement. Beta/gamma radionuclides will make a contribution to the dose experienced by a member of the public but this contribution will be very small (for the critical group the For the appropriate generic categories from contribution from combined aqueous and gaseous the EP-RSR Pollution Inventory to limit any beta/gamma discharges will be in the order of 10-5 µSv/y radionuclides not otherwise covered by the (see Section 7 of the EP-RSR Application). Additionally, limits set on the above criteria. these doses are also below the threshold of 1 µSv/y used as one of the criteria for setting discharge limits [RD32]. It is therefore proposed that discharge limits are not set for either of these radionuclide groupings but that BAT be applied to minimise their discharge to the environment (see Section 7 of the EP-RSR Application). Design elements employed which contribute to the application of BAT in this area are discussed in the relevant sections above.

3.6.1.3 Evidence: Pollution Inventory

Pollution inventory reporting on annual emissions to air and water is required if an operator wishes to discharge radioactive waste to air, water or sewers covered by a permit issued under the EPR16. This helps the NRW to © Horizon Nuclear Power Wylfa Limited 185

Wylfa Newydd Project – Best Available Techniques (BAT) Case meet its national and international environmental reporting commitments such as the UK Radioactive Discharge Strategy [Ref-110].

The pollution inventory can be completed using measurements, calculations and estimations. Measurements are direct monitoring results of a radionuclide from a given discharge route, including, spot or flow-proportional sampling of aqueous waste in a tank and subsequent analysis. Calculations can be made based on specific plant data or OPEX. Estimates can be made based on best estimates for example, arriving at figures for releases to the air based on previous experience of the likely percentage of activity released from a particular facility in a year [Ref-138].

As a part of the limit setting process one of the criterion used to determine the key radionuclides was those that are included in the appropriate generic categories from the EP-RSR Pollution Inventory (e.g. “alpha particulate and “beta/gamma particulate for discharges to air). The applicable radionuclide groups that were found to fit this criteria were beta/gamma particulate for gaseous discharges and beta/gamma radionuclides for liquid discharges. However, as stated in Table 3.6.1.2-1, discharges of beta/gamma emitters is expected to be very low such that it is proposed that limits are not set against this group.

For both liquid and gaseous discharges it was likely that alpha radionuclides will be below the limit of detection [Ref-171].

3.6.2 Argument 6b: Identification of Final Radioactive Discharge Locations and their Sampling Points The final discharge points at the end of each of the gaseous and aqueous radioactive waste discharge pathways have been identified. The discharge points for gaseous radioactive waste is via the two main stacks which are located on the top of the R/Bs (3.5.1 Argument 5a: Gaseous Discharge System - Main Stack). Feeds to the main stacks are: • Gaseous effluent from the HVAC systems, with the exception of the S/B HVAC as this will have its own minor discharge route (see below) (3.2.3 Argument 2c: Heating, Ventilation and Air Conditioning System). Note: Rw/B HVAC will be discharged exclusively via the unit 1 stack. • Gaseous effluent from the OG (3.2.1 Argument 2a: Off-Gas Waste Treatment Facility) which includes effluent from the TGS (3.2.5 Argument 2e: Optimisation of the Turbine Gland Seal). Final monitoring will be conducted from the sampling point at an elevated location on each stack. At these points all gaseous radioactive waste feeds will have converged and mixed allowing representative samples to be taken before final discharge to atmosphere (3.6.2.1 Evidence: Final Gaseous Discharge Point and Discharge Pathway). Gaseous radioactive waste discharges from the S/B HVAC are considered to be a minor source and are not routed through the main stacks. Instead they are discharged through a local vent situated on the S/B roof; this discharge will be monitored separately. The precise plant and equipment to be housed in the S/B, which will ultimately influence the generation of gaseous radioactive waste, have not yet been confirmed [Ref-39]. These monitoring arrangements will be decided by Horizon following confirmation of the plant and equipment to be employed (See Section 9 of the EP-RSR Application). The HCW and CAD components of the liquid effluent management system are designed to enable aqueous waste to be discharged to the Irish Sea (3.5.2 Argument 5b: Aqueous Waste Management). The HCW and CAD are connected to a common discharge pipe and will be monitored prior to discharge to the marine environment (3.6.2.2 Evidence: Final Liquid Discharge Point and Discharge Pathway).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

In choosing the location of the sampling points for discharges, consideration will be given to access requirements (3.6.5 Argument 6e: Space, capability and flexibility for maintenance activities and servicing). The distance between the sample extraction point and the location of any sampling trains or analytical instruments will be minimised to enable representative monitoring of radionuclides (Arguments 3.6.3 and 3.6.4). Horizon has identified the most appropriate locations for taking samples of gaseous radioactive discharges in the main stacks and of aqueous radioactive waste prior to discharge into the Irish Sea.

3.6.2.1 Evidence: Final Gaseous Discharge Point and Discharge Pathway

Gaseous effluent, with the exception of HVAC air from the S/B, during normal operation is discharged through the main stacks. The main stacks are cylindrical structures with a height of 75m and a consistent inner diameter of 3.1m. Gaseous effluent from the HVAC and OG enters the main stacks through two lines located at the base of the stacks [Ref-39].

Feeds to both stacks will be the same with the exception of the Rw/B HVAC which will be discharged solely through the unit 1 stack. This was decided because the Rw/B is required to be operational from unit 1 COD and at this point unit 2 will not be complete [Ref-102]. Revised human dose assessments have been undertaken to reflect the revised feeds and these have demonstrated that, from a dose perspective, the concept of ALARA is not challenged by the decision [Ref-172].

Samples of discharges are extracted at locations near the top of the stacks. Sampling equipment including sampling pipework, pitot tubes and the regulator’s independent monitoring equipment (3.6.6.1 Evidence: Independent Sampling of Gaseous Waste Discharges) will be installed in these locations. The operators monitoring equipment will be located within the reactors, in close proximity to the corresponding main stack. Figures 3.6.2.1-1 and 3.6.2.1-1 summarise the feed systems and discharge pathways for gaseous effluent at each unit [Ref-39].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.6.2.1-1: Overview of the Gaseous Discharge Pathways at Wylfa Newydd ABWR Unit 1

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.6.2.1-1: Overview of the Gaseous Discharge Pathways at Wylfa Newydd ABWR Unit 2

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.6.2.2 Evidence: Final Liquid Discharge Point and Discharge Pathway

The CAD and HCW sub-systems of the liquid effluent management system [Ref-110] are connected to a common discharge line. Final monitoring of aqueous radioactive waste will be conducted by sampling instrumentation located after the two sub-systems converge, at a location close to the end of the discharge pipe, where they are ultimately discharged into the Irish Sea (illustrated in Figure 3.6.2.2-1). The liquid effluent within the LCW system is not discharged and is instead transferred to the CST for recycling (3.1.8.4 Evidence: LCW Treatment System).

Figure 3.6.2.2-1: Overview of Liquid Discharge Pathway (Blue Circle Illustrates Monitoring of liquid waste before discharge)

3.6.3 Argument 6c: Sampling Methods and Analysis Techniques for Determination of Gaseous Radionuclides Activity Concentrations and Volumetric Flow Rates The gaseous sampling and monitoring system is designed to be operable whenever gaseous wastes are being discharged to the environment [Ref-39]. Since gaseous discharges are continuous, the sampling and monitoring system will continuously sample for the parameters identified in Section 3.6.1 (Argument 6a: Identification of those Parameters in the Gaseous and Aqueous Discharges to be Measured). To be able to accurately report the total discharge of radioactive material from discharge points, the volumetric flow of gaseous effluent streams will need to be continuously

Wylfa Newydd Project – Best Available Techniques (BAT) Case

monitored. This will be achieved using appropriate techniques to measure the flow rate and discharge time which will be compliant with the latest appropriate standard e.g. MCERTS. The gaseous discharge sampling system extracts samples from two vertical stretches of pipe, one in each main stack, which avoids gravitational influences that may cause larger particles to deposit on a horizontal section. 3.6.3.1 (Evidence: Sampling Point within the Main Stack). The sampling points are additionally located away from bends, branches, obstructions, fans and leaks, all of which can cause undesirable variations in the velocity profiles (3.6.3.2 Evidence: Air Velocity Profile). The gaseous discharge monitoring system will contain gas sampling probes to obtain accurate flowrates from the gaseous waste stream to ensure representative samples are taken (3.6.3.3 Evidence: Obtaining Representative Samples). Individual elements of the gaseous discharge monitoring system will be configured in such a way to ensure the most representative samples are obtained (3.6.3.4 Evidence: Main Stack Sampling System Configuration). The following techniques will be employed to capture the radionuclides of interest, subject to confirmation prior to procurement that these techniques remain BAT (See Section 9 of the EP-RSR Application): • Filters to capture particulates, which will then be analysed (3.6.3.5 Evidence: Sampling Methods); • Solid adsorbent material to capture iodine (charcoal filters) which will be analysed (3.6.3.5 Evidence: Sampling Methods); • Fixed volume calibrated chambers and appropriate detector systems to measure for noble gases (3.6.3.5 Evidence: Sampling Methods); and • Bubblers to collect samples of tritium and carbon-14 which will be analysed (3.6.3.5 Evidence: Sampling Methods). In addition to the main sampling location, noble gases, excluding Ar-41, are measured within the OG lines before they enter the main stacks (3.6.3.6 Evidence: Sampling Off-Gas for Noble Gases). Sampling lines have been designed to be as short as possible with minimal bends (3.6.3.7 Evidence: Sampling Lines). There will be a single sampling line within each of the two main stacks which will be used to monitor the samples. The sampling systems each have a duplicate which will operate in parallel providing adequate contingency should one system fail. A redundant system will not be permanently installed within the stack for monitoring flow (3.6.3.8 Evidence: Redundancy of Systems). The employment of the following analytical techniques is proposed for determining the activity concentration or flow of samples. These techniques have been selected to meet the minimum EC detection limits (3.6.3.9 Evidence: Detection Limits for Analytical Techniques and 3.6.3.10 Evidence: Proposed Analytical Methods for Key Radionuclides): • Tritium - Liquid scintillation counting; • Carbon-14 - Liquid scintillation counting; • Cobalt-60 - Gamma spectroscopy; • Strontium-90 - Liquid scintillation counting; • Caesium-137 - Gamma spectroscopy;

Wylfa Newydd Project – Best Available Techniques (BAT) Case

• Iodine-131 - Gamma spectroscopy; • Noble gases - NaI(Tl) scintillator; and • Total-alpha - Alpha spectroscopy or Inductively Coupled Plasma – Mass Spectrometry. It is expected that methods and techniques will develop over time and the most suitable technique will be driven by the latest standard e.g. MCERTS (or equivalent) and the available offerings in the sampling and analytical market at the time of commissioning. These proposals will therefore be reviewed to ensure that the most up to date guidance and standards are taken into account and that the techniques and instrumentation to be used are BAT (See Section 9 of the EP-RSR Application). This will be done early enough to enable the design of the laboratory and the procurement of the equipment prior to commissioning of the Power Station. Horizon will deploy sampling and analytical techniques for gaseous radioactive discharges that comply with applicable standards, reflect RGP and fulfil specified regulatory requirements.

3.6.3.1 Evidence: Sampling Point within the Main Stacks

Each main stack will have a sampling point within which will be located approximately 10 hydraulic diameters (10D) downstream from the last input (the upper edge of the HVAC duct) and approximately 3 hydraulic diameters (3D) upstream from the exit of the stack. This arrangement ensures that the air within the stacks is well mixed ensuring that any sample collected is representative of the final discharge, as specified in BS ISO 2889 [Ref-176]. The flow will also be measured in the same plane [Section 8.2.1 of [Ref-39]]. The exact number and locations of stack flow measurement instruments will be determined during commissioning and it is expected that instruments such as pitot tubes will be used for this purpose (See Section 9 of the EP-RSR Application) [Ref-39]. Figure 3.6.3.1-1 shows an overview of the location of the sampling point within the main stack.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.6.3.1-1: Location of Sampling Equipment within the Main Stack

3.6.3.2 Evidence: Air Velocity Profile

The air velocity profile will be determined during commissioning, as described within BS ISO 2889 [Ref-176], to show that the coefficient of variance is less than 20% across the centre two thirds of the stack [Section 8.2.1 of [Ref-39]. Testing and commissioning reports will record and baseline the air velocity profiles at points within the system before operation (See Section 9 of the EP-RSR Application).

3.6.3.3 Evidence: Obtaining Representative Samples

In each main stack a representative sample is continuously extracted through an isokinetic probe. The sample passes through filters or adsorbers in a dedicated sampling equipment room and is then returned to the main stack. The isokinetic probe, sampling pipe, and return pipe are common for both systems [Section 7.3 of [Ref-39].

There are additional pumps which control the total sampling flows to ensure they are constant in order to ensure isokinetic sampling conditions. Within each stack, when one sampling line is isolated for sample collection or maintenance, the flow rate on the non-sampling line is increased to keep the total sampling flow constant [Section 8.2.1 of [Ref-39]].

The sampling lines will be made of stainless steel with smooth inner surfaces to minimise deposition and ensure longevity of the lines. In order to avoid vapour condensation, the sample temperatures are kept equal to or above that of the corresponding stack flow. Almost all sections of the pipework are located within the inside of the stacks to make sure the sample temperature is equal to that of the corresponding stack flow. Where the pipes are located outside of the stacks, the pipes are heated to above the dew point [Ref-39].

Commissioning the systems will require HVAC specialists to perform velocity measurements at appropriate cross section points to determine that laminar flow is sufficient for air sample collection at the pitot tubes (See Section 9

Wylfa Newydd Project – Best Available Techniques (BAT) Case

of the EP-RSR Application). The system’s Commissioning Report will record air velocity and will be used to set the air sample extract flow rate at the pitot tubes as an identical extract flow rate is necessary for isokinetic sampling.

3.6.3.4 Evidence: Main Stack Sampling System Configuration

The samples are monitored in an order that ensures that a suitable sample is obtained by each instrument (see Figure 3.6.3.4-1). Particulates are collected first to minimise losses through plating out. Once particulates have been removed the sample is passed through an appropriate iodine adsorber. Typically these samples only contain several tens of millimetres of charcoal resulting in there being only a short delay through them before the sample is passed into the gas chamber for noble gases analysis. This arrangement complies with BS EN 60761-3 [Ref-177]. Tritium and carbon-14 are collected on a different line and the order will be determined by the specific instrument purchased prior to commissioning [Ref-39] (See Section 9 of the EP-RSR Application). The sampling pumps are located downstream of all sampling and monitoring equipment [Ref-39].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Figure 3.6.3.4-1: Main Stack Sampling System Configuration

It is good practice to return sampled gases downstream of the sample extraction locations to prevent either double counting or dilution of the samples [Ref-178]. However, the return points are proposed to be located upstream of the sample extraction points as shown in Figure 3.6.3.4-1. This arrangement could save over 30m of

Wylfa Newydd Project – Best Available Techniques (BAT) Case

pipe per stack [Ref-39], which in turn will reduce the volume of potentially radioactively contaminated steel that needs to be disposed of at decommissioning.

The impact of returning the gases upstream of the sampling locations has been determined to be negligible. During normal operations, the gaseous flow rates through the unit 1 and 2 main stacks will be 760,321 m3/h and 584,774 m3/h respectively, while the maximum sampling flow rate is assumed as 24 m3/h (this is a maximum value to be conservative, with the exact value be determined at a site license permission stage). This gives mixing ratios of 1:32, 500 and 1:26,000 which would have insignificant impacts in terms of both dilution of sample and double counting. It is therefore concluded that it is BAT to have the sample returned upstream of the sampling location as the benefit of saving pipework outweighs the impact on the representativeness of the sample [Ref-39].

3.6.3.5 Evidence: Sampling Methods

Sample collection allows samples to be accumulated over a time period which produces samples that have a higher likelihood of producing a detectable result than continuous monitoring achieves. Exact timescales for sample collection will depend on the equipment available at the time of purchase, in conjunction with the analytical method employed [Ref-39]. The Radiation Protection procedure on the Collection and Monitoring of Samples will specify the periodicity for Health Physics to change particulate air sample papers and collect samples. Samples filters - Particulates Particulate material will be collected on appropriate filter media for laboratory analysis. Particulate sampling equipment will be situated at a location as close as is possible to the HVAC and OG inputs to the main stacks to minimise plate out. Any other equipment (e.g. flow meters) will be placed downstream of the particulate sampling equipment except for valves needed for isolation [Ref-39]. The sample flow sensors will be placed downstream of the particulate filters.

Solid adsorbent - Iodines

In addition to any particulate iodine, other chemical forms of iodine will be collected using appropriate solid adsorbent material. This is expected to be in the form of a charcoal filters [Ref-39].

Gas chambers - Noble gases In each main stack, noble gases will be continuously monitored through the use of a fixed volume calibrated chamber and appropriate detection system(s). The radiation detector assemblies will consist of a detector housed in a shielded gas chamber. Checking sources will be contained as necessary. A radiation monitoring unit in the MCR will analyse and visually display the measured radiation levels. If the system detects a high radiation level, it will activate an alarm in the MCR to warn operators. The gas chambers will be purged with ambient air when a background level measurement is made [Ref-39]. The gases will be collected and analysed after the removal of particulate material and iodine [Ref-177]. Tritium and carbon-14 will not be removed from the sampled gases as it does not affect the gamma measurement because of their low energy beta emissions; however, they will be collected on a separate sample pipeline using different instruments [Ref-39].

Bubblers - Tritium and carbon-14 Within each main stack, on a separate line, bubblers will be employed to collect samples for analysis by an appropriate analytical technique for the determination of tritium and carbon-14. Bubblers will be employed in series to capture of as much analyte as is possible to ensure accurate measurement. The exact composition of the bubbler solutions will be determined at a later date by Horizon to ensure that BAT is being applied [Ref-39].

Sample flow Each sample pipeline will be set to an optimum sample flow rate value during commissioning. The sampling flows will be constant to ensure efficient sample collection and sufficiently high to ensure the required detection limits are achieved [Ref-39]. An acceptable range will also be set for the sample flow rates to vary within, which will not be detrimental to the performance of sampling techniques in achieving true and accurate measurements. If the

Wylfa Newydd Project – Best Available Techniques (BAT) Case

sample flow rate of a line drifts outside the lower or upper threshold values, the system will automatically alarm at the local panel and in the MCR. Volumetric flow will be measured for each sampling line to calculate activity concentration for each radionuclide. Calibration of samplers, detectors and flow meters will be routinely carried out at a suitable periodicity (likely to be annual) that adheres to the recommendations of the manufacturer’s specifications and commissioning report for the system. Servicing and calibration of equipment will be routinely carried out at the frequency and periodicity specified in the Maintenance Schedule. The maintenance regime for sampling and monitoring instrumentation will be based on recommendations from the manufacturer and in the commissioning report. Servicing and calibration of instrumentation will be in accordance with Horizon’s internal procedures or the Horizon verified procedures of the contracted servicing organisation.

3.6.3.6 Evidence: Sampling Off-Gas for Noble Gases

The only source of noble gases discharged via the main stacks is from the OG systems [Ref-179]. As such, noble gas sampling equipment will be placed within the OG systems before they join the main stacks, to prevent the dilution of material by the HVACs and to allow the detection of the noble gases at the appropriate level. The air at these points will be well mixed due to passing through the charcoal adsorber and filter and because after these points there are no additional air inputs. This ensures representative noble gas samples can be collected [Ref-39].

3.6.3.7 Evidence: Sampling Lines

Pipework is required to be of minimum length, containing minimal bends between the sampling points and sampling equipment [Ref-176] and [Ref-178]. The sampling equipment rooms are therefore placed at the nearest possible location to the main stacks [Ref-39]. The sampling lines have been designed to meet the standards required by the BS ISO standards. This is described in Table 3.6.3.7-1. [Ref-39].

Table 3.6.3.7-1. Standards Alignment in plant design

Reference Requirement/recommendation Wylfa Newydd Power Station

BS ISO 2889:2010 It is necessary that the flow through each An isokinetic probes will be used. [Ref-176] nozzle be proportional to the local velocity, so as to make the combined sample representative, and to make sampling nearly isokinetic.

BS ISO 2889:2010 The sampling nozzles should be checked The isokinetic probes can be [Ref-176] periodically for alignment, presence of extracted to the outside of the deposits of foreign materials and other factors main stacks for regular visual that can degrade the performance of the inspection and cleaning. sampling system.

BS ISO 2889:2010 The performance of the sampling system shall Modelling studies shows [Ref-176] be considered sufficient under normal, off- penetration value of particles normal and anticipated accidental conditions, exceeds 50%. if a test with near monodisperse particles yields a penetration value above 50%.

BS ISO 2889:2010 The straight sections of transport tubes, The best available location was [Ref-176] particularly horizontal tubing sections, should selected with shortest pipework be kept as short as possible, and the number and minimum bends.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

of bends should be minimised within the geometrical constraints of the application.

BS ISO 2889:2010 Bends should have a curvature ratio of at From the isokinetic probes to the [Ref-176] least 3. particulate filters, the bends have a curvature ratio over 4.

BS ISO 2889:2010 Materials recommended for the nuclear The design of the sampling pipes [Ref-176] industry are stainless steel for general is stainless steel. applications.

BS ISO 2889:2010 Sample transport lines, collectors and The sampling lines will be located [Ref-176] analysers should be designed to avoid inside the main stacks as far as condensation of vapour. possible and when they exits the stacks they will be heated.

BS EN 60761-1:2004 The pumps shall be placed downstream from All sampling pumps are located [Ref-180] any filters or activity measuring units. downstream of all sampling and monitoring equipment.

BS EN 60761-1:2004 The pumps shall be capable of continuous There is redundancy built into the [Ref-180] operation between scheduled maintenance system to allow continuous operations. operations during maintenance.

BS EN 60761-1:2004 The pumps shall allow total air flow-rates The total air flow rates are [Ref-180] which are adequate for the measurement controlled to be constant in order method. to keep the isokinetic sampling condition.

BS EN 60761-3:2004 An appropriate device shall be placed at each Noble gas is measured [Ref-177] of the sampling circuit inlets to remove any downstream of the particulate and particulates and iodine, when necessary, from iodine sampling (removal). the air.

3.6.3.8 Evidence: Redundancy of Systems

In each main stack, a duplicate gaseous sampling system will be installed alongside the main sampling system. Having two sampling systems running in parallel in the design mitigates against the loss or destruction of samples. Furthermore, the operation of the two sampling systems (A and B) will be staggered; if the sampling periods are set at two week intervals, each system will be offset in turns, e.g. System A will sample weeks 1 and 2 and System B will sample weeks 3 and 4 [Ref-39].

To provide back up for the gaseous flow measurements a secondary identical flow monitoring system will be stored outside of each stack. A redundant system will not be permanently installed within the stack to prevent both systems becoming damaged in the event of a lightning strike. It is estimated that the time taken to install the spare system would be a no more than two days. There are additional flow measurements recorded for each main stack within both the HVAC and OG as shown in Figure 3.6.2.1-1. These could be used as supplementary information during periods where the main flow readings are unavailable. During normal operations it is proposed that the operator records these values to show how they compare to the main flow measurements, which will create a robust predictive model to calculate the volumetric flow rates if the primary systems fail.

It has also been shown that the flow rate during normal operations is relatively stable, so it is concluded that the maximum downtime of two days with the supplementary information and stability of flow would be acceptable [Ref-39].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

3.6.3.9 Evidence: Detection Limits for Analytical Techniques

Detection limits of analytical tests are dependent on a wide range of variables. For gaseous monitoring specifically, sensitive parameters include the flow rates and sampling periods.

All samples will be analysed using a laboratory (whether internal or supply chain) that meets the relevant quality standards and holds the necessary environmental permits (See Section 9 of the EP-RSR Application). The laboratory will demonstrate that they are able to meet the required detection limits and ensure the use of a robust and consistent methodology for the estimation of detection limits. The required detection limits are provided in Table 3.6.3.9-1 [Ref-39]. The detection limit of a technique can be affected by a range of parameters, to mitigate this Horizon will use an MCERTS accredited (or equivalent) method where available.

Laboratory reports presenting radionuclide measurements will need to specify the detection limits for the analytical tests employed. The Company method of recording and reporting radionuclide measurements of gaseous discharges would be set out in the Gaseous Discharge Sampling & Monitoring procedure.

If the radionuclide specific measurement is below the detectable limits of the instruments, the activity will be calculated by measuring the total activity of the discharge and applying a suitable fingerprint (such as that of the source term) [Ref-39].

Table 3.6.3.9-1: Key Nuclides and Requirements for the Detection Limit in EU 2004 (Gaseous Discharges)

Requirement for the detection Category Key nuclides limit (in Bq/m3)

Noble gases Kr-85 1E+04*

Co-60 1E–02

Sr-90 2E–02

Cs-137 3E–02 Particulates (excluding iodines) Pu-239 + Pu-240† 5E–03

Am-241 5E–03

Total-alpha 1E–02

Iodines I-131 2E–02

H-3 1E+03 Non-particulates C-14 1E+01

3.6.3.10 Evidence: Proposed Analytical Methods for Key Radionuclides

Specific techniques have not been decided on at this point in the Wylfa Newydd Power Station project as it is recognised that as technology will evolve. The exact techniques to be employed will be decided by Horizon prior to commissioning to ensure that any methods used are considered BAT at the time of commissioning taking into

* Can normally be obtained by beta-measurement after decay of short-lived isotopes.

† Total-alpha should only be reported if nuclide-specific information on alpha-emitters is not available.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

account any technical improvements that may occur in the intervening time (See Section 9 of the EP-RSR Application). The selection of techniques will take into account BS EN ISO/IEC 17025 [Ref-181] and the availability of MCERTS (or equivalent) accredited equipment on the analytical market at the time of commissioning [Ref-39]. At this stage it is important not to constrain future selection of analytical techniques. Therefore the design incorporates a suitably equipped and sized room to house sensitive analytical equipment and allow flexibility of techniques (3.6.5 Argument 6e: Space, capability and flexibility for maintenance activities and servicing).

The methods listed in Table 3.6.3.10-1 are the methods currently proposed for analysing gaseous discharge samples for the key radionuclides identified.

Table 3.6.3.10-1: Currently Proposed Analytical Methods to Determine the Activity Concentration of Key Nuclides in Gaseous Discharges

Key nuclides Analytical method

H-3 Liquid scintillation counting

C-14 Liquid scintillation counting

Co-60 Gamma spectroscopy

Sr-90 Liquid scintillation counting

Cs-137 Gamma spectroscopy

Iodine Gamma spectroscopy

Noble gases (including Ar-41) NaI(Tl) scintillator

Alpha spectroscopy or Inductively Coupled Plasma – Total-alpha* Mass Spectrometry

More information on analytical techniques and estimated activity of samples is provided in [Ref-39].

3.6.4 Argument 6d: Sampling methods and analysis techniques for determination of liquid radionuclide activity concentrations and volumetric flow rate The liquid effluent management system comprises sub-systems; LCW, HCW and CAD, with water from the LCW being reused on site. Liquid radioactive waste that is due to be discharged will be held in the sample tanks. A representative sample will be collected from the tanks for the purposes of reassurance monitoring to ensure the waste meets the criteria specified prior to discharge. Following reassurance monitoring the treated liquid radioactive waste from the HCW and CAD will be discharged by a pump on a batch wise basis.

* Can normally be obtained by beta-measurement after decay of short-lived isotopes.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Final discharges of aqueous waste shall be subject to proportional sampling with an appropriate flow meter and flow proportional sampler arrangement (3.6.4.1 Evidence: Flow Proportional Sampler). Discharges are sampled downstream of the confluence of two discharge lines to ensure that measurements taken are representative of the final discharge and that no additional radioactive material can enter the system downstream of the sampling location (3.6.4.2 Evidence: Discharging in Batches). To be able to accurately report the total discharge of radioactive material from release points, the volumetric flow of aqueous waste will need to be continuously monitored. The exact volume of sample collected per unit volume of discharge will be determined by two factors; the total volume discharged and the laboratory volume requirements for sampling. A sampler that is capable of obtaining a range of volumes will therefore be selected (3.6.4.3 Evidence: Discharge Flow Measurement). The following analytical techniques are proposed for determining the activity concentrations of the key radionuclides identified. The techniques have been selected to ensure that the minimum detection limits outlined in the European Commission Recommendation [Ref-173] are achieved (3.6.4.4 Evidence: Detection Limits for Analytical Techniques and 3.6.4.5 Evidence: Proposed Analytical Methods for Key Radionuclides): • Tritium - Liquid scintillation counting • Cobalt-60 - Gamma spectroscopy • Strontium-90 - Liquid scintillation counting • Caesium-137 - Gamma spectroscopy • Plutonium-239 + Plutonium-240 - Alpha spectroscopy • Americium-241 - Alpha spectroscopy • Total-alpha - Alpha spectroscopy It is recognised that sampling and analytical techniques may develop over time and therefore what is considered to be BAT may change. Those techniques that are chosen will be considered to be BAT at the time of commissioning taking into account MCERTS (or equivalent) available offerings in the sampling and analytical market at the time. The proposals outlined in the previous paragraph will therefore be reviewed prior to commissioning of the Power Station to ensure that the most up to date guidance and standards are taken into account and that the techniques and instrumentation to be used are BAT (See Section 9 of the EP-RSR Application). Horizon will deploy sampling and analytical techniques for aqueous radioactive discharges that comply with applicable standards, reflect RGP and fulfil specified regulatory requirements.

3.6.4.1 Evidence: Flow Proportional Sampler

Samples are collected from the final discharge line during a discharge using a proportional sampler to give an accurate record of what is finally discharged. In order to do this the flow of the discharge at the sample location will also be measured [Ref-39]. There will be two flow proportional samplers in the system to allow for independent sampling and to provide a contingency.

The liquid waste will be discharged by pump which will class the flow proportional sampling system as pressurised. At present, pressurised systems are not covered by the MCERTS, but this is likely to be brought

Wylfa Newydd Project – Best Available Techniques (BAT) Case

within the scope of MCERTS in the near future. Future engagement with manufacturers will ensure that equipment will have the capacity to take a representative sample from a pressurised system [Ref-39].

3.6.4.2 Evidence: Discharging in Batches

HCW and CAD are discharged in batches. Both subsystems will have a pair of holding tanks. Once a holding tank is ready for discharge it will be isolated to prevent further ingress of liquid waste [Ref-39].

Each holding tank has a recirculation system which is used to ensure the contents are well mixed and a representative grab sample can be collected for compliance purposes, ensuring the waste meets the criteria for discharge prior to authorisation. Agitating the contents of the holding tanks through recirculation is a pre-requisite to collecting a grab sample and will be prescribed in arrangements for managing aqueous wastes. Aqueous waste will be subjected to additional monitoring at the final discharge point [Ref-39].

Following verification that the grab sample results are compliant with the Discharge Criteria, the holding tank will be emptied to the Seal Pit for mixing with the large volume of water from the CWS, TSW and RSW and from there it is discharged to the Irish Sea [Ref-165]. Discharges are subject to proportional flow sampling and continuous activity monitoring. Should the continuous monitor detect radiation exceeding a predetermined level, the isolation valve will automatically close, activating an alarm in the MCR, resulting in the cessation of discharges [Ref-39].

3.6.4.3 Evidence: Discharge Flow Measurement

A flow meter will be situated downstream of the confluence of the two discharge lines (CAD and HCW), but prior to the discharge point, to ensure that the measurement is representative of the final discharge. The discharge flow rate will be measured using an MCERTS (or equivalent) approved flow meter. There are currently nine different suppliers providing MCERTS certified flow meters for liquid discharges [Ref-182]. Whilst the specific instrument to be employed has not been specified at this stage, in recognition of the fact that BAT continually evolves as technology does, a commitment has been made to use an MCERTS (or equivalent) accredited technique [Ref- 13]. Using an MCERTS (or equivalent) accredited instrument gives confidence in the device in terms of its accuracy, precision and tolerances [Ref-39].

3.6.4.4 Evidence: Detection Limits for Analytical Techniques

The European Basic Safety Standards [Ref-183] references EU 2004 [Ref-173], [Ref-184] and provides recommendations on the standardised information which should be reported when making gaseous and liquid waste discharges to the environment from nuclear power stations and reprocessing plants under normal operations. Within the recommendations, a number of key nuclides and requirements for their detection limits are listed; these are reproduced in Table 3.6.4.4-1 [Ref-39]. The laboratory chosen to carry out analysis must be able to demonstrate that they can meet these limits of detection.

Table 3.6.4.4-1: Key Nuclides in Liquid Discharges and Requirements for the Detection Limit in EU 2004

Requirement for the detection Category Key nuclides limit (Bq/m3)

Tritium H-3 1E+05

Co-60 1E+04

Other radionuclides Sr-90 1E+03

(excluding H-3) Cs-137 1E+04

Pu-239 + Pu-240 6E+03

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Requirement for the detection Category Key nuclides limit (Bq/m3)

Am-241 5E+01

Total-alpha 1E+03

3.6.4.5 Evidence: Proposed Analytical Methods for Key Radionuclides

Specific techniques have not been decided on at this point in the Wylfa Newydd Power Station project as it is recognised that BAT evolves with technological progression. The exact BAT techniques to be employed will be decided upon at an appropriate date in the future that will enable procurement and installation before the start of the inactive commissioning Hold Point. It is important not to constrain future selection of analytical techniques; therefore the design incorporates a suitably equipped and sized room to house sensitive analytical equipment and allow flexibility of techniques (3.6.5 Argument 6e: Space, capability and flexibility for maintenance activities and servicing). However, Table 3.6.4.5-1 lists the currently proposed techniques for analysing liquid discharge samples for various radionuclides.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Table 3.6.4.5-1: Currently Proposed Analytical Methods to Determine the Activity Concentration of Key

Nuclides in Liquid Discharges

Key nuclides Analytical method

H-3 Liquid scintillation counting

Co-60 Gamma spectroscopy

Sr-90 Liquid scintillation counting

Cs-137 Gamma spectroscopy

Pu-239 + Pu-240 Alpha spectroscopy

Am-241 Alpha spectroscopy

Total-alpha Alpha spectroscopy

Liquid scintillation counting was proposed for measurement of H-3 as this method is already routinely used. The detection limits are affected by the type of instrument used as well as the count time and choice of scintillant, but an example of a detection limit given in [Ref-185] is 4E+04 Bq/m3 with a 20 minute count time which would meet the requirement for the detection limit in Table 3.6.4.4-1 [Ref-39].

It is proposed that Co-60 and Cs-137 activities be determined by gamma spectroscopy. The detection limits for this technique depend on a range of factors such as count time, sample size and geometry and background as well as the size and configuration of the crystal but an example of a detection limit given in [Ref-185] is 1E+04 Bq/m3 for both nuclides with a Ge semiconductor detector with a 2,000 second count time which would meet the requirement for the detection limit in Table 3.6.4.4-1 [Ref-39].

It is proposed that Sr-90 activity will be analysed using liquid scintillation counting as a range of chemical extraction methods are available for measuring radio-strontium, which allows for low detection limits to be achieved in conjunction with an appropriate scintillation cocktail. An example of detection limit given in [Ref-185] is 7E+02 Bq/m3 with a proportional counter and a 10 minute count time which would meet the requirement for the detection limit in Table 3.6.4.4-1 [Ref-39].

It is proposed that alpha emitters be measured using alpha spectroscopy as it is already commonly used and is capable of achieving low detection limits. An example of a detection limit using a silver activated zinc sulphide (ZnS(Ag)) scintillator with a 600 second count time given in [Ref-185] was 4E+03 Bq/m3. This value is slightly higher than the detection limit in Table 3.6.4.4-1 but the detection limit can be improved if the counting time is increased sufficiently [Ref-39]. As such it is envisaged that the necessary detection limits will be achieved using alpha spectroscopy.

Laboratory reports presenting radionuclide measurements will need to specify the detection limits for the analytical techniques employed. The Company method of recording and reporting radionuclide measurements of gaseous discharges would be set out in the Aqueous Discharge Sampling & Monitoring procedure.

3.6.5 Argument 6e: Space, capability and flexibility for maintenance activities and servicing Sufficient space and capability is provided in the design to ensure that maintenance and servicing can be undertaken efficiently and effectively. On each of the main stacks a sampling platform will be provided for such activities to be undertaken (3.6.5.1 Evidence:

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Maintenance and Servicing Activities). The size of these platforms has been designed to comply with regulatory requirements (3.6.5.2 Evidence: Size of Sampling Platform). The proposed gaseous discharge monitoring system design contains, for each main stack, a single sample collection line from the sample extraction point within the stack to a stack equipment room. This room has been designed to provide sufficient access for workers to collect samples on a periodic basis and also contains enough space for maintenance and servicing of equipment (3.6.5.3 Evidence: Size of Sampling Room). The design of Wylfa Newydd provides sufficient space to allow Horizon to deploy and maintain a range of sampling and analytical techniques for discharges of radioactivity to the environment.

3.6.5.1 Evidence: Maintenance and Servicing Activities

On each main stack the isokinetic probe can be extracted to the platform outside of the stack for visual inspection and cleaning. Performance and leak checks will be undertaken after servicing in accordance with the Surveillance Test Maintenance procedure, to ensure effective operation [Ref-39].

The sampling platforms will be designed to comply with M1 [Ref-186] so that workers will have safe access for periodic inspection and maintenance.

3.6.5.2 Evidence: Size of Sampling Platforms

The minimum requirement within M1 [Ref-186] for stacks of the proposed size is a one sided platform with a working space of about 4.60 m. However, a full circumference platform is also being explored which would have a working area of about 3.05 m. The final choice will be dependent on the supplier of the isokinetic probes and will be made prior to construction [Ref-39].

3.6.5.3 Evidence: Size of Sampling Room

The dedicated sampling equipment rooms are planned to be approximately 90m2. This may be subject to modification due development of any design changes to work space area affecting sample collection and maintenance accessibility will be subject to review on Human Factors aspects to determine acceptability [Ref-39].

3.6.6 Argument 6f: Independent Sampling Arrangements Within the Wylfa Newydd Power Station, independent sampling systems are provided at the final discharge points for use by the regulator or their representatives. No independent continuous monitoring system is provided, but all data and quality control information will be made available. The use of independent sampling systems by the regulators will not interrupt Horizon’s own sampling or off-set the quality of that measurement (3.6.6.1 Evidence: Independent Sampling of Gaseous Waste Discharges) and (3.6.6.2 Evidence: Independent Sampling of Liquid Waste Discharges). With regards to gaseous discharges, for each main stack there are two identical sampling systems allowing NRW (or their representative) to use of one of these systems during periods when they wish to conduct independent monitoring. The isokinetic probe, sampling pipe, and return pipe are common for both systems and all sampling and monitoring equipment will be located in a common room. The only part of the system within each stack which isn’t duplicated for independent monitoring purposes is the noble gas monitoring system which is situated in the OG system before it joins the main stack (3.6.6.1 Evidence: Independent Sampling of Gaseous Waste Discharges).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Within the liquid discharge sampling system there will be two flow proportional samplers, both of which will be MCERTS (or equivalent) accredited. During normal operation only one sampler will be in active use, unless NRW requires an independent sample to be taken, in which case they (or their representatives) will have exclusive access to the second sampler. Both systems will be maintained and calibrated by the operating personnel in line with manufacturer’s requirements. They will also be tested during commissioning to ensure they are both collecting a representative Sample (3.6.6.2 Evidence: Independent Sampling of Liquid Waste Discharges). The design of Wylfa Newydd incorporates features that will allow independent verification of sampling and analytical techniques for discharges of radioactivity to the environment.

3.6.6.1 Evidence: Independent Sampling of Gaseous Waste Discharges

For each stack there will be two identical sampling systems for gaseous monitoring. The isokinetic probe, sampling pipe, and return pipe are common for both systems and all the sampling and monitoring equipment is located in the same room. It is proposed that appropriate measures are put in place for NRW (or their representatives) to secure use of one of the systems during periods when it wishes to conduct independent sampling. This would be in the form of sealing off the system with tamperproof seals so that filters, cartridges etc. cannot be accessed by Horizon operators, unless in an emergency and with prior consent from NRW. In addition, tamper proof seals will be placed on associated valves. This would allow NRW the ability to collect independent samples for particulate, iodine, tritium, and carbon-14. There is no requirement for NRW to have independent access to the noble gases monitoring system [Ref-39].

Both systems will be run at all times when NRW (or their representatives) does not require access. The system will be maintained and calibrated by Horizon. When NRW (or their representatives) is using one of the sampling systems, there will be no impact on the Horizon’s ability to collect samples from their active system. Furthermore, the total sampling flow rate will be kept constant in order to maintain isokinetic sampling when the sampling lines are isolated [Ref-39].

In addition to the sampling equipment in the stack equipment room, a sampling port on the main stack will also be provided for independent flow measurement which will be situated as close to the sample extraction point as is feasibly possible. The design of the access ports will be consistent with the requirements laid out in M1 [Ref-186]. Appropriate access arrangements will be provided and shared with the standard access to the sample extraction point.

3.6.6.2 Evidence: Independent Sampling of Liquid Waste Discharges

There will be two flow proportional samplers provided for sampling of liquid discharges. One will be used continuously by Horizon and the other will be provided to NRW (or their representatives) for use when NRW require independent samples to be collected.

The independent flow proportional sampler will use tamperproof seals and will have the ability to vary the amount of sample collected, as NRW’s (or their representatives) laboratory requirements may be different to Horizon’s such that they may require different volumes of samples to be collected. The collected sample will allow for all analytes to be determined and the system will not be accessible to the operators, unless in an emergency situation and with prior consent from NRW. Management controls will be established for the operator to have extended access during emergency events [Ref-39].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

4 Radionuclide Specific BAT Route Map Table 4-1 and Table 4-2 provide a list of the main radionuclides for gaseous and liquid releases. For each radionuclide, detail is provided on its production mechanism, route to the environment, the techniques used to eliminate or reduce generation at source and the techniques used to minimise the impacts on the environment. The techniques are described in the Claims, Arguments and Evidence set out within this report. The expected Annual Discharges are also provided for information (further details on discharges are provided in Section 5 of the EP-RSR application). Table 6-3 maps the same information for each solid waste category. It provides links to the relevant Claims, Arguments and Evidence and also provides a summary of how each waste is characterised and subsequently managed. Table 4-1: Summary table of Main Radionuclides for Gaseous releases

Discharge at height via main stack (Argument 5a: Gaseous Discharge System - Main Stack). Krypton-85 1.3E+09 FPs from fuel and structural Migration into reactor water (direct or Minimise fuel cladding failures (grid-to-rod Off-gas treatment system and uranium. through pin failure) → 100% migration fretting, corrosion and crud, debris, PCI and charcoal adsorbers (Argument 2a: Krypton-85m 1.0E+10 into steam Separation at condenser manufacturing QA) (Argument 1a: Design, Off-Gas Waste Treatment Facility Radioactive noble gases are → Krypton-87 9.6E+03 Discharge from stack via OG. Manufacture and Management of Fuel). and Argument 2b: Charcoal formed by fission. They are → Adsorbers for Noble Gases and Krypton-88 9.1E+08 usually confined within the fuel but High standards of fuel design and fabrication Iodine). in the event of fuel leaks, they can (Argument 1a: Design, Manufacture and

Krypton-89 0.0E+00 pass into the coolant via defects in Management of Fuel). Discharge at height via main stack the fuel cladding. Their presence Xenon-131m 2.9E+09 Minimise “tramp uranium” (Argument 1a: (Argument 5a: Gaseous Discharge in the coolant is also due to the Design, Manufacture and Management of System - Main Stack). Xenon-133 2.0E+11 occurrence of traces of uranium Fuel). Xenon-133m 1.7E+07 (“tramp” uranium) on the surface of fuel assemblies following the Minimisation of crud formation and optimal Xenon-135 0.0E+00 manufacturing process. water chemistry (Argument 1f: Water Xenon-135m 0.0E+00 Chemistry). Xenon-137 0.0E+00 An efficient anti debris device is provided for fuel assemblies (Argument 1a: Design, Xenon-138 0.0E+00 Manufacture and Management of Fuel). The fuel performance - minimising the number of fuel assemblies used minimises the probability for cladding leakage of FPs into the coolant (Argument 1c: Efficiency of Fuel Use). Identifying and isolating fuel leaks (Argument 1d: Detection and Management of Failed Fuel).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

uranium contamination on fuel stream via HVAC system. Management of Fuel). Discharge at height via main stack surface (“tramp” uranium) within Discharge of iodine in gaseous Minimise “tramp uranium” (Argument 1a: (Argument 5a: Gaseous Discharge the reactor which can also be radioactive waste discharged to main Design, Manufacture and Management of System - Main Stack). found in the coolant. stack from the gland steam exhauster. Fuel). Minimisation of crud formation and optimal water chemistry (Argument 1f: Water Chemistry). An efficient anti debris device is implemented for fuel assemblies (Argument 1a: Design, Manufacture and Management of Fuel). The fuel performance - minimising the number of fuel assemblies used minimises the probability for cladding leakage of FPs into the coolant (Argument 1c: Efficiency of Fuel Use), Identifying and isolating fuel leaks (Argument 1d). Minimise leaks (Argument 1j: Leak Tightness of Liquid, Gas and Mixed Phase Systems).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

as a result of fission. They are High standards of fuel design and fabrication usually confined in the fuel but, in (Argument 1a: Design, Manufacture and the event of fuel leaks, they can Discharge of scattered particulate in Management of Fuel). pass into the coolant via defects in aqueous stream via HVAC system Discharge at height via main stack the fuel cladding. Their presence Minimise “tramp uranium” (Argument 1a: (Argument 5a: Gaseous Discharge

in the coolant is also due to the Design, Manufacture and Management of System - Main Stack). occurrence of traces of uranium Discharge of evaporated particulate in Fuel). (“tramp” uranium) that can never aqueous stream via TGS system. Minimisation of crud formation and optimal be completely removed on new water chemistry (Argument 1f: Water fuel assemblies following the Chemistry). manufacturing process. An efficient anti debris device is implemented for fuel assemblies (Argument 1a: Design, Manufacture and Management of Fuel). Strontium-90 1.3E+03 The fuel performance - minimising the number Caesium-137 2.8E+03 FPs from fuel, structural uranium. of fuel assemblies used minimises the probability for cladding leakage of FPs into the coolant (Argument 1c: Efficiency of Fuel Use). Identifying and isolating fuel leaks (Argument 1d: Detection and Management of Failed Fuel). Minimise leaks (Argument 1j: Leak Tightness of Liquid, Gas and Mixed Phase Systems). Cobalt-60 7.4E+04 Cobalt-59 (n,g) cobalt-60. Entrainment of aerosol into steam lines Minimisation of crud formation and optimal Filters to remove particulate Discharge via condenser and stack. water chemistry (Argument 1f: Water material (including filters on the Activation of reactor components. → (negligible) Chemistry). HVAC) (Argument 2d: Filtration of Activation of insoluble and soluble Airborne Particulate Matter). metal crud and particulate in Aerosol generation in tanks, pools and reactor water. release via HVAC system. Specification of low cobalt content materials

(Argument 1g: Specification of Materials).

Discharge of evaporated particulate in Discharge at height via main stack aqueous stream via TGS system. Minimise leaks (Argument 1j: Leak Tightness (Argument 5a: Gaseous Discharge of Liquid, Gas and Mixed Phase Systems). System - Main Stack). Tritium 5.5E+12 Ternary fission in fuel Migration into reactor water (direct or No boron usage in the water chemistry Gaseous tritium present within the Boron-10 (n,2a).Tritium from boron through pin failure or diffusion through (Argument 1b: Reactivity Control). off-gas is removed by the off-gas in control rods. pin cladding) Entrainment of aerosol recombiner and off-gas condenser. → Use of hafnium control rods (Argument 1b: into steam lines discharge via The off-gas Recombiner → Reactivity Control). condenser and stack. recombines hydrogen and oxygen

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Best Estimate 12 Month rolling Discharge (whole Technique to minimise the Technique to eliminate or reduce the Radionuclide power station) Production mechanism Route to the environment impacts on the environment generation at source (Claim 1) (Bq/y) (see Section (Claim 2-5) 5 of the EP-RSR Application) Hydrogen-2 (n,g) H-3 from and the off-gas condenser cools Adventitious discharges from steam Use of gadolinium as a burnable poison rather Hydrogen-2 in reactor water. and condenses the hydrogen leaks → than boron (Argument 1b: Reactivity Control). depleted off-gas to separate any Discharge via HVAC of aqueous Minimise fuel cladding failures (grid-to-rod moisture and return it to the main vapour. fretting, corrosion and crud, debris, PCI, and condenser. Following treatment Evaporative losses from sources of manufacturing upsets) (Argument 1a: Design, by these two components of the tritiated water→ Discharge of aqueous Manufacture and Management of Fuel). OG the hydrogen concentration is vapour via HVAC system. High standards of fuel design and fabrication minimised in the off-gas. As Discharge of tritium in gaseous (Argument 1a: Design, Manufacture and tritium is a hydrogen compound; radioactive waste discharged to main Management of Fuel). the performance of the off-gas stack from the gland steam exhauster. recombiner and off-gas condenser Minimisation of crud formation and optimal therefore also removes tritium from water chemistry (Argument 1f: Water the off-gas. The hydrogen and Chemistry). therefore any tritium is converted to An efficient anti debris device is implemented water and is returned to the CST for fuel assemblies (Argument 1a: Design, where it is reused within the plant. Manufacture and Management of Fuel). (Argument 2a: Off-Gas Waste Treatment Facility). The fuel performance - minimising the number of fuel assemblies used minimises the Discharge at height via main stack probability for cladding leakage of FPs into the (Argument 5a: Gaseous Discharge coolant (Argument 1c: Efficiency of Fuel Use). System - Main Stack). Identifying and isolating fuel leaks (Argument 1d: Detection and Management of Failed Fuel) Minimise leaks (Argument 1j: Leak Tightness of Liquid, Gas and Mixed Phase Systems). Condense 98% of TSG steam minimising residual steam discharge and tritium (Argument 2e: Optimisation of the Turbine Gland Seal). Carbon-14 1.8E+12 Neutron activation of Nitrogen-14 Carbon-14 is always carried by stable None None. and Oxygen-17results in Carbon- carbon compounds. The air entrained The principle source of Carbon-14 is the (Argument 2a: Off-Gas Waste 14 both from fuel and reactor in the coolant is ejected from the main thermal neutron reaction with Oxygen-17 in the Treatment Facility). water. condenser. This off-gas is reactor water (H2O). Therefore, there are no fundamentally air, and therefore carbon, measures for reducing the generation of as carbon dioxide, exists in the similar Carbon-14. Discharge at height via main stack Another minor mechanism ratio to other constituents as it does in (Argument 5a: Gaseous Discharge contributing to Carbon-14 is the air and discharged via OG system. System - Main Stack). reaction Carbon-13 (n, γ) -> Carbon-14, which occurs due to the presence of dissolved carbon in the coolant.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Table 4-2: Summary table for Main Radionuclides for liquid releases

Best Estimate 12 Month rolling Discharge Production Technique to eliminate or reduce the generation at source Technique to minimise the impacts on Radionuclide (whole power Route to the environment mechanism (Claim 1) the environment (Claim 2-5) station) (Bq/y) (see Section 5 of the EP-RSR Application) Strontium-89 4.2E+03 FPs from fuel, Migration into reactor water (direct or Minimise fuel cladding failures (grid-to-rod fretting, corrosion CUW system structural through pin failure) → and crud, debris, PCI, and manufacturing upsets) (Argument (Argument 1h: Recycling of Water to uranium 1a: Design, Manufacture and Management of Fuel). Partial migration into steam → Prevent Discharges) High standards of fuel design and fabrication (Argument 1a: Build-up in reactor, fuel pool water, Design, Manufacture and Management of Fuel). etc. → HCW evaporator Minimise “tramp uranium” (Argument 1a: Design, Manufacture Liquid waste gathered in each sump and Management of Fuel). HCW demineraliser. → Minimisation of crud formation and optimal water chemistry (Argument 2i: Evaporation of HCW and Discharge via HCW sample tank- (Argument 1f: Water Chemistry). Argument 2h: Demineralisers for Distillates occasional from the HCW Evaporator) An efficient anti debris device is implemented for fuel assemblies (Argument 1a: Design, Manufacture and Management of Fuel). The fuel performance - minimising the number of fuel assemblies used minimises the probability for cladding leakage of FPs into the coolant (Argument 1c: Efficiency of Fuel Use). Strontium-90 2.1E+03 Identifying and isolating fuel leaks (Argument 1d: Detection and Management of Failed Fuel). Iodine-131 7.0E+04 FPs from fuel, Migration into reactor water (direct or Minimise leaks (Argument 1j: Leak Tightness of Liquid, Gas structural through pin failure) → and Mixed Phase Systems). uranium. Partial migration into steam → Build-up in reactor, fuel pool water, etc. → Liquid waste gathered in each sump → Discharge via HCW sample tank- occasional

Cesium-137 3.1E+03 FPs from fuel, Migration into reactor water (direct or Minimise fuel cladding failures (grid-to-rod fretting, corrosion CUW system structural through pin failure) → and crud, debris, PCI, and manufacturing upsets) (Argument (Argument 1h: Recycling of Water to uranium. 1a: Design, Manufacture and Management of Fuel). Partial migration into steam → Prevent Discharges) High standards of fuel design and fabrication (Argument 1a: Build-up in reactor, fuel pool water, Design, Manufacture and Management of Fuel). etc. → HCW evaporator Minimise “tramp uranium” (Argument 1a: Design, Manufacture Liquid waste gathered in each sump and Management of Fuel). HCW demineraliser. → Minimisation of crud formation and optimal water chemistry (Argument 1f: Water Chemistry).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Best Estimate 12 Month rolling Discharge Production Technique to eliminate or reduce the generation at source Technique to minimise the impacts on Radionuclide (whole power Route to the environment mechanism (Claim 1) the environment (Claim 2-5) station) (Bq/y) (see Section 5 of the EP-RSR Application) Discharge via HCW sample tank- An efficient anti debris device is implemented for fuel (Argument 2i: Evaporation of HCW occasional assemblies (Argument 1a: Design, Manufacture and Argument 2h: Demineralisers for Distillates Management of Fuel). from the HCW Evaporator) The fuel performance - minimising the number of fuel assemblies used minimises the probability for cladding leakage of FPs into the coolant (Argument 1c: Efficiency of Fuel Use). Identifying and isolating fuel leaks (Argument 1d: Detection and Management of Failed Fuel). Minimise leaks (Argument 1j: Leak Tightness of Liquid, Gas and Mixed Phase Systems). Cobalt-60 2.1E+05 Cobalt-59 (n,g) Migration into reactor water (direct or Minimisation of crud formation and optimal water chemistry CUW system Cobalt-60 through pin failure) → (Argument 1f: Water Chemistry). (Argument 1h: Recycling of Water to Activation of Partial migration into steam → Prevent Discharges) reactor components, Build-up in reactor, fuel pool water, Specification of low cobalt content materials (Argument 1g: HCW evaporator etc. → Specification of Materials). insoluble and HCW demineraliser soluble metal Liquid waste gathered in each sump (Argument 2i: Evaporation of HCW and crud and → particulate in Minimise leaks (Argument 1j: Leak Tightness of Liquid, Gas Argument 2h: Demineralisers for Distillates reactor water. Discharge via HCW sample tank- and Mixed Phase Systems). from the HCW Evaporator) occasional Tritium 4.0E+11 Ternary fission Migration into reactor water (direct or No boron usage in the water chemistry (Argument 1b: None in fuel. through pin failure) → Reactivity Control).

Boron-10 (n,2a) Partial migration into steam → Use of hafnium control rods (Argument 1b: Reactivity Control). (Argument 2f: Configuration of Tritium (from Build-up in reactor, fuel pool water, Use of gadolinium as a burnable poison rather than boron Liquid Management Systems) boron in control etc. → (Argument 1b: Reactivity Control). rods). Liquid waste gathered in each sump Minimise fuel cladding failures (grid-to-rod fretting, corrosion Hyrogen-2 (n,g) → and crud, debris, PCI and manufacturing upsets) (Argument tritium (from 1a: Design, Manufacture and Management of Fuel). Hydrogen-2 in Discharge via HCW sample tank- reactor water). occasional High standards of fuel design and fabrication (Argument 1a: Design, Manufacture and Management of Fuel). Minimisation of crud formation and optimal water chemistry (Argument 1f: Water Chemistry). An efficient anti debris device is implemented for fuel assemblies (Argument 1a: Design, Manufacture and Management of Fuel).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Table 6-3: Summary table of solid waste arisings

Key Annual or radionuclides periodic Technique to eliminate or Technique to minimise the impacts Waste Waste management and specific Description Source disposals volume reduce the generation at of disposal on the environment Characterisation category route activity [Ref- (per reactor unit) source (Claims 2-5) 116] [Ref-116]

VLLW Dependent Miscellaneous Maintenance 13.8 m3/yr Prevent or reduce the Segregation of waste to ensure the Anticipated waste Incineration (3.3.3.2 upon combustible: operations generation of FPs, activation optimised treatment, storage and characteristics Evidence: Incineration combustible operational paper, products and CPs which disposal option is selected (3.3.2.1 determined at work and 3.4.2.1 Evidence: source; mainly polythene, subsequently lead to the Evidence: Segregation of Waste). planning phase (data Waste Processing steel activation cloth etc. 3 generation of combustible and collection to support Techniques). 3.4 m /yr non- Minimising the number of operator products. non-combustible VLLW during history and provenance combustible visits into RCAs and reducing the maintenance activities (3.1 arguments). volume of consumables that are taken Claim 1: Eliminate or Reduce Measurement at into RCAs reduces the potential to the Generation of Radioactive source. Potential for generate maintenance wastes (3.3.2.2 Waste). sampling and off-site Evidence: Locate Offices Outside of detailed Controlled Areas). characterisation. Minimising the amount of Gamma assay systems maintenance equipment and tools that installed in solid LLW are taken into RCA (3.3.2.3 Evidence: facility. Application of Storage Facilities for Tools and Other source area / plant Maintenance Equipment). fingerprint [Ref-116]. Provision of facilities to undertake the characterisation, sorting, treatment and storage of waste prior to consignment to an appropriately permitted waste management service supplier (3.4.1 Argument 4a: Provision of Solid Waste Management Facilities). Effective preventative maintenance schedules; predict, prepare and avoid (where practicable) leaks and spillages and associated clean-up activities (3.3.2.5 Evidence: Maintenance Philosophy and [Ref- 116]). Volume reduction treatment processes (3.3.3.2 Evidence: Incineration and 3.3.3.3 Evidence: Solid Waste Compaction). Decontamination where practicable to reduce waste classification and / or

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Dry-Solid iron-55, cobalt- HVAC filters. HVAC 25 m3/yr (HVAC + Prevent or reduce the HEPA filters will be changed, where Same as for VLLW. Incineration (3.4.2.1 LLW 60, zinc-65, system Misc. comb) generation of FPs, activation practicable, based on performance Evidence: Waste manganese- products and CPs which could determined using continuous Processing Techniques 54, cesium- then be filtered by the HVAC measurement of differential pressures and BAT options 137, strontium- system leading to the or as a result of manufacturer’s assessment report [Ref- 90, antimony- contamination of HVAC filters guidance (3.2.4 Argument 2d: 115]). 125 (3.1 Claim 1: Eliminate or Filtration of Airborne Particulate Reduce the Generation of Matter and 3.3.2.5 Evidence: Radioactive Waste). Maintenance Philosophy). HVAC filters will be segregated from other waste streams to ensure appropriate maintenance (3.3.2.1 Evidence: Segregation of Waste). Volume reduction treatment processes (3.3.3.2 Evidence: Incineration and 3.3.3.3 Evidence: Solid Waste Compaction).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

iron-55, cobalt- Miscellaneous As for VLLW 30.6 m3/yr Prevent or reduce the See VLLW Same as for VLLW. Incineration (3.4.2.1 60, combustible: combustible generation of FPs, activation Evidence: Waste manganese- paper, plus LCW. products and CPs which Processing Techniques 54, nickel-63 polythene, subsequently lead to the and BAT options cloth, LCW generation of combustible LLW assessment report [Ref- filter during maintenance activities 115]). membrane, (3.1 Claim 1: Eliminate or spent activated Reduce the Generation of carbon. Radioactive Waste)

iron-55, cobalt- Recyclable 2.3 m3/yr Prevent or reduce the See VLLW. Same as for VLLW. Off-site recycling (3.4.2.1 60, metals. generation of FPs, activation Evidence: Waste magnesium-54 products and CPs which Processing Techniques subsequently lead to the and BAT options generation of metal LLW during assessment report [Ref- maintenance activities (3.1 115]) Claim 1: Eliminate or Reduce the Generation of Radioactive Waste)

iron-55, cobalt- Non- As for VLLW 7.6 m3/yr Prevent or reduce the See VLLW. Same as for VLLW. Disposal at an 60, combustible non- generation of FPs, activation appropriately permitted magnesium-54 and non- combustible products and CPs which site (3.4.2.1 Evidence: compactable plus CF subsequently lead to the Waste Processing waste system. generation of non-combustible Techniques) e.g. LLWR (including LLW during maintenance (BAT options metals activities (3.1 Claim 1: Eliminate assessment report [Ref- unsuitable for or Reduce the Generation of 115]). recycling and Radioactive Waste). CF filters).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Wet - iron-55, cobalt- Organic bead Condensate, 13.2 m3/yr (before Prevent or reduce the Segregated between source systems Anticipated waste Disposal at an solid 60, demineraliser LCW, HCW. cementation) generation of FPs, activation (3.2.6 Argument 2f: Configuration of characteristics appropriately permitted LLW magnesium- resin products and CPs which Liquid Management Systems). determined at work site (3.4.2.1 Evidence: 54, nickel-63 (condensate, subsequently enter liquid planning phase (data Waste Processing Segregation of waste to ensure the LCW, HCW), systems and require treatment collection to support Techniques). optimised treatment, storage and HCW leading to the generation of history and provenance disposal option is selected (3.3.2.1 evaporator resin, sludge and granular arguments). Sampling Evidence: Segregation of Waste). sludge. activated carbon (GAC) waste and off-site detailed (3.1 Claim 1: Eliminate or Cementation, in batch campaigns, characterisation of raw Reduce the Generation of prior to disposal (3.4.1.3 Evidence: waste (including Radioactive Waste). Waste processing and Packaging cement formulation Before start-up, removing crud Facilities). trials). In-line prior to activation reduces the Replacement of pre-coated filters with measurements in the radioactivity deposited on the HFF or pleated filters (3.3.1.5 processing plant. demineraliser resins (3.1.5.4 Evidence: Replacement of Pre-coated Monitoring of final Evidence: Water Conditioning). Filters). waste package and application of During an outage the CD is isolated radioactivity fingerprint and is stored in demineralised water [Ref-116]. to prevent degradation of the resin (3.1.5.4 Evidence: Water Conditioning). Selection of resin media that can be suitably disposed (3.1.8.6 Evidence: Demineraliser Media). Allowing solid and aqueous radioactive waste to undergo radioactive decay before disposing of it to the environment or another premises will reduce the amount of radioactivity that is disposed of in the waste (3.2.10 Argument 2j: Radioactive Decay of Solid and Liquid Wastes).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Dry-Solid cobalt-60, Activated Control rods: 500 The design of the Wylfa Newydd Segregation of waste at source and Anticipated waste Disposal at GDF (3.4.4.2 ILW nickel-63 Metals: control units per reactor Power Station has evolved to separation (LLW: ILW) after decay characteristics Evidence: Disposability (californium- rods, reactor (51.75t) (see reduce the quantities of solid storage (3.3.2.1 Evidence: determined at work Assessment – 252 in neutron components Section 5 of the radioactive waste that will be Segregation of Waste) and (3.4.1.2 planning phase (data Intermediate Level source) (e.g. neutron EP-RSR generated during its life-cycle Evidence: Segregation and Sorting collection to support Waste and Options for source unit). Application) (3.3.1 Argument 3a: Design to Capabilities). history and provenance the management of Dry Minimise the Volumes of arguments). Solid ILW [Ref-187]. HAW metals: 7.6t / Operational and Provision of a dedicated facility to Measurement and 60 years (see Decommissioning Waste process and treat Dry-Solid ILW (3.4.1 assay following decay Section 5 of the Arisings). Argument 4a: Provision of Solid storage [Ref-116]. EP-RSR Waste Management Facilities). Application) Use of hafnium control rods Fuel channels to which have a longer operational Decay storage in dry casks to reduce be disposed of with life and therefore require less activity levels (3.2.10 Argument 2j: SF. frequent disposal (3.1.2 Radioactive Decay of Solid and Liquid Argument 1b: Reactivity Wastes). Control). Size reduction in order to aid optimal Implementation of disposal (3.3.3.1 Evidence: Size commissioning, start-up, Reduction of Control Rods). shutdown and outage processes to prevent the deposition of Optimised disposal (3.4.2 Argument radioactivity on reactor 4b: Optimal Disposal Route Selection, components which will become 3.4.4 Argument 4d: Disposability waste during maintenance and Assessments for Higher Activity decommissioning (3.1.5 Wastes and 3.4.5 Argument 4e: Argument 1e: Commissioning, Compatibility of Existing UK Waste Start-up, Shutdown and Outage BAT Studies). Procedures).

Selection of materials and water chemistry to reduce the activation of metals (3.1.6 Argument 1f: Water Chemistry and 3.1.7 Argument 1g: Specification of Materials).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

Wet- iron-55, cobalt- Organic CUW, FPC, 4.4 m3/yr resin, Prevent or reduce the See wet solid LLW. Anticipated waste Disposal at GDF (3.4.4.2 Solid 60, powder CF & LCW. 90m3 / 60 years generation of FPs, activation characteristics Evidence: Disposability ILW magnesium- demineraliser sludge (before products and CPs which determined at work Assessment – 54, zinc-65 resin (CUW cementation). subsequently enter liquid planning phase (data Intermediate Level and FPC systems and require treatment collection to support Waste). systems), leading to the generation of history and provenance sludge (crud) resin, sludge and GAC waste arguments). Sampling from CF and (3.1 Claim 1: Eliminate or and off-site detailed LCW filters. Reduce the Generation of characterisation of raw Radioactive Waste). waste (including cement formulation Before start-up, removing crud trials). In-line prior to activation reduces the measurements in the radioactivity deposited on the processing plant. demineraliser resins (3.1.5.4 Monitoring of final Evidence: Water Conditioning). waste package and application of radioactivity fingerprint [Ref-116].

Fuel FPs, activation Fuel Fuel Approximately 150 The efficiency with which the Segregation of fuel from other waste Calculated waste Disposal at GDF (3.4.4.1 products and assemblies of assemblies / year nuclear fuel is used in the Wylfa streams (3.3.2.1 Evidence: characteristics based Evidence: Disposability actinides GE14 design: Newydd Power Station and the Segregation of Waste). on materials and Assessment – Spent uranium frequency with which it is operations. Fuel). dioxide pellets changed will influence the Decay storage in SFP followed by dry within Zircaloy amount of SF and HAW that is cask storage in SFIS [Ref-116]. cladding; fuel generated during operations rods held in (3.1.3 Argument 1c: Efficiency of bundles. Fuel Use).

The generation of fuel waste is inevitable however there are a number of practices which ensure subsequent optimal handling, treatment and disposal (3.1.1 Argument 1a: Design, Manufacture and Management of Fuel).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[This page is intentionally blank]

Wylfa Newydd Project – Best Available Techniques (BAT) Case

5 Forward Action Plan

No. Action

[FAbc-1] Horizon shall develop a strategy for the management of failed fuel in the SFP (expected events only). As part of this, Horizon must substantiate the assumption that failed fuel rods will not release fission products once in the SFP leading to an increase in radioactivity in the SFPs. Source: 3.1.4.3 (Evidence: Management of Failed Fuel).

[FAbc-2] Horizon must define schemes, operating regimes and maintenance instructions to ensure adequate isolation (in regards to supporting leak tightness) is achieved. Source: 3.1.10.8 (Evidence: Design Policies on Isolation).

[FAbc-3] Horizon will establish a contract with an offsite active laundry service provider for its reusable coveralls. Source: 3.3.2.4 (Evidence: Re-usable protective clothing in the RCA).

[FAbc-4] Horizon will set up commercial service-level agreements with those permitted waste service providers that have been identified as representing an optimum waste route. This will be implemented to facilitate off-site transfer of radioactive waste. Source: 3.4.3.2 (Evidence: Waste Contracts).

[FAbc-5] Horizon must review, determine and underpin its selection of, if any, pre-treatment/commissioning techniques which may be employed at the Wylfa Newydd Power Station. Source: 3.1.5.1 (Evidence: Alkali Pre-Filming Technique)

[FAbc-6] Horizon must undertake a revised direct shine dose assessment of the combined SF storage facility to ensure that it does not challenge, from a dose perspective, the concept of ALARA. 3.4.1.4 (Evidence: Waste Storage Capacity)

Wylfa Newydd Project – Best Available Techniques (BAT) Case

No. Action

[FAbc-7] Horizon must undertake further work to consider optimisation of the use of the ILW Storage Facility for the storage of decommissioning wastes. It is currently envisaged that a separate ILW Storage Facility for decommissioning waste arisings will be built at the end of the plant’s life but this must be confirmed. 3.4.1.4 (Evidence: Waste Storage Capacity)

Wylfa Newydd Project – Best Available Techniques (BAT) Case

6 Conclusion The Claims and Arguments for the Wylfa Newydd Power Station Demonstration of BAT have been developed using information held by Hitachi-GE and Horizon. The evidence provided herein reflects the information which will be adopted from GDA Step 3 and information developed by Horizon in support of the Wylfa Newydd Power Station design where the design has changed from that presented at GDA. This information has then been used to develop the Claims and Arguments. Gaps and uncertainties identified during the development of the arguments have been subject to additional assessment or recorded as FA’s to ensure that they are closed out at the most appropriate time in the project lifecycle. Collectively the Claims, Arguments, Evidence model will support the demonstration that BAT has been applied to the Wylfa Newydd Power Station design, allowing examination and challenge and where applicable identifying key gaps or uncertainties. During evidence gathering Horizon have consistently asked the questions: • Can anything else be done to reduce activity of discharges, minimise volumes of solid waste or reduce impacts from discharges? • Is the time, trouble and money associated with implementing changes grossly disproportionate to the potential benefits gained? • Does this contribute to ensuring that the dose experienced by a member of the public demonstrates ALARA? The demonstration of BAT is an iterative process that feeds back to the design; if during the process any areas of insufficient evidence remain, design changes may be made to support the application of BAT. Horizon believe that the arguments set out in this Demonstration of BAT report and the substantiation provided by the existing evidence base combined with the resolution of outstanding FA’s demonstrate that the Wylfa Newydd Power Station has been environmentally optimised as prescribed in the conditions of the Environmental Permitting (England and Wales) Regulations 2016 [Ref-4].

Wylfa Newydd Project – Best Available Techniques (BAT) Case

7 References

[Ref-1] Natural Resources Wales, “Application for an environmental permit Part RSR-B3 – New bespoke radioactive substances activity permit (nuclear site – open sources and radioactive waste)”, NRW-EP-RSR-B3 Version 1, December 2015.

[Ref-2] Horizon Nuclear Power, “EP-RSR BAT Case Development: Strategy for the Development of the Wylfa Newydd Project EP-RSR BAT Case”, WN01.10.02-S3- DAP-STR-0001, Revision 1.0, December 2015

[Ref-3] Hitachi-GE Nuclear Energy, Ltd., Methodology for Expected Event Selection, GA91-9201-0003-00353, Jan 2015

[Ref-4] The Environmental Permitting (England and Wales) Regulations 2016, 2016 No. 1154, 11th December 2016

[Ref-5] Environment Agency, “How to comply with your environmental permit for radioactive substances on a nuclear licensed site”, GEHO0812BUSS-E-E, 478_10, Version 2, 21 August 2012.

[Ref-6] Environment Agency, “RSR: Principles of optimisation in the management and disposal of radioactive waste”, Issue 2, April 2010, Environment Agency.

[Ref-7] Natural Resource Wales, “Radioactive Substances Regulation Environmental Principles”, Version 3, September 2014.

[Ref-8] Technical Working Group on behalf of the Nuclear Safety Directors Forum, “Best Available Techniques (BAT) for the Management of the Generation and Disposal of Radioactive Wastes”, A Nuclear Industry Code of Practice, Issue 1, December 2010.

[Ref-9] Safety Directors Forum BAT/BPM Working Group on behalf of the Nuclear Safety Directors Forum, “Operational Experience implementing the “Nuclear Industry Code of Practice on Identifying and Implementing Best Available Techniques (BAT), Issue 1, December 2010”

[Ref-10] Nuclear Energy Agency, Organisation for Economic Co-Operation and Development (OECD), “Effluent Release Options from Nuclear Installations”, ISBN 92-64-02146-9, 2003.

[Ref-11] DEFRA, “Environmental Permitting Guidance Radioactive Substances Regulation for the Environmental Permitting (England and Wales) Regulations 2010”, Version 2.0, September 2011.

[Ref-12] Hitachi-GE Nuclear Energy, Ltd., “Approach to Optimisation”, GA91-9901-0021- 00001 (XE-GD-0096), Rev. E, February 2016.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[Ref-13] Horizon Nuclear Power, “Nuclear Safety, Security and Environmental Principles; Expectations for Nuclear Safety, Security and Environmental Protection”, HG-M- 05-POL-01-1151, Revision 3.0, February 2017.

[Ref-14] Horizon Nuclear Power, “Management of EP-RSR Best Available Techniques (BAT)”, HG-D-01-PSD-02-1241, Revision 1.0, August 2016.

[Ref-15] Horizon Nuclear Power, “EP-RSR Best Available Techniques (BAT) Evidence Management”, HG-D-01-PRC-03-1244, Revision 1.0, August 2016.

[Ref-16] Horizon Nuclear Power, “EP-RSR Best Available Techniques (BAT) Case Amendment Management”, HG-D-01-PRC-03-1242, Revision 1.0, August 2016.

[Ref-17] Horizon Nuclear Power, “Optioneering”, HG-D-01-PRC-02-795, Revision 2.0 (March 2016).

[Ref-18] Hitachi-GE Nuclear Energy, Ltd., “UK ABWR GDA (Generic Design Assessment): Demonstration of BAT”, GA91-9901-0023-00001, Rev F, July 2016.

[Ref-19] Horizon Nuclear Power, “Adoption of GDA Documentation”, HG-D-01_PRO-02- 1175, Revision 2.0, April 2017.

[Ref-20] Horizon Nuclear Power, “Management of Environmental Functions and SSC”, Revision 0.1, August 2016

[Ref-21] Horizon Nuclear Power, “Environmental Functions and Categorisation”, Revision 0.1, August 2016

[Ref-22] Horizon Nuclear Power, “Identification of Environmental SSC”, Revision 0.1, August 2016

[Ref-23] Horizon Nuclear Power, “Management of GDA Assessment Findings”, HG-D-01- PRC-03-1354, Revision 0.1, August 2016.

[Ref-24] Horizon Nuclear Power, “Design Change Control”, HG-D-01-PRO-03-940, Revision 1.0, May 2016.

[Ref-25] Hitachi-GE Nuclear Energy, Ltd., “Generic PCSR Sub-chapter 31: Decommissioning”, GA91-9101-0101-31000 (DCE-GD-0007), Rev. B, October 2015.

[Ref-26] Hitachi-GE Nuclear Energy, Ltd., “Topic Report on Decommissioning: Design for Decommissioning”, GA91-9201-0001-00172, February 2016.

[Ref-27] Hitachi-GE Nuclear Energy, Ltd., “Topic Report on Decommissioning: Decommissioning Waste Management”, GA91-9201-0001-00173, Rev. 3, July 2016.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[Ref-28] Hitachi-GE Nuclear Energy, Ltd., “Basis of Safety Cases on Fuel Handling Systems and Overhead Crane Systems”, GA91-9201-0002-00056 (M1D-UK-0006) Rev.0, January 2015.

[Ref-29] US Department of Energy, “Evaluation of Tritium Removal and Mitigation Technologies for Wastewater Treatment”, 2009.

[Ref-30] Hitachi-GE Nuclear Energy, Ltd., “Preliminary Safety Report on Reactor Core and Fuels”, GA91-9901-0046-00001 (XE-GD-0156), Rev. B, March 2014.

[Ref-31] Hitachi-GE Nuclear Energy, Ltd., “Environmental impacts of Gadolinia (Response to RQ-ABWR-0241)”, GA91-9201-0003-00445 (UE-GD-0312), Rev.0, December 2014.

[Ref-32] Hitachi-GE Nuclear Energy, Ltd., “Generic PCSR Chapter 11: Reactor Core”, GA91-9101-0101-11000 (UE-GD-0182), Rev. B, October 2015.

[Ref-33] Hitachi-GE Nuclear Energy, Ltd., “Use of hafnium control rods (Response to RQ- ABWR-0222)”, GA91-9201-0003-00446 (UE-GD-0311), Rev.1, February 2015.

[Ref-34] Hitachi-GE Nuclear Energy, Ltd., Disposability Assessment Submission (ILW), WE- GD-0002, Rev.4, December 2014, Sent via Letter No. HGNE-REG-0071N.

[Ref-35] Hitachi-GE Nuclear Energy, Ltd., “Potential for failure/rupture of boron carbide control rods (Response to RQ-ABWR-0245)”, GA91-9201-0003-00466 (HE-GD- 0067), Rev.1, March 2015.

[Ref-36] Hitachi-GE Nuclear Energy, Ltd., “Design Basis for B4C Control Rod Lifetimes (Response to Query 2 and 4 of RQ-ABWR-0469)”, GA91-9201-0003-00694 (HE- GD-0084), Rev.0, May 2015.

[Ref-37] Hitachi-GE Nuclear Energy, Ltd., “Quantification of Discharges and Limits”, GA91- 9901-0025-00001 (HE-GD-0004), Rev. F, July 2016.

[Ref-38] Areva/EDF, Pre-Construction Safety Report, Chapter 11.3 - Outputs for the operating installation, UKEPR-0002-113, Issue 05, August 2012.

[Ref-39] Hitachi-GE Nuclear Energy, Ltd., “Approach to Sampling and Monitoring” GA91- 9901-0029-00001 (3E-GD-K002), Rev. G, July 2016.

[Ref-40] IAEA Nuclear Energy Series, NF-T-3.6, “Management of Damaged Spent Nuclear Fuel”, June 2009.

[Ref-41] N. Suzuki et al., 9th Int. Conf. on Water Chemistry in Nuclear Reactor System, Avignon (2002).

[Ref-42] M. Nagase et al., J. Nucl. Sci. and Technol., Vol.38 [12], p.1090-1096 (2001).

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[Ref-43] K. Ishida et al., J. Nucl. Sci and Technol., Vol.39 [9], p.941-949 (2002).

[Ref-44] Hitachi-GE Nuclear Energy, Ltd., “Preliminary Safety Report on Reactor Chemistry”, GA91-9901-0041-00001 (XE-GD-0152), Rev. B, March 2014.

[Ref-45] Hitachi-GE Nuclear Energy, Ltd., “Topic Report on Reduction of Source Terms by Operating Practices”, Rev. 0, GA91-9201-0001-00165.

[Ref-46] M. Aizawa et al., J. Nucl. Sci. and Technol., Vol.39 [10], p.1051-1059 (2002).

[Ref-47] Institute of Nuclear Power Operations, “Guidelines for Achieving Excellence in Foreign Material Exclusion (FME)”, INPO 07-008, Revision 1, 2011.

[Ref-48] International Atomic Energy Agency, “Commissioning for Nuclear Power Plants”, Specific Safety Guide No. SSG-28, May 2014.

[Ref-49] Hitachi-GE Nuclear Energy, Ltd., “Generic PCSR Chapter 29: Commissioning”, GA91-9101-0101-29000 (QGI-GD-0011), Revision B, March 2014.

[Ref-50] Horizon Nuclear Power, “Commissioning Strategy – Construction Test Phase”, HNP-S3-CMG-STR-00003, Revision 2, January 2016.

[Ref-51] Horizon Nuclear Power, “Pre-operational Test Phase Commissioning Strategy”, HNP-S3-CMG-STR-00002, Revision 2, January 2016.

[Ref-52] Horizon Nuclear Power, “Commissioning Strategy - Start Up Test Phase”, HNP-S3- CMG-STR-00004, Revision 2, January 2016.

[Ref-53] Hitachi-GE Nuclear Energy, Ltd., “Water Chemistry in ABWRs”, WPE-GD-0002, Rev.0, July 2013.

[Ref-54] Hitachi-GE Nuclear Energy, Ltd., “Reagent Addition (BAT aspects) (Response to RQ-ABWR-0363)”, GA91-9201-0003-00606 (WPE-GD-0117), Rev.0, March 2015.

[Ref-55] Hitachi-GE Nuclear Energy, Ltd., “UK ABWR Reactor Chemistry Safety Case: Demonstration that the Primary Cooling System Operating Chemistry reduces risks SFAIRP”, GA91-9201-0003-00455 (WPE-GD-0071), Rev. 2, September 2015.

[Ref-56] Hitachi-GE Nuclear Energy, Ltd., “Topic Report on Reduction of Source Terms by Operating Chemistry”, GA91-9201-0001-00166, Rev 1.0, December 2016.

[Ref-57] Hitachi-GE Nuclear Energy Ltd, “Topic Report of Radioactivity Behaviour in UK ABWR, GA91-9201-0001-00248, Revision 0, DATE TBC.

[Ref-58] EPRI, BWRVIP-137: BWR Vessel and Internals Project Optimizing Coolant Chemistry for Radiation Control Using Depleted Zinc with Noble Metal Chemical Application; Final Report, 1009569, November 2004

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[Ref-59] EPRI, BWRVIP-159: BWR Vessel and Internals Project – HWC/NMCA Experience Report and NMCA Applications Guidelines, 1013397, November 2006

[Ref-60] Hitachi-GE Nuclear Energy, Ltd., “Specification of Low Cobalt Material (Response to RQ-ABWR-0227)”, GA91-9201-0003-00428 (WPE-GD-0093), Rev.0, December 2014.

[Ref-61] Hitachi-GE Nuclear Energy, Ltd., “Submission of the Establishment permit application for Nuclear Reactor Installation Appendix 8 of Japanese ABWR”, SE- UK-0017, Rev. 0, April 2013.

[Ref-62] Hitachi-GE Nuclear Energy, Ltd., “Improvements applied to ABWR (Response to RQ-ABWR-0227)”, GA91-9201-0003-00347 (XE-GD-0262), Rev. 0, October 2014.

[Ref-63] Hitachi-GE Nuclear Energy Ltd, UK ABWR GDA (Generic Design Assessment) Material Selection Report, GA11-1001-0002-00001, Rev 2, December 2015

[Ref-64] Hitachi-GE Nuclear Energy Ltd, Basis of Safety Case Reactor Water Clean-up System, Document Number TBC, Date TBC

[Ref-65] Hitachi-GE Nuclear Energy, Ltd., “Water Quality Specification”, GA24-1001-0001- 00001 (WPE-GD-0016), Rev.1, August 2015.

[Ref-66] Hitachi-GE Nuclear Energy, Ltd., “Generic PCSR Chapter 19: Fuel Storage and Handling”, GA91-9101-0101-19000 (M1D-UK-0004), Rev. B, October 2015.

[Ref-67] Hitachi-GE Nuclear Energy, Ltd., “Generic PCSR Sub-chapter 18.2: Liquid Radioactive Waste Management System”, GA91-9101-0101-18002 (WE-GD- 0019), Rev. B, October 2015.

[Ref-68] Hitachi-GE Nuclear Energy, Ltd., “Selection of the treatment technology for LCW system, HCW system and LD system”, GA91-9201-0003-00346 (WE-GD-0023), Rev.1, January 2016.

[Ref-69] Environment Agency, Generic Design Assessment UK EPR nuclear power plant design by AREVA NP SAS and Electricité de France SA, Consultation Document, GEHO0510BRUV-E-P, June 2010.

[Ref-70] Horizon Nuclear Power Ltd., “Integrated Waste Strategy Wylfa Newydd Power Station”, HNP-S3-EWM-STR-00006, Revision 1.0, June 2016.

[Ref-71] Hitachi-GE Nuclear Energy, Ltd., “Design Policy for Prevention of Leakage”, GA91- 9201-0003-00256 (WJ-GD-0052), Rev. 1, February 2015.

[Ref-72] Hitachi-GE Nuclear Energy, Ltd., “Design Specification for Penetration for Wall and Floor”, GA40-1001-0003-00001 (8D-GD-M0003), Rev.0, July 2014.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[Ref-73] Hitachi-GE Nuclear Energy, Ltd., “Piping Design Specification”, GA31-1001-0004- 00001 (PD-GD-0007), Rev.1, October 2015.

[Ref-74] Hitachi-GE Nuclear Energy, Ltd., “Piping Joint Design Specification”, GA31-1001- 0006-00001 (PD-GD-0008), Rev.1, October 2015.

[Ref-75] Hitachi-GE Nuclear Energy, Ltd., “Off-Gas System – System Design Description”, GN62-1001-0001-00001 (GD-GD-0001), Rev. 0, July 2013.

[Ref-76] Hitachi-GE Nuclear Energy, Ltd., “Generic PCSR Sub-chapter 18.3: Off-Gas Radioactive Waste Management System”, GA91-9101-0101-18003 (GE-GD- 0022), Rev B, October 2015.

[Ref-77] Hitachi-GE Nuclear Energy, Ltd., “Generic PCSR Sub-chapter 28: ALARP Evaluation”, Rev B, GA91-9101-0101-28000, October 2015.

[Ref-78] Hitachi-GE Nuclear Energy, Ltd., “Generic PCSR Sub-chapter 16.3: Heating Ventilating and Air Conditioning System”, GA91-9101-0101-16003 (HPE-GD- H018), Rev. B, October 2015.

[Ref-79] GE Energy, “ABWR Plant General Description”, December 2006.

[Ref-80] Hitachi-GE Nuclear Energy, Ltd., “Generic PCSR Sub-chapter 12.1: Reactor Coolant Systems and Associated Systems”, GA91-9101-0101-12001 (SE-GD- 0134), Rev. B, October 2015.

[Ref-81] Hitachi-GE Nuclear Energy, Ltd., “Improvements of leak tightness”, GA91-9201- 0003-00551 (SE-GD-0011), Rev.3, March 2015.

[Ref-82] Hitachi-GE Nuclear Energy, Ltd., “Assessment of Treatment Techniques for Carbon-14 and Tritium”, GA91-9201-0003-00197 (WE-GD-0026), Rev.2, February 2016.

[Ref-83] Hitachi-GE Nuclear Energy, Ltd., “Off-gas system operational regime (Response to RQ-ABWR-0244)”, GA91-9201-0003-00417 (GE-GD-0029), Rev.0, January 2015.

[Ref-84] International Atomic Energy Agency, “Management of Waste Containing Tritium and Carbon-14”, Technical Reports Series No. 421, 2004.

[Ref-85] Horizon Nuclear Power Ltd., Assessment of Treatment Techniques for Carbon-14 and Tritium at Wylfa Newydd, DCRM Ref Number: TBC, Revision, TBC.

[Ref-86] Hitachi-GE Nuclear Energy, Ltd., “OG Charcoal Adsorber Design Description (Response to RQ-ABWR-0240)”, GA91-9201-0003-00309 (GE-GD-0026), Rev.1, May 2015.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[Ref-87] Hitachi-GE Nuclear Energy, Ltd., “OG Charcoal Adsorber efficiency for 60 years of operation (Response to RQ-ABWR-0193)”, GA91-9201-0003-00430 (GE-GD- 0031), Rev.1, December 2015.

[Ref-88] Environment Agency, “Abatement of Radioactive Releases to Water from Nuclear Facilities”, Technical Guidance Note (Abatement), A6, 1999.

[Ref-89] Hitachi-GE Nuclear Energy, Ltd., “Performance of OG Charcoal Adsorber for radioactive decay”, GA91-9201-0003-00115 (GE-GD-0006), Rev. 5, November 2015.

[Ref-90] Hitachi-GE Nuclear Energy, Ltd., “OG Charcoal Adsorber performance for Iodine” GA91-9201-0003-00098 (GE-GD-0005), Rev. 1, May 2014.

[Ref-91] International Atomic Energy Agency, “Treatment of Radioactive Gaseous Waste”, IAEA-TECDOC-1744, July 2014.

[Ref-92] The Nuclear Ventilation Forum on behalf of the Nuclear Industry Safety Directors Forum, “An Aid to the Design of Ventilation of Radioactive Areas”, NVF/DG001, 2009.

[Ref-93] Hitachi-GE Nuclear Energy, Ltd., “HVAC Radionuclide Concentrations (Response to RQ-ABWR-0243)”, GA91-9201-0003-00536 (HE-GD-0073), Rev.2, June 2016.

[Ref-94] Hitachi-GE Nuclear Energy, Ltd., “Iodine Abatement Techniques for the HVAC (Response to RQ-ABWR-0424)”, GA91-9201-0003-00667 (XE-GD-0349), Rev.1, February 2016.

[Ref-95] Hitachi-GE Nuclear Energy, Ltd., “Topic Report on ALARP Assessment for R/A, T/B, Rw/B and S/B HVAC Systems”, GA91-9201-0001-00196 (HPE-GD-H050), Rev.0, February 2016.

[Ref-96] Menter Newydd, Horizon Nuclear Power Wylfa Ltd, Wylfa Newydd Nuclear Power Plant (ECE), Heating, Ventilation and Air Conditioning System, - System Design Description, HPD-UW-H001, Rev 0, March 2017

[Ref-97] Environment Agency, Abatement of Atmospheric Radioactive releases from Nuclear facilities, Technical Guidance Note A5, ISBN 0 11 310134 1, August 1998.

[Ref-98] Hitachi-GE Nuclear Energy, Ltd., “Improvement of ABWR to Reduce Environmental Impact – HVAC”, GA91-9201-0003-00257 (HPE-UK-H003), Rev. 0, April 2013

[Ref-99] Hitachi-GE Nuclear Energy, Ltd., “Prospective Dose Modelling”, GA91-9901-0026- 00001 (HE-GD-0005), Rev. F, July 2016.

[Ref-100] Hitachi-GE Nuclear Energy, Ltd., “Turbine Gland Steam System: Demonstration of BAT”, GA91-9201-0003-01362, Rev. 0, July 2016.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[Ref-101] Horizon Nuclear Power, Radioactive Waste Processing and Storage Facilities, WN01.05.01-S3-EWM-REP 00001, Rev 1.0, July 2015.

[Ref-102] Horizon Nuclear Power, Justification for Design Changes to the Radwaste Processing, Storage and Disposal Facilities Following Site Optimisation, Document Number TBC.

[Ref-103] Hitachi-GE Nuclear Energy, Ltd., “Latest Liquid Waste and Solid Waste Treatment System and Development”, GA91-9901-0049-0001 (WJ-GD-0017), Rev. 1, April 2014.

[Ref-104] Holtec International, Central Interim Storage Facility for Spent Fuel and HLW (HI- STORE), NRC DSFM REG CON 2015, November 19th 2015

[Ref-105] Hitachi-GE Nuclear Energy, Ltd., “Implementation of review for “Design Policies” (Response to RQ-ABWR-0238)”, GA91-9201-0003-00552 (WE-GD-0053), Rev.0, February 2015.

[Ref-106] Hitachi-GE Nuclear Energy, Ltd., “Available option for processed water unsuited for reuse and discharge”, GA91-9201-0003-00324 (WE-GD-0024), Rev.1, January 2016.

[Ref-107] Hitachi-GE Nuclear Energy, Ltd., “Evaluation of the water balance of the plant and the RW system operations to avoid discharge from HCW system (Response to RQ-ABWR-0233)”, GA91-9201-0003-00342 (WE-GD-0042), Rev.0, January 2015.

[Ref-108] Hitachi-GE Nuclear Energy, Ltd., “Discharge frequency from HCW system”, GA91- 9201-0003-00341 (WE-GD-0041), Rev.0, December 2014.

[Ref-109] Hitachi-GE Nuclear Energy, Ltd., “Other Environmental Regulations”, GA91-9901- 0027-00001 (XE-GD-0098), Rev. F, July 2016.

[Ref-110] Hitachi-GE Nuclear Energy, Ltd., “Liquid/Solid Waste Management System; Proposed WMS for Wylfa Newydd NPP”, WJ-UK-0601, Version 2, September 2016.

[Ref-111] Hitachi-GE Nuclear Energy, Ltd., “Liquid waste treatment system tank capacity calculations”, GA91-9201-0003-00305 (WE-GD-0025), Rev. 0, October 2014.

[Ref-112] Environment Agency/SEPA, “Pollution Prevention Guidelines: Above Ground Oil Storage Tanks” (PPG 2), August 2011.

[Ref-113] International Atomic Energy Agency, “Handling and Processing of Radioactive Waste from Nuclear Applications”, Technical Reports Series No. 402, 2001.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[Ref-114] “Assessment of the 4th Round of Reporting on the Implementation of PARCOM Recommendation 91/4 on Radioactive Discharges”, OSPAR Commission document 351/2008, 2008.

[Ref-115] Hitachi-GE Nuclear Energy, Ltd., “BAT Optioneering Report”, GA91-9201-0003- 00325 (WE-GD-0027), Rev. 1, January 2016.

[Ref-116] Hitachi-GE Nuclear Energy, Ltd., “Radioactive Waste Management Arrangements”, GA91-9901-0022-00001 (WE-GD-0001), Rev. G, July 2016.

[Ref-117] Hitachi-GE Nuclear Energy, Ltd., “Design approach to avoid the necessity of discharges from the HCW (Response to RQ-ABWR-0238)”, GA91-9201-0003- 00367 (SBE-GD-0020), Rev.0, December 2014.

[Ref-118] Horizon Nuclear Power Ltd, Philosophy for the Operational Management of Lower Activity Radioactive Waste (LAW), HNP-S3-EWM-STR-00013, Revision 1.0, November 2016.

[Ref-119] Hitachi-GE Nuclear Energy, Ltd., “Measurement and monitoring of HCW system for normal operation (Response to RQ-ABWR-0239)”, GA91-9201-0003-00340 (WE- GD-0040), Rev.1, January 2016.

[Ref-120] Nuclear Decommissioning Authority, Interim Storage of Higher Activity Waste Packages – Integrated Approach, Issue 3, December 2016

[Ref-121] Department for Environment, Food and Rural Affairs, Radioactive Waste Management Advisory Committee, “Current Arrangements and Requirements for the Conditioning, Packaging and Storage of Intermediate Level Radioactive Waste”, 2002.

[Ref-122] Cavendish Nuclear, “Horizon Operational Waste Optioneering Study”, P0153- 10059, Issue 2, March 2015.

[Ref-123] Horizon Nuclear Power, “Technical Specification for the Spent Nuclear Fuel, High Level & Intermediate Level Radioactive Waste Storage Facilities”, HNP-S3-EWM- SPE-00036, Draft 0.1, Oct 2016.

[Ref-124] Hitachi-GE Nuclear Energy, Ltd., “Description of Improvements of ABWR System”, SE-UK-0013, Rev. 1, July 2013.

[Ref-125] Hitachi-GE Nuclear Energy, Ltd., “Estimation of PCV volume for BWR-5 and ABWR (Response to RQ-ABWR-0232 Query 3)”, GA91-9201-0003-00461 (DCE- GD-0022), Rev. 1, February 2015.

[Ref-126] Hitachi-GE Nuclear Energy, Ltd., “The Development of Primary Containment Vessel Design”, SE-GD-0018, Rev. 1, January 2014.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[Ref-127] Hitachi-GE Nuclear Energy, Ltd., “Overview of ongoing researches and developments for minimising environmental impacts”, SE-GD-0016, Rev. 0, July 2013.

[Ref-128] Office for Nuclear Regulation, “Civil Engineering; Nuclear Safety Technical Assessment Guide”, NS-TAST-GD-017, Revision 3, May 2013.

[Ref-129] Nuvia, “Optioneering Study to Report for Hitachi-GE ABWR Piping Systems”, 80734/DE/TR/004, Issue C, March 2016.

[Ref-130] Hitachi-GE Nuclear Energy Ltd., “Radioactive Solid Wastes Monitoring Requirements”, GA91-9201-0003-00629, Rev 1, March 2016.

[Ref-131] Horizon Nuclear Power, “Wylfa Newydd Plot Plan”, 310QC85-330, Rev 2 Draft B, November 2015.

[Ref-132] Health and Safety Executive, “Work with Ionising Radiation; Ionising Radiations Regulations; Approved Code of Practice and guidance”, ISBN 978 0 7176 1746 3 2012.

[Ref-133] Horizon Nuclear Power, “Design option 2: Site Laundry for Radiation Protection Clothing – Executive Summary”, HNP-S3-EWM-REP-00010, Rev 1, October 2015.

[Ref-134] Horizon Nuclear Power, “Maintenance Philosophy”, HNP-S10-OPS-REP-00012, Revision 0.1, August 2016.

[Ref-135] Environment Agencies Requirements Working Group (EARWG), “Best Practice in Waste Minimisation”.

[Ref-136] Nuclear Decommissioning Authority, Integrated Waste Management, Overview, 11636153, December 2009.

[Ref-137] Department of Energy and Climate Change, The Energy Act 2008: Funded Decommissioning Programme Guidance for New Nuclear Power Stations, December 2011.

[Ref-138] Office for Nuclear Regulation, “Joint Regulatory Guidance - The Management of Higher Activity Radioactive Waste on Nuclear Licensed Sites”, ONR/EA/SEPA/NRW, Revision 2, February 2015.

[Ref-139] Horizon Nuclear Power Ltd, “Characterisation Requirements for Radioactive Wastes”, April 2016.

[Ref-140] Nuclear Decommissioning Authority, “Geological Disposal – Generic Waste Package Specification”, NDA/RWMD/067, March 2012.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[Ref-141] LLWR Ltd, “Waste Services Contract, Waste Acceptance Criteria – Low Level Waste Disposal”, WSC-WAC-LOW, Version 5.0, July 2016.

[Ref-142] LLWR Ltd, “Waste Services Contract, Waste Acceptance Criteria –Combustible Waste Treatment, WSC-WAC-COM”, Version 3.0, April 2012.

[Ref-143] LLWR Ltd, “Waste Services Contract, Waste Acceptance Criteria – Metallic Waste Treatment”, WSC-WAC-MET, Version 3.0, April 2012.

[Ref-144] LLWR Ltd, “Waste Services Contract, Waste Acceptance Criteria – Supercompactable Waste Treatment”, WSC-WAC-SUP, Version 3.0, April 2012.

[Ref-145] LLWR Ltd, “Waste Services Contract, Waste Acceptance Criteria – Very Low Level Waste Disposal”, WSC-WAC-VER, Version 3.0, April 2012.

[Ref-146] Cavendish Nuclear, “Horizon Dry Solid HAW Optioneering Study”, P0153-10065, Issue 003, March 2015.

[Ref-147] Hitachi-GE Nuclear Energy, Ltd., “Generic PCSR Sub-chapter 18.4: Solid Waste Management System”, GA91-9101-0101-18004 (WE-GD-0020), Rev. B, October 2015.

[Ref-148] Hitachi-GE Nuclear Energy, Ltd., “System Design Description (Dry-Solid ILW Management)”, GA91-9201-0003-00330 (WE-GD-0033), Rev.1, February 2016.

[Ref-149] Hitachi-GE Nuclear Energy, Ltd., “System Design Description (Wet-Solid ILW Treatment Facility)”, GA91-9201-0003-00328 (WE-GD-0031), Rev.1, February 2016.

[Ref-150] Hitachi-GE Nuclear Energy, Ltd., “System Design Description (Wet-Solid LLW Treatment Facility)”, GA91-9201-0003-00327 (WE-GD-0030), Rev.1, February 2016.

[Ref-151] Department for Environment, Food and Rural Affairs, DTI and the Devolved Administrations, “Policy for the Long Term Management of Solid Low Level Radioactive Waste in the United Kingdom”, 2007.

[Ref-152] International Atomic Energy Agency, “The Management System for the Processing, Handling and Storage of Radioactive Waste; Safety Guide” No. GS-G-3.3, June 2008.

[Ref-153] Hitachi-GE Nuclear Energy, Ltd., “Waste Enquiry Form”, XE-GD-0209, Rev.0, HIT- ABW-WEN-L-0001, 7 July 2014.

[Ref-154] LLW Repository Ltd, “Agreement in Principle for Wylfa Newydd Power Station”, HIT-ABW-WEN-L-0001 AIP, 23 July 2014.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[Ref-155] Hitachi-GE Nuclear Energy, Ltd., “Compatibility Assessment of LLW to LLWR Waste Acceptance Criteria”, GA91-9201-0003-00332 (WE-GD-0035), Rev.1, January 2016.

[Ref-156] Hitachi-GE Nuclear Energy, Ltd., “Disposability Assessment Submission (Spent Fuel)”, XE-GD-0140, Rev.0, June 2014, Sent via Letter No. HGNE-REG-0052N.

[Ref-157] Hitachi-GE Nuclear Energy Ltd., “Response to RWM assessment report on UK ABWR waste and spent fuel disposability”, GA91-9201-0003-01150, (not issued).

[Ref-158] Hitachi-GE Nuclear Energy, Ltd., “High Level Optioneering on Spent Fuel Storage”, GA91-9201-0003-00458 (FRE-GD-0019), Rev.0, March 2015.

[Ref-159] Horizon Nuclear Power, “Design Option 131: Scheduling of Cask Supplier Decision”, Ref: HNP-S3-DA-REP-00034, Revision 0.1, August 2016.

[Ref-160] Nuclear Decommissioning Authority, “Strategic BPEO for Metallic Waste Management – Options Evaluation”, 2006.

[Ref-161] LLW Repository Ltd., “Strategic BPEO Study in the Management of Combustible Low Level Radioactive Waste”, 2008.

[Ref-162] LLW Repository Ltd., “Strategic BPEO Study for Very Low Level Waste”, 2009.

[Ref-163] Hitachi-GE Nuclear Energy, Ltd., “Plot Plan”, 310QC85-330, Rev.2, March 2016.

[Ref-164] Horizon Nuclear Power, “Wylfa Newydd Reactor Stack Height, Strategic Optioneering Report”, WD03.03.01-S5-PDC-REP-00002, Rev 0.4, May 2015.

[Ref-165] Horizon Nuclear Power, “Radioactive Effluent and Circulating Water Discharge Best Available Technique (BAT) Assessment”, HNP-S3-EWM-REP-00016, Revision 1.0, April 2016

[Ref-166] Horizon Nuclear Power, “Technical Query; Operation – Abstraction”, MPTQ_155_1, Revision 1, April 2016.

[Ref-167] SD-UW-0017, Rev 1, Date TBC.

[Ref-168] Horizon Nuclear Power, “Response to L&P PS06 Cooling Water Systems”, WN009-S3-ME-TEC-00001, Revision 2.0, August 2016.

[Ref-169] Environment Agency, “Initial Radiological Assessment Methodology – Part 1 User Report Science Report: SC030162/SR1”, May 2006.

[Ref-170] Environment Agency, “Initial Radiological Assessment Methodology – Part 2 Methods and Input Data Science Report: SC030162/SR2”, May 2006.

Wylfa Newydd Project – Best Available Techniques (BAT) Case

[Ref-171] Horizon Nuclear Power “Environmental Permit Application - Radioactive Substances Activity”, HNP-S5-PDC-REP-00005, (not issued).

[Ref-172] Horizon Nuclear Power, Report on Normal Operations – Near and Far Field Human Impacts”, HNP-S5-PDC-REP-00007 Rev 2.0, Date TBC

[Ref-173] Euratom, “Commission Recommendation of 18 December 2003 on Standardised Information on Radioactive Airborne and Liquid Discharges into the Environment from Nuclear Power Reactors and Reprocessing Plants in Normal Operation”, 2004 (2004/2/Euratom).

[Ref-174] Hitachi-GE Nuclear Energy, Ltd., “Nuclide Selection by End User requirement”, GA91-9201-0003-00941, HE-GD-0116, Rev. 0, September 2015.

[Ref-175] Environment Agency, “Criteria for setting limits on the discharge of radioactive waste from nuclear sites”, Version 1.0, June 2012.

[Ref-176] British Standards Institution, “Sampling Airborne Radioactive Materials from the Stacks and Ducts of Nuclear Facilities”, BS ISO 2889:2010.

[Ref-177] British Standards Institution, “Equipment for continuous monitoring of radioactivity in gaseous effluents – Part 3: Specific Requirements for Radioactive Noble Gas Monitors”, BS EN 60761-3:2004.

[Ref-178] Environment Agency, “Monitoring of Radioactive Releases to Atmosphere from Nuclear Facilities”, Technical Guidance Note (Monitoring) M11, 1999.

[Ref-179] Hitachi-GE Nuclear Energy, Ltd., “Quantification of Discharges and Limits”, GA91- 9901-0025-00001, HE-GD-0004, Rev. E, February 2016.

[Ref-180] British Standards Institution, “Equipment for continuous monitoring of radioactivity in gaseous effluents – Part 1: General requirements”, BS EN 60761-1:2004.

[Ref-181] British Standards Institution, “General Requirements for the Competence of Testing and Calibration Laboratories2, BS EN ISO/IEC 17025:2005.

[Ref-182] Sira Certification, “MCERTS Certified Products: Continuous Water Monitoring System (CWMS) Part 3 – Water Flowmeters”, Issue 60, http://www.csagroupuk. org/wp-content/uploads/2015/11/MCERTSCertifiedProductsCWMSPart3.pdf, (as of November 2015).

[Ref-183] Euratom, “Directives laying down basic safety standards of protection against the dangers arising from exposure to ionising radiation”, 2013 (2013/59/EURATOM).

[Ref-184] Euratom, “Corrigendum to Commission Recommendation 2004/2/Euratom of 18 December 2003 on standardised information on radioactive airborne and liquid

Wylfa Newydd Project – Best Available Techniques (BAT) Case

discharges into the environment from nuclear power reactors and reprocessing plants in normal operation”, 2004.

[Ref-185] Nuclear Safety Commission, “Guideline for Measurement of Radioactive Materials Released from Commercial Light Water Reactor Facilities”, March 2001.

[Ref-186] Environment Agency, “Sampling Requirements for Stack Emission Monitoring, Technical Guidance Note (Monitoring) M1”, March 2016.

[Ref-187] Hitachi-GE Nuclear Energy, Ltd., “Options for the Management of Dry Solid ILW”, GA91-9201-0003-00326 (WE-GD-0028), Rev.0, April 2015.

CONTACT US: Horizon Nuclear Power Sunrise House If you have any questions or feedback 1420 Charlton Court regarding the Wylfa Newydd Project you can Gloucester Business Park contact us on our dedicated Wylfa Newydd Gloucester, GL3 4AE freephone hotline and email address, T +44 (0)1242 508508 by calling on 0800 954 9516 or emailing [email protected] www.horizonnuclearpower.com

All material in this document is, unless specified otherwise, copyright of Horizon Nuclear Power Wylfa Ltd and may not be reproduced without prior permission. Any unauthorised use or copying of the material may violate trademark, copyright and other proprietary rights and civil and criminal statutes. The material shall not be used in any manner that infringes any trademark, copyright or other proprietary rights and Horizon Nuclear Power Wylfa Ltd reserves all rights with respect to such unauthorised use.

HNP-S3-DA-REP-00013