NUCLEAR WASTE MANAGEMENT TECHNICAL SUPPORT DOCUMENT NEW NUCLEAR - DARLINGTON ENVIRONMENTAL ASSESSMENT NK054-REP-07730-00027 Rev 000
Prepared By: Nuclear Waste Management Division Ontario Power Generation Inc.
August 2009
New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
EXECUTIVE SUMMARY
This Technical Support Document (TSD) relates to Nuclear Waste Management for the New Nuclear – Darlington Project (NND) and has been prepared to support the NND Environmental Assessment. It describes available technologies and assesses the radiological effects on members of the public and Nuclear Energy Workers (NEWs) on the Darlington Nuclear (DN) site from operating the on-site radioactive waste storage systems. It has been prepared by the Nuclear Waste Management Division (NWMD) of Ontario Power Generation Inc. (OPG), which carries technical responsibility for the Nuclear Waste Management System for NND. This TSD is one of a series of related documents describing different aspects of the overall effects assessment, one for each environmental component. More details on the basis for the EA are given in Appendix C.
The radioactive low and intermediate level waste (L&ILW) produced during the operation, maintenance, refurbishment, and decommissioning of the reactors will be managed in a similar manner regardless of the reactor design selected. L&ILW will be managed either on the DN site in a L&ILW management facility or transported off-site to be managed at an appropriately licensed facility.
On-site used nuclear fuel storage facilities (both wet and dry) will be part of each of the reactor designs considered. The on-site dry storage of used fuel proposed for all three reactor designs is expansion of the current at the Darlington Waste Management Facility (DWMF). It is assumed that the used fuel will be stored on-site until the federally mandated Nuclear Waste Management Organization (NWMO) takes responsibility for the long-term management of the used fuel. It is assumed that the NWMO long-term management facility will be available within the operating lifespan of NND.
ES-1 Waste Management Concepts
The range of options presented for management of the radioactive wastes that will be generated by the proposed NND is intended to provide bounding conditions for the EA. No decision has yet been taken on which waste management system will be used.
The EA will consider two options for storage of low and intermediate level waste (L&ILW): on-site, using compaction, packaging, and a modular storage building; and off-site, transporting un-processed L&ILW to an appropriately licensed facility. Storage is assumed to be in “standard” L&ILW Storage Buildings (SBs).
The Atomic Energy of Canada Limited (AECL) MACSTOR (Modular Air Cooled STORage) system is the standard used fuel dry storage (UFDS) system offered by AECL for the proposed reactor ACR-1000, consisting of concrete storage cells that provide shielding and convective air cooling. Another option for the ACR-1000 is OPG’s Dry Storage Container (DSC) system - a proven system that has been used for CANada DeuteriumUranium (CANDU) used fuel since 1995.
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Three basic technologies are widely used around the world for dry storage of used Pressurized Water Reactor (PWR) fuels: solid metal casks suitable for transport or storage; concrete canisters consisting of an outer vertical concrete shield with an inner steel liner; and concrete modules, consisting of an outer horizontal concrete shield vault with an inner steel liner.
Long-term management and eventual disposal of used fuel in Canada is the responsibility of the NWMO.
Refurbishment wastes are conservatively assumed to be stored on-site until the stations are decommissioned, at which point they will be transported off-site to a suitably licensed repository. Steam generators may eventually need to be segmented for off-site shipment.
The majority of the decommissioning wastes will be generated at the time the station is dismantled. It is assumed that on-site storage of these wastes will not be required and that they will be sent directly to a suitable repository.
Long term site planning for NND will need to consider space for three L&ILW SBs (4,500 m2), three UFDS buildings (16,000 m2), one UFDS processing building (2,000 m2), one steam generator storage building (4,550 m2), and one refurbishment waste storage building (3,150 m2). The total area should include a minimum 5 m buffer between the storage buildings and the waste management facility fence, plus an additional buffer for security around the UFDS buildings. While there is no reason to believe that a Safety Assessment could not demonstrate that a location north of the CN rail line is feasible, the safety assessment used in this TSD assume that any waste processing or storage building are built south of the CN rail line and no closer than 150 m to the site perimeter fence. However, for EA planning purposes, this TSD has accepted this analysis to demonstrate that the UFDS can be located anywhere on the site. Should the Vendor require the UFDS buildings to be located north of the CN rail line, or any waste processing or storage building to be located closer than 150 m to the site perimeter fence, OPG has committed to updating safety assessment for this location as part of the licensing process.
An EA for the existing DWMF was completed [OPG, 2003] and approved in 2004 for the construction and operation of the processing building for used fuel. The DWMF has been in-service since January 2008.
ES-2 Waste Forecasts
Operational Wastes will include both L&ILW and used nuclear fuel.
L&ILW is assumed to be largely similar to wastes from OPG's current reactor fleet, both in its physical characteristics and in its radiological activity levels. The volumes and types of L&ILW for each reactor are summarized, both annually and the expected lifetime arisings for a nominal 60 years operation: for both the ACR-1000 and AP1000
ES-2 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document designs, the expected lifetime arisings are just below 10,000 m3; for the Areva US EPR (EPR) design, approximately 13,500 m3 lifetime arisings are expected. An important “new” waste stream for the AP1000 and Evolutionary Pressurized Reactor (EPR) is related to the boric acid system used for reactivity control in light water reactors.
All three reactor technologies under consideration for use at NND use enriched fuel, although the degree of enrichment varies by reactor type. The dimensions and configuration of ACR-1000 fuel is quite similar to conventional CANDU fuel. The fuel used by the AP1000 and EPR are of an entirely different design.
The ACR-1000 is operated with on-power refuelling: fuel bundles are discharged in a regular stream to the fuel bay and stored in baskets designed to be compatible with the AECL MACSTOR dry storage system. PWRs are operated in batch cycles of 18 months to 2 years, after which the reactor is shut down and a portion of the core removed and replaced with new fuel. The expected used fuel arisings (in tonnes of uranium over the lifetime) from each of the reactor designs are: ACR-1000, 5,246; AP1000, 1,400; and EPR, 2,712.
For refurbishment wastes, this study assumes that mid-life refurbishment will be required. For the ACR-1000, this would consist of replacing the steam generators, fuel channels, calandria tubes, and feeders. For the AP1000 and EPR, refurbishment would consist of replacing the steam generators and reactor vessel heads.
ES-3 Waste Management Considerations
ES-3.1 Waste Characteristics
L&ILW from the light water reactors are expected to have much less tritium and C-14 than the current CANDU reactors. Tritium and C-14 from the ACR-1000 are expected to be comparable to current operations if no tritium removal facility is in operation. An important “new” waste stream to Canadian power reactors is related to the boric acid system used for reactivity control in light water reactors and criticality control in the used fuel bays.
The fuel from the new-build reactors will have higher enrichment than current CANDU fuels. This introduces elements of criticality control requirements for storage as well as potential heat load issues for dry storage and eventual disposal.
The light water reactor fuel assemblies are much larger and heavier than the traditional CANDU fuel bundle.
The steam generators for the reactor types considered are larger and heavier than those used in OPG’s existing reactor fleet, making procedures surrounding their eventual replacement more complex. The radioactivity in a steam generator is expected to be similar for all reactor types.
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ES-3.2 Processing
There are two processing scenarios being considered for LLW in the NND environmental assessment: on-site processing, consisting of compaction and storage; and transport to an off-site facility. On-site processing using compaction will reduce the volume for eventual off-site transportation but is the not the most effective technique to minimize overall storage requirements. Off-site processing can further reduce disposal volumes by the use of more advanced processing technologies at the expense of increased initial transportation.
Processing of used fuel refers to preparation for dry storage, which is well-developed both within OPG and internationally. Some modifications to available designs will be required for all reactor types due to the new fuel dimensions and for the ACR-1000 higher burnup and heat load.
Refurbishment waste may require decontamination and/or size reduction. Apart from the large size and weight of the objects, no technical issues are expected with processing refurbishment waste.
ES-3.3 Storage
The storage facilities must be designed to meet the regulatory dose rate limits of an averaged 0.5 µSv/hr at the facility fence and 1 mSv/yr at the station boundary.
All three reactor types have 10 to 15 years of wet bay storage and assume that older fuel will be transferred to dry storage as the bay fills up. The timing of dry storage operation depends on the fuelling cycle. For the purposes of this report, it is assumed that 50% of the used fuel during the reactor lifetime will require on-site dry storage. After this, it is assumed that the NWMO will take ownership of the fuel, and that a long term waste management facility will be in place.
Storage of refurbishment wastes is required for fuel channel components (ACR-1000), reactor vessel heads (AP1000 and EPR) and steam generators (all reactor types). It is assumed that refurbishment wastes will be stored on-site in a dedicated storage building. Fuel channel components and reactor vessel heads will require shielded storage. Currently, OPG uses shielded containers in storage buildings as the reference concept for future reactor refurbishments.
ES-3.4 Long-term Management
Operational L&ILW will eventually require transfer to a suitably licensed, long-term management facility.
Long-term management of refurbishment wastes faces similar issues as operational L&ILW. The most likely destination for these wastes would be the decommissioning waste repository due to the similarity to anticipated decommissioning wastes.
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The responsibility for long-term used fuel management lies with the Nuclear Waste Management Organization (NWMO). Its Adaptive Phased Management approach has been approved by the federal government and is now entering the siting phase. The assumed in-service date of a long-term management facility for used fuel is 2035. Some preliminary implications of NND on the NWMO have been assessed.
ES-3.5 Off-site Transportation
Future transportation of L&IL radioactive materials for NND to an off-site licensed facility will be conducted under OPG’s Radioactive Material Transportation (RMT) Program. If additional transportation packages are required for the transport of specific radioactive wastes from NND, these packages would be designed, certified as necessary, procured according to OPG's existing processes, and will comply with Canadian Packaging and Transport of Nuclear Substances Regulations.
The regulatory requirements on the design of transportation packages used to move L&IL waste between sites, OPG’s existing well-developed transportation program, the many years of experience in transporting radioactive materials, and the training required for personnel involved with transportation and the Transportation Emergency Response Plan would ensure that future transportation incidents remain rare.
The bounding scenario for off-site shipment of LLW is the EPR which assumed that all their generated radioactive waste will be shipped off-site for processing and storage. This bounding scenario would result in a 38,700 m3 lifetime arising of LLW for the EPRs, which requires approximately 1,935 truck shipments of 20 m3 each over a 60 year period of NND, or about two to three truck shipments per month.
For ILW, the lifetime generation from the AP1000 is the bounding quantity of approximately 688 m3 per reactor. For the four AP1000 reactors, the lifetime volume generated would also result in two to three truck shipments per month during the operating period. Note that the peak shipping rates may be higher during outage campaigns, but the lifetime average shipping rate is still very low.
Other shipments of radioactive materials, contaminated equipment and contaminated clothing would also periodically occur. An example of this might be the shipment of tritiated heavy water for off-site upgrading and detritiation.
ES-4 Malfunctions and Accidents
Bounding accidents for the different reactor designs for L&ILW types and used fuel loading, storage and transfer were identified. The hypothetical public and NEW radiation doses were calculated for these bounding accidents.
The overall bounding accident is a drop of a PWR dry fuel storage cask containing 40 used fuel assemblies with a 30% failure of the fuel elements. Using a slightly more
ES-5 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document conservative source term than the vendor supplied information, the hypothetical public doses to a member of the public are 0.237 mSv (237 µSv) to an adult and 0.240 mSv (240 µSv) to an infant. The corresponding hypothetical radiation dose to a NEW is 33.9 mSv. The hypothetical radiation doses are below their respective annual radiation dose limits of 1 mSv to a member of the public and 50 mSv to a NEW.
Ensuring criticality safety for dry fuel storage under all credible circumstances will be achieved as a design requirement for all three reactor types.
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TABLE OF CONTENTS
1.0 Introduction...... 1-1 1.1 Background...... 1-1 1.1.1 The New Nuclear – Darlington Project ...... 1-1 1.1.2 The New Nuclear – Darlington Environmental Assessment ...... 1-2 1.2 Technical Support Document ...... 1-2 1.3 Nuclear Waste Management Technical Support Document...... 1-3 1.3.1 Storage of NND Operational L&IL Wastes...... 1-4 1.3.2 On-site Storage of NND Used Fuel ...... 1-4 1.4 Scope of TSD in relation to Environmental Assessment...... 1-4 2.0 Waste Management Concepts...... 2-1 2.1 Operational Wastes – Low Level Waste...... 2-1 2.2 Operational Wastes - Intermediate Level Waste ...... 2-2 2.3 Used Fuel...... 2-2 2.3.1 ACR-1000...... 2-2 2.3.2 AP1000 and EPR ...... 2-3 2.3.3 Facility Examples...... 2-9 2.3.4 Used Fuel Disposal ...... 2-13 2.4 Refurbishment Wastes ...... 2-15 3.0 Waste characteristics and Volume Forecasts...... 3-1 3.1 L&ILW Operational Waste Characteristics...... 3-1 3.2 L&ILW Projected Volumes...... 3-2 3.3 L&ILW Projected Storage Requirements...... 3-3 3.3.1 LLW Projected Storage Requirement...... 3-3 3.3.2 ILW Projected Storage Requirements ...... 3-4 3.3.3 Overall Storage Building Requirements for L&ILW ...... 3-5 3.4 Used Fuel...... 3-5 3.4.1 Used Fuel Characteristics and Volumes ...... 3-5 3.4.2 ACR-1000 Bundle Storage System Options...... 3-8 3.4.3 Used Fuel Projected Storage Requirements...... 3-10 3.5 Refurbishment Wastes ...... 3-12 3.6 Summary of Waste Management Needs...... 3-14 4.0 Other Waste Management Considerations...... 4-1 4.1 Processing ...... 4-1 4.1.1 Operational L&ILW...... 4-1 4.1.2 Used Fuel...... 4-1 4.1.3 Refurbishment Wastes ...... 4-2 4.2 Operational Points...... 4-2 4.2.1 Station Storage Provisions ...... 4-2 4.2.2 Used Fuel Management ...... 4-2 4.2.3 Refurbishment Wastes ...... 4-3 4.3 Long-Term Waste Management ...... 4-3 4.3.1 Operational L&ILW...... 4-3 4.3.2 Used Fuel...... 4-4 4.3.3 Refurbishment Wastes ...... 4-5 4.4 Off-Site Transportation and Accidents ...... 4-5 i New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
4.4.1 Overview of Current Radioactive Material Transportation Program ...... 4-5 4.4.2 Transportation of L&IL Radioactive Waste to an Off-site Licensed Facility ...... 4-6 4.4.3 Transportation Package Requirements for Transportation of L&IL Waste...... 4-7 4.4.4 Shipments to an Off-site Licensed Facility...... 4-9 4.4.5 Transportation Emergency Response Plan ...... 4-9 4.4.6 Transportation Summary ...... 4-10 5.0 Normal Operations...... 5-1 5.1 Used Fuel Operations – Radiological Impacts...... 5-1 5.1.1 Introduction...... 5-1 5.1.2 ACR-1000 Normal Operation...... 5-1 5.1.3 EPR Normal Operation...... 5-3 5.1.4 AP1000 Normal Operation ...... 5-4 5.2 Routine Radiological Emissions from Radioactive Waste Processing...... 5-5 5.3 Conventional Emissions...... 5-6 5.4 Impact on the Environment...... 5-7 6.0 Radiological Impact of Malfunctions and Accidents...... 6-1 6.1 Approach to Identifying Malfunctions and Accidents...... 6-1 6.2 Assessment Methodology...... 6-1 6.3 L&ILW - Malfunctions and Accidents...... 6-3 6.3.1 Screening of Malfunctions and Accidents for L&ILW ...... 6-3 6.3.2 Bounding Case for Low Level Waste – Pool Fire Beside Stacked Waste Containers ...... 6-3 6.3.3 Bounding Case for Intermediate Level Waste - Pool Fire Involving Resin Liner...... 6-8 6.4 Refurbishment Waste Storage and Handling...... 6-8 6.4.1 Screening of Malfunctions and Accidents for Refurbishment Waste...... 6-8 6.4.2 Bounding Case - Drop of a Retube Waste Container ...... 6-12 6.4.3 Bounding Case - Drop of a Steam Generator ...... 6-12 6.5 Used Fuel Dry Storage – Assessment of Bounding Accident ...... 6-12 6.6 ACR-1000 Malfunctions and Accidents During Dry Storage of Used Fuel ..... 6-13 6.6.1 Screening of Malfunctions and Accidents for ACR-1000...... 6-13 6.6.2 ACR-1000 – Assessment of Bounding Accident...... 6-14 6.6.3 Criticality Assessment for ACR-1000 ...... 6-14 6.7 EPR Malfunctions and Accidents During Dry Storage of Used Fuel...... 6-20 6.7.1 Screening of Malfunctions and Accidents for EPR ...... 6-20 6.7.2 EPR – Assessment of Bounding Accident...... 6-24 6.7.3 Criticality Assessment for EPR ...... 6-24 6.8 AP1000 Malfunctions and Accidents During Dry Storage of Used Fuel...... 6-25 6.8.1 Screening of Malfunctions and Accidents for AP1000 ...... 6-25 6.8.2 AP1000 – Assessment of Bounding Accident...... 6-29 6.8.3 Criticality Assessment for AP1000...... 6-29 7.0 Summary...... 7-1 8.0 References...... 8-1
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LIST OF APPENDICES
Appendix A: Operational Waste Details...... A8-1 Appendix B: Refurbishment Waste Details...... B-1 Appendix C: New Nuclear - Darlington - Basis for EA ...... C-1
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LIST OF FIGURES
Figure 2.1-1: Typical Storage Building ...... 2-1 Figure 2.3-1: Typical MACSTOR 200 Modules at Gentilly-2...... 2-4 Figure 2.3-2: MACSTOR 200 Module Cross-section ...... 2-4 Figure 2.3-3: OPG Dry Storage Container ...... 2-5 Figure 2.3-4: Typical Metal Cask Container for PWR Fuel...... 2-6 Figure 2.3-5: Concrete Canister for PWR Fuel ...... 2-8 Figure 2.3-6: NUHOMS Concrete Module for PWR Fuel...... 2-9 Figure 2.3-7: ZWILAG Interim Storage for Used Fuel...... 2-10 Figure 2.3-8: Lingen Interim Storage for Used Fuel ...... 2-11 Figure 2.3-9: Typical US Concrete Cask Storage...... 2-11 Figure 2.3-10: Typical US NUHOMS Concrete Vault Storage ...... 2-12 Figure 2.3-11: Typical US Metal Cask Storage...... 2-12 Figure 2.3-12: Typical On-Site Metal Cask Transporter...... 2-13 Figure 2.3-13: Swedish Used Fuel Disposal System...... 2-14 Figure 2.4-1: Conceptual ACR-1000 Steam Generator Storage...... 2-16 Figure 2.4-2: Conceptual ACR-1000 Fuel Channel Waste Storage ...... 2-17 Figure 2.4-3: Bruce A Steam Generator Storage at WWMF...... 2-18 Figure 2.4-4: Palo Verde Steam Generator Storage...... 2-18 Figure 2.4-5: RWC Storage at WWMF ...... 2-19 Figure 3.4-1: EPR Fuel ...... 3-6 Figure 3.4-2: ACR-1000 Fuel...... 3-8 Figure 3.4-3: ACR-1000 Fuel Storage Basket...... 3-8 Figure 3.4-4: ACR-1000 Fuel Basket Stacking Frame...... 3-9 Figure 3.4-5: 60-Bundle MACSTOR Fuel Basket ...... 3-10 Figure 3.5-1: ACR-1000 Fuel Channel...... 3-15 Figure 3.5-2: ACR-1000 Steam Generator ...... 3-16 Figure 3.5-3: AP1000 Reactor Vessel ...... 3-17 Figure 3.5-4: AP1000 Steam Generator...... 3-18 Figure 3.5-5: EPR Reactor Vessel ...... 3-19 Figure 3.5-6: EPR Steam Generator ...... 3-20
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LIST OF TABLES
Table 2.3-1: Examples of Metallic PWR Dry Storage Casks...... 2-7 Table 3.1-1: Summary of Operational L&ILW Characteristics ...... 3-1 Table 3.2-1: Summary of L&ILW Annual Arisings...... 3-2 Table 3.2-2: Summary of L&ILW Lifetime Arisings...... 3-3 Table 3.3-1: Summary of Storage Requirements for Operational LLW ...... 3-4 Table 3.3-2: Summary of Storage Requirements for Operational ILW...... 3-4 Table 3.3-3: On-Site L&ILW Storage Building Expansions...... 3-5 Table 3.4-1: Summary of Fuel Characteristics and Arisings...... 3-7 Table 3.4-4: Summary of Dry Storage Space Requirements...... 3-11 Table 3.4-5: On-Site Used Fuel Storage Building Expansions...... 3-11 Table 3.5-1: On-Site Refurbishment Waste Storage...... 3-13 Table 5.1-1: Properties of 10 Year Cooled ACR-1000 Fuel ...... 5-1 Table 5.1-2: Properties of 10 Year Cooled EPR Fuel...... 5-3 Table 6.3-1: Screening of Malfunctions and Accidents for L&IL Waste...... 6-5 Table 6.4-1: Screening of Malfunctions and Accidents for Refurbishment Waste ...... 6-9 Table 6.6-1: Screening of Malfunctions and Accidents for ACR-1000 ...... 6-15 Table 6.7-1: Properties of 10 Year Cooled EPR Fuel...... 6-20 Table 6.7-2: Screening of Malfunctions and Accidents for EPR...... 6-21 Table 6.8-1: Properties of 10 Year Cooled AP1000 Fuel...... 6-25 Table 6.8-2: Screening of Malfunctions and Accidents for AP1000...... 6-26 Table A-1: Summary of Average L&ILW Specific Activity ...... A-1 Table A-2: Typical EPR L&ILW Source Term Details ...... A-2 Table A-3: Details of Processed L&ILW Volumes...... A-3 Table A-4: Annual Waste Forecasts ...... A-4 Table B-1: Summary of Darlington Fuel Channel Component Specific Activity...... B-1 Table B-2: Summary of Steam Generator Activity ...... B-2 Table C-1: New Nuclear – Darlington – Basis for EA ...... C-1
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SPECIAL TERMS
Units
Bq becquerel C Celsius GBq gigabecquerel kg kilogram km kilometre m metre mm millimetre mSv millisievert MW megawatt m2 square meter m3 cubic meter Sv sievert TBq terabecquerel Tonne a unit of weight equivalent to 1000 kilograms µSv microsievert
Abbreviations and Acronyms
ACR-1000 Advanced CANDU 1000, reactor design by Atomic Energy of Canada Ltd. AECL Atomic Energy of Canada Limited ALARA As Low As Reasonably Achievable AP1000 Advance Passive (AP1000) Reactor design offered by Westinghouse APM Adaptive Phase Management CANDU Canada Deuterium Uranium (trademark of Atomic Energy of Canada Limited) CEAA Canadian Environmental Assessment Act CNSC Canadian Nuclear Safety Commission DFO Department of Fisheries and Ocean DGR Deep Geologic Repository DN Darlington Nuclear DNGS Darlington Nuclear Generating Station DWMF Darlington Waste Management Facility DSC OPG’s Dry Storage Container EA Environmental Assessment EF End Fittings EIS Environmental Impact Statement EPR Areva US EPR IAEA International Atomic Energy Agency IC In-ground Container. Current design size is 18 m3, the IC-18. ILW Intermediate-Level Waste JRP Joint Review Panel L&ILW Low and Intermediate Level Waste LLW Low-Level Waste vi New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
LVRF Low Void Reactivity Fuel MACSTOR Modular Air Cooled Used fuel storage system offered by AECL NBC National Building Code of Canada NEW Nuclear Energy Worker NFC National Fire Code of Canada NND New Nuclear – Darlington NU Natural Uranium NWMD Nuclear Waste Management Division NWMO Nuclear Waste Management Organization OD Outer Diameter OL Overall Length OPEX Operating Experience OPG Ontario Power Generation Inc. PNGS Pickering Nuclear Generating Station PPE Plant Parameter Envelope PT Pressure Tube PWR Pressurized Water Reactor PWMF Pickering Waste Management Facility RWC Retube Waste Container RWS Refurbishment Waste Storage SB Storage Building SG Steam Generator TSD Technical Support Document UFDS Used Fuel Dry Storage US DOE U.S. Department of Energy WWMF Western Waste Management Facility
Glossary of Terms
Term Definition ALARA A principle in radiation protection according to which radiation exposures are kept as far below the regulatory limits as reasonable, taking into account social and economic factors. These factors could include, for example, the financial impact of protection measures as balanced against the benefit obtained. Becquerel The standard international unit of radioactivity equal to one radioactive disintegration per second. Darlington Waste The DWMF provides dry fuel storage for the Darlington Management Facility reactors. (DWMF) Dry Storage Placement of used fuel in an engineered, dry environment for storage, such as in concrete dry storage containers. Fuel is cooled for at least 10 years in the Irradiated Fuel Bay before transfer to dry storage. vii New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
Term Definition In-Ground Storage Storage of radioactive waste in in-ground storage containers (ICs) generally used for intermediate-level waste. All ICs with the exception of those used for heat exchangers consist of steel liners fixed with concrete inside boreholes in the ground. Irradiated Fuel Bay Water-filled pool-type storage (also called “wet bay”, or “wet storage”), located at reactor sites, in which used nuclear fuel is stored, cooled and shielded. Intermediate Level Waste Consists mostly of used reactor components, as well as the resins and filters used to keep reactor water systems clean. These items, which cannot be handled without shielding, are stored in steel-lined in-ground storage structures. Low-Level Storage Building A storage building manufactured from prefabricated, pre- stressed concrete, used for low-level waste. Processed and non-processed wastes are stored in a variety of stackable metal containers. Low Level Waste Consists of minimally radioactive materials such as mop- heads, rags, paper towels, floor sweepings and protective clothing used in the nuclear stations during routine operation and maintenance. This waste does not require shielding and, after any processing, is stored in Low Level Storage Buildings. Nuclear Energy Worker A worker who might receive as a result of their work or occupation a radiation dose greater than the dose limit for the general public. Nuclear Waste Management The NWMO was established in 2002 by Ontario Power Organization (NWMO) Generation Inc., Hydro-Québec and New Brunswick Power Corporation. This organization was formed to assume responsibility for the long-term management of Canada’s used nuclear fuel. The NWMO operates in accordance with the Nuclear Fuel Waste Act. Pickering Waste The PWMF provides dry fuel storage for the Pickering Management Facility reactors. (PWMF) Refurbishment Waste Radioactive waste produced from the refurbishment and life extension of reactors including retubing (fuel channel replacement); steam generator replacement (large heavy object wastes, i.e. steam generators); and/or feeder pipe replacement. Repository Facility for the long-term management of waste materials. Sievert A measurement unit of radiation dose. Frequently expressed as millisievert (mSv), equal to one-thousandth of a sievert, or as a microsievert (PSv), equal to one- millionth of a sievert.
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Term Definition Storage The placement of waste in a nuclear facility where isolation, environmental protection and human control, i.e., monitoring, are provided with the intent that the waste will be retrieved for processing and/or transfer to a long- term repository at a later time. Tritiated Heavy Water Results from the substitution of deuterium in heavy water (D2O) with tritium. Used Fuel When a fuel bundle no longer contains enough fissionable uranium to heat water efficiently, the fuel is considered “used” and is then replaced by a new fuel bundle. The used fuel contains more than 99% of the radioactive by- products of nuclear reactors. Waste Management All activities, administrative and operational, that are involved in the handling, pre-treatment, treatment, conditioning, transportation, storage and long-term management of waste from a nuclear facility. Western Waste The WWMF is a centralized processing and storage Management Facility facility for OPG’s low and intermediate- level radioactive (WWMF) wastes, and dry storage for used fuel from the Bruce nuclear generating stations.
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LIST OF TECHNICAL SUPPORT DOCUMENTS (TSDs)
Atmospheric Environment Existing Environmental Conditions TSD – SENES Consultants Limited Atmospheric Environment Assessment of Environmental Effects TSD – SENES Consultants Limited Surface Water Environment Existing Environmental Conditions TSD – Golder Associates Limited Surface Water Environment Assessment of Environmental Effects TSD – Golder Associates Limited Aquatic Environment Existing Environmental Conditions TSD – SENES Consultants Limited and Golder Associates Limited Aquatic Environment Assessment of Environmental Effects TSD - SENES Consultants Limited and Golder Associates Limited Terrestrial Environment Existing Environmental Conditions TSD – Beacon Environmental Terrestrial Environment Assessment of Environmental Effects TSD – Beacon Environmental Geological and Hydrogeological Environment Existing Environmental Conditions TSD – CH2M HILL Canada Limited and Kinectrics Incorporated Geological and Hydrogeological Environment Assessment of Environmental Effects TSD – CH2M HILL Canada Limited Land Use Existing Environmental Conditions TSD – MMM Group Limited Land Use Assessment of Environmental Effects TSD – MMM Group Limited Traffic and Transportation Existing Environmental Conditions TSD – MMM Group Limited Traffic and Transportation Assessment of Environmental Effects TSD – MMM Group Limited Radiation and Radioactivity Environment Existing Environmental Conditions TSD – AMEC NSS Radiation and Radioactivity Environment Assessment of Environmental Effects TSD – SENES Consultants Limited and AMEC NSS Socio-Economic Environment Existing Environmental Conditions TSD - AECOM Socio-Economic Environment Assessment of Environmental Effects TSD - AECOM Physical and Cultural Heritage Resources Existing Environmental Conditions TSD – Archaeological Services Incorporated Physical and Cultural Heritage Resources Assessment of Environmental Effects TSD – Archaeological Services Incorporated Ecological Risk Assessment and Assessment of Effects on Non-Human Biota TSD – SENES Consultants Limited Scope of Project for EA Purposes TSD – SENES Consultants Limited Emergency Planning and Preparedness TSD – SENES Consultants Limited and KLD Associates Incorporated Communications and Consultation TSD – Ontario Power Generation Incorporated Aboriginal Interests TSD – Ontario Power Generation Incorporated Human Health TSD – SENES Consultants Limited Malfunctions, Accidents and Malevolent Acts TSD – SENES Consultants Limited Nuclear Waste Management TSD – Ontario Power Generation Incorporated
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1.0 INTRODUCTION
1.1 Background
Ontario Power Generation Inc. (OPG) was directed by the Ontario Minister of Energy in June 2006 to begin the federal approvals process, including an environmental assessment (EA), for new nuclear units at an existing site. OPG initiated this process and in September 2006 submitted an application for a Licence to Prepare Site to the Canadian Nuclear Safety Commission (CNSC) for a new nuclear power generating station at the Darlington Nuclear site (DN site), located in the Municipality of Clarington on the north shore of Lake Ontario in the Region of Durham. The DN site is currently home to the Darlington Nuclear Generating Station (DNGS), a 4-unit plant, the first unit of which was commissioned by OPG in 1990. It remains under OPG’s ownership and operational control.
Before any licensing decision can be made concerning the new nuclear generating station, an EA must be performed to meet the requirements of the Canadian Environmental Assessment Act (CEAA) and be documented in an Environmental Impact Statement (EIS). An EIS is a document that allows a Joint Review Panel, regulators, members of the public and Aboriginal groups to understand the Project, the existing environment and the potential environmental effects of the Project. Guidelines for the preparation of the EIS were prepared by the Canadian Environmental Assessment Agency (the CEA Agency) and the CNSC (in consultation with Department of Fisheries and Oceans Canada (DFO), the Canadian Transportation Agency and Transport Canada). The Guidelines require that the proponent prepare the EIS and support it with detailed technical information which can be provided in separate volumes. Accordingly, OPG has conducted technical studies that will serve as the basis for the EIS. These technical studies are documented in Technical Support Documents (see Section 1.2 below). The basis for the NND project is included in Appendix C.
1.1.1 The New Nuclear – Darlington Project
New Nuclear – Darlington (NND), a new generating station, is proposed to be located primarily on the easterly one-third (approximately) of the DN site, with reactor buildings and other related structures located south of the CN rail line. The proposed New Nuclear – Darlington Project involves the construction and operation of up to four nuclear reactor units supplying up to 4,800 MW of electrical capacity to meet the base load electrical requirements of Ontario. The proposed Project will include:
x Preparation of the DN site for construction of the new nuclear facility; x Construction of the NND nuclear reactors and associated facilities; x Construction of the appropriate nuclear waste management facilities for storage and volume reduction of waste; x Operation and maintenance of the NND nuclear reactors and associated facilities for approximately 60 years of power production (i.e, for each reactor); x Operation of the appropriate nuclear waste management facilities; and
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x Development planning for decommissioning of the nuclear reactors and associated facilities, and eventual turn-over of the site to other uses.
For EA planning purposes, the following temporal framework has been adopted for the Project:
Project Phase Start Finish Site Preparation and Construction 2010 2025 Operation and Maintenance 2016 2100 Decommissioning and Abandonment 2100 2150
1.1.2 The New Nuclear – Darlington Environmental Assessment
The EA considers the three phases of the NND Project (i.e., Site Preparation and Construction, Operation and Maintenance, and Decommissioning and Abandonment) extending over approximately 140 years. In doing so, it addresses:
x The need for, and purpose of the Project; x Alternatives to the Project; x Alternative means of carrying out the Project that are technically and economically feasible, and the environmental effects of such alternatives; x The environmental effects of the Project including malfunctions, accidents and malevolent acts, and any cumulative effects that are likely to result from the Project in combination with other projects or activities that may be carried out; x Measures to mitigate significant adverse environmental effects that are technically and economically feasible; x The significance of residual (after mitigation) adverse environmental effects; x Measures to enhance any beneficial environmental effects; x The capacity of renewable resources that are likely to be significantly affected by the project, to meet the needs of the present and the future; x The requirements of a follow-up program in respect of the Project; x Consideration of community knowledge and Aboriginal traditional knowledge; and x Comments that are received during the EA.
1.2 Technical Support Document
The EA studies were carried out and are documented within a framework of individual aspects or “components” of the environment. The environmental components are:
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x Atmospheric Environment; x Surface Water Environment; x Aquatic Environment; x Terrestrial Environment; x Geological and Hydrogeological Environment; x Land Use; x Traffic and Transportation; x Radiation and Radioactivity Environment; x Socio-Economic Environment; x Physical and Cultural Heritage Resources; x Aboriginal Interests; x Health - Human; and x Health – Non-Human Biota (Ecological Risk Assessment).
Other EA-related studies addressed subjects not associated with a specific environmental component, however, were necessary to support the EA program. These included:
x Scope of the Project for EA Purposes; x Emergency Planning and Preparedness; x Communications and Consultation; x Malfunctions, Accidents and Malevolent Acts; and x Nuclear Waste Management.
The various EA studies are documented in individual Technical Support Documents (TSDs). In most cases where the TSDs relate to environmental components, separate documents were prepared to describe: i) existing environmental conditions; and ii) likely environmental effects. In other cases, the subject of the study is included in a single TSD.
This TSD relates to Nuclear Waste Management. It has been prepared by the Nuclear Waste Management Division of OPG.
1.3 Nuclear Waste Management Technical Support Document
This Technical Support Document (TSD) describes available radioactive waste system and assesses the radiological effects of the Project due to the management of solid radioactive wastes on members of the public and on Nuclear Energy Workers (NEWs). This TSD also describes the malfunction and accident scenarios related to used fuel, and low and intermediate level waste, and identifies and further assesses the bounding scenario for each.
It has been prepared in support of the EA by the Nuclear Waste Management Division of OPG, which carries responsibility for the Nuclear Waste Management System for NND. This TSD is one of a series of related documents describing different aspects of the overall effects assessment, one for each environmental component.
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1.3.1 Storage of NND Operational L&IL Wastes
The radioactive low and intermediate level waste (L&ILW) produced during the day-to-day operation and maintenance of the reactors will be handled in a manner that attempts to minimize the volume produced. Typical examples of L&ILW include ion exchange resins, filters, rags, mops, floor sweepings, tools and clothing that have become contaminated as part of operation and maintenance activities. L&ILW will be managed in a similar manner regardless of the reactor design selected.
For EA purposes, two alternative means of managing L&ILW are proposed: the L&ILW will be managed on-site with an expansion of the existing Darlington Waste Management Facility (DWMF); or transported off-site to be managed at an appropriately licensed facility. The specific types, volumes and characteristics of L&ILW produced during the life of the facility will be described for each reactor class.
1.3.2 On-site Storage of NND Used Fuel
On-site wet storage of used fuel will be part of each of the reactor designs considered. For each reactor design, the facility will provide transfer systems that carry the used fuel from the reactor to an irradiated fuel bay in which the used fuel is stored and cooled. The used fuel will be stored in an irradiated fuel bay until it has cooled sufficiently for dry storage.
The NND Project proposes on-site storage through expansion of the DWMF. For EA planning purposes, the volumes and characteristics of used fuel waste arising from the operation of each reactor will vary depending on the reactor technology.
For EA planning purposes, it is also assumed that the used fuel will continue to be stored on-site until the federally mandated Nuclear Waste Management Organization (NWMO) takes responsibility for the long-term management of the used fuel as directed by the federal government. The site study, construction, and operation of this long-term used fuel management facility will be the subject of its own separate environmental assessment process.
1.4 Scope of TSD in relation to Environmental Assessment
At the time of completing this TSD, three vendors were being considered by the Province of Ontario for supplying and installing the reactors and associated equipment for the Project. Accordingly, the specific reactor to be constructed and operated had not yet been determined. Therefore, for purposes of the EA, the Project was defined in a manner that effectively incorporated the salient aspects of all of the considered reactors. Similarly, the existing environmental conditions and the likely environmental effects of the Project were also determined in a manner that considered the range of reactor types and number of units that may comprise the Project.
The essential aspect of the method adopted for defining the “Project for EA Purposes” is the use of a bounding framework that brackets the variables to be assessed. This bounding framework is defined within a Plant Parameter Envelope (PPE). The PPE is a set of design parameters that
1-4 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document delimit key features of the Project. The bounding nature of the PPE allows for appropriate identification of a range of variables within a project for the purpose of the environmental assessment while also recognizing the unique features of each design. For further information concerning the use of the PPE for this EA, the reader is directed to Section 2.1 of the EIS.
The information presented in this TSD is deemed to be appropriately bounding so as to facilitate the assessment of environmental effects that may be associated with any of the considered reactors. As both the EA studies and the vendor selection programs continue, it may be that aspects of this TSD are updated to respond to these evolving programs, in which case the updated information will be presented in an addendum to this TSD or in the EIS. The EIS itself will remain subject to edits until it has been accepted by the Joint Review Panel (JRP) as suitable for the basis of the public hearing that will be convened to consider the Project.
This TSD is a document prepared in support of the EIS. Where there may be differences in the information presented in the two documents, the EIS will take precedence for the reasons noted above.
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2.0 WASTE MANAGEMENT CONCEPTS
The following sections present information on potential management options for the various waste types. The range of options is intended to provide bounding conditions for the environmental assessment and licensing. No decision has been taken at this time on which options will be used in practice.
The storage facilities must be designed to meet the regulatory dose rate limits of an averaged 0.5 µSv/hr at the facility fence and 1 mSv/yr at the station boundary.
2.1 Operational Wastes – Low Level Waste
The EA will consider two options for the storage of low-level waste (LLW):
a) On-site: consisting of compaction of a portion of the LLW, combined with appropriate packaging and interim storage in a modular storage building on the Darlington site. Eventually, the waste would be transported to an appropriate facility off-site for long- term management.
b) Off-site: consisting of transporting the un-processed LLW to an appropriately licensed facility, such as the Western Waste Management Facility (WWMF), for processing, packaging and storage. Processing would consist of incineration and/or compaction of appropriate portions of the LLW. Eventually, the waste would be transported or transferred to an appropriate facility for long-term management.
In either option, interim storage is assumed to be in “standard” storage buildings (SBs), with a nominal capacity of 7,000 m3 each, as shown in Figure 2.1-1.
FIGURE 2.1-1: TYPICAL STORAGE BUILDING
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Waste processing in option (a) would consist of low-force compaction of the incinerable and compactable portions of the LLW with a box compactor, similar to the one currently used at the WWMF, with an average volume reduction of 4:1. The wastes are packaged into 2.5 m3 steel boxes. Non-processible wastes would be packaged into steel containers, similar to the ones currently used at the WWMF, with an average volume increase of 25% due to the void spaces and other container stacking inefficiencies. Any miscellaneous radioactive liquid wastes would either be solidified or processed (typically by incineration) at a suitably licensed off-site facility.
Only a small processing area is required for compaction of waste and the location of the compactor would be determined later. The compactor could be located within the new powerhouse, within the dry fuel processing building, or if required, a small area for compaction could be set up inside the first storage building. Any location must ensure that the compactor exhaust is both filtered and goes past a monitoring point before discharge.
Waste processing in option (b) would consist of incineration and/or compaction, with an average volume reduction of 40:1 for incinerable wastes and 4:1 for compactable wastes. Non- processible wastes would be packaged with an average volume increase of 25%, as per above.
2.2 Operational Wastes - Intermediate Level Waste
As with the LLW, two options are being considered for ILW:
a) On-site: consisting of self-shielded packaging and interim storage in a modular storage building on the Darlington site. Eventually, the waste would be transported to an appropriate facility off-site for long-term management.
b) Off-site: consisting of shielded transportation to an appropriately licensed off-site facility, such as the WWMF, for storage. Eventually, the waste would be transported or transferred to an appropriate facility for long-term management. Storage could either be in above ground facilities using shielded packages, or in in-ground facilities (such as In- Ground Containers (IC) such as the IC-18) using un-shielded packages within a shielded shipping flasks.
On-site storage facilities might consist of a shared storage building between LLW and intermediate-level waste (ILW), with a segregated area for the ILW. ILW waste might also be in a smaller separate building. Above-ground storage of ILW with LLW would require that all waste be in container and there be supplementary fire protection/fire detection features.
2.3 Used Fuel
2.3.1 ACR-1000
There are two potential systems for the dry storage of ACR-1000 used fuel:
a) AECL MACSTOR: (Modular Air-Cooled STORage) consisting of above-ground, air cooled storage modules, with the fuel placed into unshielded canisters and transferred
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from the reactor to the storage facility in a reusable shielding cask. This is similar to systems that AECL has deployed at Gentilly 2, Korea, and Romania.
b) OPG shielded dry storage container: a free-standing reinforced concrete container, with an inner steel liner and an outer steel shell, which can hold up to 384 fuel bundles. The dry storage containers would be stored in a warehouse-type building, similar to the current operations at DWMF, PWMF and WWMF.
The AECL MACSTOR system is the “standard” dry storage system offered by AECL for the ACR-1000 and uses a 60 bundle storage basket. It is available in several configurations: the MACSTOR 200, shown in Figures 2.3-1 and 2.3-2, holds 200 baskets per module for a total of 12,000 bundles per module. The newer MACSTOR/KN 400 holds 400 baskets per module for a total of 24,000 bundles per module.
Loaded fuel casks are hoisted to the top of the structure by the traveling bridge crane, and the module is then lowered into the storage cell. The concrete storage cells provide shielding and convective air cooling.
OPG’s DSC system for used CANDU fuel is a proven system in use since 1995 at Pickering and more recently at Western and Darlington Waste Management Facilities. The DSCs are engineered to have a life of 50 years but with monitoring and maintenance that they receive, it is likely that they will last long. As shown in Figure 2.3-3, the DSC will hold 384 fuel bundles in 4 standard OPG type fuel modules.
A loaded DSC weighs some 75 tonnes. It is loaded while submerged in the fuel bay.
Prior to transferring fuel to a DSC, the fuel must be loaded into modules, if it is not already stored in that configuration. The loaded module is then placed in the DSC. After seal welding the DSC, it is transferred to a storage building, which is designed to hold a nominal 500 DSCs.
2.3.2 AP1000 and EPR
The used fuels from the Westinghouse AP1000 and Areva US EPR (EPR) are very similar to each other both physically and radiologically. Several basic technologies are in wide use around the world for dry storage of PWR fuels. They are all licensed for this purpose in a number of jurisdictions.
a) Metal casks: Consisting of solid metal casks suitable for transport or storage. The casks typically hold 24 to 40 PWR fuel assemblies. The casks may be stored either indoors (common in Europe) or outdoors on a simple concrete pad (common in the US). For the purposes of this report, a standard cask size of 32 fuel assemblies has been assumed as a reasonable average to calculate storage space requirements. An example of this would be the Transnuclear TN 32 cask. It is also assumed that the casks would be stored inside a building, similar to the European practice. The casks are typically loaded in the fuel bay or through a docking port with the fuel handling system. Lids are typically bolted in
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FIGURE 2.3-1: TYPICAL MACSTOR 200 MODULES AT GENTILLY-2
FIGURE 2.3-2: MACSTOR 200 MODULE CROSS-SECTION
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FIGURE 2.3-3: OPG DRY STORAGE CONTAINER
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FIGURE 2.3-4: TYPICAL METAL CASK CONTAINER FOR PWR FUEL (Dimensions are in mm)
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place. A typical metal cask is shown in Figure 2.3-4, including the impact limiters for transportation. Examples of some available PWR casks are given in Table 2.3-1. (Note that these cask designs are based on current generation PWR fuel lengths (~4.1 m Overall Length (OL)).Versions for the new longer fuels used in the AP1000 and EPR (~4.8 m OL) are currently under development by several of the vendors).
b) Concrete canister: Consisting of an outer vertical concrete shield, with an inner steel liner. The steel liner is loaded in the fuel bay, and then transferred to the canister in a re- usable shielding cask. The canister is generally located outdoors, on a concrete pad and is not moved. The inner steel liner typically has a welded closure. The concrete shield has integral air channels for convective cooling. An example of this is the BNFL/Sierra VSC-24, which holds 24 PWR fuel assemblies, shown in Figure 2.3-5.
c) Concrete module: Consisting of an outer horizontal concrete shield vault, with an inner steel liner. The steel liner is loaded in the fuel bay, and then transferred to the canister in a re-usable shielding cask. The canister is generally located outdoors, on a concrete pad and is not moved. The inner steel liner typically has a welded closure. The concrete shield is typically pre-fabricated off-site for easy installation and has integral air channels for convective cooling. The modules are ganged together to improve the shielding efficiency. An example of this is the NUHOMS 32, which holds 32 PWR fuel assemblies per liner, one liner per shield vault, shown in Figure 2.3-6.
d) Modular Vault: Other vault storage systems are also used, such as the MVDS (modular vault dry storage system), which consists of a large shielded building with individual in- floor tubes for fuel storage. The tubes are surrounded by a passive convective cooling system. The system is capital intensive, and despite its name, is less modular than the three main systems described above.
Forecasted expansion dates for dry storage are summarized in Table 2.3-1. Further details on the forecasts can be found in Appendix A.
TABLE 2.3-1: EXAMPLES OF METALLIC PWR DRY STORAGE CASKS
CASK CAPACITY MAXIMUM AVERAGE FUEL GROSS DIMENSIONS (M) (# PWR FUEL HEAT BURNUP LIMIT MASS ASSEMBLIES) LOAD (GW D/ TONNE LOADED (KW) U) (TONNE) Castor V/21A 24 34 60 108 2.4 m OD x 4.9 m OL Castor X33F 33 16.6 60 96 2.4 m OD x 4.8 m OL NAC 128 S/T 28 17.4 35 94 2.4 m OD x 4.6 m OL Transnuclear TN24 24 24 35 95 2.3 m OD x 5.1 m OL Transnuclear TN32 32 32.7 40 105 2.5 m OD x 5.1 m OL Transnuclear TN40 40 27 45 103 2.5 m OD x 5.1 m OL
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FIGURE 2.3-5: CONCRETE CANISTER FOR PWR FUEL
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FIGURE 2.3-6: NUHOMS CONCRETE MODULE FOR PWR FUEL
2.3.3 Facility Examples
Examples of some existing facilities for the dry storage of PWR fuel are described below.
2.3.3.1 ZWILAG
The ZWILAG facility is located in Switzerland. It is a centralized facility for the processing and storage of wastes from the country’s nuclear power plants, including used fuel. The used fuel is stored in containers within a storage building, shown in Figure 2.3-7, measuring 68 m long x 41 m wide x 18 m high. When fully occupied, this hall can store around 200 containers standing on end. Containers are handled by an overhead crane. The containers are transported horizontally by rail or road and are rotated to the vertical as they are off-loaded for storage.
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FIGURE 2.3-7: ZWILAG INTERIM STORAGE FOR USED FUEL
2.3.3.2 Lingen
The Lingen facility is located in Emsland, Germany. It is a power plant facility for the storage of used fuel originating from that power plant (1300 MW PWR). The storage building hall, shown in 2.3-8, is designed for 130 containers. Containers are handled by an overhead crane. The containers are transported horizontally by rail or road and are rotated to the vertical as they are off-loaded for storage. The building includes forced ventilation and passive convective cooling.
2.3.3.3 Other Facilities
Used fuel dry storage facilities in the United States (US) are typically constructed in the form of an outdoor concrete storage pad. The pad is surrounded with a security fence, and may also be surrounded by an earthen berm for shielding and/or visual screen purposes. There are currently approximately 30 used fuel dry storage facilities in the US, with the oldest ones dating back to the 1980’s. Many of them store several different cask designs (e.g. purchased from different vendors, or upgraded to newer models of casks as designs have evolved). Typical installations and a transporter are shown below in Figures 2.3-9 through 2.3-12.
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FIGURE 2.3-8: LINGEN INTERIM STORAGE FOR USED FUEL
FIGURE 2.3-9: TYPICAL US CONCRETE CASK STORAGE
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FIGURE 2.3-10: TYPICAL US NUHOMS CONCRETE VAULT STORAGE
FIGURE 2.3-11: TYPICAL US METAL CASK STORAGE
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FIGURE 2.3-12: TYPICAL ON-SITE METAL CASK TRANSPORTER
2.3.4 Used Fuel Disposal
Long-term management and eventual disposal of used fuel in Canada is the responsibility of the Nuclear Waste Management Organization (NWMO). While their program is based primarily on CANDU fuels, they have the mandate to manage all used fuel in Canada. The disposal portion of the NWMO Adaptive Phased Management plan is similar to those developed in other countries PWR reactor fuel, such as Sweden (shown in Figure 2.3-13). The primary difference with the NWMO concept is in the detailed design of the disposal canister (to accommodate the physical dimensions of the PWR fuel) and the spacing of the canisters (to accommodate the higher expected heat load from the enriched fuels).
The other factor to consider for new-build reactors is the timing of the disposal operation. The new reactors will be in operation until 2100, long past the shutdown dates of the existing fleet. Therefore, long-term used fuel management must be available for an extended period of time.
Once a reactor design has been selected, the NWMO will need to modify its concept to include the fuel from the new-build reactors. Some preliminary studies have been done (Russell, 2008).
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FIGURE 2.3-13: SWEDISH USED FUEL DISPOSAL SYSTEM
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2.4 Refurbishment Wastes
The reference assumptions for management of refurbishment wastes from the current reactor fleet include on-site storage, followed by shipment off-site for disposal at a later date. For the purposes of this report, similar assumptions are made for refurbishment waste from NND.
All wastes are assumed to be prepared for storage by the reactor refurbishment project organization and delivered to the hand-over point at the waste management facility. For steam generators, this would include draining, drying, sealing up of all openings with welded plates, and applying a suitable corrosion protection to the sealed surfaces. Steam generators would be transported on multi-axle heavy load vehicles. Figures 2.4-1 and 2.4-2 show some conceptual layouts for the refurbishment waste.
In the longer term, the steam generators would likely need to be segmented prior to off-site shipment for disposal. This could be done at the time of decommissioning the existing Darlington reactors, when equipment to segment large objects from decommissioning is assumed to be on-site. This would occur in the 2050 time frame (assuming that the existing reactors are not refurbished) to 2080 timeframe (assuming the existing reactors are refurbished and their life extended). Alternatively, the steam generators would be stored until the new reactors are decommissioned. In both cases, it is assumed that a suitable repository is available at that time.
Figures 2.4-3 and 2.4-4 show examples of steam generator storage at the WWMF and at Palo Verde in the US (which has SGs comparable in size to the ones being considered for NND).
For fuel channel wastes, the preparation for storage includes the packaging of the wastes into “Retube Waste Containers” (RWCs). The filled and sealed RWCs would be transported to the storage building, and handled by heavy forklift, as shown in Figure 2.4-5.
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70 m
65 m
7 m
24 m
FIGURE 2.4-1: CONCEPTUAL ACR-1000 STEAM GENERATOR STORAGE
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70 m
45 m
(Heavy forklift maneuvering area)
(Door)
Note: For 4 Darlington B ACR units (520 fuel channels each), assuming Bruce A style RWCs: - 260 endfitting boxes, 1.7 m x 3.35 m, stacked 3 high, assuming 16 EFs per box - 140 PT/CT boxes, 1.85 m x 1.85 m, stacked 2 high, assuming PTs & CTs are volume reduced
FIGURE 2.4-2: CONCEPTUAL ACR-1000 FUEL CHANNEL WASTE STORAGE
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FIGURE 2.4-3: BRUCE A STEAM GENERATOR STORAGE AT WWMF
FIGURE 2.4-4: PALO VERDE STEAM GENERATOR STORAGE
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FIGURE 2.4-5: RWC STORAGE AT WWMF
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3.0 WASTE CHARACTERISTICS AND VOLUME FORECASTS
3.1 L&ILW Operational Waste Characteristics
The operational solid L&ILW from the new reactors is expected to have physical and radiological characteristics similar to the waste from OPG’s existing CANDU fleet. However, waste from the light water reactors is expected to have much less tritium and C-14 than the current CANDU reactors. A summary of the specific activity data is given in Table 3.1-1. The waste characteristics of the existing OPG CANDU wastes are also included for comparison purposes. Further details on the radiological characteristics can be found in Appendix A. Note that AECL has not reported values for tritium and C-14 in ACR-1000 wastes in their data [Candesco, 2008]. Based on existing CANDU experience, these can be expected to be higher than the PWR reactors.
TABLE 3.1-1: SUMMARY OF OPERATIONAL L&ILW CHARACTERISTICS
ACR-1000 AP1000 EPR OPG CANDU Average specific activity for 9.4 E+10 2.5 E+11 3.9 E+11 2.4 E+11 total L&ILW (Bq/m3) Estimated average specific 6.7 E+10 2.9E+10 4.5 E+10 1.7 E+11 activity for LLW (Bq/m3) Estimated average specific 3.9 E+11 5.1 E+12 7.9 E+12 1.0 E+12 activity for ILW (Bq/m3) Comments LLW / ILW split LLW / ILW split LLW / ILW split LLW / ILW split based on existing based on ratios to based on waste based on OPG waste OPG CANDU split EPR split stream specific data stream specific data Does not include supplied by Areva H-3 or C-14 References [Candesco, 2008] [Candesco, 2008] [Areva, 2007], [Rodrigues, 2008] [Candesco 2008]
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3.2 L&ILW Projected Volumes
Operational L&ILW is largely similar to wastes from the OPG current reactor fleet. LLW will consist of used protective equipment, paper, plastic, contaminated components, sludges, etc. ILW will consist mostly of IX resins and filters. The annual volumes and types for each reactor are summarized in Table 3.2-1, while Table 3.2-2 summarizes the expected lifetime arisings for a nominal 60 year reactor operating life. The characteristics and activity levels of the wastes are generally similar to existing OPG CANDU wastes. Further details can be found in Appendix A.
TABLE 3.2-1: SUMMARY OF L&ILW ANNUAL ARISINGS
AS GENERATED (M3/YR) PER REACTOR
WASTE TYPE ACR-1000 AP1000 EPR LLW Incinerable1 111 106 150 Compactable 30 28 40 Non-processible 14 7 2 Sludge 1 23 TOTAL 155 142 215 ILW Ion Exchange (IX) resins 7 11 7 Filters 3 0.2 3 TOTAL 9 11 10 GRAND TOTAL L&ILW 164 154 225
References [AECL, 2007], [Candesco, 2008], [Areva, 2007], [Candesco, 2008] [Westinghouse, 2007] [Candesco, 2008]
For the AP1000 and EPR, a “new” waste stream to Canadian power reactor operations is related to the boric acid system used for reactivity control in light water reactors. The primary coolant, used fuel wet bay and other liquid streams will contain boric acid. These liquid streams are mostly processed and recycled in-plant. Liquid boric acid wastes are eventually concentrated by evaporation (EPR), resulting in a sludge or concentrate that is solidified, or by ion-exchange (AP1000), resulting in borated spent resins. The AP1000 also has the flexibility to use temporary, mobile liquid waste processing, such as reverse osmosis equipment. This will also result in the production of concentrate and sludge.
Because boric acid is highly corrosive to carbon steel, special precautions will be taken to ensure that the wastes are neutralized and packaged in suitable containers.
1 The vendors of the ACR-1000 and AP1000 did not separate out “incinerable” wastes as a separate category. These had been included in the “compactible” category for these reactors. If incineration is available (e.g. at the WWMF), then a large fraction of the “compactible” wastes may in fact be incinerated. For the purposes of this report, it has been assumed that the fraction of incinerable waste for ACR-1000 and AP1000 is the same as for the EPR.
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TABLE 3.2-2: SUMMARY OF L&ILW LIFETIME ARISINGS
AS GENERATED (M3) PER REACTOR
WASTE TYPE ACR-1000 AP1000 EPR LLW Incinerable2 6669 6368 8,999 Compactable 1779 1698 2,403 Non-processible 864 406 122 Sludge 63 1,352 TOTAL 9,312 8,534 12,876 ILW IX resins 393 679 391 Filters 153 9 204 TOTAL 546 688 594 GRAND TOTAL L&ILW 9,858 9,222 13,470
References [AECL, 2007], [Candesco, 2008], [Areva, 2007], [Candesco, [Candesco, 2008] [Westinghouse, 2007] 2008]
3.3 L&ILW Projected Storage Requirements
3.3.1 LLW Projected Storage Requirement
Table 3.3-1 summarizes the number of SBs required to store the lifetime arisings of LLW from the various reactor configurations. The limiting case for EA purposes is 2 SBs for on-site storage of LLW. This will store all of the LLW from any of the reactor configurations being considered. Processing and storage of operational LLW is required to be in operation by the time the first unit is “radioactive” (~2017), and must continue until the end of life of the last unit.
For off-site processing and storage, one SB will be required. However, it is likely that no additional SBs will need to be constructed at the WWMF, since the bulk of the wastes will be generated after 2018 when the L&ILW Deep Geologic Repository (DGR) is assumed to be in operation. As the existing SBs are emptied and the contents transferred to the DGR, the freed up space can be used to store the wastes from new-build reactors if this waste is not destined for the DGR.
2 The vendors of the ACR-1000 and AP1000 did not separate out “incinerable” wastes as a separate category. These had been included in the “compactible” category for these reactors. If incineration is available (e.g. at the WWMF), then a large fraction of the “compactible” wastes may in fact be incinerated. For the purposes of this report, it has been assumed that the fraction of incinerable waste for ACR-1000 and AP1000 is the same as for the EPR.
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TABLE 3.3-1: SUMMARY OF STORAGE REQUIREMENTS FOR OPERATIONAL LLW
ACR-1000 AP1000 EPR Maximum number of reactor units 4 4 3 On-site: Lifetime LLW on-site stored with compaction only 3,192 2,586 4,354 (m3 per reactor) # of SBs required per unit (7,000 m3 each) < 0.5 < 0.5 < 1 # of SBs required for maximum number of units < 2 < 2 < 2 Off-Site: Lifetime LLW stored off-site 1,691 1,154 2,329 with incineration + compaction (m3 per reactor) # of SBs required per unit (7,000 m3 each) < 0.25 < 0.25 < 0.5 # of SBs required for maximum number of units < 1 < 1 < 1
3.3.2 ILW Projected Storage Requirements
The required ILW storage space is summarized in Table 3.3-2.
TABLE 3.3-2: SUMMARY OF STORAGE REQUIREMENTS FOR OPERATIONAL ILW
ACR-1000 AP1000 EPR Maximum number of reactor units 4 4 3 On-site: Lifetime Total ILW stored (m3 per reactor) 546 688 594 Lifetime IX resins stored (m3 per reactor) 393 679 391 Lifetime filters & misc ILW stored (m3 per reactor) 153 9 204 # of SBs required for on-site storage per unit < 0.1 < 0.1 < 0.1 (7,000 m3 each) # of SBs required for on-site storage for maximum number of < 0.5 < 0.5 < 0.5 units Off-site: # of ICs required for off-site storage per unit 31 39 33 (18 m3 each) # of ICs required for off-site storage for maximum number of 122 153 99 units
The limiting case for EA purposes is one SB for on-site storage of ILW. For off-site storage, up to approximately 150 in-ground containers (ICs) would be required. However, it is likely that no additional ICs will need to be constructed at the WWMF, since the bulk of the wastes will be generated after 2018 when the L&ILW DGR is assumed to be in operation. As the existing ICs
3-4 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document are emptied and the contents transferred to the DGR, the freed up space can be used to store the wastes from new-build reactors if this waste is not destined for the DGR.
3.3.3 Overall Storage Building Requirements for L&ILW
The total number of on-site L&ILW storage buildings required and the projected need date are summarized in Table 3.3-3. Further details on annual waste arisings can be found in Appendix A.
TABLE 3.3-3: ON-SITE L&ILW STORAGE BUILDING EXPANSIONS
ACR-1000 AP1000 EPR Maximum number of reactor units 4 4 3 Total lifetime stored volume for maximum number of 14,952 13,098 14,845 reactor units (m3) Total number of SBs required 3 2 3 Forecast In-service Dates: SB 1 2017 2017 2017 SB 2 2046 2049 2045 SB 3 2074 N/A 2074
3.4 Used Fuel
3.4.1 Used Fuel Characteristics and Volumes
The fuel from all of the new-build reactors will have higher enrichment and burnup than current CANDU fuels. This introduces elements of criticality control requirements for storage as well as potential heat load issues for dry storage and eventual disposal. The high burnup will also affect the source term of radionuclides in the used fuel. Typical source term values for PWR fuels are summarized in Appendix A, extracted from reference [US DOE, 2008].
The light water reactor fuel assemblies as shown in Figure 3.4.1 are physically much different from the traditional CANDU fuel bundle. They are much larger and heavier (~4.8 m long and ~800kg total weight) and the fuels also have integral control rods and burnable poison rods.
PWRs are operated in batch cycles of 18 months to 2 years, when the reactor is shut down and a portion of the core is removed and replaced with new fuel (typically 40% to 60%, depending on the length of the operating cycle, degree of enrichment, final burnup, etc). The fuel rods are handled vertically when placed into or removed from the core. They are also stored vertically in the used fuel bay.
Table 3.4-1 summarizes the fuel characteristics and expected volumes for the three reactor designs.
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FIGURE 3.4-1: EPR FUEL
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TABLE 3.4-1: SUMMARY OF FUEL CHARACTERISTICS AND ARISINGS
PARAMETER ACR-1000 AP1000 EPR Description 43 element CANFLEX-ACR bundle (0.5 Conventional PWR: Conventional PWR: m L x 0.1 m OD) (264 rods, 17x17 array (265 rods, 17x17 array, ~ 20 kg/bundle 4.8 m L x 0.21m sq 4.8 m L x 0.21m sq 12 bundles per channel, 520 channels ~538 kg U / assembly, ~535 kg U / assembly, (6,240 bundles per reactor) ~786 kg total wt) ~785 kg total wt) 157 assemblies in core 241 assemblies in core Enrichment Up to 2.5% for equilibrium core 2.4-4.5% avg initial core Up to 5% for equilibrium core 4.8% avg for reloads Burnup (MWday/tonne U) 20,000 60,000 62,000 Refuelling cycle On-power, ~ 12 bundles per day at Shutdown and change 64 assemblies every Shutdown and change 40% to 60% of fuel equilibrium 18 months at equilibrium in 18 to 24 month cycle at equilibrium Lifetime fuel arisings 5,246 1,400 2,712 (tonnes U) Lifetime fuel arisings ~262,300 bundles ~2,600 assemblies ~5,100 assemblies (bundles or fuel assemblies) Average annual fuel arisings 4,372 43 85 (bundles or fuel assemblies) Comments Decay heat estimated to be 11.8 watts per Fuel assembly also contains various Fuel assembly also contains various bundle at 10 years for 20,500 burnable poison elements and control burnable poison elements and control MWday/tonne U burnup [Pontikakis et al elements to compensate for fresh fuel elements to compensate for fresh fuel 2005] reactivity reactivity References [AECL, 2007], [Candesco, 2008] [Candesco, 2008], [IAEA, 2004], [Areva, 2007], [Candesco, 2008], [Westinghouse, 2007] [IAEA, 2004]
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3.4.2 ACR-1000 Bundle Storage System Options
The ACR-1000 fuel bundle (shown in Figure 3.4-2) looks similar to conventional CANDU fuel bundle in terms of dimensions and physical configuration. Like other CANDUs, the ACR-1000 is operated with on-power refuelling, resulting in a regular stream of fuel bundles discharged and transferred to the wet used fuel bay. The default storage system in the wet bay for ACR-1000 fuel is 36 bundle “baskets”, as shown in Figure 3.4-3. These in term are loaded into stacking frames in the wet fuel bay as shown in Figure 3.4-4. In the AECL MACSTOR dry storage system, a 60 bundle basket as shown in Figure 3.4-5 is used for storage.
FIGURE 3.4-2: ACR-1000 FUEL
FIGURE 3.4-3: ACR-1000 FUEL STORAGE BASKET
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FIGURE 3.4-4: ACR-1000 FUEL BASKET STACKING FRAME
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FIGURE 3.4-5: 60-BUNDLE MACSTOR FUEL BASKET
The alternative storage system option would be to adapt OPG’s current used wet/dry system to the storage of used fuel that was initially enriched. The dry storage canister has been described earlier.
The MACSTOR system components are not compatible with OPG’s DSC system. Therefore, the fuel would need to be re-packaged into standard modules prior to storage in DSCs if this was adopted as the preferred dry storage option.
3.4.3 Used Fuel Projected Storage Requirements
One planning assumption is that only 50% of the used fuel requires an interim dry storage facility on site. Table 3.4-4 summarizes the amount of used fuel produced and the number of individual storage containers/cask required for the different reactor designs. Table 3.4-5 summarizes the number of future storage buildings and their required dates for the various reactor designs.
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TABLE 3.4-4: SUMMARY OF DRY STORAGE SPACE REQUIREMENTS
PER REACTOR ACR-1000 AP1000 EPR Lifetime fuel generated (bundles or fuel assemblies) 262,300 2,602 5,069 Fuel dry stored (bundles or fuel assemblies) 131,150 1,301 2,535 # of DSCs required (384 bundles per DSC) 342 N/A N/A # of AECL fuel baskets (36 bundles per basket) 3,644 N/A N/A # of MACSTOR/KN-400 modules 10 N/A N/A (400 baskets/module) # of 24 fuel assembly PWR casks N/A 55 106 # of 32 fuel assembly PWR casks N/A 41 80 # of 40 fuel assembly PWR casks N/A 33 64
TABLE 3.4-5: ON-SITE USED FUEL STORAGE BUILDING EXPANSIONS
ACR-1000 AP1000 EPR Maximum number of reactor units 4 4 3 Fraction of total fuel dry stored 50% 50% 50% Total used fuel dry stored for maximum number of 524,600 5,204 7,604 reactor units (bundles or fuel assemblies) Option A – Cask/DSC in Building: Total number of dry storage casks/DSCs required3 1,366 163 238 Total number of dry storage buildings required4 3 1 1 Forecast In-service Dates: SB1 2028 2028 2028 SB2 2043 N/A N/A SB3 2054 N/A N/A Option B – MACSTOR / NUHOMS modular vault: Total number of dry storage modules required 40 163 238 Forecast In-service Dates: Stage 15 2028 2028 2028 Stage 23 2038 2038 2036 Stage 33 2046 2047 2041 Stage 43 2054 2056 2047 Stage 53 N/A N/A 2053
3 Each DSC contains 384 CANDU fuel bundles. Each PWR cask contains 32 fuel assemblies. 4 Each dry storage building holds 500 DSCs or 300 PWR casks. 5 Each MACSTOR stage consists of 10 modules, each module with 400 baskets. Each NUHOMS stage consists of 48 vaults, each holding 1 canister with 32 fuel assemblies.
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3.5 Refurbishment Wastes
For the purposes of this study, it is assumed that mid-life refurbishment of the reactors will be required. Plant specific radiological source term information has not been generated for refurbishment waste yet.
For the ACR-1000, the principal refurbishment wastes would include fuel channels, calandria tubes, feeders, and/or steam generators, depending on the required scope of the refurbishment. The fuel channel materials (zirconium alloys and stainless steel) are similar to those used in existing CANDU reactors. Therefore, the expected activation products will be similar, although the absolute amounts of given nuclides may differ due to the design specific neutron flux spectrum and location of the various materials in the core. The pressure tubes, calandria tubes, calandria tube inserts, end fittings and shield plugs would likely be classified as ILW, while the feeder pipes would be classified as LLW. A typical fuel channel source term based on Darlington A is summarized in Appendix B.
Steam generator contamination occurs through the deposition of materials, such as oxides, from the primary coolant, mainly on the inner surfaces of the steam generator tubes (primary side). Both activated corrosion products and fission products may be present. Leakages in the tubes may result in contamination on the secondary side. All of the reactor types under consideration use light water as the primary coolant with similar primary chemistry. Therefore, the type of radioactivity deposited is expected to be similar for all of the reactor types. The absolute amount of the fission product and transuranic contamination will depend largely on the amount of these materials released from the fuel during operation (e.g. via pinhole leaks in the fuel cladding) and on the efficiency of the reactor coolant cleanup system.
Typical expected steam generator contamination levels for Pickering B, Darlington A, and Ringhals 3 (a Swedish PWR) [Vattenfall 2007] are summarized in Appendix B. The total radioactivity in a steam generator appears to be relatively consistent for the two reactor types, with both being in the range of 1 to 3 TBq per steam generator, or about 2 to 5 GBq/tonne. The steam generators are expected to be classified as LLW.
For the AP1000 and EPR, refurbishment would consist of steam generator and reactor vessel head replacement. Note that these are conservative assumptions for EA planning purposes, and may not in fact be required. The outage duration for a reactor unit is conservatively assumed to be two years for the ACR-1000 and one year for the AP1000 and EPR designs. (In practice, PWR steam generator replacement outages have typically been as short as several months).
PWR reactor vessel heads are expected to be activated stainless and carbon steels, with some small amounts of fission product and other surface contamination on the inside portions of the head. The heads are likely to be classified as ILW due to high concentrations of long-lived nickel isotopes and high dose rates.
Tables 3.5-1 summarizes the sizes and required storage space to refurbishment waste.
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TABLE 3.5-1: ON-SITE REFURBISHMENT WASTE STORAGE
ACR-1000 AP1000 EPR Refurbishment major LLW 4 SGs per unit 2 SGs per unit 4 SGs per unit (16 total) (8 total) (12 total) 5.5 m OD x 24.2 m L 5.6 m OD x 22.5 m L 5.2 m OD x 24.3 m L 472 tonne each 665 tonne each 550 tonne each Assumed packaging Sealed SG shell Sealed SG shell Sealed SG shell Storage method Intact in SG storage Intact in SG storage Intact in SG storage building building building ~65 m x 70 m footprint ~32 m x 70 m footprint ~65 m x 55 m footprint In-service date 2048 2048 2048 Refurbishment major ILW 520 fuel channels per 1 reactor pressure vessel 1 reactor pressure vessel unit (2080 total) head (4 total) head (3 total) 4.8 m OD x 2.0m H 5.75m OP x 3.2 H 116 tonne Assumed packaging Bruce style retube waste Sealed shell Sealed shell containers: 16 end fittings per RWC-EF container; 28 pressure tubes per RWC-PT; 40 calandria tubes per PWC-PT; 3 RWC-PTs per unit for misc components. Total 4 units: 260 RWC-EF 140 RWC-PT Storage method RWC storage building Included in SG storage Included in SG storage ~45 m x 70 m footprint building building In-service date 2048 2048 2048 References [AECL, 2007], [Candesco, 2008], [Areva, 2007], [Candesco, 2008] [IAEA, 2004], [Candesco, 2008], [Westinghouse, 2007] [IAEA, 2004]
As summarized in Table 3.5-1, the steam generators for all of the new reactor types are larger and heavier than those used in OPG’s existing reactor fleet. (By way of comparison, the Bruce A steam generators are 2.6 m max OD x 10.4 m OL with a weight of ~110 tonnes, while the Pickering B SGs are 2.5 m max OD x 14.3 m OL with a weight of 87 tonnes. Darlington SGs, which are the largest in the OPG fleet, are 4.7 m max OD x 22.2 m OL and 340 tonnes.) If they are to be replaced, the large size and weight would introduce complexities related to the handling and movement of the components, as well as the preparation of the components for eventual disposal. However, steam generators of this size have been successfully replaced in a number of plants in the US and Europe.
Figures 3.5-1 through 3.5-6 depict the major refurbishment components for each reactor type.
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For both LLW and ILW, the limiting case for refurbishment waste storage is the 4 unit ACR- 1000 station. Conceptual storage building layouts were shown earlier in Figures 2.4-1 and 2 4-2.
The additional operational LLW is projected at 500 m3 per reactor per year of refurbishment outage.
3.6 Summary of Waste Management Needs
Suitable land for the expansion of the DWMF will need to be reserved on the Darlington site. For long term DN site planning involving NND, this should include space for:
a) three low level storage buildings (30 m x 50 m each – 4,500 m2 total) b) three used fuel dry storage buildings (70 m x 76 m each – 16,000 m2 total) c) one used fuel dry storage processing building (40 m x 50 m – 2,000 m2) d) one steam generator storage building (65 m x 70 m – 4,550 m2) e) one retube waste storage building (45 m x 70 m – 3,150 m2)
Some of the future buildings might be located separately on-site from the current DWMF.
The waste management area(s) should include a minimum 5 m buffer between the storage buildings and the waste management facility fence. Additional security perimeter clearances are also required for the used fuel dry storage buildings. While there is no reason to believe that a Safety Assessment could not demonstrate that a location north of the CN rail line is feasible, the safety assessment used in this TSD assume that any waste processing or storage building are built south of the CN rail line and no closer than 150 m to the site perimeter fence. However, for EA planning purposes, this TSD has accepted this analysis to demonstrate that the UFDS can be located anywhere on the site. Should the Vendor require the UFDS buildings to be located north of the CN rail line, or any waste processing or storage building to be located closer than 150 m to the site perimeter fence, OPG has committed to updating safety assessment for this location as part of the licensing process.
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FIGURE 3.5-1: ACR-1000 FUEL CHANNEL
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FIGURE 3.5-2: ACR-1000 STEAM GENERATOR
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FIGURE 3.5-3: AP1000 REACTOR VESSEL6
6 The reactor vessel head is the portion above the “upper support plate”.
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FIGURE 3.5-4: AP1000 STEAM GENERATOR
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FIGURE 3.5-5: EPR REACTOR VESSEL7
7 The reactor vessel head is the portion above the “o-ring seal”.
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FIGURE 3.5-6: EPR STEAM GENERATOR
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4.0 OTHER WASTE MANAGEMENT CONSIDERATIONS
4.1 Processing
4.1.1 Operational L&ILW
Operational solid L&ILW is very similar physically and radiologically to wastes currently handled by OPG stations. Therefore, no issues are expected with the processing of these wastes.
The total storage / disposal volume of wastes will depend on the chosen processing and packaging method. For example, incineration will offer a higher volume reduction (generally > 40:1) compared to simple compaction (4:1). Advanced compaction techniques, such as supercompaction, can process a wider range of wastes than simple compactors and can offer high volume reduction (up to 10:1), resulting in lower overall stored volume.
There are two processing scenarios being considered for the environmental assessment that bound the processing options:
a) On-site processing (consisting of compaction) and storage.
b) Transport of unprocessed waste to an off-site facility (such as WWMF) for incineration, compaction and storage
On-site processing by compaction will reduce the number of eventual off-site transportation trips because some of the waste has been compacted. But the volume of stored waste at NND will be greater than if the waste was further processed off-site by better volume reduction techniques such as incineration for low level waste.
4.1.2 Used Fuel
Processing of used fuel refers to the preparation for dry storage. In all cases, this typically involves drying of the fuel, sealing of the dry storage container (either by welding or bolting), backfilling with inert gas, decontamination of the container and transferring it from the fuel bay or processing area to the storage area. These tasks are all well developed, both within OPG (for CANDU type fuels) and internationally for PWR type fuels.
The ACR-1000 fuel bundles are stored in 36 bundle baskets, which are optimized for the AECL MACSTOR dry storage system. If OPG decides to use a DSC-type of dry storage, then a mechanism must be fabricated to transfer the fuel from these baskets to an OPG style fuel module.
Except for matching design details of the plant to the dry storage system (e.g. physical space available in the bays for cask handling, crane capacities, etc), no major issues are expected for the processing of used fuels from the PWR reactors.
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4.1.3 Refurbishment Wastes
Processing of refurbishment wastes refers to the decontamination and/or size reduction of replaced steam generators and other large components.
All three reactor types have large and heavy steam generators (> 450 tonnes). These will likely require segmentation prior to disposal. Although there is no Operating Experience (OPEX) within OPG for handling steam generators of this size, there is extensive international experience in Europe and the US, both in the handling, decontamination and segmentation of SGs in this size range.
Aggressive decontamination techniques can be used, such as grit blasting, since the equipment will not be re-used. Secondary wastes are generally limited to spent grit (which is recycled in the process to a large degree), which can be packaged in conventional containers, such as drums.
Apart from the need to design for handling of the large size and weight of the objects, no technical issues are expected with the processing of this waste.
4.2 Operational Points
4.2.1 Station Storage Provisions
The AP1000 and EPR provide a small amount of buffer storage for packaged operational L&ILW prior to off-site shipment. The basic ACR-1000 design includes a storage building for all operational L&ILW.
All three reactor types provide storage for wet operational wastes, such as IX resins. However, periodic emptying of the storage tanks and transferring to external storage is still required.
4.2.2 Used Fuel Management
All three reactor designs only provide 10 to 15 years of wet storage for used fuel. They all assume that older fuel will be transferred to dry storage as the bay fills up.
The timing of dry storage operation depends on the fuelling cycle. The ACR-1000 has continuous on-power refuelling at a rate of about 4,400 to 4,500 bundles per year. This rate requires the equivalent of about 12 DSCs per year to be removed from the fuel bay to maintain steady state. If the AECL MACSTOR system is used, then approximately 120 fuel baskets need to be removed from the bay per year.
The AP1000 and EPR have batch refuelling campaigns every 18 to 24 months. A typical refuelling outage would replace 60 to 100 fuel assemblies (18 month cycle). A typical PWR dry storage cask will hold 24 to 40 fuel assemblies, which results in 1.5 to 4 casks being filled every 18 months. This is a far lower rate than the ACR-1000, and would more than likely be done in batch campaigns every few years.
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For the purposes of this report, it is assumed that 50% of the lifetime used fuel will require on- site dry storage. After this, it is assumed that the NWMO will take ownership of the fuel under its Adaptive Phased Management (APM) program.
Note that the higher enrichment and burnup of the fuels in the new reactor designs will have an impact on the minimum cooling period prior to transfer to dry storage (i.e. the higher the burnup, the higher the initial decay heat and the longer the required cooling period before dry storage). The practical cooling time will be limited by the available storage space in the wet bays. This will drive the design requirements for the storage casks.
PWR dry storage casks are typically designed for much higher heat loads than OPG’s DSCs (up to about 40 kW, compared to 2.2 kW for a DSC), allowing shorter cooling times [IAEA, 2007].
4.2.3 Refurbishment Wastes
Storage of refurbishment wastes is required for fuel channel components (ACR-1000), reactor vessel heads (AP1000 and EPR) and steam generators (all reactor types). Dedicated storage facilities have been constructed in a number of countries (e.g. US, Germany, Japan, Slovenia, etc) for full sized steam generators.
For the purposes of this report, it is assumed that refurbishment wastes are stored on-site in a dedicated storage facility (i.e. separate from any operational waste storage) until the stations are decommissioned, at which point they will be transported off-site to a suitably licensed repository.
For the fuel channel components and reactor vessel heads, the primary hazard is related to high dose rates. Therefore, shielded storage will be required for these wastes, either in the form of shielded containers, a shielded storage structure or some combination thereof. Storage for fuel channel wastes has been constructed at WWMF, based on a “shielded container in a storage building” concept, and at Pt Lepreau, based on an “unshielded container in a shielded vault” concept. Currently, OPG has adopted the “shielded container in a storage building” as the reference concept for future reactor refurbishments for NEW As Low as Reasonably Achievable (ALARA) purposes and repository readiness.
4.3 Long-Term Waste Management
4.3.1 Operational L&ILW
For the purpose of the NND EA, the long-term management facility is not specifically defined, other than it must be a suitably licensed facility. The potential options would include:
a) Revising the Hosting Agreement and allowing the proposed L&ILW DGR to fill up to its current design capacity, then conducting a further EA in the future for an expansion, if required. For example, if not all of the existing reactors are refurbished and life- extended, then the wastes from the existing fleet will be less than the design capacity of the DGR allowing room for some additional wastes from new-build.
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b) Maintaining the wastes in interim storage (either at the DMWF or the WWMF) until such time as a decommissioning waste repository or other facility is available. The decommissioning waste repository has not yet been designed, so its size can easily be adjusted to accommodate extra wastes from NND.
4.3.2 Used Fuel
The responsibility for long-term used fuel management lies with the Nuclear Waste Management Organization (NWMO). The NWMO was established in 2002 by OPG, New Brunswick Power and Hydro Quebec and has the legal obligation to provide long-term used fuel management services to all used fuel owners in Canada. Currently, OPG holds some 90% of the used fuel inventory in Canada.
Its “Adaptive Phased Management” approach [NWMO, 2005] has been approved by the federal government and is now entering the siting phase. For planning purposes, the assumed in-service date of a long-term management facility is 2035. While the focus of the NWMO program has been on conventional CANDU fuels, it is recognized that there are other fuels that need to be managed, such as from research reactors as well as those from various experimental programs.
The ACR-1000 fuel is similar in configuration to conventional CANDU fuels, so should be easily accommodated, although some adjustments may be required to parameters such as disposal canister spacing in order to compensate for the higher enrichment and the higher burnup of the ACR-1000 fuel.
The PWR fuels from the AP1000 and EPR are physically much different from the CANDU fuels. However, there is international OPEX from countries such as Finland, Sweden and others, which can be adapted to the Canadian context. Except for the exact configuration of the fuel canisters, the Finnish and Swedish repository concepts for the long-term management of used fuel are very similar to the Canadian one.
The main technical points that need to be addressed for fuels from NND reactors are:
a) Effect of different physical configuration (e.g. longer, heavier fuels from PWRs) b) Effect of higher burnup (e.g. heat load and required storage/cooling times prior to emplacement) c) Effect of higher initial enrichment (e.g. criticality issues) d) Capacity of the repository to handle the additional fuel
A preliminary assessment of these points has been done (Russell, 2008).
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4.3.3 Refurbishment Wastes
Long-term management of refurbishment wastes faces many of the same issues as for operational wastes. The most likely destination for these wastes would be the decommissioning waste repository.
An additional consideration is the size and weight of the steam generators and other large components. These would require substantial size reduction to be handled by the proposed DGR (e.g. 35 tonne hoist limit would require each SG to be cut into about 15 pieces). A future decommissioning repository could be designed to handle much larger and/or heavier packages since these may be commonplace from decommissioning.
4.4 Off-Site Transportation and Accidents
This section discusses OPG’s Radioactive Material Transportation (RMT) program, the types of the transportation packages used, and OPG’s Transportation Emergency Response Plan. Transportation accident is described in the Malfunctions, Accidents and Malevolent Acts TSD.
The transportation of radioactive material is regulated by the Canadian Nuclear Safety Commission (CNSC) under the Nuclear Safety and Control Act (NSCA) and the Packaging and Transport of Nuclear Substance Regulations (PTNSR). In the regulations, there is a graded approach to the packaging requirements that corresponds to the hazard level of the radioactive material to be transported. The most hazardous materials must be shipped in the most robust packages. The packages are designed to withstand tests representing different severity levels of transport conditions. The severest level includes tests for accidents such as collisions and fire. The PTNSR also specifies the requirements for transport of nuclear substances, including the production, use, inspection, maintenance and repair of packaging and the preparation, consigning, handling, loading, carriage, storage during transport, receipt at final destination and unloading of packages.
The transportation of radioactive material is also regulated under the Transportation of Dangerous Goods Act, Class 7 Radioactive, and must also comply with the Highway Traffic Act(s) of the provinces through which the material travels. In this highly regulated environment, a robust program for procurement, maintenance, documentation, staff training and oversight has been developed.
4.4.1 Overview of Current Radioactive Material Transportation Program
The Nuclear Waste Management Division (NWMD) of OPG has the overall accountability for the transportation of radioactive material. It operates a Radioactive Material Transportation (RMT) program that provides a fleet of tractors, trailers and specialized packaging, a maintenance facility and support staff. RMT also provides quality-assurance oversight and verification for higher-risk (Type A and Type B) radioactive shipments originating from OPG. This existing program will be expanded as required to meet the needs of NND.
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OPG has an excellent radioactive materials transportation safety record. In an average year for the overall OPG RMT program, over 900 shipments of radioactive materials are consigned, and/or carried by OPG, traveling approximately 500,000 kilometers. Materials shipped include contaminated tools and equipment, low and intermediate level radioactive waste, solid and liquid samples, used fuel, and tritiated heavy water which is currently transported off-site from Pickering and Bruce Power NGS for processing to remove tritium.
All radioactive material shipments are logged into a computerized database that records information about the type of material being transported, point of origin, destination, shipper, and carrier. RMT also transports non-radioactive material, such as work clothing to and from the Bruce Power laundry facility.
In the more than 35 years that Ontario Power Generation has been transporting radioactive materials, and more than 11.5 million kilometres travelled, only five (5) shipments have been involved in traffic accidents. Three (3) accidents involved trucks transporting low level waste and two (2) involved the transportation of heavy water. There were no releases to the environment as a result of these accidents.
OPG’s radioactive transportation program is supported by:
(a) Packaging designed, fabricated, and tested in accordance with applicable regulations and standards.
(b) Regular audits and reviews of transportation procedures.
(c) An on-going Transportation of Dangerous Goods Class 7 (radioactive materials) training program.
(d) Rigourous transportation package inspection and maintenance; long service life packages are also subject to an aging management program.
(e) Oversight of high-hazard and non-routine shipments.
(f) Procurement and engineering support for tractors and trailers.
(g) A Transportation Emergency Response Plan that is audited both internally and externally by authorities like Transport Canada.
4.4.2 Transportation of L&IL Radioactive Waste to an Off-site Licensed Facility
Future transportation of L&IL radioactive materials for NND to an off-site licensed facility will be conducted under the RMT program as outlined in Section 4.4.1. The timing of shipments will depend on the final decision on whether the L&IL waste will be stored on-site versus off-site, the waste forms, and the availability of an alternate off-site licensed facility for interim storage, long-term storage or disposal.
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If additional transportation packages are required for the transport of specific radioactive wastes from NND, these packages would be designed, certified as necessary, and procured according to OPG's existing processes.
4.4.3 Transportation Package Requirements for Transportation of L&IL Waste
The PTNSR provide comparable levels of safety for radioactive materials of different radiotoxicities and different quantities by relating the nature and amount of the contents with graded packaging integrity and performance requirements. The relevant categories of packages established in the regulations are presented below in increasing order of physical integrity and performance:
x Excepted packaging for contents such as empty containers with traces of radioactivity from previous usage; x Industrial Packaging (Type IP-1, Type IP-2, Type IP-3) for LLW; x Type A for contents such as contaminated inspection equipment; and x Type B for ILW such as filters and ion exchange resins, and non-waste radioactive materials such as tritiated heavy water.
The graded approach to the packaging requirements in the Regulations addresses three general severity levels for transport conditions:
(a) routine conditions of transport (incident free); (b) normal conditions of transport (minor mishaps such as rough handling); (c) accident conditions of transport (including collisions and fire).
Excepted packages are packages in which the allowed radioactive content is restricted to such low levels that the potential hazards are insignificant and therefore no testing is required to demonstrate containment or shielding integrity for routine conditions of transport.
The Industrial package types are required for transport of surface contaminated objects (where relatively low quantities of radioactive material is distributed over the surface of a non- radioactive entity) or low specific activity materials (where the radioactive material is distributed at relatively low concentration within a non-radioactive material). The three Industrial package types have different safety functions. Type IP-1 packages simply contain their radioactive contents under routine conditions of transport, Type IP-2 and IP-3 packages protect against loss or dispersal of their contents, and loss of shielding under normal conditions of transport. There are tests consisting of a free drop test and stacking test for demonstrating the ability of these packages to withstand the normal conditions of transport. The free drop test simulates the type of shock that a package would experience if it were in a vehicle that braked suddenly, or if it was lifted or lowered abruptly during handling. In most cases packages would still be fit for transport after experiencing such shocks. Since heavier packages are less likely to be exposed to large drop heights during normal handling, the free drop distance for this test is graded (between 0.3 and 1.2 metres) according to the package mass. The stacking test is designed to simulate the
4-7 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document effect of loads pressing on a package over a prolonged period of time to ensure that the effectiveness of the shielding and containment systems will not be impaired.
A Type IP-3 package must also have a demonstrated ability to withstand a penetration test (a 6 kg steel bar dropped onto the package from 1 metre). This test is intended to ensure that the contents will not escape from the containment system or that the shielding would not be damaged if a slender object such as a length of metal tubing should strike and penetrate the outer layers of the packaging.
A Type A package is required when the quantity of radioactive material exceeds the limits for the Industrial package types. Type A packages also protect against loss or dispersal of their contents, and loss of shielding under normal conditions of transport. Type A packages for liquids and gases have additional requirements and must withstand more severe tests that consist of:
1. A 9 metre free drop test onto an unyielding surface. This test represents a major accidental impact or collision. 2. A penetration test similar to the one for a Type IP-3 package but with an increased drop height (1.7 m).
The purpose of these tests is to ensure the package has stronger integrity to counteract the greater ability of the contents to escape from a damaged package.
The design requirement for a Type B package is that it is capable of withstanding severe accident conditions in transport without a loss of containment, or an increase in external radiation level to an extent which would endanger the general public or those involved in rescue or cleanup operations. It should be safely recoverable after an accident but it would not necessarily be capable of being reused. The tests for demonstrating the ability to withstand accident conditions of transport include:
1. A 9 metre free drop test onto an unyielding surface. This test represents a major accidental impact or collision. 2. A 1 metre free drop onto a steel penetrator bar. This test represents a collision with a pointed object. 3. A thermal test of 800°C for 30 minutes. This test represents a fully engulfing fire occurring after an accident that ignited liquid, solid or gaseous combustible materials in the vicinity of the package. 4. A water immersion test under 15 metres of water for at least eight hours. As a result of transport accidents near or on a river, lake or sea, a package could be subjected to an external pressure from submersion under water. To simulate the equivalent damage from this low probability event, the transportation regulations require that a package be able to withstand external pressures resulting from submersion at reasonable depths.
The design and intended operations of a Type B package must be reviewed, and a design approval certificate issued by the CNSC prior to first use of the package. For the other package types, CNSC design approval is not required but the consignor of any shipment using one of
4-8 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document these packages must have documented evidence readily available for inspection by the CNSC, indicating that the package design complies with the applicable regulatory requirements.
LLW will usually be transported in Type A or lesser package types, while ILW will transported in Type B packages. OPG currently uses and will continue to use packages that are fully compliant with the transportation regulations.
4.4.4 Shipments to an Off-site Licensed Facility
The bounding scenario for shipments of LLW is the EPR with all of the generated radioactive waste sent for off-site processing and storage. This scenario would result in a 12,876 m3 lifetime arising of LLW which would require approximately 640 truck shipments of 20 m3 each over a 60 year period, or about one per month. Thus, for the case of 3 EPRs, the lifetime volume of low- level radioactive waste is estimated at approximately 38,700 m3 which would result in approximately 1,935 truck shipments of 20 m3 each, or two to three truck shipments per month during the 60-year operating life of NND.
For ILW, the lifetime generation from the AP1000 is the bounding quantity of approximately 688 m3 per reactor. For the case of four AP1000 reactors, the lifetime volume generated would also result in two to three truck shipments per month during the operating period. Note that the peak shipping rates may be higher during outage campaigns, but the lifetime average shipping rate is still very low.
During the refurbishment year for a reactor, approximately two additional shipments per day would be required for the refurbishment waste. This transport of L&ILW would be done along routes similar to those currently used. There are two options available for the handling of large components generated from refurbishment activities that require storage as ILW. The first option is that the large objects can be transported intact as special shipments. While radioactive objects of this size have been transported in other jurisdictions around the world, they have usually been done by rail or by barge. Alternately these large objects could be segmented first to simplify transportation and to meet waste acceptance criteria at an off-site licensed facility.
4.4.5 Transportation Emergency Response Plan
OPG has the capability of responding to a transportation incident (including accidents) involving radioactive material through its Radioactive Material Transportation Emergency Response Plan (TERP). The TERP identifies OPG's responsibilities during a transportation incident involving an OPG shipment of radioactive material, and identifies the liaison and potential interface with external emergency response organizations. This plan also includes requirements for personnel training, procedures and equipment, a mutual aid agreement (Mutual Initial Response Assistance Agreement) with other nuclear facilities and a service agreement with an external spills contractor.
Under the Transportation of Dangerous Goods (TDG) Regulations, the Shipper is required to have emergency response capability, and to file an emergency response plan with the Director General, Transport Canada, when transporting quantities which exceed a threshold value.
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Transport Canada assesses the acceptability of the identified response capability and confirms the feasibility of the outlined emergency response plan. The TERP program is tested annually using drills and exercises to practice emergency response capability, and to provide the means to test the effectiveness of different aspects of emergency response capability and identify areas for improvement.
4.4.6 Transportation Summary
In summary, the regulatory requirements on the design of transportation packages used to move L&IL waste between sites, OPG’s existing well-developed transportation program, the many years of experience in transporting radioactive materials, and the training required for personnel involved with transportation and the TERP are in place to prevent a release of radioactivity resulting from a transportation accident involving a shipment of low or intermediate level waste.
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5.0 NORMAL OPERATIONS
5.1 Used Fuel Operations – Radiological Impacts
5.1.1 Introduction
Normal operating releases or accidental releases from the fuel are due to volatile gases that escape through small defects in the fuel cladding. Any mechanical handling during storage and processing will introduce some mechanical shock to the system that may increase the percentage of defects in the fuel and lead to some releases.
The release scenarios (either normal operations or accidents) for used fuel storage are similar for the three reactor designs. After the minimum 10 years of wet storage, the source term of volatile radionuclides available for release is reduced to the amount of Kr-85 and tritium that can escape through a defect in the fuel cladding occurs. Available for release refers to the fractions of Kr-85 and tritium that have migrated to the gap between the fuel and the cladding, and to the grain boundaries within the fuel pellet. These gap fractions vary for the different fuel types.
5.1.2 ACR-1000 Normal Operation
As described earlier, there are differences in the amount of used fuel in the different used fuel storage options. In general, the DSC approach will always be more conservative than the MACSTOR approach simply because more used fuel is involved and the DSC will be used for this assessment.
Table 5.1-1gives the properties for 10 year cooled ACR-1000 fuel [AMEC, 2008].
TABLE 5.1-1: PROPERTIES OF 10 YEAR COOLED ACR-1000 FUEL
Key Parameter ACR-1000 Kr-85 per assembly (Bq) 2.12E+12 Kr-85 Gap Fraction 0.0617 H-3 per assembly (Bq) 1.18E+11 H-3 Gap Fraction 0.0617 Estimated Annual Inventories 4,372 (Bundles) Fuel Defect Rates 0.10%
Under normal operating conditions, minimal airborne emissions are expected from used fuel dry storage operations from the station fuel bay and at the Darlington Waste Management Facility (DWMF). This is because the uranium dioxide matrix, the used fuel sheath and the transfer clamp elastomeric seal (used in conjunction with OPG DSC) provide multiple barriers toward preventing the release of radioactive materials. The final vacuum drying step for a dry storage container will have some low levels of emissions.
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Airborne releases are also unlikely to arise under normal operating conditions during storage of seal-welded or bolted containers/casks. There is a small potential for airborne emissions resulting from processing operations such as sealing and vacuum drying.
Surface contamination on containers exterior surfaces will be controlled through prevention measures and decontamination at NND, as well as through routine surface contamination checks and monitoring at the DWMF as per current successful practice.
Radioactive contamination could be present on the outside of the fuel cladding. Although this contamination is expected to adhere to the fuel during storage, there is some potential for it to become airborne during vacuum drying of the container/cask cavity. A dedicated hose will be used for vacuum drying operations, to prevent the spread of such contamination to other workshop systems. Vacuum pump discharge will be directed to an active ventilation system, where particulate contamination is removed by filters. The stack filters will be monitored routinely for particulate contamination.
It is expected that a very small quantity of fuel elements may have minor defects in the cladding. Cladding defects are present in less than 0.1% of fuel bundles (representing < 0.01% of fuel elements) based on current 37-element natural uranium operating experience. Fuel bundles that are known to be defective will not be loaded into containers; releases from defective fuel have been conservatively assessed as described below.
For the purpose of evaluating the potential emissions under normal operating conditions, the following conservative assumptions are used to obtain an upper bound estimate for airborne emissions: a) one fuel element in 1% of fuel bundles is damaged during handling (4 elements per container), and for each failed fuel element, the free inventory of Kr-85 and tritium is released into the container cavity; b) the container seal is ignored and these radionuclides are released into the environment.
These assumptions are deemed conservative for the following reasons:
x The postulated defect rate is about three times higher on a per element basis than OPG fuel performance experience. x Fuel element defects occur primarily in the bundle manufacturing process or resulting from debris fretting in the reactor core. At high fuel temperatures during irradiation, the free inventory of Kr-85 and tritium in fuel elements with cladding defects would have been released within the reactor core. Upon cooling release rates drop due to mechanical factors and thermal factors. x Used fuel is stored for at least 10 years in wet storage prior to transfer to a container for dry storage. Leaching of grain-boundary inventory and release of gap inventory would have occurred over this period for bundles with minor cladding defects. x Should free inventory remain in the fuel-sheath gap or grain boundaries subsequent to wet storage, its release would have occurred during initial vacuum drying.
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x Dry storage system is designed and operated to ensure fuel integrity is maintained to the highest degree practical during storage.
Assuming the facility will process 12 DSCs per year containing 384 bundles, or 128 AECL fuel storage baskets per year containing 36 bundles, it is postulated that a total of 48 fuel elements (4 elements in each DSC, or approximately 1 element for every 3 canisters) fail during one year under normal operating conditions (a very conservative scenario).
The chronic off-site radiation dose consequence from this scenario, for a member of the public at the DN site boundary, is described in the Radiation and Radioactivity Effects Assessment TSD.
5.1.3 EPR Normal Operation
Table 5.1-2 gives the properties for 10 year cooled EPR fuel [AMEC, 2008].
TABLE 5.1-2: PROPERTIES OF 10 YEAR COOLED EPR FUEL
Key Parameter EPR Kr-85 per assembly (Bq) 1.69E+14 Kr-85 Gap Fraction 0.1 H-3 per assembly (Bq) 9.03E+12 H-3 Gap Fraction 0.05 Estimated Annual Inventories 85 (Assemblies) Fuel Defect Rates 0.25%
Under normal operating conditions, no airborne emissions are expected from a storage cask during transfer from the NND to the DWMF. This is because the uranium dioxide matrix, the used fuel sheath and the container shall provide multiple barriers toward preventing the release of radioactive materials. A number of the PWR cask design options are bolted prior to transport and thereby ensure an air tight seal during transfer from the station to the DWMF. Table 2.1-4 summarizes some current PWR dry storage casks.
Airborne releases are also unlikely to arise under normal operating conditions during storage of seal-welded containers or bolted containers. There is a small potential for airborne emissions resulting from container processing operations such as sealing and vacuum drying.
It is expected that a very small quantity of fuel elements may have minor defects in the cladding. A design basis defect rate of 0.25% should be applied for all fuel assemblies for EPR [AMEC, 2008]. Fuel assemblies known to be defective will not be loaded in containers; releases from defective fuel have been conservatively assessed as described below.
For the purpose of evaluating the potential emissions under normal operating conditions, the following conservative assumptions are used to obtain an upper bound estimate for airborne emissions for EPR:
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a) 0.25% of all fuel assemblies are damaged during handling and the free inventory of H-3 and Kr-85 is released into the container cavity; b) the container containment seal is ignored and these radionuclides are released into the environment.
These assumptions are deemed conservative for the following reasons:
x The postulated defect rate is likely to be over conservative. x Fuel element defects occur primarily in the bundle manufacturing process or in the reactor core. At high fuel temperatures during irradiation, the free inventory of H-3 and Kr-85 in elements with cladding defects would have been released within the reactor core. x Used fuel is stored for at least 10 years in the station irradiated fuel bays prior to transfer to a container. Leaching of grain-boundary inventory and release of gap inventory would have occurred over this period for bundles with minor cladding defects. x Should free inventory remain in the fuel-sheath gap or grain boundaries subsequent to in- bay storage, its release would have occurred during initial vacuum drying process (if applicable). x The containers are designed and operated to ensure fuel integrity is maintained to the highest degree practical during storage. x A maximum PWR fuel bundle radionuclide characteristics are used which is more conservative than the vendor specific information.
The chronic dose is dependent on the annual fuel throughput. The EPR is expected to have an annual throughput of 85 assemblies per year.
For the EPR, the maximum chronic off-site radiation dose consequence from this scenario, for a member of the public at the DN site boundary is described in the Radiation and Radioactivity Effects Assessment TSD.
5.1.4 AP1000 Normal Operation
Under normal operating conditions, no airborne emissions are expected from a storage cask during transfer from the NND to the DWMF. This is because the uranium dioxide matrix, the used fuel sheath and the container shall provide multiple barriers toward preventing the release of radioactive materials. A number of the PWR cask design options are bolted prior to transport and thereby ensure an air tight seal during transfer from the station to the DWMF. Table 2.3-1 summarizes some current PWR dry storage casks.
Airborne releases are also unlikely to arise under normal operating conditions during storage of seal-welded containers or bolted containers. There is a small potential for airborne emissions resulting from container processing operations such as sealing and vacuum drying.
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It is expected that a very small quantity of fuel elements may have minor defects in the cladding. A design basis defect rate of 0.25% should be applied for all fuel assemblies for AP1000 [AMEC, 2008]. Fuel assemblies known to be defective will not be loaded in containers; releases from defective fuel have been conservatively assessed as described below.
For the purpose of evaluating the potential emissions under normal operating conditions, the following conservative assumptions are used to obtain an upper bound estimate for airborne emissions for the AP1000:
x 0.25% of all fuel assemblies are damaged during handling and the free inventory of H-3 and Kr-85 is released into the container cavity; x the container containment seal is ignored and these radionuclides are released into the environment.
These assumptions are deemed conservative for the following reasons:
x The postulated defect rate is likely to be over conservative. x Fuel element defects occur primarily in the bundle manufacturing process or in the reactor core. At high fuel temperatures during irradiation, the free inventory of H-3 and Kr-85 in elements with cladding defects would have been released within the reactor core. x Used fuel is stored for at least 10 years in the station irradiated fuel bays prior to transfer to a container. Leaching of grain-boundary inventory and release of gap inventory would have occurred over this period for bundles with minor cladding defects. Should free inventory remain in the fuel-sheath gap or grain boundaries subsequent to in- bay storage, its release would have occurred during initial vacuum drying process (if applicable). x The containers are designed and operated to ensure fuel integrity is maintained to the highest degree practical during storage. x A maximum PWR fuel bundle radionuclide characteristics are used which is more conservative than the vendor specific information.
The chronic dose is dependent on the annual fuel throughput. The AP1000 is expected to have an annual throughput of 43 assemblies per year. The radiation dose methodology is described in more detail in [AMEC, 2008]
For the AP1000, the maximum chronic off-site dose consequence from this scenario, for a member of the public at the DN site boundary, is described in the Radiation and Radioactivity Effects Assessment TSD.
5.2 Routine Radiological Emissions from Radioactive Waste Processing
Radioactive waste processing might be done in a single building or several. Low level waste processing by a compactor would make up the majority of routine emissions from radioactive waste processing. Operational experience at OPG’s waste management facilities can be used to
5-5 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document estimate potential radiological emissions from a future waste processing involving a compactor for low level waste and steps involving final drying, helium back-filling and welding of a used fuel storage container or cask.
The exhaust from a compactor would go to the processing area ventilation exhaust where it would be filtered and monitored before exhaust. Using historical data from the WWMF’s Waste Volume Reduction Building ventilation exhaust combined with an estimate of future low level waste volumes, a conservative estimate can be prepared for general building emission plus compactor exhaust emissions. Tritium is a major component of routine emissions and hypothetical radiation doses in an accident. The ACR-1000 becomes a bounding case for all reactor types. Because projected airborne emissions from an ACR-1000 without a tritium removal facility are comparable or less than current airborne emissions on a unit basis, current low level waste characteristics are used.
Radiological emissions at each of the three used fuel storage facilities have traditionally been very low. Only particulate emissions are measured. Future emissions using enriched fuel and a slightly higher fuel defect rate is expected to be higher. Note that some losses of radioactive gases will occur in the fuel bay during loading and would not appear at the used fuel storage processing and storage buildings.
A conservative estimate for an annual radiological waste processing building from general building exhaust, compaction and used fuel container processing would be:
Tritium 1.2E+11 Bq Particulate 7.5E+05 Bq
These would represent about 1% of DN site annual airborne particulate emissions and 0.1% of airborne tritium emissions over the last three years [OPG, 2006, 2007, 2008].
Operating experience at WWMF has shown that LLW Storage Building sumps and building footing drainage may contain tritium but few other radionuclides. Sumps will have to be sampled and if required drainage taken to the station for treatment and monitoring before discharge. A sampling station to measure and take samples of any footing drainage would be a probable requirement if the ACR-1000 design was chosen.
5.3 Conventional Emissions
Conventional emissions from waste processing and storage on-site would be expected to be small due to the passive nature of the operations. Potential sources for non-radiological airborne emissions will include emissions from vehicles and material handling equipment , operation of a diesel generator (for equivalent) for use during loss of normal electrical power (safeguards and security systems associated with dry fuel storage require backup power), and use of welding equipment. A fire suppression system using carbon dioxide for Storage Buildings containing L&ILW is a possibility. If so, routine testing of the fire suppression system will release some carbon dioxide.
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5.4 Impact on the Environment
The impact of waste storage (both construction of storage facilities and normal waste operations) on the environment at NND can be inferred from experience at the larger WWMF. Over the period from 2000 to 2005, an extensive Environmental Assessment Follow-Up Program was undertaken to assess the effect of construction and operations of new buildings at the WWMF. The conclusion was that were no unreasonable adverse effects due to the construction and operations of the new storage structures [Nash, 2005; Klassen, 2006]. Similarly no unreasonable adverse impacts on the environment are anticipated for normal waste storage and processing at NND.
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6.0 RADIOLOGICAL IMPACT OF MALFUNCTIONS AND ACCIDENTS
6.1 Approach to Identifying Malfunctions and Accidents
The CEAA requires that every screening or comprehensive study of a project include consideration of the environmental effects of malfunctions or accidents that may occur in connection with the project. Furthermore CEAA also requires considerations of measures provided or intended to mitigate such effects. This section provides a summary description of radiological malfunctions and accidents identified with on-site storage of L&IL radioactive waste and used fuel. Non-radiological (or conventional) events are also possible but would be covered under and bounded by similar events for the project that are discussed further in the Malfunctions, Accidents and Malevolent Acts TSD.
The focus of the EA is on those events that are considered credible in the context of the proposed project. It is not the intent to address all conceivable abnormal occurrences, but rather to address those that may reasonably occur considering the specific aspects of site conditions and the project design. Within credible accidents, the intent is to focus on the bounding accident scenario.
The assessment also acknowledges that malfunctions and accidents (i.e. upset conditions) may be precipitated by external factors, either natural or anthropogenic. In the context of this assessment, external factors that lead to upset conditions are considered “initiating events”.
Initiating events can be external to normal nuclear waste operations. An extreme weather condition is an example of an external event. Initiating events represent either the failure of or damage to the systems and components of the radioactive waste on-site storage operations. A meaningful assessment requires a full consideration of the likelihood of initiating events (both due to nuclear waste operations related events and non-nuclear waste related but with nuclear) waste implication) as well as the consequence of such events.
The screening approach taken was to identify those events that may reasonably occur and then establish if they result in a radiological consequence that warrant further consideration. Where it was determined, on the basis of screening, that the event could result in a radiological consequence, that event was advanced for subsequent evaluation. The process is intended to identify both typical and the bounding (limiting) credible malfunctions and accident from the radiological perspective.
6.2 Assessment Methodology
The radioactive waste on-site storage philosophy embodies the defence-in-depth approach to keep radionuclide emissions within regulatory limits and at levels that are ALARA. The defence-in-depth approach is achieved by using multiple barriers between the nuclear waste and the environment.
The assessment of nuclear malfunctions and accidents is divided into general areas according to the waste type.
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Low and Intermediate Level Waste. Refurbishment Waste Used Fuel - ACR-1000 used fuel processing and storage using a modification of the current DSC design - AP1000 used fuel processing and storage using a shielded cask design - EPR used fuel processing and storage using a shielded cask design
For each of these waste types, release of radiation can occur due to the failure of the system and components. There are two general categories of initiating events resulting in abnormal conditions or accident, as follows:
Internal events, which are abnormal conditions generated within the radioactive waste on-site storage systems as a result of equipment failure or human error; and External events, which are natural and human-made phenomena originating outside the project that has the potential of leading to wide-spread, multiple internal events.
The internal and external initiating events are screened later in a series of tables (Tables 6.3-1, 6.4-1, 6.6-1, 6.7-2 and 6.8-2).
Nuclear criticality is also reviewed for the three reactor designs to ensure that technology to ensure safety is available and can be adapted.
Each event was screened to establish if it could result in any radiological impact to the public and a NEW. Events with a frequency of less than 10-7 events per year are considered “incredible” and are not considered further. Design provisions, procedural measures and worker training that could prevent or mitigate its consequences were also considered. After the screening of all initiating events, the events with the worse consequence to the public were chosen as the bounding event for that waste type and phase of operations. The hypothetical radiation doses to a member of the public and a NEW were then calculated for that waste type and phase of operation. While there is no reason to believe that a Safety Assessment could not demonstrate that a location north of the CN rail line is feasible, the safety assessment used in this TSD assume that any waste processing or storage building are built south of the CN rail line and no closer than 150 m to the site perimeter fence. However, for EA planning purposes, this TSD has accepted this analysis to demonstrate that the UFDS can be located anywhere on the site. Should the Vendor require the UFDS buildings to be located north of the CN rail line, or any waste processing or storage building to be located closer than 150 m to the site perimeter fence, OPG has committed to updating safety assessment for this location as part of the licensing process.
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6.3 L&ILW - Malfunctions and Accidents
6.3.1 Screening of Malfunctions and Accidents for L&ILW
Malfunctions and accidents with a radioactive release involving low and intermediate level waste basically fall into two categories. A handling accident involving some spill of the contents and a handling accident that also involves a fire situation are the categories. It is assumed that all waste will be in metal containers. Some new container designs would be required for above- ground storage of intermediate level waste such as filters. These new intermediate level waste containers would have to provide shielding and be robust enough to contain waste in a fire scenario.
In general an accident involving fire is much more significant as package releases in a respirable size would be higher resulting in higher radiation dose to the public and to the worker. Table 6.3-1 summarizes the accident scenarios for LLW. Two cases involving fire represent the bounding case for low and intermediate level waste respectively.
6.3.2 Bounding Case for Low Level Waste – Pool Fire Beside Stacked Waste Containers
Future L&ILW waste storage at NND is expected to be in storage buildings similar to what is being used today at WMMF. It is assumed that all waste will be in containers that are either similar or more robust that what are used today. Fire detection systems at WWMF are being upgraded and equivalent or better systems would be anticipated for NND waste storage structures. Because all waste will be in containers, the safety approach developed by the US Department of Energy for Transuranic (TRU) Waste Facilities [US DOE, 2007] was considered suitable for this application. This approach has been used in the US Waste Isolation Pilot Project for storage in an underground storage room [US DOE, 2006]. The bounding accident scenario is a “pool fire” near a stack of waste containers. The basic scenario is leakage from the gasoline or diesel fuel tank of a forklift or material handling vehicle catches fire. The conservative assumption is that the forklift is immediately adjacent to a stack of waste that extends from the floor to the ceiling of a low level storage building. For example the forklift might have just placed a container into the top row of waste containers inside the building. In US DOE methodology, a “pool fire” is considered to be the more severe case than a normal combustible fire because the more intense heat causes lid loss on the top row of containers and waste is physically ejected from the containers. Particulate release fraction that is respirable from ejected waste are higher than the remaining waste that burns within the container itself. Waste within 4 meters of the forklift is assumed to be affected by the pool fire. The heat from this fire is considered intense enough to give a plume rise. The hypothetical radiation dose to a member of the public from this fire was calculated to be 14 µSv which is between 1% to 2% of the regulatory limit for a member of the public.
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At the start of the fire, it is assumed that the NEW will immediately leave to get a fire extinguisher, return and empty the extinguisher at the fire. He/she will then leave the area. The time that a NEW would be exposed to the fire scenario is expected to be no more than three minutes. Either the NEW or an Emergency Response Team member will then return later but wearing full respiratory protection and some plastic protection against tritium uptake through the skin. Nearly all the radiation dose to the NEW will come from the inhalation dose component in those first three minutes. Depending on the fire situation, the decision to use carbon dioxide fire suppression to the storage building may be taken. The hypothetical radiation dose to the NEW in this fire scenario is 14.2 mSv which is about 28% of the regulatory annual dose to a worker.
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TABLE 6.3-1: SCREENING OF MALFUNCTIONS AND ACCIDENTS FOR L&IL WASTE
Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities Natural Initiating Events Earthquake The processing or storage building will be Operations and No further assessment is designed in accordance with the National Maintenance required. Building Code (NBC) and/or the National Fire Code (NFC). They are expected to remain standing. Flood Given the Darlington site characteristic, Operations and No further assessment is extensive flooding affecting the processing Maintenance required. or storage buildings is not a credible event. Thunderstorm Thunderstorms can potentially involve Operations and No further assessment is lightning striking either the processing or a Maintenance required. storage building. No public dose consequences are expected from this event, as the buildings will have appropriate grounding provisions.
Procedures will be in place not to allow waste transfer operations between buildings during inclement weather.
Tornado The processing or storage building will be Operations and No further assessment is designed in accordance with the National Maintenance required. Building Code (NBC) or and the National Fire Code (NFC). They are expected to remain standing.
Procedures will be in place not to allow waste transfer operations between buildings
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Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities during inclement weather.
Low and intermediate level waste containers would be largely protected by the heavy shielding walls of the storage buildings. Any consequences due to tornado missiles sticking a container would be bounded by the fire scenario. Aircraft Crash Aircraft Crash The probability of an aircraft strike is Operations and No further assessment is proportional to the target area. Therefore the Maintenance required postulated frequency is less than 10-7 events per year making this an incredible event [AMEC, 2008]. Therefore no further assessment is required.
Conventional Accidents Conventional Accidents Similar to and covered by the more general Site Preparation and Refer to Malfunctions, - spills cases for the overall EA. Refer to Construction Accidents and - release of chemicals Malfunctions, Accidents and Malevolent Malevolent Acts - fall of heavy equipment Acts Technical Support Document Operations and Technical Support - fire and explosion accidents Maintenance Document - releases of gases Low and Intermediate Level Waste – Radiological Accidents Vehicle/Package Accident A handling accident involving either low or Operations and No further assessment is During Transfer to intermediate waste is possible during a Maintenance required. Processing/Storage Building transfer to either a processing or storage (no fire) building. It is anticipated that the This will be bounded by consequences would be limited to a small an intermediate waste spill of package contents. This single package fire
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Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities contamination would be cleaned up as part of normal operations. It is assumed that any highly dispersible waste forms containing higher levels of radioactivity would be in a more robust container. Fire During Placement of a A possible accident here is a pool fire (spill Operations and Further assessment is Waste Container on the top row of gasoline or diesel fuel from a material Maintenance required. in a LLSB. handling vehicle that catches fire beside a stack of waste containers). This is the bounding case for low level waste Fire During Transfer to or A possible accident here is a pool fire (spill Operations and Further assessment is Storage of an Intermediate of gasoline or diesel fuel from a material Maintenance required. Level Waste Package in a handling vehicle that impacts on the waste) Storage Building involving transfer of an intermediate level This is the bounding waste form such as a 3 m3 resin liner. case for intermediate Intermediate waste packaging is assumed to level waste. be robust enough/response time soon enough that only a “confined” * burn occurs.
* In a confined burn, waste is not ejected from the container. Gaskets may fail and internal gases are allowed to escape from the container. A confined burn requires a container with secure fastenings and depending on contents some type of venting provision.
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6.3.3 Bounding Case for Intermediate Level Waste - Pool Fire Involving Resin Liner
A fire could also occur during intermediate waste handling particularly during material transfer to an above ground storage structure. The basic accident scenario again is a pool fire that occurs during transfer to an above-ground storage facility using equipment such as a forklift. The vehicle’s gasoline or diesel fuel tank has ruptured and the fuel has caught on tire. This material handling accident assumes only a relatively short distance between the forklift’s fuel tank and the waste container. The accident assumes a “confined burn”. This means that any intermediate level waste would be in a robust container with design provisions that in a fire scenario, the lid would stay intact. This could be done in a variety of means such as over-packing, bolting and venting provisions. A 3 m3 resin liner in an overpacked arrangement is considered to be a representative bounding case for intermediate level waste. Because the fire would involve less combustible material, it is modeled as a ground-level release. It is assumed that the majority of the dose from this postulated scenario comes from Carbon-14 and tritium. Because a future ACR-1000 may not have a tritium removal facility, the specific activity of tritium is increased by a factor of three from current values. The hypothetical radiation dose to a member of the public was calculated to be 83 µSv for a pool fire involving intermediate level waste which is about 8% of the regulatory limit to a member of the public. At the start of the fire, it is assumed that the NEW will immediately leave to get a fire extinguisher, return and empty the extinguisher at the fire. He/she then will leave the area. The time that a NEW would be exposed to the fire scenario is expected to be no more than three minutes. Either the NEW or an Emergency Response Team member will then return later but wearing full respiratory protection and some plastic protection against tritium uptake through the skin. Nearly all the NEW dose will come from the inhalation dose component in those first three minutes. The hypothetical dose to a NEW is 1.43 mSv which is about 3% of the regulatory annual dose to a worker.
6.4 Refurbishment Waste Storage and Handling
6.4.1 Screening of Malfunctions and Accidents for Refurbishment Waste
Malfunctions and accidents involving refurbishment waste represent a special case of intermediate level waste handling. It is assumed that because of the potentially much higher specific activity of refurbishment waste that it would be stored separately from other waste packages containing potential combustible waste. Therefore the accident scenarios are reduced to material handling accidents involving a partial spill from a container and release of contents. Table 6.4-1 summarizes the possible scenarios.
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TABLE 6.4-1: SCREENING OF MALFUNCTIONS AND ACCIDENTS FOR REFURBISHMENT WASTE
Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities Natural Initiating Events Earthquake The processing or storage building will be Operations and No further assessment is designed in accordance with the National Maintenance required. Building Code (NBC) and/or the National Fire Code (NFC). They are expected to remain standing and be structurally acceptable. Flood Given the Darlington site characteristic, Operations and No further assessment is extensive flooding affecting any processing Maintenance required. or storage buildings is not a credible event. Thunderstorm Thunderstorms can potentially involve Operations and No further assessment is lightning striking either the processing or a Maintenance required. storage building. No public dose consequences are expected from this event, as the buildings will have appropriate grounding provisions.
Procedures will be in place not to allow waste transfer operations between buildings during inclement weather.
Tornado The processing or storage building will be Operations and No further assessment is designed in accordance with the National Maintenance required. Building Code (NBC) or and the National Fire Code (NFC). They are expected to remain standing and be structurally acceptable.
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Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities Procedures will be in place not to allow waste transfer operations between buildings during inclement weather.
Refurbishment waste (such as refurbishment waste containers or steam generators) would also be expected to be largely unaffected by tornado missiles due to the containers’ design and the storage building design. Aircraft Crash Aircraft Crash The probability of an aircraft strike is Operations and No further assessment is proportional to the target area. Therefore the Maintenance required postulated frequency is less than 10-7 events per year making this an incredible event [AMEC, 2008]. Therefore no further assessment is required.
Conventional Accidents Conventional Accidents Similar to and covered by the more general Site Preparation and Refer to Malfunctions, - spills cases for the overall EA. Refer to Construction Accidents and -release of chemicals Malfunctions, Accidents and Malevolent Malevolent Acts - fall of heavy equipment Acts Technical Support Document Operations and Technical Support - fire and explosion accidents Maintenance Document - releases of gases Refurbishment Waste – Radiological Accidents Drop of a refurbishment waste During material handling or storage Operations and Further assessment is container activities with intermediate waste from Maintenance required. refurbishment, the container falls. It is assumed that the container is a robust container similar to the retube waste
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Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities containers being stored at WWMF. Drop of a steam generator Any openings in a steam generator would Operations and Further assessment is either be bolted or welded shut. During Maintenance required. waste moving and loading into a storage building, any potential drop would be quite short. Because the steam generator is a heavy pressure vessel, any damage would be limited to rupture of the bolted or welded cover. Drop of a Rector Head During storage operations a reactor head is Operations and No further assessment is sealed and then a drop of the reactor head Maintenance required. occurs during transfer to storage. It is assumed that steps would be taken before This accident would be moving the reactor head to either provide a bounded by the steam degree of containment or to fix potentially generator case. loose contamination in place. Note the hazard here comes from the thin corrosion film on the reactor head and not the reactor head itself. Fire Refurbishment waste is almost entirely non- Refurbishment No further assessment is combustible. Also some of the waste forms required such as Retube Waste Containers would have a relatively large thermal inertia. Current and future practices are that this type of waste is stored separately from the low level combustible waste. Releases from a fire event would be considered to be small.
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6.4.2 Bounding Case - Drop of a Retube Waste Container
During refurbishment of a reactor, there may be a need to remove some core components and store them. Because of the very high specific activity of core components there will be a need to store the waste in a very robust (can survive anticipated waste handling accidents) and heavily shielded container. Of the three reactor types, the ACR-1000 is expected to have the highest volume of refurbishment waste and probably the highest gamma hazard.
While the hypothetical radiation doses are very dependent on the time from reactor shutdown to removal of the core component, the maximum hypothetical radiation dose from the drop of retube waste container is 0.7 µSv to a member of the public which is less than 1% 0f the regulatory annual limit for radiation dose to a member of the public. The dose to a NEW due this postulated scenario is 4.1 mSv which is about 8% of the regulatory radiation dose limit for radiation dose to a worker.
6.4.3 Bounding Case - Drop of a Steam Generator
During refurbishment of all reactor types, replacement of steam generators is a possibility. A material handling accident is assumed where there is a short drop of a steam generator. All openings on the steam generator would have been previously welded or bolted shut. The accident assumes that there is a breech in one of these openings. The drop would jar loose some of the particulate on the steam generator internal surfaces. Some of this particulate would be in the respiratory range and would escape through the breech(es).
For the Continued Operation of Pickering B EA, a very conservative modelling was done for a steam generator drop. No credit was given for the point that the majority of particulate that would become airborne would be inside the steam generator tubes and would have little chance of escape. Despite the NND steam generators being larger, modelling of a NND steam generator drop using the methodology [US DOE, 2007] would be expected to give a lower result. The conservative modelling for the PNGS B refurbishment case of a steam generator drop with a hypothetical public radiation dose of <0.1 µSv to a member of the public (less than 0.01% of regulatory annual radiation dose limit to a member of the public) and 609 µSv for a NEW (about 1% of the regulatory annul radiation dose limit to a worker) would still represent a reasonable upper limit for the larger steam generators.
6.5 Used Fuel Dry Storage – Assessment of Bounding Accident
These scenarios for all three reactor start with the transfer of the loaded cask/container to the storage or processing building from the irradiated fuel bay area.
Normal operating releases or accidental releases from the fuel are due to volatile gases that escape through small defects in the fuel cladding. Accident conditions introduce a mechanical shock to the system that increase the percentage of defects in the fuel and lead to higher releases. The bounding case for mechanical shock is the design basis accident where a significant percentage (30%) of the fuel elements becomes damaged
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The accident scenarios for used fuel storage are similar for the three reactor designs. After the minimum 10 years of wet storage, the source term of volatile radionuclides available for release is reduced to the inventory of Kr-85 and tritium if an accident causing a defect in the fuel cladding occurs. Available for release refers to the fractions of Kr-85 and tritium that have migrated to the gap between the fuel and the cladding, and to the grain boundaries within the fuel pellet. These gap fractions vary for the different fuel types.
There are two cases that need to be considered for all three reactor types:
Acute Release – Malfunctions and Accidents: An acute release from a credible accident scenario in which used fuel bundles/assembles are damaged and the volatile inventories in the gap are released directly into the environment. The failure of 30% of fuel elements within the storage container/cask is assumed to occur. Realistically, fuel sheath failure is not expected to result from any postulated credible scenario. This case represents the bounding acute release assessment.
Criticality – There should be both design and operating provisions for used fuel handling and storage to ensure that nuclear criticality (the point in which a nuclear reaction is self-sustaining) cannot occur.
IAEA Safety Series document (IAEA, 2008) lays out the basic principles for safe storage of used nuclear fuels. It states that the storage facility should be designed to fulfill the fundamental safety functions including control of sub-criticality. It continues to state that a fundamental safety objective of all designs for used fuel storage facility is to ensure sub-criticality of the entire system under all credible circumstances.
The sub-criticality of the used fuel may be ensured or influenced by a number of design features and precautions.
x Material – Mass, Element, Enrichment, Heterogeneity x Shape – Geometry, Volume, Concentration, Density x Poison – Solid, Liquid x Others – Reflection, Moderator, Unit Interaction.
6.6 ACR-1000 Malfunctions and Accidents During Dry Storage of Used Fuel
6.6.1 Screening of Malfunctions and Accidents for ACR-1000
As described earlier, there are differences in the amount of used fuel in the different used fuel storage options. In general, the DSC approach will always be more conservative than the MACSTOR approach simply because more used fuel is involved and the DSC will be used for this assessment.
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Malfunctions and accidents involving the bounding used fuel dry storage canister are summarized in Table 6.6-1. The hypothetical radiation dose for the bounding case needs to be further assessed.
6.6.2 ACR-1000 – Assessment of Bounding Accident
To provide an upper bound estimate of a stylized acute release resulting from a worst-case accident during dry storage operations, public and occupational doses were estimated assuming failure of 30 percent of the used fuel in a container containing a total of 384 used fuel bundles. It is assumed that the container seal weld fails (or loss of seal and slightly negative atmospheric pressure inside a container) and the available free radionuclide inventory in the container is immediately released to the atmosphere.
The acute off-site dose consequences resulting from this stylized scenario, for a member of the public at the DN fenced site boundary, are estimated to be 21.6 µSv for an adult or 21.0 µSv for an infant. The adult dose estimate is about 2% of the regulatory dose limit of 1 mSv/year (1,000 µSv/year). The estimated acute maximum dose to a worker in the vicinity is 3.2 mSv. Because of the conservatism of the assumptions, these dose estimates are considered bounding for container handling accidents in dry storage operations at the DWMF.
6.6.3 Criticality Assessment for ACR-1000
The general principles for criticality control were discussed earlier in Section 6.5.
Criticality assessments previously carried out for natural uranium (NU) reference fuel bundles at DWMF concluded that the used fuel stored in DSCs would remain sub-critical under all normal and abnormal storage conditions, as well as under any credible accident conditions. Similar criticality assessments have also been done for the slightly enriched CANFLEX-LVRF (Low Void Reactivity Fuel). However, neither the conclusions drawn for the NU DWMF reference fuel bundles or the slightly enriched CANFLEX-LVRF necessarily extend to the storage of used CANFLEX–ACR fuel in similar containers due to the use of higher enriched uranium in the new fuel. In extending the application of the current or modified DSC design to used fuel from ACR- 1000 it must be demonstrated that used CANFLEX–ACR fuel remains sub-critical throughout the container loading, transfer, processing and storage operations, and under any credible accident scenario at the DWMF.
6.6.3.1 Normal Operations
Under normal operating conditions the fuel is held in a secure configuration, therefore, a change in configuration of the material is not possible. Thus ensuring sub-criticality under normal operating condition for CANFLEX-ACR used fuel is based on demonstrating
a) sub-critical conditions in each individual container, and b) sub-critical conditions inside the storage facility once all containers are loaded.
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TABLE 6.6-1: SCREENING OF MALFUNCTIONS AND ACCIDENTS FOR ACR-1000
Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities Natural Initiating Events Earthquake The processing or storage building will be Operations and No further assessment is designed in accordance with the National Maintenance required. Building Code (NBC) and/or the National Fire Code (NFC). They are expected to remain standing and be structurally acceptable. Flood Given the Darlington site characteristic, Operations and No further assessment is extensive flooding affecting the processing Maintenance required. or storage buildings is not a credible event. Thunderstorm Thunderstorms can potentially involve Operations and No further assessment is lightning striking either the processing or a Maintenance required. storage building. No public dose consequences are expected from this event, as the buildings will have appropriate grounding provisions.
Procedures will be in place not to allow waste transfer operations between buildings during inclement weather. Tornado The processing or storage building will be Operations and No further assessment is designed in accordance with the National Maintenance required. Building Code (NBC) or and the National Fire Code (NFC). They are expected to remain standing and be structurally acceptable.
Procedures will be in place not to allow
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Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities waste transfer operations between buildings during inclement weather.
Either DSCs or the MACSTOR approach would also be expected to be largely unaffected by tornado missiles because of heavy shielding and/or thick walls of the storage building. Aircraft Crash Aircraft Crash The probability of an aircraft strike is Operations and No further assessment is proportional to the target area. Therefore the Maintenance required postulated frequency is less than 10-7 events per year making this an incredible event [AMEC, 200]. Therefore no further assessment is required. Conventional Accidents Conventional Accidents Similar to and covered by the more general Site Preparation and Refer to Chapter 3 in - spills cases for the overall EA. Refer to Chapter 3 Construction Malfunctions, Accidents -release of chemicals in Malfunctions, Accidents and Malevolent and Malevolent Acts - fall of heavy equipment Acts Assessment of Environmental Effects Operations and Assessment of - fire and explosion accidents Technical Support Document Maintenance Environmental Effects - releases of gases Technical Support Document Nearby Fire to Storage Container Fire near the dry storage The heavily shielded containers for dry fuel Operations and No further assessment is container storage would have considerable thermal Maintenance required. inertia. Also it would be operating practice to store them away from any appreciable amount of combustible material. It is highly unlikely that the containers/casks would see
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Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities any fire strong enough to lead to a release.
ACR-1000 – Bounding Accident Bounding case is postulated to Gaseous radionuclides (Tritium, Kr-85) will Operations and Further assessment is be a drop of a loaded canister be released. Maintenance required causing damage to 30% of the fuel pencils. ACR-1000 Criticality A criticality incident might Design features should ensure that a Operations and No further assessment is occur during used fuel criticality event would not occur. Maintenance required. operations,
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For storage of containers containing used fuel, ANS 8.7 Section 4.2.1, states that the “limits for storage of fissile material shall be based on experimental data or on the results of calculations made through the use of validated computational techniques”. This requirement could set limits on the size and the configuration of the array in which containers are arranged in the storage buildings, since an increased number of containers involves a larger mass of fissile material, and hence implies a greater potential for neutron interaction between containers. However, the concrete of the container walls effectively shields out neutrons originating outside the container, so that there is essentially no neutron interaction between containers. There are therefore no constraints, from a criticality perspective, on the number of containers that can be stored or the configuration in which they are stored.
6.6.3.2 Abnormal Operating Conditions and Credible Accidents
Under abnormal operating conditions or credible accidents two possible factors which could influence the potential for criticality must be considered. These underlying changes in process conditions involve:
a) configuration of the material, and b) addition of a moderator.
None of the previously considered accident scenarios could lead to a criticality incident as there is no case in which the lid of the container fails, leading to potential loss of configuration of the fuel bundles, nor is there a pathway created for ingress of a moderator.
There are, however, other abnormal operating conditions which could be postulated, that may not previously have been considered, as they would not give rise to a release of radioactive material. They could however be significant from a criticality perspective. These identified conditions are:
inadequate/incomplete drainage/drying of the container prior to transfer to the DWMF – the water remaining in the container would act as a moderator, inadvertent addition of a moderator into the container during processing at the DWMF, and loss of fuel/module integrity during long term storage, resulting in a change of configuration of the fuel bundles.
6.6.3.2.1 Inadequate Drainage of the Container
Sub-criticality of the contents of a container must be guaranteed for the complete range of conditions from the loaded container filled with water to being completely drained. The most limiting configuration for criticality safety is the situation where there is water retained within the storage module tubes in and around bundles, but reduced density or no water between the tubes, as this condition maximizes the neutron coupling between the bundles. Nonetheless, neutron absorption in the stainless steel module tubes ensures that, even in this limiting configuration, substantial sub-criticality margin remains.
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While there is potential for some water to remain in the container after draining and vacuum drying, it is not likely that the container, as readied for transfer to the DWMF, will retain sufficient quantities to mirror the conditions for the most limiting configuration. Any residual water would be expected to condensate in cooler areas such as the bottom of the container away from the fuel. Furthermore, upon arrival at the DWMF, the containers undergo additional draining and vacuum drying steps. During further processing and subsequent storage, therefore, it is extremely unlikely that the containers will retain any water at all.
It can thus be concluded that containers containing CANFLEX–ACR bundles as transferred from the NND to the DWMF are unlikely to contain substantial quantities of water as a result of inadequate drainage or drying of the container interior cavity. However, in the event that some water is present, the configuration is bounded by the limiting condition addressed above. Therefore, this scenario is not assessed further.
6.6.3.2.2 Addition of a Moderator during Container Processing at the DWMF
The potential for a moderator, other than water, to enter the container during the processing operations is discussed in this section. The most likely point in the operations for this to happen is while the container is under vacuum during the final drying and helium backfill process. This operation is carried out via the drain port located at the bottom of the container. This port has a non-return valve. A pump is connected to the drain port and the vacuum generated. The pump is then disconnected and the vacuum maintained by the action of the non-return valve. The helium is delivered to the container via a tube which incorporates a tool to release the non-return valve. Ingress of a moderator, for example, oil from a malfunctioning pump, is therefore prevented by the non-return valve or the physical attachment of the helium delivery tube. Hence, even if the non-return valve failed in such a way that a route into the container was possible, no significant quantities of fluid are available to be drawn into the container. The other fluid with potential for ingress to the container is fire fighting water or foam.
In order for fire fighting liquid to get into the container it would need to have a point of ingress. At the only point in the operations where the container is held under vacuum, entry of water or foam is prevented by the non-return valve on the drain port, as discussed above. The other entry point to the container is the vent port, which is sealed via a welded plug at the time the lid is welded in place. The vent port is located on the side of the container lid, so there is no potential for sufficient liquid to accumulate on a surface of the container and, as a consequence, drip through the port. Furthermore, prior to the final seal welding of the port, a transfer plug is in place. This scenario is therefore not credible given the lack of an entry route for the liquid into the container and is not assessed further.
6.6.3.2.3 Loss of Container or Storage Module Integrity during Long Term Storage
The long term integrity of the container is maintained during storage. The integrity of the concrete and steel components (including welds) will be adequate to provide at least 50 years of service if stored indoors with a maintenance program. The modules are made of stainless steel and, as such, are not susceptible to the corrosion/oxidation which could affect the carbon steel
6-19 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document components of the container. The integrity of the modules will, therefore, be at least similar to that of the containers.
The only scenario which could adversely affect the criticality safety of the container would be a loss of configuration of the fuel bundles, which would require a massive failure of the structure of the modules or the container. The integrity of all components ensures that such a failure, within the 50 year storage period considered is not credible and is not assessed further.
6.6.3.3 Criticality Safety Assessment Summary
Prior to considering use of either the current OPG design of DSC, a modified version of the DSC, or indeed any alternative storage option such as the MACSTOR system, for dry storage of used CANFLEX-ACR fuel, the system design will first be assessed against all applicable design requirements, including those related to criticality safety. The design of the container, as well as the entire used fuel management process, must ensure that used CANFLEX-ACR fuel remains adequately sub-critical throughout the proposed container loading, transfer, processing and storage operations, and under any credible accident scenarios. The assessment presented above indicates that, given a used fuel storage container and process which has been suitably qualified for this design basis, there are no anticipated scenarios for dry storage of used fuel from ACR- 1000 under which criticality safety objectives would not be met.
6.7 EPR Malfunctions and Accidents During Dry Storage of Used Fuel
6.7.1 Screening of Malfunctions and Accidents for EPR
For malfunctions and accidents assessment, a more conservative value (Maximum PWR, [AMEC, 2008]) will be used for the source term for hypothetical radiation dose assessments. Table 6.7-1 gives the properties for 10 year cooled EPR fuel [AMEC, 2008] and for Maximum PWR.
TABLE 6.7-1: PROPERTIES OF 10 YEAR COOLED EPR FUEL
Key Parameter EPR Maximum PWR Kr-85 per assembly (Bq) 1.69E+14 1.55E+14 Kr-85 Gap Fraction 0.1 0.1 H-3 per assembly (Bq) 9.03E+12 1.39E+13 H-3 Gap Fraction 0.05 0.05 Estimated Annual Inventories 85 85 (Assemblies) Fuel Defect Rates 0.25% 0.25%
Malfunctions and accidents involving used fuel casks are summarized in Table 6.7-2. The hypothetical radiation dose for the bounding case needs to be further assessed.
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TABLE 6.7-2: SCREENING OF MALFUNCTIONS AND ACCIDENTS FOR EPR
Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities Natural Initiating Events Earthquake The processing or storage building will be Operations and No further assessment is designed in accordance with the National Maintenance required. Building Code (NBC) and/or the National Fire Code (NFC). They are expected to remain standing and be structurally acceptable. Flood Given the Darlington site characteristics, Operations and No further assessment is extensive flooding affecting the processing Maintenance required. or storage buildings is not a credible event. Thunderstorm Thunderstorms can potentially involve Operations and No further assessment is lightning striking either the processing or a Maintenance required. storage building. No public dose consequences are expected from this event, as the buildings will have appropriate grounding provisions.
Procedures will be in place not to allow waste transfer operations between buildings during inclement weather.
Tornado The processing or storage building will be Operations and No further assessment is designed in accordance with the National Maintenance required. Building Code (NBC) or and the National Fire Code (NFC). They are expected to remain standing and be structurally acceptable.
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Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities Procedures will be in place not to allow waste transfer operations between buildings during inclement weather.
Because of thick walls of the storage building and the heavy shielding of the cask itself, dry fuel storage casks would also be expected to be largely unaffected by tornado missiles. Aircraft Crash Aircraft Crash The probability of an aircraft strike is Operations and No further assessment is proportional to the target area. Therefore the Maintenance required postulated frequency is less than 10-7 events per year making this an incredible event [AMEC, 2008]. Therefore no further assessment is required.
Conventional Accidents Conventional Accidents Similar to and covered by the more general Site Preparation and Refer to Malfunctions, - spills cases for the overall EA. Refer to Construction Accidents and -release of chemicals Malfunctions, Accidents and Malevolent Malevolent Acts - fall of heavy equipment Acts Technical Support Document Operations and Assessment Technical - fire and explosion accidents Maintenance Support Document - releases of gases Nearby Fire to Storage Cask Fire near the dry storage cask. The heavily shielded casks for dry fuel Operations and No further assessment is storage would have considerable thermal Maintenance required. inertia. Also it is operating practice to store them away from any appreciable amount of combustible material. It is highly unlikely
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Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities that the containers/casks would see any fire strong enough to lead to a release. EPR – Bounding Accident Bounding case is the drop of a Gaseous radionuclides (Tritium, Kr-85) will Operations and Further assessment is cask with a failure of 30% of be released. Maintenance required. the fuel elements. EPR - Criticality A criticality incident might Design features should ensure that a Operations and No further assessment is occur during used fuel criticality event would not occur. Maintenance required. operations,
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6.7.2 EPR – Assessment of Bounding Accident
To provide an upper bound estimate of an acute hypothetical release resulting from a worst-case accident during used fuel dry storage operations, public and occupational doses were estimated assuming failure of 30% of used fuel in a cask. It is not expected that significant fuel failure will occur under any of the postulated handling and storage events given the robust design of the cask and fuel assemblies themselves. It is assumed for this assessment that the seal weld or bolted containment boundary of the cask will be lost and the free radionuclide inventory in the container is immediately released into the atmosphere.
The acute off-site dose consequences resulting from this hypothetical scenario, for a member of the public at the DN site boundary will be dependent on whether a 24 or 40 assembly cask is involved. The hypothetical radiation doses for a 24 assemblies cask are 142 µSv for an adult and 144 µSv for an infant. The hypothetical radiation dose for a 40 assemblies cask are 237 µSv for and adult and 240 µSv for an infant. The adult dose estimate is about 24% percent of the regulatory dose limit of 1 mSv/year (1,000 µSv/year) for the 40 assemblies cask bounding accident.
Assuming that the worker is immersed in a cloud consisting of the available free radionuclide inventory, the estimated acute maximum hypothetical radiation doses to a worker in the vicinity are 20.4 mSv from a 24 used fuel assemblies cask and 33.9 mSv, from a 40 used fuel assemblies cask. This is about 68% of the one year radiation dose limit of 50 mSv for a NEW.
6.7.3 Criticality Assessment for EPR
The general principles for criticality control were discussed earlier in Section 6.5.
PWR containers for an EPR will be designed to remain sub-critical. However sub-criticality can be influenced by internal or external hazards which have the potential to reconfigure the pre- existing used fuel assembly array in such a way as to increase the potential for criticality. Erosion of the criticality safety margin can occur if any of the factors listed above are changed. Therefore those accident scenarios which may result in a change in any of these factors must be considered, and the storage facility designed in such a way as to either make these accidents incredible or to maintain sufficient criticality safety margin that the resulting changes are not able to cause criticality. Fuel baskets and containers for used fuel storage should be designed in such a way as to ensure that the used fuel will remain in a configuration which has been determined to be sub-critical during loading, transport, storage, and retrieval. The used fuel dry storage facilities will be designed in such a way that consequences likely to result from the redistribution or the introduction of a moderator as a consequence of an internal or external event can be accommodated.
Criticality events are prevented by the absence of a moderator and provision of sufficient poison in the basket even if the internal fuel orientations are changed by events. The lack of a moderator in the cask is significant since, at enrichments below 5%, un-moderated criticality is not physically possible under any conditions in the absence of other neutron sources (NRC, 2007). In the absence of moderation, experiments and calculations have demonstrated
6-24 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document that criticality is not possible at the enrichments currently used in light-water reactor (LWR) fuel (NRC, 2007). Furthermore, prior to dry storage, the fuel assemblies would previously have been stored in the fuel pool with neutron poison in the racks to ensure sub-criticality. The smaller subset of fuel elements placed in the dry storage cask may be unable to become critical even with moderator (water) in cask (EPRI, 2004). As such, assessment of the PWR container must demonstrate no credible accident can lead to ingress of a moderator.
Alternatively, if a design is chosen which cannot meet this objective, additional neutron absorbing material may be required, such as a boron poison in the form of boral plating. Boral plating on the basket is an effective means of criticality control. With this poison present, criticality is not credible whether the assembly is moderated or unmoderated. No credible scenario which would eliminate the boral and leave the geometry of the fuel intact has been identified. Even if the boral plates were to separate from the basket structure, they would have limited room to move and, once the basket is sealed, there is no credible means by which they could fall out of the basket. If the contents of the cask were to somehow relocate, the boral plating would relocate with the debris. In all likelihood, this postulated configuration would be highly sub-critical because of the boral plating and because a bed of debris is not an optimal geometry [(NRC, 2007).
Ensuring criticality safety under all credible circumstances will be achieved as a design requirement of the technology option selected for dry storage of used EPR fuel.
6.8 AP1000 Malfunctions and Accidents During Dry Storage of Used Fuel
6.8.1 Screening of Malfunctions and Accidents for AP1000
For malfunctions and accidents assessment, a more conservative value (Maximum PWR, [AMEC, 2008]) will be used for the source term for hypothetical radiation dose assessments. Table 6.8-1 gives the properties for 10 year cooled AP1000 fuel and for Maximum PWR.
TABLE 6.8-1: PROPERTIES OF 10 YEAR COOLED AP1000 FUEL
Key Parameter AP1000 Maximum PWR Kr-85 per assembly (Bq) 1.31E+14 1.55E+14 Kr-85 Gap Fraction 0.1 0.1 H-3 per assembly (Bq) 9.03E+12 1.39E+13 H-3 Gap Fraction 0.05 0.05 Estimated Annual Inventories 43 43 (Assemblies) Fuel Defect Rates 0.25% 0.25%
Malfunctions and accidents involving used fuel casks are summarized in Table 6.8-2. The hypothetical radiation dose for the bounding case needs to be further assessed.
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TABLE 6.8-2: SCREENING OF MALFUNCTIONS AND ACCIDENTS FOR AP1000
Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities Natural Initiating Events Earthquake The processing or storage building will be Operations and No further assessment is designed in accordance with the National Maintenance required. Building Code (NBC) or and the National Fire Code (NFC). They are expected to remain standing and be structurally acceptable. Flood Given the Darlington site characteristics, Operations and No further assessment is extensive flooding affecting the processing Maintenance required. or storage buildings is not a credible event. Thunderstorm Thunderstorms can potentially involve Operations and No further assessment is lightning striking either the processing or a Maintenance required. storage building. No public dose consequences are expected from this event, as the buildings will have appropriate grounding provisions.
Procedures will be in place not to allow waste transfer operations between buildings during inclement weather.
Tornado The processing or storage building will be Operations and No further assessment is designed in accordance with the National Maintenance required. Building Code (NBC) or and the National Fire Code (NFC). They are expected to remain standing and be structurally acceptable.
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Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities Procedures will be in place not to allow waste transfer operations between buildings during inclement weather.
Because of thick walls of the storage building and the heavy shielding of the cask itself, dry fuel storage casks would also be expected to be largely unaffected by tornado missiles. . Aircraft Crash Aircraft Crash The probability of an aircraft strike is Operations and No further assessment is proportional to the target area. Therefore the Maintenance required postulated frequency is less than 10-7 events per year making this an incredible event [AMEC, 2008]. Therefore no further assessment is required.
Conventional Accidents Conventional Accidents Similar to and covered by the more general Site Preparation and Refer to Chapter 3 in - spills cases for the overall EA. Refer to Chapter 3 Construction Malfunctions, Accidents -release of chemicals in Malfunctions, Accidents and Malevolent and Malevolent Acts - fall of heavy equipment Acts Assessment of Environmental Effects Operations and Assessment of - fire and explosion accidents Technical Support Document Maintenance Environmental Effects - releases of gases Technical Support Document Nearby Fire to Storage Cask Fire near the dry storage cask. The heavily shielded casks for dry fuel Operations and No further assessment is storage would have considerable thermal Maintenance required. inertia. Also it is operating practice to store them away from any appreciable amount of
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Potential Malfunction or Preventative and Mitigative Measures/ Project Phase/Works Screening Decision Accident Scenario Screening Evaluation and Activities combustible material. It is highly unlikely that the containers/casks would see any fire strong enough to lead to a release. AP1000 – Bounding Accident Bounding case is a drop of a Gaseous radionuclides (H-3, Kr-85) will be Operations and Further assessment is loaded cask and it is assumed released. Maintenance required. that 30% of the fuel pencils fail. AP1000 - Criticality A criticality incident might Design features should ensure that a Operations and No further assessment is occur during used fuel criticality event would not occur. Maintenance required. operations,
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6.8.2 AP1000 – Assessment of Bounding Accident
To provide an upper bound estimate of an acute hypothetical release resulting from a worst-case accident during used fuel dry storage operations, public and occupational doses were estimated assuming failure of 30% of used fuel in a cask. It is not expected that significant fuel failure will occur under any of the postulated handling and storage events given the robust design of the cask and fuel assemblies themselves. It is assumed for this assessment that the seal weld or bolted containment boundary of the cask will be lost and the free radionuclide inventory in the cask is immediately released into the atmosphere.
The acute off-site dose consequences resulting from this hypothetical scenario, for a member of the public at the DN site boundary will be dependent on whether a 24 or 40 assembly cask is involved. The hypothetical radiation dose for a 24 assemblies cask are 142 µSv for and adult and 144 µSv for an infant. The hypothetical radiation dose for a 40 assemblies cask are 237 µSv for and adult and 240 µSv for an infant. The adult dose estimate is less than 25% percent of the regulatory dose limit of 1 mSv/year (1,000 µSv/year) for the 40 assemblies cask case.
Assuming that the worker is immersed in a cloud consisting of the available free radionuclide inventory, the estimated acute maximum hypothetical radiation doses to a worker in the vicinity are 20.4 mSv from a 24 used fuel assemblies cask and 33.9 mSv, from a 40 used fuel assemblies cask. The hypothetical radiation to a NEW for a 40 assemblies flask bounding accident is 68% of the annual radiation dose limit of 50 mSv.
6.8.3 Criticality Assessment for AP1000
The general principles for criticality control were discussed earlier in Section 6.5.
PWR containers for an AP1000 will be designed to remain sub-critical. However sub-criticality can be influenced by internal or external hazards which have the potential to reconfigure the pre- existing used fuel assembly array in such a way as to increase the potential for criticality. Erosion of the criticality safety margin can occur if any of the factors listed above are changed. Therefore those accident scenarios which may result in a change in any of these factors must be considered, and the storage facility designed in such a way as to either make these accidents incredible or to maintain sufficient criticality safety margin that the resulting changes are not able to cause criticality. Fuel baskets and containers for used fuel storage should be designed in such a way as to ensure that the used fuel will remain in a configuration which has been determined to be sub-critical during loading, transport, storage, and retrieval. The used fuel dry storage facilities will be designed in such a way that consequences likely to result from the redistribution or the introduction of a moderator as a consequence of an internal or external event can be accommodated.
Criticality events are prevented by the absence of a moderator and provision of sufficient poison in the basket even if the internal fuel orientations are changed by events. The lack of a moderator in the cask is significant since, at enrichments below 5%, un-moderated criticality is not physically possible under any conditions in the absence of other neutron sources (NRC, 2007). In the absence of moderation, experiments and calculations have demonstrated
6-29 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document that criticality is not possible at the enrichments currently used in light-water reactor (LWR) fuel (NRC, 2007). Furthermore, prior to dry storage, the fuel assemblies would previously have been stored in the fuel pool with neutron poison in the racks to ensure sub-criticality. The smaller subset of fuel elements placed in the dry storage cask may be unable to become critical even with moderator (water) in cask (EPRI, 2004). As such, assessment of the PWR container must demonstrate no credible accident can lead to ingress of a moderator.
Alternatively, if a design is chosen which cannot meet this objective, additional neutron absorbing material may be required, such as a boron poison in the form of boral plating. Boral plating on the basket is an effective means of criticality control. With this poison present, criticality is not credible whether the assembly is moderated or unmoderated. No credible scenario which would eliminate the boral and leave the geometry of the fuel intact has been identified. Even if the boral plates were to separate from the basket structure, they would have limited room to move and, once the basket is sealed, there is no credible means by which they could fall out of the basket. If the contents of the cask were to somehow relocate, the boral plating would relocate with the debris. In all likelihood, this postulated configuration would be highly sub-critical because of the boral plating and because a bed of debris is not an optimal geometry (NRC, 2007).
Ensuring criticality safety under all credible circumstances will be achieved as a design requirement of the technology option selected for dry storage of used AP1000 fuel.
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7.0 SUMMARY
This TSD relates to Nuclear Waste Management, providing an assessment of any radiological effects on NEWs, and members of the public that may result from operating the proposed waste management facility to support NND. It also considered the malfunctions and accident scenarios related to the different waste types.
The EA considers two options for storage of L&ILW; on-site, using compaction on LLW where possible, packaging, and a modular storage building; and off-site, transporting un-processed LLW to an appropriately licensed facility. Storage is assumed to be in appropriate licenced containers and placed “standard” SBs. Refurbishment wastes are conservatively assumed to be stored on-site until the stations are decommissioned, at which point they will be transported off- site to a suitably licensed repository.
L&ILW from the light water reactors are expected to have much less tritium and C-14 than the current CANDU reactors. Tritium and C-14 from the ACR-1000 will also be reduced. An important “new” waste stream is related to the boric acid system used for reactivity control in light water reactors and criticality control in the used fuel bays.
Steam generator radioactivity is expected to be similar for all reactor types.
The fuel from all of the new-build reactors will have higher enrichment than current CANDU fuels. This introduces elements of criticality control requirements for storage as well as potential heat load issues for dry storage and eventual disposal.
All three reactor types have 10 to 15 years of wet bay storage and assume that older fuel will be transferred to dry storage as the bay fills up. The timing of dry storage operation depends on the fuelling cycle. It is assumed that 50% of the lifetime used fuel will require on-site dry storage and that the NWMO will take ownership of the fuel.
Long term site planning will need to include space for three L&ILW SBs (4,500 m2), three UFDS buildings (16,000 m2), one UFDS processing building (2,000 m2), one steam generator storage building (4,550 m2), and one refurbishment waste storage building (3,150 m2). The total area should include a minimum 5 m buffer between the storage buildings and the waste management facility fence, plus an additional buffer for security around the UFDS buildings. While there is no reason to believe that a Safety Assessment could not demonstrate that a location north of the CN rail line is feasible, the safety assessment used in this TSD assume that any waste processing or storage building are built south of the CN rail line and no closer than 150 m to the site perimeter fence. However, for EA planning purposes, this TSD has accepted this analysis to demonstrate that the UFDS can be located anywhere on the site. Should the Vendor require the UFDS buildings to be located north of the CN rail line, or any waste processing or storage building to be located closer than 150 m to the site perimeter fence, OPG has committed to updating safety assessment for this location as part of the licensing process.
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Hypothetical radiation doses to a member of the pubic and to a NEW for malfunctions and accident situations were calculated and were found to be within their respective dose limits. The worst case is the bounding accident to a 40 assemblies cask. Using a slightly more conservative source term than the vendor supplied information, the hypothetical dose to a member of the public are 237 µSv to an adult and 240 µSv to an infant. This corresponds to slightly less than 25% of the regulatory annual radiation dose limit to a member of the public. The corresponding hypothetical NEW dose is 33.9 mSv. This corresponds to about 68% of the regulatory annual radiation dose limit to a worker.
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8.0 REFERENCES
(AECL) Atomic Energy of Canada Ltd, 2007. “ACR-1000 Technical Description”, AECL Report 10820-01371-TED-001, Rev 1, June 2007. (AMEC) AMEC NSS Limited, 2008., “New Nuclear – Darlington ACR-1000, AP1000, EPR Used Fuel Dry Storage Option”, AMEC Report P1114/RP/001, October 2008. (AMEC) AMEC NSS Limited, 2008a. “Darlington Retube Waste Radionuclide Characterization”, Report P1047/RP/001 prepared for Nuclear Waste Management Division, March 13, 2008. Areva, 2007. U.S. EPR Final Safety Analysis Report. Areva Report, Rev 0. Candesco, 2008. “The Use of Plant Parameter Envelopes to Assess the Reactor Designs Being Considered for the Darlington Site”, Candesco Report OPG-PPE-00025-0 prepared for Ontario Power Generation, March 2008. Issued as OPG Report CD# N-REP-01200- 10000, April 2008. (EPRI) Electric Power Research Institute, 2004. “Probabilistic Risk Assessment (PRA) of Bolted Storage Casks. Updated Quantification and Analysis Report”, 1009691, EPRI Technical Report December 2004. (IAEA) International Atomic Energy Agency, 2008. “Storage of Spent Fuel”, IAEA Safety Standards, Draft Safety Guide DS 371, February 11 2008. (IAEA) International Atomic Energy Agency, 2007. Operation and Maintenance of Spent Fuel Storage and Transportation Casks/Containers. IAEA TECDOC Series 1532, January (IAEA) International Atomic Energy Agency, 2004. Status of Advanced Light Water Reactor Designs. IAEA TECDOC Series 1391, July 2004. Klassen, K. J., 2006. CNSC Letter to K. E. Nash of OPG, “Western Waste Management Facility – Integrated EA Follow-up Program Reports”, File: CD# 01098-CORR-00531- 00341, February 15, 2006. Nash, K.E., 2005. OPG Letter to K. Klassen of the CNSC, “Western Waste Management Facility – Integrated EA Follow-up Program Reports”, File: CD# 01098-CORR-00531- 00314, June 30, 2005 (includes four reports). (NRC) U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Washington, 2007. “A Pilot Probabilistic Risk Assessment Of a Dry Cask Storage System At a Nuclear Power Plant”, DC 20555-0001, NUREG 1864, March 2007. (NWMO) Nuclear Waste Management Organization, 2005, Report, “Choosing a Way Forward The Future Management of Canada’s Used Nuclear Fuel Final Study”, November 2005. (OPG) Ontario Power Generation Inc., 2008. OPG Report, “2007 Results of Radiological Environmental Monitoring Programs”, CD# N-REP-03481-10006, April 2008. (OPG) Ontario Power Generation Inc., 2007. OPG Report, “2006 Results of Radiological Environmental Monitoring Programs”, CD# N-REP-03481-10005-R01, December 2007.
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(OPG) Ontario Power Generation Inc., 2006. OPG Report, “Annual Summary and Assessment of Environmental Radiological Data for 2005”, CD# N-REP-03481-10004-R01, November 2006. (OPG) Ontario Power Generation Inc., 2003. “Darlington Used Fuel Dry Storage Project Environmental Assessment”, March 2003. Pontikakis, N., C.R. Boss, K.F. Hau, and K. Wittann, 2005. Improved Design Features for ACR- 700 Radioactive Waste Management Systems. Paper presented at Canadian Nuclear Society conference on Waste Management, Decommissioning and Environmental Restoration for Canada’s Nuclear Activities: Current Practices and Future Needs, Ottawa, Ontario Canada May 8-11 2005. Rodrigues, F., 2008. Reference Low and Intermediate Level Waste Inventory For The Deep Geologic Repository, OPG Report, CD# 00216-REP-03902-00003-R001, Ontario Power Generation, August 2008. Russell, Sean, 2008, “Preliminary Assessment of Potential Technical Implications of Reactor Refurbishment and New Nuclear Build on Adaptive Phased Management”, NWMO Report TR-2008-10 November 2008. (US DOE) U.S. Department of Energy, 2008. “Final Supplemental Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High- Level Radioactive Waste at Yucca Mountain, Nye County, Nevada”, Report DOE/EIS- 0250FS1, June 2008. (US DOE) U.S. Department of Energy, 2007. “Preparation of Safety Basis Documents for Transuranic (TRU) Waste Facilities”, DOE Standard File DOE-STD-5506-2007, April 2007 (US DOE) U.S. Department of Energy, 2006, DOE Report, “Waste Isolation Pilot Plant Contact Handled (CH) Waste Documented Safety Analysis”,: File DOE/WIPP-95-2065 Revision 10, November 2006. Vattenfall, 2007. :Aktivitetsinnehåll i skrotad ånggenerator från Ringhals 3 (Radioactivity in scrap steam generator from the Ringhals 3), Vattenfall Report 1945206 rev 4.0, June 29, 2007. Westinghouse Electric Company (Westinghouse), 2007. “UK AP1000 Safety, Security and Environmental Report”, Westinghouse Report UKP-GW-GL-700, Rev 1, August 2007.
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APPENDIX A
OPERATIONAL WASTE DETAILS
New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
TABLE A-1: SUMMARY OF AVERAGE L&ILW SPECIFIC ACTIVITY
ACR AP1000 USEPR OPG CANDU Avg L&ILW Avg L&ILW Avg L&ILW Avg L&ILW Avg LLW Avg ILW Radionuclide Quantity Specific Quantity Specific Quantity Specific Specific Specific Specific (Bq/yr) Activity (Bq/yr) Activity (Bq/yr) Activity Activity Activity Activity (Bq/m3) (Bq/m3) (Bq/m3) (Bq/m3) (Bq/m3) (Bq/m3) Ag-108m 7.7E+08 3.4E+06 1.5E+04 2.0E+03 1.5E+05 Ag-110m 7.5E+10 4.6E+08 1.7E+09 1.1E+07 4.1E+12 1.8E+10 Am-241 6.4E+06 4.6E+06 2.6E+07 Ba-137m 1.3E+13 5.7E+10 Ba-140 3.2E+09 2.1E+07 7.1E+09 3.2E+07 Be-7 1.6E+08 7.3E+05 C-14 1.1E+10 6.8E+07 1.0E+10 4.5E+07 5.4E+10 3.2E+08 6.3E+11 Cl-36 7.6E+03 3.5E+02 8.5E+04 Ce-141 5.0E+09 3.0E+07 2.7E+09 1.2E+07 Ce-144 5.0E+09 3.0E+07 2.9E+10 1.3E+08 Cm-244 4.6E+06 1.8E+06 3.4E+07 Co-58 2.3E+12 1.5E+10 4.2E+12 1.9E+10 Co-60 6.1E+11 3.7E+09 1.1E+13 6.9E+10 8.7E+12 3.9E+10 9.5E+09 6.7E+09 3.9E+10 Cr-51 1.6E+12 9.6E+09 1.1E+10 7.0E+07 9.1E+10 4.0E+08 Cs-134 5.0E+09 3.0E+07 6.8E+12 3.0E+10 6.3E+08 6.3E+07 6.7E+09 Cs-137 1.3E+11 7.9E+08 1.3E+13 5.7E+10 1.5E+09 2.5E+08 1.4E+10 Eu-152 1.1E+08 1.7E+04 1.3E+09 Eu-154 5.2E+07 2.7E+07 3.2E+08 Eu-155 1.8E+06 2.2E+05 1.9E+07 Fe-55 1.1E+12 6.9E+09 1.2E+13 7.5E+10 1.8E+13 7.8E+10 2.6E+10 1.8E+10 1.1E+11 Fe-59 2.0E+10 1.2E+08 6.3E+10 2.8E+08 Gd-153 1.1E+11 6.4E+08 H-3 5.9E+10 3.9E+08 3.2E+09 1.4E+07 1.5E+11 1.4E+11 1.6E+11 I-129 3.4E+07 1.5E+05 1.3E+03 6.8E+01 1.5E+04 I-131 2.8E+12 1.7E+10 3.4E+08 1.5E+06 I-133 1.6E+11 9.4E+08 La-140 1.5E+09 9.6E+06 7.1E+09 3.2E+07 Mn-54 2.0E+10 1.2E+08 8.3E+11 5.4E+09 1.3E+13 5.9E+10 Na-24 1.5E+10 9.1E+07 Nb-93m 5.1E+06 6.0E+07 Nb-94 1.1E+06 2.1E+05 1.0E+07 Nb-95 5.6E+12 3.4E+10 1.2E+10 7.8E+07 2.8E+11 1.3E+09 Ni-59 4.3E+08 1.9E+06 4.1E+06 2.1E+05 4.5E+07 Ni-63 1.2E+13 7.6E+10 1.6E+11 7.3E+08 7.0E+08 5.7E+07 7.5E+09 Pr-144 2.8E+10 1.2E+08 Pt-193 8.2E+04 9.5E+05 Pu-238 1.7E+06 1.5E+06 4.3E+06 Pu-239 3.2E+06 2.7E+06 8.5E+06 Pu-240 4.8E+06 4.1E+06 1.2E+07 Pu-241 4.2E+09 2.7E+07 1.3E+10 5.6E+07 2.1E+08 1.0E+08 1.3E+09 Rh-103m 4.2E+11 1.9E+09 Rh-106 1.8E+11 8.0E+08 Ru-103 7.5E+10 4.6E+08 4.5E+11 2.0E+09 Ru-106 4.5E+10 2.7E+08 7.7E+11 3.4E+09 9.7E+08 5.9E+08 5.0E+09 Sb-124 3.9E+11 2.4E+09 4.2E+08 1.9E+06 Sb-125 1.4E+09 6.2E+06 4.1E+09 7.2E+07 4.7E+10 Sn-113 5.8E+07 2.6E+05 Sn-121m 1.4E+07 1.6E+08 Sr-89 1.6E+11 7.0E+08 Sr-90 5.0E+09 3.0E+07 9.7E+10 4.3E+08 1.0E+09 5.4E+08 6.3E+09 Tc-99 8.8E+03 6.9E+01 1.0E+05 Te-127m 7.5E+11 3.3E+09 Te-129 1.3E+10 5.6E+07 Te-129m 7.0E+10 3.1E+08 Y-90 9.4E+10 4.2E+08 Y-91 3.4E+10 1.5E+08 Zn-65 3.5E+12 1.5E+10 Zr-93 7.6E+06 3.0E+00 8.8E+07 Zr-95 2.6E+12 1.6E+10 2.7E+09 1.7E+07 1.4E+11 6.0E+08 Others 1.1E+12 7.2E+09 2.7E+04 3.3E+03 2.8E+05 Total 1.5E+13 9.4E+10 3.8E+13 2.5E+11 8.8E+13 3.9E+11 2.4E+11 1.7E+11 1.0E+12
Volume (m3/yr) 164 154 225 315 300 15 per unit
A-1 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
TABLE A-2: TYPICAL EPR L&ILW SOURCE TERM DETAILS
Bq/m3 LLW ILW Demin / COMBINED Non- AVERAGE Evaporator Storage tank AVERAGE AVERAGE Primary Radwaste AVERAGE Nuclide Incinerable Compacible Centrifuge Mixed waste Filters L&ILW processible DAW concentrate sludge Sludge LLW Coolant IX Demin IX ILW sludge Ag-108m 1.8E+07 1.9E+07 4.0E+06 8.6E+07 3.6E+06 3.4E+06 Ag-110m 6.3E+07 2.9E+07 2.4E+07 5.6E+07 2.1E+10 2.1E+10 1.1E+10 2.1E+10 1.4E+08 2.2E+09 1.4E+12 2.1E+10 3.7E+11 1.8E+10 Ba-137m 8.8E+10 8.8E+10 4.4E+10 8.7E+10 9.1E+09 4.1E+12 8.8E+10 1.1E+12 5.7E+10 Ba-140 4.4E+07 1.1E+07 1.1E+07 3.7E+07 5.4E+07 3.3E+07 3.2E+07 Be-7 3.9E+06 3.9E+06 8.5E+05 1.8E+07 7.7E+05 7.3E+05 C-14 5.5E+07 4.2E+07 5.0E+07 5.2E+07 2.0E+08 4.7E+07 4.5E+07 Ce-141 1.0E+09 2.7E+08 1.2E+07 Ce-144 1.7E+07 1.7E+07 3.8E+06 1.4E+08 1.4E+08 6.9E+07 1.4E+08 8.1E+07 1.8E+07 9.5E+09 1.4E+08 2.5E+09 1.3E+08 Co-58 1.1E+09 7.4E+08 7.3E+08 1.0E+09 1.6E+10 1.6E+10 7.8E+09 1.5E+10 3.5E+09 2.5E+09 6.2E+11 5.9E+11 1.6E+10 3.7E+11 1.9E+10 Co-60 1.2E+09 7.7E+08 8.0E+08 1.1E+09 1.5E+10 1.5E+10 7.6E+09 1.5E+10 3.6E+09 2.5E+09 1.4E+12 1.3E+12 1.5E+10 8.3E+11 3.9E+10 Cr-51 8.3E+07 1.3E+08 1.3E+08 9.4E+07 4.0E+08 4.0E+08 2.0E+08 3.9E+08 6.3E+08 1.3E+08 1.1E+10 1.0E+10 4.0E+08 6.4E+09 4.0E+08 Cs-134 1.8E+08 1.9E+08 3.9E+07 5.1E+10 5.1E+10 2.5E+10 5.0E+10 8.4E+08 5.3E+09 2.1E+12 5.1E+10 5.7E+11 3.0E+10 Cs-137 5.9E+08 6.0E+08 1.3E+08 8.8E+10 8.8E+10 4.4E+10 8.7E+10 2.8E+09 9.3E+09 4.1E+12 8.8E+10 1.1E+12 5.7E+10 Fe-55 3.9E+09 2.0E+09 1.3E+09 3.5E+09 3.1E+10 3.1E+10 1.6E+10 3.1E+10 9.2E+09 6.3E+09 2.7E+12 2.6E+12 3.1E+10 1.6E+12 7.8E+10 Fe-59 7.7E+06 7.7E+06 1.7E+06 3.0E+08 3.0E+08 1.5E+08 2.9E+08 3.6E+07 3.2E+07 9.5E+09 9.1E+09 3.0E+08 5.7E+09 2.8E+08 H-3 7.5E+07 7.5E+07 1.6E+07 3.5E+08 1.5E+07 1.4E+07 I-129 7.9E+05 7.8E+05 1.7E+05 3.7E+06 1.6E+05 1.5E+05 I-131 8.0E+06 8.0E+06 1.8E+06 3.8E+07 1.6E+06 1.5E+06 La-140 4.4E+07 1.1E+07 1.3E+07 3.7E+07 5.3E+07 3.3E+07 3.2E+07 Mn-54 2.1E+08 5.4E+07 5.6E+07 1.8E+08 2.9E+10 2.9E+10 1.5E+10 2.9E+10 2.5E+08 3.2E+09 2.1E+12 2.0E+12 2.9E+10 1.3E+12 5.9E+10 Nb-95 1.2E+08 1.2E+08 1.3E+08 1.2E+08 2.1E+09 2.1E+09 1.1E+09 2.1E+09 5.8E+08 3.3E+08 7.9E+10 2.1E+09 2.1E+10 1.3E+09 Ni-59 1.0E+07 9.3E+06 2.2E+06 4.8E+07 2.0E+06 1.9E+06 Ni-63 9.5E+08 5.2E+08 4.7E+08 8.6E+08 2.5E+09 7.7E+08 7.3E+08 Pr-144 1.4E+08 1.4E+08 6.9E+07 1.4E+08 1.4E+07 9.5E+09 1.4E+08 2.5E+09 1.2E+08 Pu-241 7.9E+07 1.8E+07 1.5E+07 6.5E+07 8.4E+07 5.8E+07 5.6E+07 Rh-103m 4.2E+09 4.2E+09 2.1E+09 4.1E+09 4.3E+08 1.2E+11 4.2E+09 3.3E+10 1.9E+09 Rh-106 8.4E+08 8.4E+08 4.2E+08 8.3E+08 8.8E+07 6.2E+10 8.4E+08 1.6E+10 8.0E+08 Ru-103 4.0E+09 4.6E+09 2.3E+09 4.0E+09 4.2E+08 1.3E+11 4.6E+09 3.6E+10 2.0E+09 Ru-106 3.9E+07 1.1E+07 1.9E+06 3.3E+07 3.6E+09 3.6E+09 1.8E+09 3.6E+09 5.2E+07 4.0E+08 2.6E+11 3.6E+09 6.9E+10 3.4E+09 Sb-124 1.0E+07 9.3E+06 2.2E+06 4.7E+07 2.0E+06 1.9E+06 Sb-125 3.3E+07 3.4E+07 7.2E+06 1.5E+08 6.5E+06 6.2E+06 Sn-113 1.4E+06 1.3E+06 3.0E+05 6.5E+06 2.7E+05 2.6E+05 Sr-89 1.2E+07 1.1E+07 2.5E+06 1.5E+09 1.5E+09 7.3E+08 1.4E+09 5.5E+07 1.5E+08 4.6E+10 1.5E+09 1.3E+10 7.0E+08 Sr-90 7.7E+07 7.7E+07 1.7E+07 3.6E+08 3.6E+08 1.8E+08 3.6E+08 3.6E+08 5.3E+07 3.3E+10 3.6E+08 8.7E+09 4.3E+08 Te-127m 5.1E+09 5.1E+09 2.5E+09 5.0E+09 5.3E+08 2.4E+11 5.1E+09 6.4E+10 3.3E+09 Te-129 4.8E+08 4.8E+08 2.4E+08 4.7E+08 5.0E+07 4.8E+08 1.9E+08 5.6E+07 Te-129m 7.3E+08 7.3E+08 3.7E+08 7.3E+08 7.6E+07 2.0E+10 7.3E+08 5.4E+09 3.1E+08 Y-90 3.6E+08 3.6E+08 1.8E+08 3.6E+08 3.8E+07 3.3E+10 3.6E+08 8.7E+09 4.2E+08 Y-91 3.0E+08 3.0E+08 1.5E+08 3.0E+08 3.1E+07 1.0E+10 3.0E+08 2.8E+09 1.5E+08 Zn-65 1.7E+07 1.7E+07 3.6E+06 8.1E+09 8.1E+09 4.1E+09 8.0E+09 7.8E+07 8.5E+08 5.6E+11 5.3E+11 8.1E+09 3.3E+11 1.5E+10 Zr-95 4.4E+07 6.5E+07 6.4E+07 4.9E+07 1.1E+09 1.1E+09 5.3E+08 1.1E+09 3.1E+08 1.5E+08 3.8E+10 1.1E+09 1.0E+10 6.0E+08 Total 7.9E+09 5.6E+09 4.9E+09 7.4E+09 3.7E+11 3.7E+11 1.9E+11 3.7E+11 2.6E+10 4.5E+10 7.5E+12 2.0E+13 3.7E+11 7.9E+12 3.9E+11
Note: Extracted from [Areva 2007] and converted to Bq/m3
A-2 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
TABLE A-3: DETAILS OF PROCESSED L&ILW VOLUMES L&ILW Operational Wastes (per reactor unit)
Annual As generated As stored (@ DWMF) As stored (@ WWMF) ACRAP1000 USEPR ACR AP1000 USEPR ACR AP1000 USEPR Waste type m3/yr ft3/yr m3/yr ft3/yr m3/yr m3/yr ft3/yr m3/yr ft3/yr m3/yr m3/yr ft3/yr m3/yr ft3/yr m3/yr LLW Incinerable 111.2 3,750 106.1 5,300 150.0 27.8 938 26.5 1,325 37.5 2.8 94 2.7 133 3.7 Compactible 29.6 1,000 28.3 1,415 40.0 7.4 250 7.1 354 10.0 7.4 250 7.1 354 10.0 Non-processible 14.4 239 6.8 70 2.0 18.0 299 8.5 88 2.5 18.0 299 8.5 88 2.5 Sludge 37 1.0 796 22.5 37 1.0 796 22.5 37 1.0 796 22.5 Misc 2 0.1 2 0.1 2 0.1 TOTAL 155.2 5,026 142.2 7,583 214.6 53.2 1,523 43.1 2,564 72.6 28.2 680 19.2 1,372 38.8
ILW IX resins 6.6 400 11.3 230 6.5 6.6 400 11.3 230 6.5 6.6 400 11.3 230 6.5 Filters 2.6 5 0.1 120 3.4 2.6 5 0.1 120 3.4 2.6 5 0.1 120 3.4 Misc TOTAL 9.1 405 11.5 350 9.9 9.1 405 11.5 350 9.9 9.1 405 11.5 350 9.9
GRAND TOTAL L&ILW 164 5,431 154 7,933 225 62 1,928 55 2,914 82 37 1,085 31 1,722 49
DWMF includes compaction @ 4:1 WWMF includes compaction @ 4:1 plus incineration at 40:1
Lifetime As generated As stored (@ DWMF) As stored (@ WWMF) ACRAP1000 USEPR ACR AP1000 USEPR ACR AP1000 USEPR Waste type m3 ft3 m3 ft3 m3 m3 ft3 m3 ft3 m3 m3 ft3 m3 ft3 m3 LLW Incinerable 6,669 225,000 6,368 318,000 8,999 1,667 56,250 1,592 79,500 2,250 167 5,625 159 7,950 225 Compactible 1,779 60,000 1,698 84,900 2,403 445 15,000 425 21,225 601 445 15,000 425 21,225 601 Non-processible 864 14,340 406 4,200 119 1,080 17,925 507 5,250 149 1,080 17,925 507 5,250 149 Sludge 2,220 63 47,760 1,352 2,220 63 47,760 1,352 2,220 63 47,760 1,352 Misc 120 3 120 3 120 3 TOTAL 9,312 301,560 8,534 454,980 12,876 3,192 91,395 2,586 153,855 4,354 1,691 40,770 1,154 82,305 2,329
ILW IX resins 393 24,000 679 13,800 391 393 24,000 679 13,800 391 393 24,000 679 13,800 391 Filters 153 312 9 7,200 204 153 312 9 7,200 204 153 312 9 7,200 204 Misc TOTAL 546 24,312 688 21,000 594 546 24,312 688 21,000 594 546 24,312 688 21,000 594
GRAND TOTAL L&ILW 9,858 325,872 9,222 475,980 13,470 3,738 115,707 3,275 174,855 4,948 2,237 65,082 1,842 103,305 2,924
LLSB volume = 7000 m3 # LLSBs required 1 1 1 1 1 1 # reactor units 4 4 3 4 4 3 Total lifetime volumes (m3) 14,952 13,098 14,845 8,949 7,367 8,771 Total # LLSBs for all reactor units 3 2 3 2 2 2
A-3 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
TABLE A-4: ANNUAL WASTE FORECASTS
2018 2019 2020 2021 2022 2023 2024 2025 2026 2027 2028 2029 2030 2031 2032 2033 2034 2035 2036 2037 2038 2039 2040 2041 2042 2043 ACR 1 unit 2 units 3 units 4 units L&ILW LLW produced (m3) 155 310 310 310 310 310 466 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 LLW stored on site (m3) 53 106 106 106 106 106 160 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 LLW stored off-site (m3) 28 56 56 56 56 56 85 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 ILW produced (m3)918181818182736363636363636363636363636363636363636 ILW stored (m3)918181818182736363636363636363636363636363636363636 Cummulative Total on-site storage (m3) 62 187 312 436 561 685 872 1,121 1,371 1,620 1,869 2,118 2,367 2,617 2,866 3,115 3,364 3,613 3,863 4,112 4,361 4,610 4,859 5,109 5,358 5,607 On-site storage expansions 7,000 On-site storage capacity 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 Cummulative total off-site storage (m3) 37 112 186 261 336 410 522 671 820 970 1,119 1,268 1,417 1,566 1,715 1,864 2,014 2,163 2,312 2,461 2,610 2,759 2,909 3,058 3,207 3,356 Used Fuel Used fuel produced (bundles) 4,415 8,830 8,830 8,830 8,830 8,830 13,244 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 Used fuel dry stored (bundles) 4,415 8,830 8,830 8,830 8,830 8,830 13,244 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 Cummulative dry storage (bundles) 0 0 0 0 0 0 0 0 0 0 4,415 13,244 22,074 30,904 39,733 48,563 61,808 79,467 97,126 114,786 132,445 150,104 167,763 185,423 203,082 220,741 Cummulative dry storage (DSCs) 0 0 0 0 0 0 0 0 0 0 11 34 57 80 103 126 161 207 253 299 345 391 437 483 529 575 DSC Dry storage expansions (DSCs) 500 500 DSC Dry storage capacity (DSCs) 500 500 500 500 500 500 500 500 500 500 500 500 1,000 1,000 1,000 1,000 MACSTOR expansions (bundles) 144,000 144,000 MACSTOR Capacity (bundles) 144,000 144,000 144,000 144,000 144,000 144,000 144,000 144,000 144,000 144,000 288,000 288,000 288,000 288,000 288,000 288,000 Refurbishment waste LLW ILW AP1000 1 unit 2 units 3 units 4 units L&ILW LLW produced (m3) 142 284 284 284 284 284 427 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 LLW stored on site (m3) 43 86 86 86 86 86 129 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 LLW stored off-site (m3) 19 38 38 38 38 38 58 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 ILW produced (m3) 11 23 23 23 23 23 34 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 ILW stored (m3) 11 23 23 23 23 23 34 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 Cummulative Total on-site storage (m3) 55 164 273 382 491 600 764 982 1,201 1,419 1,637 1,856 2,074 2,292 2,510 2,729 2,947 3,165 3,384 3,602 3,820 4,039 4,257 4,475 4,693 4,912 On-site storage expansions 7,000 On-site storage capacity 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 Cummulative total off-site storage (m3) 31 92 153 215 276 338 430 553 675 798 921 1,044 1,166 1,289 1,412 1,535 1,658 1,780 1,903 2,026 2,149 2,272 2,394 2,517 2,640 2,763 Used Fuel Used fuel produced (assemblies) 43 87 87 87 87 87 130 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 Used fuel dry stored (assemblies) 43 87 87 87 87 87 130 173 173 173 173 173 173 173 173 173 Cummulative dry storage (assemblies) 0 0 0 0 0 0 0 0 0 0 43 130 217 304 390 477 607 781 954 1,128 1,301 1,475 1,648 1,822 1,995 2,169 Cummulative dry storage (casks - 32) 0 0 0 0 0 0 0 0 0 0 1 4 7 9 12 15 19 24 30 35 41 46 52 57 62 68 Dry storage expansions 300 Dry storage capacity 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 Nuhoms expansions (canisters) 48 48 Nuhoms Capacity (canisters) 48 48 48 48 48 48 48 48 48 48 96 96 96 96 96 96 Refurbishment waste LLW ILW USEPR 1 unit 2 units 3 units L&ILW LLW produced (m3) 215 215 429 429 429 429 429 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 LLW stored on site (m3) 73 73 145 145 145 145 145 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 LLW stored off-site (m3) 39 39 78 78 78 78 78 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 ILW produced (m3) 10 10 20 20 20 20 20 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 ILW stored (m3) 10 10 20 20 20 20 20 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 Cummulative Total on-site storage (m3) 82 165 330 495 660 825 990 1,237 1,485 1,732 1,979 2,227 2,474 2,722 2,969 3,216 3,464 3,711 3,959 4,206 4,454 4,701 4,948 5,196 5,443 5,691 On-site storage expansions 7,000 On-site storage capacity 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 7,000 Cummulative total off-site storage (m3) 49 97 195 292 390 487 585 731 877 1,023 1,169 1,316 1,462 1,608 1,754 1,900 2,046 2,193 2,339 2,485 2,631 2,777 2,924 3,070 3,216 3,362 Used Fuel Used fuel produced (assemblies) 84 84 169 169 169 169 169 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 Used fuel dry stored (assemblies) 84 84 169 169 169 169 169 253 253 253 253 253 253 253 253 253 Cummulative dry storage (assemblies) 0 0 0 0 0 0 0 0 0 0 84 169 338 507 676 845 1,014 1,267 1,521 1,774 2,028 2,281 2,535 2,788 3,041 3,295 Cummulative dry storage (casks - 32) 0 0 0 0 0 0 0 0 0 0 3 5 11 16 21 26 32 40 48 55 63 71 79 87 95 103 Dry storage expansions 300 Dry storage capacity 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 Nuhoms expansions (canisters) 48 48 48 Nuhoms Capacity (canisters) 48 48 48 48 48 48 48 48 96 96 96 96 96 144 144 144 Refurbishment waste LLW ILW
A-4 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
2044 2045 2046 2047 2048 2049 2050 2051 2052 2053 2054 2055 2056 2057 2058 2059 2060 2061 2062 2063 2064 2065 2066 2067 2068 2069 U1 + U2 U2 + U4 ACR U1 refurb U2 refurb U3 refurb U4 refurb refurb refurb L&ILW LLW produced (m3) 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 621 LLW stored on site (m3) 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 213 LLW stored off-site (m3) 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 113 ILW produced (m3) 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 ILW stored (m3) 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 Cummulative Total on-site storage (m3) 5,856 6,105 6,355 6,604 6,853 7,102 7,351 7,601 7,850 8,099 8,348 8,597 8,847 9,096 9,345 9,594 9,843 10,093 10,342 10,591 10,840 11,089 11,339 11,588 11,837 12,086 On-site storage expansions 7,000 On-site storage capacity 7,000 7,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 Cummulative total off-site storage (m3) 3,505 3,654 3,804 3,953 4,102 4,251 4,400 4,549 4,698 4,848 4,997 5,146 5,295 5,444 5,593 5,743 5,892 6,041 6,190 6,339 6,488 6,638 6,787 6,936 7,085 7,234 Used Fuel Used fuel produced (bundles) 17,659 17,659 17,659 17,659 19,484 15,070 13,244 17,659 17,659 17,659 19,484 15,070 13,244 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 Used fuel dry stored (bundles) 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 19,484 15,070 13,244 8,830 Cummulative dry storage (bundles) 238,401 256,060 273,719 291,379 309,038 326,697 344,357 362,016 379,675 397,334 414,994 432,653 450,312 467,972 487,456 502,526 515,770 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 Cummulative dry storage (DSCs) 621 667 713 759 805 851 897 943 989 1,035 1,081 1,127 1,173 1,219 1,269 1,309 1,343 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 DSC Dry storage expansions (DSCs) 500 DSC Dry storage capacity (DSCs) 1,000 1,000 1,000 1,000 1,000 1,000 1,000 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 MACSTOR expansions (bundles) 144,000 144,000 MACSTOR Capacity (bundles) 288,000 288,000 432,000 432,000 432,000 432,000 432,000 432,000 432,000 432,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 Refurbishment waste LLW 2,000 3,600 1,600 2,000 3,600 1,600 ILW 1,000 1,000 1,000 1,000 AP1000 U1 refurb U2 refurb U3 refurb U4 refurb L&ILW LLW produced (m3) 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 569 LLW stored on site (m3) 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 172 LLW stored off-site (m3) 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 77 ILW produced (m3) 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 ILW stored (m3) 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 46 Cummulative Total on-site storage (m3) 5,130 5,348 5,567 5,785 6,003 6,222 6,440 6,658 6,876 7,095 7,313 7,531 7,750 7,968 8,186 8,405 8,623 8,841 9,059 9,278 9,496 9,714 9,933 10,151 10,369 10,588 On-site storage expansions 7,000 On-site storage capacity 7,000 7,000 7,000 7,000 7,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 Cummulative total off-site storage (m3) 2,886 3,008 3,131 3,254 3,377 3,499 3,622 3,745 3,868 3,991 4,113 4,236 4,359 4,482 4,605 4,727 4,850 4,973 5,096 5,218 5,341 5,464 5,587 5,710 5,832 5,955 Used Fuel Used fuel produced (assemblies) 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 Used fuel dry stored (assemblies) 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 173 86 Cummulative dry storage (assemblies) 2,342 2,515 2,689 2,862 3,036 3,209 3,383 3,556 3,730 3,903 4,077 4,250 4,424 4,597 4,771 4,944 5,118 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 Cummulative dry storage (casks - 32) 73 79 84 89 95 100 106 111 117 122 127 133 138 144 149 155 160 163 163 163 163 163 163 163 163 163 Dry storage expansions Dry storage capacity 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 Nuhoms expansions (canisters) 48 48 Nuhoms Capacity (canisters) 96 96 96 144 144 144 144 144 144 144 144 144 192 192 192 192 192 192 192 192 192 192 192 192 192 192 Refurbishment waste LLW 1,600 1,600 1,600 1,600 ILW 37 37 37 37 USEPR U1 refurb U2 refurb U3 refurb L&ILW LLW produced (m3) 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 644 LLW stored on site (m3) 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 218 LLW stored off-site (m3) 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 116 ILW produced (m3) 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 ILW stored (m3) 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 Cummulative Total on-site storage (m3) 5,938 6,185 6,433 6,680 6,928 7,175 7,423 7,670 7,917 8,165 8,412 8,660 8,907 9,155 9,402 9,649 9,897 10,144 10,392 10,639 10,886 11,134 11,381 11,629 11,876 12,124 On-site storage expansions 7,000 On-site storage capacity 7,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 Cummulative total off-site storage (m3) 3,508 3,654 3,801 3,947 4,093 4,239 4,385 4,531 4,678 4,824 4,970 5,116 5,262 5,409 5,555 5,701 5,847 5,993 6,139 6,286 6,432 6,578 6,724 6,870 7,016 7,163 Used Fuel Used fuel produced (assemblies) 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 Used fuel dry stored (assemblies) 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 253 Cummulative dry storage (assemblies) 3,548 3,802 4,055 4,309 4,562 4,816 5,069 5,323 5,576 5,830 6,083 6,336 6,590 6,843 7,097 7,350 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 Cummulative dry storage (casks - 32) 111 119 127 135 143 150 158 166 174 182 190 198 206 214 222 230 238 238 238 238 238 238 238 238 238 238 Dry storage expansions Dry storage capacity 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 Nuhoms expansions (canisters) 48 48 Nuhoms Capacity (canisters) 144 144 144 192 192 192 192 192 192 240 240 240 240 240 240 240 240 240 240 240 240 240 240 240 240 240 Refurbishment waste LLW 2,565 2,565 2,565 ILW 85 85 85
A-5 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
2070 2071 2072 2073 2074 2075 2076 2077 2078 2079 2080 2081 2082 2083 2084 2085 2086 2087 2088 2089 2090 TOTAL ACR U1 SD U2 SD U3 SD U4 SD L&ILW LLW produced (m3) 621 621 621 621 621 621 621 621 466 310 310 310 310 310 155 0 0 0 0 0 0 37,248 LLW stored on site (m3) 213 213 213 213 213 213 213 213 160 106 106 106 106 106 53 0 0 0 0 0 0 12,768 LLW stored off-site (m3) 113 113 113 113 113 113 113 113 85 56 56 56 56 56 28 0 0 0 0 0 0 6,765 ILW produced (m3) 36 36 36 36 36 36 36 36 27 18 18 18 18 18 9 0 0 0 0 0 0 2,184 ILW stored (m3) 36 36 36 36 36 36 36 36 27 18 18 18 18 18 9 0 0 0 0 0 0 2,184 Cummulative Total on-site storage (m3) 12,335 12,585 12,834 13,083 13,332 13,581 13,831 14,080 14,267 14,391 14,516 14,641 14,765 14,890 14,952 14,952 14,952 14,952 14,952 14,952 14,952 On-site storage expansions 7,000 On-site storage capacity 14,000 14,000 14,000 14,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 Cummulative total off-site storage (m3) 7,383 7,532 7,682 7,831 7,980 8,129 8,278 8,427 8,539 8,614 8,688 8,763 8,838 8,912 8,949 8,949 8,949 8,949 8,949 8,949 8,949 Used Fuel Used fuel produced (bundles) 17,659 17,659 17,659 17,659 17,659 17,659 17,659 17,659 13,244 8,830 8,830 8,830 8,830 8,830 4,415 0 0 0 0 0 0 1,049,200 Used fuel dry stored (bundles) 524,600 Cummulative dry storage (bundles) 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 524,600 Cummulative dry storage (DSCs) 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 1,366 DSC Dry storage expansions (DSCs) DSC Dry storage capacity (DSCs) 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 1,500 MACSTOR expansions (bundles) MACSTOR Capacity (bundles) 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 576,000 Refurbishment waste LLW 14,400 ILW 4,000 AP1000 U1 SD U2 SD U3 SD U4 SD L&ILW LLW produced (m3) 569 569 569 569 569 569 569 569 427 284 284 284 284 284 142 0 0 0 0 0 0 34,137 LLW stored on site (m3) 172 172 172 172 172 172 172 172 129 86 86 86 86 86 43 0 0 0 0 0 0 10,346 LLW stored off-site (m3) 77 77 77 77 77 77 77 77 58 38 38 38 38 38 19 0 0 0 0 0 0 4,615 ILW produced (m3) 46 46 46 46 46 46 46 46 34 23 23 23 23 23 11 0 0 0 0 0 0 2,752 ILW stored (m3) 46 46 46 46 46 46 46 46 34 23 23 23 23 23 11 0 0 0 0 0 0 2,752 Cummulative Total on-site storage (m3) 10,806 11,024 11,242 11,461 11,679 11,897 12,116 12,334 12,498 12,607 12,716 12,825 12,934 13,043 13,098 13,098 13,098 13,098 13,098 13,098 13,098 On-site storage expansions On-site storage capacity 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 14,000 Cummulative total off-site storage (m3) 6,078 6,201 6,324 6,446 6,569 6,692 6,815 6,938 7,030 7,091 7,152 7,214 7,275 7,337 7,367 7,367 7,367 7,367 7,367 7,367 7,367 Used Fuel Used fuel produced (assemblies) 173 173 173 173 173 173 173 173 130 87 87 87 87 87 43 0 0 0 0 0 0 10,409 Used fuel dry stored (assemblies) 5,204 Cummulative dry storage (assemblies) 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 5,204 Cummulative dry storage (casks - 32) 163 163 163 163 163 163 163 163 163 163 163 163 163 163 163 163 163 163 163 163 163 Dry storage expansions Dry storage capacity 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 Nuhoms expansions (canisters) Nuhoms Capacity (canisters) 192 192 192 192 192 192 192 192 192 192 192 192 192 192 192 192 192 192 192 192 192 Refurbishment waste LLW 6,400 ILW 148 USEPR U1 SD U2 SD U3 SD L&ILW LLW produced (m3) 644 644 644 644 644 644 644 644 429 429 215 215 215 215 215 0 0 0 0 0 0 38,628 LLW stored on site (m3) 218 218 218 218 218 218 218 218 145 145 73 73 73 73 73 0 0 0 0 0 0 13,062 LLW stored off-site (m3) 116 116 116 116 116 116 116 116 78 78 39 39 39 39 39 0 0 0 0 0 0 6,988 ILW produced (m3) 30 30 30 30 30 30 30 30 20 20 10 10 10 10 10 0 0 0 0 0 0 1,783 ILW stored (m3) 30 30 30 30 30 30 30 30 20 20 10 10 10 10 10 0 0 0 0 0 0 1,783 Cummulative Total on-site storage (m3) 12,371 12,618 12,866 13,113 13,361 13,608 13,856 14,103 14,268 14,433 14,515 14,598 14,680 14,763 14,845 14,845 14,845 14,845 14,845 14,845 14,845 On-site storage expansions 7,000 On-site storage capacity 14,000 14,000 14,000 14,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 21,000 Cummulative total off-site storage (m3) 7,309 7,455 7,601 7,747 7,894 8,040 8,186 8,332 8,430 8,527 8,576 8,624 8,673 8,722 8,771 8,771 8,771 8,771 8,771 8,771 8,771 Used Fuel Used fuel produced (assemblies) 253 253 253 253 253 253 253 253 169 169 84 84 84 84 84 0 0 0 0 0 0 15,207 Used fuel dry stored (assemblies) 7,604 Cummulative dry storage (assemblies) 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 7,604 Cummulative dry storage (casks - 32) 238 238 238 238 238 238 238 238 238 238 238 238 238 238 238 238 238 238 238 238 238 Dry storage expansions Dry storage capacity 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 300 Nuhoms expansions (canisters) Nuhoms Capacity (canisters) 240 240 240 240 240 240 240 240 240 240 240 240 240 240 240 240 240 240 240 240 240 Refurbishment waste LLW 7,695 ILW 255 Note: Colours indicate special dates such as start of a unit, new storage buildings going into service, or unit refurbishment. A-6 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
APPENDIX B
REFURBISHMENT WASTE DETAILS
New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
TABLE B-1: SUMMARY OF DARLINGTON FUEL CHANNEL COMPONENT SPECIFIC ACTIVITY8
NUCLIDE TBq/Kg (AFTER 270 DAYS OF DECAY) Pressure Tubes Calandria Tubes End fittings Shield Plugs Calandria Tube Inserts C-14 1.32E-03 2.19E-04 4.42E-05 1.01E-08 2.37E-04 Cr-51 1.32E-05 9.92E-05 1.37E-03 3.42E-07 6.52E-03 Mn-54 2.52E-04 2.32E-04 5.04E-02 4.43E-03 1.84E-01 Fe-55 5.76E-02 7.10E-02 2.21E+00 1.67E-01 1.15E+01 Fe-59 3.11E-04 4.18E-04 1.24E-03 1.04E-04 7.22E-03 Co-60 8.28E-03 1.24E-02 2.23E-01 9.24E-03 4.67E-01 Ni-59 5.97E-07 1.47E-05 2.24E-05 8.32E-05 Ni-63 2.37E-04 5.82E-03 2.71E-03 1.36E-14 1.28E-02 Nb-94 1.06E-02 1.63E-06 5.05E-07 1.65E-06 Nb-95 1.23E+00 1.17E+00 6.80E-07 4.12E-06 Nb-95m 6.27E-03 6.58E-03 2.63E-09 9.96E-09 Zr-93 3.47E-04 3.68E-04 1.27E-10 5.58E-10 Zr-95 5.33E-01 5.59E-01 2.24E-07 8.47E-07 In-113m 2.82E-05 2.56E-02 4.63E-05 1.35E-04 In-114 2.93E-04 1.89E-03 5.68E-07 3.56E-06 Sn-113 2.82E-05 2.56E-02 4.63E-05 1.35E-04 Sn-119m 6.71E-04 3.45E-01 7.13E-04 1.87E-03 Sn-123 1.57E-05 1.46E-02 1.03E-05 4.28E-05 Sb-125 1.59E-04 1.37E-01 2.57E-04 7.12E-04 Te-123m 1.22E-04 9.62E-03 8.44E-06 6.43E-05 Te-125m 3.91E-05 3.35E-02 6.28E-05 1.74E-04 All Others 7.96E-03 1.13E-02 5.02E-04 1.34E-05 7.00E-03 TOTAL (TBq/kg) 1.86E+00 2.43E+00 2.49E+00 1.81E-01 1.22E+01 Material Zr-2.5% Nb Zircalloy-2 SS403 A295 ductile iron SS410
8 Extracted from [AMEC, 2008a] B-1
New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
TABLE B-2: SUMMARY OF STEAM GENERATOR ACTIVITY
NUCLIDE TOTAL PER SG (TBq) Ringhals 3 (Sweden)9 Pickering B H-3 – 1.2E+00 C-14 9.0E-04 – Mn-54 – 4.7E-03 Fe-55 – 1.2E+00 Fe-59 – 9.3E-03 Co-60 7.4E-01 1.1E-01 Ni-59 3.5E-03 – Ni-63 3.3E-01 – Zn-65 – 8.0E-04 Sr-90 3.6E-03 – Zr-95 – 3.3E-02 Nb-94 – 7.3E-04 Nb-95 – 8.3E-02 Ru-103 – 2.0E-02 Ru-106 – 9.4E-03 Sn-113 – 2.1E-03 Sb-124 – 3.2E-02 Sb-125 – 1.1E-02 Ce-141 – 9.8E-03 Ce-144 – 9.2E-03 Am-241 1.9E-06 3.6E-04 Cm-242 8.1E-13 – Cm-244 2.4E-06 9.6E-05 Pu-238 1.3E-05 9.6E-05 Pu-239/240 1.2E-05 5.5E-04 Pu-241 5.6E-04 – Total 1.1E+00 2.7E+00
9 Extracted from [Vattenfall 2007] B-2
New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
APPENDIX C
NEW NUCLEAR - DARLINGTON - BASIS FOR EA New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
TABLE C-1: NEW NUCLEAR – DARLINGTON – BASIS FOR EA
Project Phase / Works Description and Activities Site Preparation Phase Mobilization and Mobilization (construction workforce and equipment): will involve mobilization of equipment and the construction workforce to the site. The Preparatory Works physical aspects of mobilization will involve the establishment of parking areas for staff and equipment, service areas for construction offices, construction phase fencing for security and safety and equipment storage; security/guardhouse and reception facilities. Clearing and Grubbing: Vegetation within areas of future construction will be removed. A variety of methods including the removal of trees by truck, chipping of smaller vegetation and grubbing with a dozer or excavator will be used to remove vegetation. Environmental effects management measures will be applied throughout the activity such as minimizing the area to be cleared to the extent feasible and complying with seasonal constraints and regulatory requirements for clearing operations. Installation of Services and Utilities: includes temporary services and utilities required during construction and permanent services and utilities required to support operations. Wherever possible, utilities and services will be installed to accommodate the needs of both construction and operation phases. Utilities and services will include: i) potable water; ii) sanitary sewage collection discharging to a municipal water pollution control plant; iii) electrical and telephone service; iv) P.A. system; v) fencing. Excavation to install services is captured by other earthmoving activities. Development of Roads and Related Infrastructure: includes improvements to access into the site and features to provide for temporary (i.e., during construction) and permanent (during operations) access, egress and parking. Onsite roads and infrastructure will include local access roads and parking facilities within the site to accommodate workforce-related and other traffic during both construction and operation phases. For EA purposes, it is assumed that off-site parking facilities may be used with workers transferred to the NND via shuttle bus. Excavation and Grading Excavation and grading will comprise all earth and rock-moving activities including earthmoving and grading, drilling and blasting. Excavation activities will be conducted in-the-dry with dewatering where required. Collected water will be managed and discharged as described in Management of Stormwater. On-Land Earthmoving and Grading: During site preparation activities, effectively all land area east of Holt Road will be disturbed to a large extent. Topsoil stripping will be by means of suitable earthmoving equipment (e.g., scrapers, excavators and trucks). Excavated soils transferred to the Northeast and Northwest Landfill Areas and lake infill will be placed using good management practices that address surface erosion, dust control and related aspects including noise and vehicle emissions. Transport of Surplus Soil to Off-site Disposal: Should it be necessary to do so, surplus soil will be transported to disposal at an off-site location(s). The destinations for this material have not been determined, however, it is intended that the material be used to rehabilitate extraction pits and quarries or other development sites, or similar beneficial use. Rock Excavation and Grading (Drilling, Blasting, Boring): will involve the excavation and grading of rock and like material, and associated activities such as drilling or blasting to facilitate its excavation and transfer to rock fill areas (i.e. lake infill) or disposal areas. Development of Construction Laydown Areas: will include specific areas identified for, and developed as, staging areas for contractor operations and storage areas for construction equipment and materials. Laydown areas will be graded, temporarily fenced, and surfaced, depending on function, with granular or asphalt.
C-1 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
Project Phase / Works Description and Activities Marine and Shoreline Marine and Shoreline Works includes all works and activities conducted within or adjacent to Lake Ontario such that they are likely to interact Works with the marine and aquatic environment. Marine and shoreline-related works and activities will include the following: Lake Infilling and Shoreline Protection: will occur throughout an area of Lake Ontario and will extend from the easterly limit of the DN site to approximately the DNGS intake channel; and about 100 m into the lake on its westerly limit to approximately 450 m on its easterly limit. Lake infilling will create a new landform of up to approximately 40 ha. The lake infill operation will begin with the construction of a low-permeability coffer dam on its outer perimeter to contain the deposit lake infill materials and isolate the area from lake water intrusion. The core would typically consist of low-permeability soils or compacted granular materials, driven or vibrated steel sheeting, or drilled caissons. The lake-facing surface of the dam will be covered with armour stone placed by crane on the lake side of the dam. Any fish within the area to be dammed will be directed out of the work area by progressive seining and other appropriate means as the dam is placed. Once the cofferdam is complete, the water contained within it will be pumped out and discharged to Lake Ontario. The material placed within the cofferdam to create the new landform will originate on-site and be placed as part of the Excavation and Grading activity. Construction of Wharf: a wharf will be developed in a portion of the lake infilled area generally in front of the Power Block. The wharf will be used during construction for off-loading oversize and over-weight components and its construction will be appropriate for this purpose. Lake Bottom Dredging: dredging activities are expected to be minimal, but may be required at the point where the cooling water intake tunnel daylights to the lake bottom. Any such minor dredging will involve conventional equipment designed and operated for the purpose (suction and/or mechanical). All dredged sediment will be placed into barges and subsequently off-loaded and disposed of in the Northeast Landfill Area or existing onsite construction landfill. Development of Administration and Support Facilities comprise various buildings housing staff, equipment and operations necessary to provide ongoing support to Administration and the NND. These will include offices, workshops, maintenance, storage and perimeter security buildings, and utilities operating centres. All such Physical Support buildings will consist of conventional steel and masonry structures. Facilities Construction Phase
For assessment purposes, it is assumed that the entire site will be prepared for construction at the outset. Construction of the nuclear power plant elements (i.e., construction phase) will begin as soon as possible into the site preparation activities and accordingly, the site preparation and construction phases will overlap in time. This is a bounding assumption since it represents the greatest amount of related work in the shortest period of time. Construction of Power The Power Block includes the reactor building, the turbine-generator building/turbine hall (powerhouse) and related structural features that are Block physically associated with them. Development of the Power Block includes the installation of all power generation equipment within it, including the reactors, primary and secondary heat transport components, and all powerhouse components including turbines, generators and heat exchangers and pumps and standby power systems. Supply of construction materials and operating equipment to the site is included in the Construction Material and Operating Equipment Supply. Foundations will extend into bedrock and may require drilling and blasting. Some elements of construction will be further supported on steel piles. Above-grade construction will involve techniques typical of heavy industrial development. Placement will involve extensive use of heavy equipment, including heavy-lift fixed and mobile cranes. Installation of operating equipment will involve movement and placement of large and specialty components using various standard and extraordinary procedures, depending on the size and weight of the component.
C-2 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
Project Phase / Works Description and Activities Construction of Intake Intake and Discharge Tunnels and Structures for Once-Through Lake Water Cooling: For EA purposes, the once-through cooling water and Discharge Structures intake and diffuser structures at NND are assumed to be similar to the existing structures at DNGS, although appropriately sized to accommodate the required water flow rates at NND. The tunnels at DNGS were constructed using typical underground mining techniques involving blasting and excavation. Tunnels for once through cooling water at NND may alternatively be constructed by boring using a purpose-built tunnel boring machine (TBM). Intake and Discharge Structures for Cooling Tower Water Makeup and Service Water: Although the water from both mechanical draft and natural draft cooling towers is recirculated, some make-up water is required to replace tower blowdown and other losses (e.g.,evaporation) and for plant service water needs. This water will be drawn from Lake Ontario via intake and discharge pipelines. The open-cut drill-and-blast method is likely to be used to excavate a trench to place the intake or outfall pipe. Pipes will be placed in trenches and backfilled with a granular material, and armour surface protection. Screens may be used to prevent debris from entering the intake structure. Both the intake and discharge structures for makeup water and service water will be substantially smaller than those required for once-through lakewater cooling due to the smaller associated water volumes. Construction of Ancillary Ancillary facilities include all features necessary to support operations of the reactors and generation of electricity, although not physically Facilities associated with the power block. Clearing and grubbing and major earthmoving and grading to accommodate development of the ancillary features are included in the Mobilization and Preparatory Works, and the Earthmoving and Grading activities, respectively. Expansion of Existing Switchyard: will involve the physical enlargement of the footprint of the existing DNGS switchyard, an increase to the electrical capacity to accommodate its use for NND, and its connection to the existing electrical grid. The switchyard expansion will effectively be as an easterly extension to the existing switchyard. Cooling Towers – Mechanical Draft: includes the towers and the associated infrastructure to support their operation. Mechanical draft cooling towers are typically shorter in height and larger in footprint than natural draft cooling towers. Construction of the towers will involve conventional techniques and materials, primarily steel framing, concrete and masonry, and mechanical and electrical components. Cooling Towers – Natural Draft: includes the towers and associated infrastructure to support their operations. Up to two natural draft towers may be constructed for each unit (depending on the design). The towers will have a hyperbolic shape. The towers will be constructed of steel reinforced concrete with structural, mechanical and electrical components and will be erected by means of traditional construction methods (e.g., slip forming, crane lifts), and conventional construction materials. Cooling Towers – Fan Assisted Natural Draft: are not included in any of the three model plant layout scenarios considered in the EA. Because they are a variation of the two cooling tower types that are considered, their potential interfaces with the environment during construction are considered to be bounded by the cooling tower options that are addressed in the EA. Fan assisted natural draft cooling towers have a slightly larger base dimension than the natural draft cooling tower, and have fans placed around the base of the tower to increase the air flow rate. These towers have a similar hyperbolic shape as a traditional natural draft tower, but approximately ѿ the height. Cooling Tower Blowdown Ponds: For each of the cooling tower options one or more blowdown ponds may be required to receive and treat blowdown from the towers. Blowdown is the portion of the circulating water flow that is removed in order to maintain the amount of dissolved solids and other impurities at acceptable levels. The ponds would be excavated into the ground surface and lined (e.g., with clay or synthetic materials) to ensure proper containment. The ponds will be sized to accommodate the required volume for the system, and the water would be appropriately treated to comply with discharge water quality criteria, prior to discharge.
C-3 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
Project Phase / Works Description and Activities Construction of Radioactive Waste Storage Facilities comprise used fuel dry storage facility to house containerized used fuel bundles following their removal from Radioactive Waste wet storage in the used fuel bays. Low and Intermediate Level Waste Storage building(s) may also be required. For EA purposes, it is assumed Storage Facilities that a used fuel dry storage building for NND will not be required until approximately 2025, though a storage building for Low and Intermediate Level Waste will likely be required starting in 2017. Common to Site Preparation and Construction – Works and Activities Management of As the site is developed, ditches and swales will be constructed to collect and convey surface water to stormwater management ponds and Stormwater ultimately to discharge to an existing drainage course or Lake Ontario. Stormwater management features will be developed to address the requirements for runoff control both during site preparation and construction (temporary) and during operations (permanent). Wherever possible, stormwater management features will consider the needs of both construction and operation phases. Supply of Construction Supply of construction materials and operating equipment includes the delivery to the site, of all necessary materials and components for Equipment, Material and construction of NND. While much of the material that will be delivered to the site will be via the road network, large components may be Operating Plant delivered by rail (to an existing rail siding on a neighbouring property and then transported overland to the site or to a new rail siding on the DN Components site), or by barge to the new wharf. Rock Delivery for Cofferdam: delivery of imported rock for cofferdam construction is estimated to be up to 200 trucks per day. Construction Equipment: comprises all mechanized and related equipment required to support construction. Heavy earthmoving equipment will be typical of large-scale construction projects (e.g., trucks, dozers, loaders, excavators, scrappers, graders, compactors). Aggregate and Concrete: For EA purposes, it is assumed that ready-mixed concrete will be provided by an offsite supplier operating on a nearby property, or is mixed on site in a concrete batch plant. Approximately 750,000 to 1,000,000 m3 of concrete will be required for 4 units. Manufactured Construction Materials: will include items associated with site preparation (e.g., precast concrete structures, culverts and utility piping, fence), structural components for buildings and other facilities (e.g., fabricated steel products, masonry), mechanical and electrical components for buildings and facilities, and various sundry items (e.g., interior finish components). All manufactured construction materials will be delivered to the site via highway-licensed trucks travelling on provincial and municipal roads, by rail, or by barge. Aside from concrete, the largest single quantity of material that will be delivered to the site will be structural steel (rebar etc). Approximately 150,000-200,000 tonnes of structural steel would be required for 4 units. Plant Operating Components: are fixtures and components associated with an operating nuclear plant. These will include conventional items (e.g., pumps, turbines, electrical power systems) as well as those that are unique to nuclear plants (e.g., calandria). Most operating components will be delivered to the site via highway-licensed trucks travelling on provincial and municipal roads. Some oversize items will require special permits and transport provisions, and others are likely to be transported to the site by rail or via barge and off-loaded at the purpose built wharf.
C-4 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
Project Phase / Works Description and Activities Management of Construction waste: will be transferred from the site to disposal or recycling at appropriately-licensed waste management facilities. This activity Construction Waste, does not include disposal of excavated spoil (see Excavation and Grading). The existing on-site DNGS construction landfill may also be reopened Hazardous Materials, for the disposal of construction waste. Fuels and Lubricants Hazardous Materials: (e.g., solvents, chemicals, compressed gases) associated with site preparation and construction will be managed, including storage, use and disposal, in compliance with applicable legislation, codes and practices. These materials will include expired chemicals, cleaners, paint, aerosol cans and electrical components. Non-radioactive oil and chemical wastes will be removed from the site for disposal. Fuels, Lubricants and Chemicals: those required for mechanical construction equipment will be delivered to the site in appropriately-qualified vehicles and/or containers, stored in purpose-built facilities, and dispensed and used, all in compliance with applicable legislation, codes and practices. Contingency plans for a detailed response system in the event of a spill will be developed. Work Force, Payroll and Site preparation and construction will require a contractor labour force that will vary in size throughout the work based on the scope and nature of Purchasing the activities underway at any given time. This activity will represent the daily transportation-related aspects of workforce commute as well as the economic aspects associated with payroll and construction-related capital purchases. The labour force will peak, in the early years of the Project, at approximately 3,800. In later years of the site preparation and construction phase, the workforce involved in the construction of units 3 and 4 will overlap with staff operating units 1 and 2 and will peak at approximately 5,200. Operation and Maintenance Phase Prior to the start of the Operation and Maintenance Phase, commissioning activities will be undertaken including the testing of systems and components. Nuclear fission reactions in the reactor core will be increased in a controlled manner until criticality is achieved. Reactor power will then be increased in a controlled manner. Steam will be admitted into the turbine and the steam and feedwater system will be placed into service. The unit’s electrical generator will be connected, or synchronized, to the electrical grid. Maintenance, both routine and major, is included in this phase of the Project. Three general areas of maintenance are performed: preventative maintenance, corrective maintenance, and improvement or upgrade activities (including during planned shutdowns and outages). Operation of Reactor The reactor consists of the reactor assembly and reactivity control devices. The reactor core is the starting point for the generation of radioactivity. Core All other systems in the nuclear power plant (NPP) work to support the reactor core. This activity includes operation, startup, shutdown, and maintenance, testing and modification of the reactor core components, including the maintenance required for refurbishment. Nuclear malfunction and accident considerations will originate here. In an ACR-1000 reactor the horizontal calandria vessel is axially penetrated by calandria tubes. The calandria tubes provide access through the calandria vessel to the fuel channel assemblies containing nuclear fuel bundles of varying fuel enrichments.
In the EPR and AP1000 reactors, a pressure vessel contains vertically oriented assemblies of fuel rods called fuel assemblies. The assemblies, containing various fuel enrichments, are configured into the core arrangement located and supported by the reactor internals. The reactor internals also direct the flow of the coolant past the fuel rods.
C-5 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
Project Phase / Works Description and Activities Operation of Primary The function of the primary heat transport system is to move heat from the reactor core into the primary side of the steam generator. This system Heat Transport System will generate L&ILW (such as filters and ion exchange resins). This is captured in the Waste Management work activity. Maintenance of this system includes periodic chemical cleaning of the steam generators and replacement of parts during refurbishment and is included in the Major Maintenance work activity. Water losses are captured under the ventilation and drainage project works and activities. For all of the technologies, the chemistry of the reactor coolant is controlled by filtering, ion exchange, and chemical addition.
In an EPR reactor, core cooling and moderation are provided by light water (H20) at high pressure. There is no separate moderator system, only a reactor coolant system. The coolant is circulated through 4 cooling loops, each containing a steam generator. A pressurizer and a chemical and volume control system are used to maintain inventory and chemical composition in the reactor coolant system. The coolant used in this system contains boron, which acts as a neutron absorber and can also result in a reaction that forms tritium in the heat transport system fluid. Unique to the AP1000 reactor is the use of 2 cooling loops instead of 4, and therefore the use of only two steam generators. The remainder of the system is similar to that of the EPR reactor. In an ACR-1000 reactor, the heat transport system circulates light water through the reactor fuel channels to remove the heat produced by the fission of uranium fuel within the fuel bundles. Coolant from the fuel channels passes to the four steam generators where the heat is transferred to the secondary side to generate steam.
The ACR-1000 reactor has a calandria filled with a heavy water (D2O) moderator. The moderator slows down neutrons from fission reactions in the fuel, increasing the opportunity for these neutrons to trigger additional fissions. The heavy water moderator is circulated and cooled. This system is separate from the primary heat transport system, and is a low pressure, low temperature closed circuit. This activity includes routine maintenance of the moderator systems and their auxiliaries. Heavy water management is only applicable to the ACR-1000. Heavy water is managed during maintenance activities and those activities connected to the movement of heavy water inventories into and out of the moderator system. Heavy water is managed in the ACR-1000 by the D2O Supply System, the D2O Vapour Recovery System and the D2O Cleanup System. Measures are taken to minimize the loss and downgrading of the heavy water, which escapes from the moderator systems. Heavy water may be transported offsite to a licensed facility for the removal of tritium. Losses from the heavy water management system are addressed under the active ventilation systems and radioactive liquid waste management activities.
C-6 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
Project Phase / Works Description and Activities Operation of Active Radioactive Liquid Waste Management: The active drainage system segregates liquid waste by the degree of contamination and directs it to the Ventilation and receiving tanks of the radioactive liquid waste management system. The system discharges treated wastes at a controlled rate to Lake Ontario after Radioactive Liquid stringent testing and treatment to maintain acceptable activity levels for release. Waste Management Tritium can be found in heavy water after contact with the reactor core, and this may be present in waterborne and airborne emissions from water Systems losses. There are cleanup (ion exchange columns and filters) and upgrading facilities for recovered heavy water that will be used if heavy water is present in the liquid waste stream. There are also heavy water vapour recovery circuits in each reactor building to dry the atmosphere in areas that are subject to heavy water leakage during operation or servicing of equipment. Tritium can also be produced through neutron capture by B-10 in the EPR and AP1000 reactors. This tritium can be found in liquid and airborne effluents due to water losses. Radioactive Gaseous Waste Management: Gaseous wastes from potentially active areas, such as reactor buildings, will be monitored for activity before release to the atmosphere. The gases from the active ventilation stacks are filtered through absolute and charcoal filters before being released, to minimize the release of radioactivity. In some cases, the release of active gaseous waste is delayed to allow for decay of short- lived radioisotopes. Operation of Safety and A multiple barrier approach has been built into the design of all of the reactors to control releases of radioactivity to the environment. Related Systems The ACR-1000 reactor has five safety systems: Shutdown System 1 (SDS1) and Shutdown System 2 (SDS2), which provide emergency safe shutdown capability for the reactors, the Emergency Core Cooling System (ECCS), the Emergency Feedwater System (EFW) and the Containment System. The EPR reactor design includes four safety systems: the Safety Injection System (SIS) which provides emergency cooling, the Rod Cluster Control Assembly (RCCA) shutdown system which provides rapid reactor shutdown, the Emergency Feedwater System (EFWS), as well as the Containment System. The AP1000 reactor includes four safety systems: the Passive Core Cooling System (PXS) which is designed to provide emergency core cooling; the Passive Containment Cooling System (PCS) which provides for the removal of heat from the containment vessel using water and airflow; the Containment System which is a steel vessel surrounded by a concrete shielding structure; and the Reactor Trip System, which acts to keep the reactor operating away from any safety limit. Operation of Fuel and Fuel and Fuel Handling includes receipt, handling and storage of fresh fuel and used fuel. Fuel Handling Systems Fuel: The reactor may be fuelled with low enriched uranium (LEU) or more highly enriched uranium, with a maximum enrichment of approximately 5% U-235. The enrichment level and configuration of the fuel differs based on the reactor class. Fuel will be delivered to the NND site in protective flame retardant containers and stored in these containers until required. Criticality safety is a concern due to the enrichment of the fuel and a criticality program will be put in place to mitigate this. Fuel Storage and Handling: The fuel handling system comprises equipment required for fuel changing, for the storage of fresh fuel, and for on- site storage of used fuel.
C-7 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
Project Phase / Works Description and Activities
New fuel storage: New fuel is stored in a high density rack which includes integral neutron absorbing material to maintain the required degree of subcriticality. The rack is designed to store fuel of the maximum design basis enrichment.
Fuelling system: In the ACR-1000 reactor, fuelling of the reactor is completed online. Fresh fuel bundles are pushed into one end of the fuel channel by a remotely operated fuelling machine. Irradiated fuel bundles are simultaneously discharged at the other end of the channel into another fuelling machine. For the EPR and AP1000 reactors, fuelling must be completed during a refuelling outage. The refuelling operation is divided into four major phases: preparation, reactor disassembly, fuel handling, and reactor assembly. Prior to refuelling, the reactor pressure vessel (RPV) cavity is flooded with borated water and the reactor internals are placed in an internals storage pool separated from the reactor cavity by a removable gate. Fuel assemblies are remotely removed from the RPV and sent to the Spent Fuel Pool (SFP) through the fuel transfer tube. Some new fuel assemblies may be stored in the SFP, from where they will move through the fuel transfer tube and be placed into the RPV by the refuelling machine. When the refuelling is complete, the RPV internals are replaced into the RPV, instrumentation, and control/shutdown rods are reconnected, and the reactor vessel head is placed and fastened back onto the RPV. The borated water is then drained from the refuelling work areas and can be reused in the IRWST. Used Fuel Handling: In every reactor technology, the used fuel storage facility will be composed of transfer systems that carry the used fuel from the reactor to a used fuel storage pool in which the fuel is stored and cooled. The used fuel will be stored in a used fuel storage bay until it has cooled sufficiently for storage using an alternative means. Used Fuel Bay and Auxiliaries: The design specifications and location of the used fuel storage pool will be determined based on the reactor technology selected and the level of enrichment of the fuel to be used. Neutron absorbing material and spacers will be used to maintain the desired degree of subcriticality. A fuel bay cooling and purification system is used to maintain chemical composition, volume, activity level and temperature of the water in the fuel bay at desired levels. Filters, ion exchange columns and heat exchangers may be used depending on the specific reactor design selected. Operation of Secondary Turbine/Generator and Auxiliaries comprise the turbine/generator, steam supply, main condenser, feedwater heating system and auxiliary Heat Transport System systems. These systems are similar for the EPR, AP1000 and ACR-1000 reactors. This system also includes the generator oil supply and the and Turbine Generators associated fire suppression systems. This activity also includes maintenance of the system components. Interactions with the environment resulting from this activity are from oil leaks and water usage. Turbine/Generator System: Each unit has one turbine/generator unit and its auxiliary systems. The EPR and ACR-1000 reactors have four steam generators, and the AP1000 has two. Steam Supply: Steam is produced in steam generators in the reactor building, and transported by pipes to each turbine/generator. The specific configuration may vary by reactor design.
C-8 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
Project Phase / Works Description and Activities Main Condenser: Steam from the turbines exhausts into the condenser shells where it is condensed using Condenser Circulating Water and collected in the hotwells. The condensate feedwater system collects the condensed steam from the turbine and supplies it to the steam generators. External makeup to the closed loop steam and feedwater system is from the demineralized water storage tank. This configuration is independent of reactor technology selected. Feedwater Heating System: The feedwater heating system supplies feedwater to the steam generators where applicable, preheats the water to achieve a good heat rate, and performs several other functions. This is generally true for all reactor technologies. Auxiliary Systems: The major turbine/generator auxiliary systems are: the sampling system, which permits sampling steam and feedwater for chemical analysis; and the chemical control system, which eliminates the residual oxygen from the deaerated feedwater and controls its pH. These systems have different names depending on which reactor is being discussed but perform the same functions. Operation of Condenser The condenser circulating water system (CCW) supplies cold water to the condenser tubes to condense the steam from the turbine exhaust. Four and Condenser options are being assessed for the CCW system. These options are: once through cooling water, natural or mechanical draft cooling towers, or fan Circulating Water, assisted natural draft cooling towers. Dependent on climate and land considerations, a combination of these technologies may be used to provide Service Water and condenser circulating water at NND. Cooling Systems The once-through CCW system draws water from Lake Ontario, pumps the water through the condenser tubes, and discharges the water back to Lake Ontario. Water will be brought into the plant through a lake bottom intake tunnel. The configuration of the intake tunnel and structure will be similar to that currently being used at DNGS, but sized to the necessary water volumes. Natural draft cooling towers are taller and have a smaller footprint than mechanical draft cooling towers, and up to two towers will be required for each reactor unit. A natural draft tower uses convection and evaporation forces to cool the condenser circulating water. Mechanical draft cooling towers use power driven fan motors to force or draw air through the tower. They are typically shorter and have a larger footprint than natural draft cooling towers. For both cooling tower technologies, makeup condenser cooling water is drawn from Lake Ontario at significantly lower rates than with once through cooling, however, a portion of the water is lost to evaporation. The blowdown flow is directed to blowdown ponds, where mineral and particulate impurities may be removed. Discharge will comply with appropriate criteria for surface water discharge to Lake Ontario. Service Water Systems: Water will be drawn from Lake Ontario and distributed to the various systems. For the once-through cooling option, service water will be combined with the CCW systems intake. For the cooling tower option, service water is drawn from the CCW closed loop circuit. Demineralized Water: NND will include two demineralized water plants to remove minerals removed from lake water prior to use in plant cooling systems. Inactive Drainage Systems: The inactive drainage system collects wastewater in various buildings (turbine building, waste treatment building, pumphouses etc.). The wastewater is collected and treated as required to comply with discharge criteria prior to discharge.
Operation of Electrical Electrical Power Systems deliver power to and from the grid, generate emergency power and distribute power throughout the station. The Power Systems Electrical Power Systems will be similar for all reactor technologies as their operation is independent of the reactor itself. Possible environmental interactions may include noise, spills or leaks from storage tanks, and air emissions from the generators.
C-9 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
Project Phase / Works Description and Activities Switchyard and Main Transformers: A switchyard is located near the station to connect the station to the grid transmission lines. The main transformers and associated service transformers are oil cooled.
On-Site Power System: Power used internally at DNGS is supplied both from the unit itself and from the grid. Several buildings largely used for administration or general support functions are supplied with electricity from the grid.
Generation of Emergency and Standby Power: On-site standby diesel generators (DGs) provide back-up power sources to specific station loads. The configuration of the diesel generators is similar for all reactor technologies. Operation of Site Domestic Water: The domestic water system will be supplied from Durham Region water mains. Services and Utilities Sewage System: The sewage system collects waste throughout the complex and discharges it into the Regional Municipality of Durham sewage mains. Stormwater Management: Stormwater management features will be developed to address the requirements for runoff control. Stormwater runoff ponds will be sufficient in number and size to provide adequate retention times following rainfall events. The pond design will incorporate an emergency overflow bypass for flows in excess of the design storage capacity. Compressed Air: The compressed air systems consist of instrument air, service air, high pressure air and breathing air. Heating and Ventilation: The heating and ventilation systems are required to provide comfort to people working inside the plant and prevent equipment and line freezing during plant shutdown in the winter. Steam, electricity, and hot water are used for heating. On-Site Transportation: There is an extensive existing road network at the DN site including the roadways and parking lots necessary to service DNGS. Further infrastructure will be developed to service NND. The roads are used by employees, contractors and visitors to drive to and from the site, as well as for the transfer of materials. Other Auxiliary Systems: Other auxiliary systems will include: communication systems; lighting systems, site security facilities, auxiliary and service buildings, and fencing. NND will also have a dedicated onsite laundry facility. Management of Management of Low and Intermediate-Level Waste (L&ILW) will be similar regardless of reactor design selected. Two options for management Operational Low and of L&ILW include storage in a modular building on the DN site, and transport to an appropriately licensed facility off-site. Low Level Storage Intermediate-Level Buildings (LLSB), constructed as required, could accommodate both Low and Intermediate Level Waste. Eventually, the waste would be Waste transported to an appropriate facility off-site for long-term management. The first LLSB will be required by approximately 2017. Transportation of Transportation of L&ILW to the WWMF or another licensed facility and transportation of other radioactive materials, such as tritiated heavy Operational Low and water, will be carried out in accordance with the NSCA and its Regulations and other applicable regulations (e.g., as made under the Intermediate-Level Transportation of Dangerous Goods Act). Waste to a Licensed Off- site Facility
C-10 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
Project Phase / Works Description and Activities Dry Storage of Used Fuel Used fuel from NND will be stored in used fuel bays for approximately ten years following removal from the reactor. After this cooling period, the fuel is moved to dry storage containers which are processed and stored in a Used Fuel Dry Storage (UFDS) Building. Storage containers differ between the ACR and the two PWR reactors due to differences in fuel characteristics. UFDS buildings will be constructed as required, and will be either an independent facility of an expansion to the existing DWMF. Management of The generation of non-radioactive wastes will be minimized to the extent practicable through re-use and recycling programs. All residual waste Conventional Waste will be collected regularly by licensed contractors and transferred to appropriately licensed off-site disposal facilities. Hazardous wastes will be handled in accordance with applicable regulations. Replacement / Major Maintenance: Some systems and components will require maintenance, replacement or upgrading. A maintenance program for the plant Maintenance of Major will be developed to address issues related to ageing, wear and degradation. A portion of this work will require the unit to be offline for these Components and Systems maintenance activities to be completed. Typically, this work is done during a maintenance or refuelling outage that occurs once every one to three years (1-2 months duration), depending on station protocols and an assessment of needs. The periodic chemical cleaning of systems and components (e.g. steam generators) is also included in this activity. Many maintenance activities do not require a unit shutdown, and will be performed with the unit in an operating state. Refurbishment: During the 60 year life of the station, specific reactor components and the steam generators, will likely require replacement. In addition to the steam generators, refurbishment of the ACR-1000 would require replacement of fuel channel assemblies, calandria tubes and feeder pipes; and the EPR and AP1000 would require replacement of the reactor pressure vessel head. Each of these activities will require the reactors being removed from service for a period of time (one to three years). The reactor will be defuelled, systems will be drained and access ways through containment created. The components will be removed by cutting or disconnecting piping and equipment. The Low and Intermediate Level Waste from refurbishment will be transported either to a purpose built facility on-site or transported a licensed facility is in accordance with CNSC transportation regulations in place at the time of refurbishment. Safe Storage: Preparation for, and safe storage of a reactor are the first two of the three-stage decommissioning program (the final stage is dismantling, disposal and site restoration). Safe storage involves removing the reactors from service for a period of time to allow for decay of radionuclides. In preparation for safe storage, the reactors will be defueled, and dewatered. During the safe storage period resident maintenance staff will perform routine inspections and carry out preventative and corrective maintenance. Physical Presence of the When complete, NND will exist as a functioning nuclear power plant comprised of up to four individual reactors. The greatest potential difference, Station in an environmental context, between the new facility and the existing station are the cooling towers that may be included as an alternative to the once-through cooling. From a physical presence perspective, natural draft cooling towers would be the more dominant of the cooling tower options, with several towers likely, each extending to a height of as much as 152.4 m above finished grade. A visible steam plume would routinely be associated with cooling tower operation. During operations, used reactor fuel will be stored onsite in water-filled bays for a period of several years, following which it will be removed from the bays, repackaged into dry storage containers and placed into on-land storage, also onsite, for a period of up to several decades.
C-11 New Nuclear – Darlington Environmental Assessment Nuclear Waste Management Ontario Power Generation Inc. Technical Support Document
Project Phase / Works Description and Activities Administration, Upon completion of the Construction Phase of the project, the maximum estimated staff required for the operation of NND is expected to be 1,400 Purchasing and Payroll for the first two units in approximately 2016, and 2,800 for four units in about 2025. During the period 2018-2024, the workforce involved in the operation of units 1 and 2 will overlap with the workforce staff associated with the construction of units 3 and 4. During these years the Project- related workforce will total approximately 5,200. The Project-related workforce will increase from the normal complement of 2,800 by a further 2,000 during NND refurbishment (approximately 2050-2055).
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