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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

ES-1

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.




ES-2

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 m

2), one UFDS processing building (2,000

m2), one steam generator storage building (4,550 m

2), 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-3

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.




ES-4

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.




ES-5

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-6

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.




i

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


4.3 Long-Term Waste Management .......................................................................... 4-3

4.3.1 Operational L&ILW................................................................................. 4-3

4.3.2 Used Fuel ................................................................................................. 4-4


4.4 Off-Site Transportation and Accidents ................................................................ 4-5




ii

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




iii

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




iv

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




v

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




vi

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




vii

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

Management Facility

(DWMF)

The DWMF provides dry fuel storage for the Darlington

reactors.

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.




viii

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

Organization (NWMO)

The NWMO was established in 2002 by Ontario Power

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

Management Facility

(PWMF)

The PWMF provides dry fuel storage for the Pickering

reactors.

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 ( Sv), equal to one-millionth of a sievert.




ix

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

Management Facility

(WWMF)

The WWMF is a centralized processing and storage

facility for OPG’s low and intermediate- level radioactive

wastes, and dry storage for used fuel from the Bruce

nuclear generating stations.




x

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




1-1

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:

Preparation of the DN site for construction of the new nuclear facility;

Construction of the NND nuclear reactors and associated facilities;

Construction of the appropriate nuclear waste management facilities for storage and

volume reduction of waste;

Operation and maintenance of the NND nuclear reactors and associated facilities for

approximately 60 years of power production (i.e, for each reactor);

Operation of the appropriate nuclear waste management facilities; and




1-2

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:

The need for, and purpose of the Project;

Alternatives to the Project;

Alternative means of carrying out the Project that are technically and economically

feasible, and the environmental effects of such alternatives;

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;

Measures to mitigate significant adverse environmental effects that are technically and

economically feasible;

The significance of residual (after mitigation) adverse environmental effects;

Measures to enhance any beneficial environmental effects;

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;

The requirements of a follow-up program in respect of the Project;

Consideration of community knowledge and Aboriginal traditional knowledge; and

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:




1-3

Atmospheric Environment;

Surface Water Environment;

Aquatic Environment;

Terrestrial Environment;

Geological and Hydrogeological Environment;

Land Use;

Traffic and Transportation;

Radiation and Radioactivity Environment;

Socio-Economic Environment;

Physical and Cultural Heritage Resources;

Aboriginal Interests;

Health - Human; and

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:

Scope of the Project for EA Purposes;

Emergency Planning and Preparedness;

Communications and Consultation;

Malfunctions, Accidents and Malevolent Acts; and

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.




1-4

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




1-5

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.




2-1

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




2-2

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




2-3

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




2-4

FIGURE 2.3-1: TYPICAL MACSTOR 200 MODULES AT GENTILLY-2

FIGURE 2.3-2: MACSTOR 200 MODULE CROSS-SECTION




2-5

FIGURE 2.3-3: OPG DRY STORAGE CONTAINER




2-6

FIGURE 2.3-4: TYPICAL METAL CASK CONTAINER FOR PWR FUEL

(Dimensions are in mm)




2-7

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

(# PWR FUEL

ASSEMBLIES)

MAXIMUM

HEAT

LOAD

(KW)

AVERAGE FUEL

BURNUP LIMIT

(GW D/ TONNE

U)

GROSS

MASS

LOADED

(TONNE)

DIMENSIONS (M)

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




2-8

FIGURE 2.3-5: CONCRETE CANISTER FOR PWR FUEL




2-9

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.




2-10

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




2-12

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).




2-14

FIGURE 2.3-13: SWEDISH USED FUEL DISPOSAL SYSTEM




2-15

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.




2-16

FIGURE 2.4-1: CONCEPTUAL ACR-1000 STEAM GENERATOR STORAGE

70 m

65 m

24 m

7 m




2-17

FIGURE 2.4-2: CONCEPTUAL ACR-1000 FUEL CHANNEL WASTE STORAGE

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

(Door)

(Heavy forklift

maneuvering

area)

70 m

45 m




2-18

FIGURE 2.4-3: BRUCE A STEAM GENERATOR STORAGE AT WWMF

FIGURE 2.4-4: PALO VERDE STEAM GENERATOR STORAGE




2-19

FIGURE 2.4-5: RWC STORAGE AT WWMF




3-1

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

total L&ILW (Bq/m3)

9.4 E+10 2.5 E+11 3.9 E+11 2.4 E+11

Estimated average specific

activity for LLW (Bq/m3)

6.7 E+10 2.9E+10 4.5 E+10 1.7 E+11

Estimated average specific

activity for ILW (Bq/m3)

3.9 E+11 5.1 E+12 7.9 E+12 1.0 E+12

Comments LLW / ILW split

based on existing

OPG CANDU split

Does not include

H-3 or C-14

LLW / ILW split

based on ratios to

EPR split

LLW / ILW split

based on waste

stream specific data

supplied by Areva

LLW / ILW split

based on OPG waste

stream specific data

References [Candesco, 2008] [Candesco, 2008] [Areva, 2007],

[Candesco 2008][Rodrigues, 2008]




3-2

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]

[Candesco, 2008],

[Westinghouse, 2007]

[Areva, 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.




3-3

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


[Candesco, 2008]

[Candesco, 2008],


[Areva, 2007], [Candesco,

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.




3-4

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

(m3 per reactor)

3,192 2,586 4,354

# 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

with incineration + compaction (m3 per reactor)

1,691 1,154 2,329

# 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


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

(7,000 m3 each)

< 0.1 < 0.1 < 0.1

# of SBs required for on-site storage for maximum number of

units

< 0.5 < 0.5 < 0.5

Off-site:

# of ICs required for off-site storage per unit

(18 m3 each)

31 39 33

# of ICs required for off-site storage for maximum number of

units

122 153 99

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




3-5

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


Total lifetime stored volume for maximum number of

reactor units (m3)

14,952 13,098 14,845

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.




3-6

FIGURE 3.4-1: EPR FUEL




3-7

TABLE 3.4-1: SUMMARY OF FUEL CHARACTERISTICS AND ARISINGS

PARAMETER ACR-1000 AP1000 EPR

Description 43 element CANFLEX-ACR bundle (0.5

m L x 0.1 m OD)

~ 20 kg/bundle

12 bundles per channel, 520 channels

(6,240 bundles per reactor)

Conventional PWR:

(264 rods, 17x17 array

4.8 m L x 0.21m sq

~538 kg U / assembly,

~786 kg total wt)

157 assemblies in core

Conventional PWR:

(265 rods, 17x17 array,

4.8 m L x 0.21m sq

~535 kg U / assembly,

~785 kg total wt)

241 assemblies in core

Enrichment Up to 2.5% for equilibrium core 2.4-4.5% avg initial core

4.8% avg for reloads

Up to 5% for equilibrium core

Burnup (MWday/tonne U) 20,000 60,000 62,000

Refuelling cycle On-power, ~ 12 bundles per day at

equilibrium Shutdown and change 64 assemblies every

18 months at equilibrium

Shutdown and change 40% to 60% of fuel

in 18 to 24 month cycle at equilibrium

Lifetime fuel arisings

(tonnes U) 5,246 1,400 2,712

Lifetime fuel arisings

(bundles or fuel assemblies)

~262,300 bundles ~2,600 assemblies ~5,100 assemblies

Average annual fuel arisings


4,372 43 85

Comments Decay heat estimated to be 11.8 watts per

bundle at 10 years for 20,500

MWday/tonne U burnup [Pontikakis et al

2005]

Fuel assembly also contains various

burnable poison elements and control

elements to compensate for fresh fuel

reactivity

Fuel assembly also contains various

burnable poison elements and control

elements to compensate for fresh fuel

reactivity

References [AECL, 2007], [Candesco, 2008] [Candesco, 2008], [IAEA, 2004],


[Areva, 2007], [Candesco, 2008],

[IAEA, 2004]




3-8

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




3-9

FIGURE 3.4-4: ACR-1000 FUEL BASKET STACKING FRAME




3-10

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.




3-11

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

(400 baskets/module)

10 N/A N/A

# of 24 fuel assembly PWR casks N/A 55 106



TABLE 3.4-5: ON-SITE USED FUEL STORAGE BUILDING EXPANSIONS

ACR-1000 AP1000 EPR


Fraction of total fuel dry stored 50% 50% 50%

Total used fuel dry stored for maximum number of

reactor units


524,600 5,204 7,604

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.




3-12

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.




3-13

TABLE 3.5-1: ON-SITE REFURBISHMENT WASTE STORAGE

ACR-1000 AP1000 EPR

Refurbishment major LLW 4 SGs per unit

(16 total)

5.5 m OD x 24.2 m L

472 tonne each

2 SGs per unit

(8 total)

5.6 m OD x 22.5 m L

665 tonne each

4 SGs per unit

(12 total)

5.2 m OD x 24.3 m L

550 tonne each

Assumed packaging Sealed SG shell Sealed SG shell Sealed SG shell

Storage method Intact in SG storage

building

~65 m x 70 m footprint

Intact in SG storage

building


Intact in SG storage

building


In-service date 2048 2048 2048

Refurbishment major ILW 520 fuel channels per

unit (2080 total)

1 reactor pressure vessel

head (4 total)

4.8 m OD x 2.0m H

1 reactor pressure vessel

head (3 total)

5.75m OP x 3.2 H

116 tonne

Assumed packaging Bruce style retube waste

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

Sealed shell Sealed shell

Storage method RWC storage building


Included in SG storage

building

Included in SG storage

building

In-service date 2048 2048 2048


[Candesco, 2008]

[Candesco, 2008],

[IAEA, 2004],


[Areva, 2007],

[Candesco, 2008],

[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.




3-14

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.




3-15

FIGURE 3.5-1: ACR-1000 FUEL CHANNEL




3-16

FIGURE 3.5-2: ACR-1000 STEAM GENERATOR




3-17

FIGURE 3.5-3: AP1000 REACTOR VESSEL6

6 The reactor vessel head is the portion above the “upper support plate”.




3-18

FIGURE 3.5-4: AP1000 STEAM GENERATOR




3-19

FIGURE 3.5-5: EPR REACTOR VESSEL7

7 The reactor vessel head is the portion above the “o-ring seal”.




3-20

FIGURE 3.5-6: EPR STEAM GENERATOR




4-1

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.




4-2

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.




4-3

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].


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.




4-4

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).




4-5


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.




4-6

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.




4-7

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:

Excepted packaging for contents such as empty containers with traces of radioactivity

from previous usage;

Industrial Packaging (Type IP-1, Type IP-2, Type IP-3) for LLW;

Type A for contents such as contaminated inspection equipment; and

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




4-8

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




4-9

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.




4-10

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.




5-1

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

(Bundles)

4,372

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.




5-2

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:

The postulated defect rate is about three times higher on a per element basis than OPG

fuel performance experience.

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.

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.

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.




5-3

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


(Assemblies)

85

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.


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.


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.



emissions for EPR:




5-4

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.


The postulated defect rate is likely to be over conservative.

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.

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


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).

The containers are designed and operated to ensure fuel integrity is maintained to the


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.


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.




5-5


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.



emissions for the AP1000:

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;

the container containment seal is ignored and these radionuclides are released into the

environment.


The postulated defect rate is likely to be over conservative.

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.

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


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).

The containers are designed and operated to ensure fuel integrity is maintained to the


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




5-6

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.




5-7

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.




6-1

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.




6-2

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.




6-3

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.




6-4

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.




6-5

TABLE 6.3-1: SCREENING OF MALFUNCTIONS AND ACCIDENTS FOR L&IL WASTE

Potential Malfunction or

Accident Scenario

Preventative and Mitigative Measures/

Screening Evaluation

Project Phase/Works

and Activities

Screening Decision

Natural Initiating Events

Earthquake The processing or storage building will be

designed in accordance with the National

Building Code (NBC) and/or the National

Fire Code (NFC). They are expected to

remain standing.

Operations and

Maintenance

No further assessment is

required.

Flood Given the Darlington site characteristic,

extensive flooding affecting the processing

or storage buildings is not a credible event.

Operations and

Maintenance


required.

Thunderstorm Thunderstorms can potentially involve

lightning striking either the processing or a

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.

Operations and

Maintenance


required.

Tornado The processing or storage building will be


Building Code (NBC) or and the National


remain standing.



Operations and

Maintenance


required.




6-6


Accident Scenario



Project Phase/Works

and Activities

Screening Decision


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

proportional to the target area. Therefore the

postulated frequency is less than 10-7

events

per year making this an incredible event

[AMEC, 2008]. Therefore no further

assessment is required.

Operations and

Maintenance


required

Conventional Accidents


- spills

- release of chemicals

- fall of heavy equipment

- fire and explosion accidents

- releases of gases

Similar to and covered by the more general

cases for the overall EA. Refer to

Malfunctions, Accidents and Malevolent

Acts Technical Support Document

Site Preparation and

Construction

Operations and

Maintenance

Refer to Malfunctions,

Accidents and

Malevolent Acts

Technical Support

Document

Low and Intermediate Level Waste – Radiological Accidents

Vehicle/Package Accident

During Transfer to

Processing/Storage Building

(no fire)

A handling accident involving either low or

intermediate waste is possible during a

transfer to either a processing or storage

building. It is anticipated that the

consequences would be limited to a small

spill of package contents. This

Operations and

Maintenance


required.

This will be bounded by

an intermediate waste

single package fire




6-7


Accident Scenario



Project Phase/Works

and Activities

Screening Decision

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

Waste Container on the top row

in a LLSB.

A possible accident here is a pool fire (spill

of gasoline or diesel fuel from a material

handling vehicle that catches fire beside a

stack of waste containers).

Operations and

MaintenanceFurther assessment is

required.

This is the bounding

case for low level waste

Fire During Transfer to or

Storage of an Intermediate

Level Waste Package in a

Storage Building

A possible accident here is a pool fire (spill

of gasoline or diesel fuel from a material

handling vehicle that impacts on the waste)

involving transfer of an intermediate level

waste form such as a 3 m3 resin liner.

Intermediate waste packaging is assumed to

be robust enough/response time soon enough

that only a “confined” * burn occurs.

Operations and


required.

This is the bounding

case for intermediate

level waste.

* 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.




6-8

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.



Ontario Power Generation Inc Technical Support Document

6-9

TABLE 6.4-1: SCREENING OF MALFUNCTIONS AND ACCIDENTS FOR REFURBISHMENT WASTE


Accident Scenario



Project Phase/Works

and Activities

Screening Decision






remain standing and be structurally

acceptable.

Operations and

Maintenance


required.


extensive flooding affecting any processing


Operations and

Maintenance


required.










Operations and

Maintenance


required.






acceptable.

Operations and

Maintenance


required.




6-10


Accident Scenario



Project Phase/Works

and Activities

Screening Decision




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




events




Operations and

Maintenance


required



- spills

-release of chemicals



- releases of gases






Construction

Operations and

Maintenance


Accidents and

Malevolent Acts

Technical Support

Document

Refurbishment Waste – Radiological Accidents

Drop of a refurbishment waste

container

During material handling or storage

activities with intermediate waste from

refurbishment, the container falls. It is

assumed that the container is a robust

container similar to the retube waste

Operations and


required.




6-11


Accident Scenario



Project Phase/Works

and Activities

Screening Decision

containers being stored at WWMF.

Drop of a steam generator Any openings in a steam generator would

either be bolted or welded shut. During

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.

Operations and


required.

Drop of a Rector Head During storage operations a reactor head is

sealed and then a drop of the reactor head

occurs during transfer to storage. It is

assumed that steps would be taken before

moving the reactor head to either provide a

degree of containment or to fix potentially

loose contamination in place. Note the

hazard here comes from the thin corrosion

film on the reactor head and not the reactor

head itself.

Operations and

Maintenance


required.

This accident would be

bounded by the steam

generator case.

Fire Refurbishment waste is almost entirely non-

combustible. Also some of the waste forms

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.

Refurbishment No further assessment is

required




6-12

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




6-13

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.

Material – Mass, Element, Enrichment, Heterogeneity

Shape – Geometry, Volume, Concentration, Density

Poison – Solid, Liquid

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.




6-14

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.




6-15

TABLE 6.6-1: SCREENING OF MALFUNCTIONS AND ACCIDENTS FOR ACR-1000


Accident Scenario



Project Phase/Works

and Activities

Screening Decision







acceptable.

Operations and

Maintenance


required.




Operations and

Maintenance


required.










Operations and

Maintenance


required.






acceptable.


Operations and

Maintenance


required.




6-16


Accident Scenario



Project Phase/Works

and Activities

Screening Decision



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




events




Operations and

Maintenance


required



- spills




- releases of gases


cases for the overall EA. Refer to Chapter 3

in Malfunctions, Accidents and Malevolent

Acts Assessment of Environmental Effects

Technical Support Document


Construction

Operations and

Maintenance

Refer to Chapter 3 in

Malfunctions, Accidents

and Malevolent Acts

Assessment of

Environmental Effects

Technical Support

Document

Nearby Fire to Storage Container

Fire near the dry storage

container

The heavily shielded containers for dry fuel

storage would have considerable thermal

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

Operations and

Maintenance


required.




6-17


Accident Scenario



Project Phase/Works

and Activities

Screening Decision

any fire strong enough to lead to a release.

ACR-1000 – Bounding Accident

Bounding case is postulated to

be a drop of a loaded canister

causing damage to 30% of the

fuel pencils.

Gaseous radionuclides (Tritium, Kr-85) will

be released.

Operations and


required

ACR-1000 Criticality

A criticality incident might

occur during used fuel

operations,

Design features should ensure that a

criticality event would not occur.

Operations and

Maintenance


required.




6-18

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.




6-19

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




6-20

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


(Assemblies)

85 85

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.




6-21

TABLE 6.7-2: SCREENING OF MALFUNCTIONS AND ACCIDENTS FOR EPR


Accident Scenario



Project Phase/Works

and Activities

Screening Decision







acceptable.

Operations and

Maintenance


required.

Flood Given the Darlington site characteristics,



Operations and

Maintenance


required.










Operations and

Maintenance


required.






acceptable.

Operations and

Maintenance


required.




6-22


Accident Scenario



Project Phase/Works

and Activities

Screening Decision




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




events




Operations and

Maintenance


required



- spills




- releases of gases






Construction

Operations and

Maintenance


Accidents and

Malevolent Acts

Assessment Technical

Support Document

Nearby Fire to Storage Cask

Fire near the dry storage cask. The heavily shielded casks for dry fuel


inertia. Also it is operating practice to store

them away from any appreciable amount of

combustible material. It is highly unlikely

Operations and

Maintenance


required.




6-23


Accident Scenario



Project Phase/Works

and Activities

Screening Decision

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

cask with a failure of 30% of

the fuel elements.

Gaseous radionuclides (Tritium, Kr-85) will

be released.

Operations and


required.

EPR - Criticality



operations,



Operations and

Maintenance


required.




6-24

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


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




6-25

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


(Assemblies)

43 43

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.




6-26

TABLE 6.8-2: SCREENING OF MALFUNCTIONS AND ACCIDENTS FOR AP1000


Accident Scenario



Project Phase/Works

and Activities

Screening Decision







acceptable.

Operations and

Maintenance


required.

Flood Given the Darlington site characteristics,



Operations and

Maintenance


required.










Operations and

Maintenance


required.






acceptable.

Operations and

Maintenance


required.




6-27


Accident Scenario



Project Phase/Works

and Activities

Screening Decision




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




events




Operations and

Maintenance


required



- spills




- releases of gases


cases for the overall EA. Refer to Chapter 3

in Malfunctions, Accidents and Malevolent

Acts Assessment of Environmental Effects

Technical Support Document


Construction

Operations and

Maintenance

Refer to Chapter 3 in

Malfunctions, Accidents

and Malevolent Acts

Assessment of

Environmental Effects

Technical Support

Document

Nearby Fire to Storage Cask

Fire near the dry storage cask. The heavily shielded casks for dry fuel


inertia. Also it is operating practice to store

them away from any appreciable amount of

Operations and

Maintenance


required.




6-28


Accident Scenario



Project Phase/Works

and Activities

Screening Decision

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

loaded cask and it is assumed

that 30% of the fuel pencils

fail.

Gaseous radionuclides (H-3, Kr-85) will be

released.

Operations and


required.

AP1000 - Criticality



operations,



Operations and

Maintenance


required.




6-29

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


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




6-30

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.




7-1

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 m

2), one steam generator

storage building (4,550 m2), and one refurbishment waste storage building (3,150 m

2). 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.




7-2

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.




8-1

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.




8-2

(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.




APPENDIX A

OPERATIONAL WASTE DETAILS




A-1

TABLE A-1: SUMMARY OF AVERAGE L&ILW SPECIFIC ACTIVITY

Quantity

(Bq/yr)

Avg L&ILW

Specific

Activity

(Bq/m3)

Quantity

(Bq/yr)

Avg L&ILW

Specific

Activity

(Bq/m3)

Quantity

(Bq/yr)

Avg L&ILW

Specific

Activity

(Bq/m3)

Avg L&ILW

Specific

Activity

(Bq/m3)

Avg LLW

Specific

Activity

(Bq/m3)

Avg ILW

Specific

Activity

(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)

per unit315 300 15

Radionuclide

OPG CANDU

164 154 225

ACR AP1000 USEPR




A-2

TABLE A-2: TYPICAL EPR L&ILW SOURCE TERM DETAILS

Nuclide Incinerable CompacibleNon-

processible

AVERAGE

DAW

Evaporator

concentrate

Storage tank

sludge

Demin /

Centrifuge

sludge

AVERAGE

SludgeMixed waste

AVERAGE

LLWFilters

Primary

Coolant IX

Radwaste

Demin IX

AVERAGE

ILW

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

Bq/m3

LLW ILW

COMBINED

L&ILW




A-3

TABLE A-3: DETAILS OF PROCESSED L&ILW VOLUMES L&ILW Operational Wastes (per reactor unit)

Annual

ACR ACR ACR

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

ACR ACR ACR

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

As generated As stored (@ DWMF)

AP1000 USEPRAP1000 USEPR AP1000 USEPR

As generated As stored (@ DWMF)

AP1000 USEPR AP1000 USEPR

As stored (@ WWMF)

AP1000 USEPR

As stored (@ WWMF)




A-4

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) 9 18 18 18 18 18 27 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36

ILW stored (m3) 9 18 18 18 18 18 27 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) 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



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



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



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



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 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


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-5

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

ACR U1 refurbU1 + U2

refurbU2 refurb U3 refurb

U2 + U4

refurbU4 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



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



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



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





Used Fuel


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


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 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



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





Used Fuel


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



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 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-6

Note: Colours indicate special dates such as start of a unit, new storage buildings going into service, or unit refurbishment.

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 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



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


On-site storage expansions



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 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



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





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 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




APPENDIX B

REFURBISHMENT WASTE DETAILS




B-1

TABLE B-1: SUMMARY OF DARLINGTON FUEL CHANNEL COMPONENT

SPECIFIC ACTIVITY8

TBq/Kg (AFTER 270 DAYS OF DECAY) NUCLIDE

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-2

TABLE B-2: SUMMARY OF STEAM GENERATOR ACTIVITY

TOTAL PER SG (TBq) NUCLIDE

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]




APPENDIX C

NEW NUCLEAR - DARLINGTON - BASIS FOR EA




C-1

TABLE C-1: NEW NUCLEAR – DARLINGTON – BASIS FOR EA

Project Phase / Works

and Activities

Description

Site Preparation Phase

Mobilization and

Preparatory Works

Mobilization (construction workforce and equipment): will involve mobilization of equipment and the construction workforce to the site. The

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-2


and Activities

Description

Marine and Shoreline

WorksMarine and Shoreline Works includes all works and activities conducted within or adjacent to Lake Ontario such that they are likely to interact

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

Physical Support

Facilities

Administration and Support Facilities comprise various buildings housing staff, equipment and operations necessary to provide ongoing support to

the NND. These will include offices, workshops, maintenance, storage and perimeter security buildings, and utilities operating centres. All such

buildings will consist of conventional steel and masonry structures.

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

BlockThe Power Block includes the reactor building, the turbine-generator building/turbine hall (powerhouse) and related structural features that are

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-3


and Activities

Description

Construction of Intake

and Discharge Structures Intake and Discharge Tunnels and Structures for Once-Through Lake Water Cooling: For EA purposes, the once-through cooling water

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

Facilities Ancillary facilities include all features necessary to support operations of the reactors and generation of electricity, although not physically

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 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-4


and Activities

Description

Construction of

Radioactive Waste

Storage Facilities

Radioactive Waste Storage Facilities comprise used fuel dry storage facility to house containerized used fuel bundles following their removal from

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

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

Stormwater As the site is developed, ditches and swales will be constructed to collect and convey surface water to stormwater management ponds and

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

Equipment, Material and

Operating Plant

Components

Supply of construction materials and operating equipment includes the delivery to the site, of all necessary materials and components for

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

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

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-5


and Activities

Description

Management of

Construction Waste,

Hazardous Materials,

Fuels and Lubricants

Construction waste: will be transferred from the site to disposal or recycling at appropriately-licensed waste management facilities. This activity

does not include disposal of excavated spoil (see Excavation and Grading). The existing on-site DNGS construction landfill may also be reopened

for the disposal of construction waste.

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

PurchasingSite preparation and construction will require a contractor labour force that will vary in size throughout the work based on the scope and nature of

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

Core

The reactor consists of the reactor assembly and reactivity control devices. The reactor core is the starting point for the generation of radioactivity.

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-6


and Activities

Description

Operation of Primary

Heat Transport System

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

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-7


and Activities

Description

Operation of Active

Ventilation and

Radioactive Liquid

Waste Management

Systems

Radioactive Liquid Waste Management: The active drainage system segregates liquid waste by the degree of contamination and directs it to the

receiving tanks of the radioactive liquid waste management system. The system discharges treated wastes at a controlled rate to Lake Ontario after

stringent testing and treatment to maintain acceptable activity levels for release.

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

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

Related Systems

A multiple barrier approach has been built into the design of all of the reactors to control releases of radioactivity to the environment.

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.

Fuel and Fuel Handling includes receipt, handling and storage of fresh fuel and used fuel.

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.

Operation of Fuel and

Fuel Handling Systems

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.




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Description

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.

Turbine/Generator and Auxiliaries comprise the turbine/generator, steam supply, main condenser, feedwater heating system and auxiliary

systems. These systems are similar for the EPR, AP1000 and ACR-1000 reactors. This system also includes the generator oil supply and the

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.

Operation of Secondary

Heat Transport System

and Turbine Generators

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.




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Description

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

and Condenser

Circulating Water,

Service Water and

Cooling Systems

The condenser circulating water system (CCW) supplies cold water to the condenser tubes to condense the steam from the turbine exhaust. Four

options are being assessed for the CCW system. These options are: once through cooling water, natural or mechanical draft cooling towers, or fan

assisted natural draft cooling towers. Dependent on climate and land considerations, a combination of these technologies may be used to provide

condenser circulating water at NND.

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

Power Systems

Electrical Power Systems deliver power to and from the grid, generate emergency power and distribute power throughout the station. The

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.




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and Activities

Description

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.

Domestic Water: The domestic water system will be supplied from Durham Region water mains.

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.

Operation of Site

Services and Utilities

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

Operational Low and

Intermediate-Level

Waste

Management of Low and Intermediate-Level Waste (L&ILW) will be similar regardless of reactor design selected. Two options for management

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

Buildings (LLSB), constructed as required, could accommodate both Low and Intermediate Level Waste. Eventually, the waste would be

transported to an appropriate facility off-site for long-term management. The first LLSB will be required by approximately 2017.

Transportation of

Operational Low and

Intermediate-Level

Waste to a Licensed Off-

site Facility

Transportation of L&ILW to the WWMF or another licensed facility and transportation of other radioactive materials, such as tritiated heavy

water, will be carried out in accordance with the NSCA and its Regulations and other applicable regulations (e.g., as made under the

Transportation of Dangerous Goods Act).




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Description

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

Conventional Waste

The generation of non-radioactive wastes will be minimized to the extent practicable through re-use and recycling programs. All residual 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.

Major Maintenance: Some systems and components will require maintenance, replacement or upgrading. A maintenance program for the plant

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

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.

Replacement /

Maintenance of Major

Components and Systems

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

Station

When complete, NND will exist as a functioning nuclear power plant comprised of up to four individual reactors. The greatest potential difference,

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.




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Description

Administration,

Purchasing and Payroll

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

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).