Executive Summary

The Society of Professional Engineers and Associates (SPEA) has examined the project documentation in detail and supports the Deep Geological Repository (DGR) project for the following reasons:

1)      We have confidence in the lengthy, robust regulatory approvals EA process and positive results from the Environmental Impact Study:

 

a)      A thorough, traceable and step-wise assessment process was applied to identify the potential environmental effects of all phases of the DRG Project (including site preparation and construction, operations, decommissioning, and abandonment and long-term performance of the DREG Project ) on the  health and safety of the public, the workers and non-human biota. 

b)     After more than four years of studies , investigations, and analyses of many components of the environment (air quality, noise levels, hydrology and water quality, aquatic and terrestrial environments, geology, radiation and radioactivity, aboriginal interests, socio-economic, human health and ecological environment) , the EA concluded, taking into consideration the proposed design and identified mitigations measures, that the DGR project is not likely to result in any significant adverse effects on the health and safety of the public, the workers or non-human biota.  It is predicted that the DGR Project will result in beneficial socio-economic effects for the local and regional communities.

 

2)      Transparency of the EA process.  The EA process was transparent through extensive Public Participation and Aboriginal Engagement program for the DRG Project which provides many opportunities for the public to become informed and updated, ask questions, provide comment and discuss concerns about the DGR project. OPG was able to address stakeholders concerns by extensive public consultation and engagement strategies (see Socio-economic Technical Information below).

 

3)      Safe long-term management of LLW and ILW. The DGR Project provides a long-term solution for low and intermediate level wastes from reactor operations, and refurbishment and decommissioning projects at many nuclear sites in a centralized repository instead of leaving the waste for future generations.

4)      Use of international best practices and expertiseThe design is based on international best practices and the safety of the long-term management of low and intermediate level wastes in the repository. 

 

5)      Extensive OPG expertise in waste management.  OPG has a long history in the safe management at the Western Waste Management facility (WWMF) that inspires confidence and trust within the local population.

 

6)      Public Acceptance. Kincardine is a willing host municipality.

 

7)      Ideal location for the repository. The location of DGR at Bruce Nuclear makes sense since most of the waste is already managed at the Western Waste Management Facility by trained Nuclear Energy Workers (NEWs).

 

We also highlight that the average background dose of radiation in Canada from natural sources is 1.8 mSv/year.  Operation of the DGR will, at most, increase radiation levels in the vicinity of the facility by 0.6% (one part in 180); which, in other words, is an insignificant amount beyond what already exists in nature.
1.  Introduction

There are many different types and classifications of radioactive waste.  To the average person, the word radioactive waste is something to be afraid of and to some others, it is synonymous with spent nuclear reactor fuel, which contains short and long live fission products as well as unused uranium, plutonium and radioactive transuranic elements. This is not the type of waste that would be sequestered in the Ontario Power Generation (OPG) deep geological repository (DGR).  Rather the wastes that will be sequestered in the OPG DGR are low level waste (LLW) and intermediate level waste (ILW). 

Section 2 addresses the subject of radiation doses in general as well as those specific to the OPG DGR and compares these with regulations as well as natural background sources.

To appreciate the nature and hazards associated with the LLW and ILW material, it is useful to characterize it in detail and that has been  done in section 3 of this submission.

Section 4 addresses the need for the DGR project and long term waste management.

Safety regulations and requirements and our assessment as to whether or not these have been met in the design of the DGR facility are addressed in section 5 of this submission.\

Finally, the socio-economic aspects of the project are addressed in section 6.

2.  Public Doses from the DGR Compared with Regulations and Natural Background Radiation

The general radiological protection requirements for the DGR are in accordance with the Radiation Protection Regulations (SOR/2000-203) promulgated under the Nuclear Safety and Control Act (NSCA97).

 

Doses resulting from the DGR operation will be within the regulatory dose limits, and

will be kept as low as reasonably achievable (ALARA). The CNSC regulatory dose limits for the public and the Nuclear Energy Workers (NEWs) are shown below.

Table 2-1 CNSC Effective Dose Limits

Person

Period

Effective Dose (mSv)

 

NEW, including a pregnant

NEW

One-year dosimetry period

Five-year dosimetry period

50

100

Pregnant NEW

Balance of the pregnancy (after the licensee is informed of the pregnancy)

4

 

A person who is not a NEW

One calendar year

1

 

The dose rate targets for the DGR pre-closure period, derived from the table above for a

member of the general public or a non-NEW, are as follows [1] :

 

·         ≤ 0.5 µSv/hr at the DGR fence, based on the CNSC annual dose limit of 1 mSv for a member of the public, over a maximum of 2000 hours per year occupancy for non-NEWs; and

 

·         ≤ 10 μSv/year at the Bruce nuclear site boundary, based on year round occupancy – this dose rate target is 1% of the CNSC annual dose limit of 1 mSv for a member of the public.

 

A maximum dose rate of 2 mSv/hr at contact from waste package exterior surface

was assumed in this assessment.  Most packages will be at or less than this dose

rate, after decay and package shielding.

 

The peak estimated doses to the most exposed public groups due to the airborne and waterborne emissions from the DGR are expected to be around  0.21 (REMP-based calculation) to 0.56 µSv per year (DRL-based calculation).  There is no contribution from gammas streaming through the shielding as there is too much shielding between the containers and the receptor.  The doses are a result of air and water particulate transported through the ventilation shaft.

The doses were estimated using 2 separate models.  The calculations use a conservative set of assumptions. 

The maximum doses occur in 2023.  Based on the current schedule, most of the wastes will have been transferred from the WWMF to the DGR by 2023, resulting in the maximum potential releases and consequentially, maximum doses to the public.

Doses at other years would be lower due to the decay of the radionuclides and due to the diminution in leakage as panels are closed off.   A slower transfer rate of packages to the DGR would lead to a lower peak dose rate from the DGR.

The REMP-based method gives a more realistic estimate of the public dose (0.21 µSv.a), which is lower than that derived from the DRL pathways method. In either case, the results indicate very low doses to the public, similar to what would be calculated for WWMF for similar LLW and ILW radionuclide inventories. Nevertheless, both models predicts maximum public doses that are less than 1 µSv/a.  These doses are far below the CNSC regulatory limit of 1 mSv/year (by a factor of at least 1000).

The assessment results indicate that there are no concerns with respect to exposure to members of the public during normal operations of the DGR.

Despite these conservative assumptions, the calculated doses are minimal compared with natural ‘background’ radiation levels. 

 

Figure 2.1  The electromagnetic spectrum [2]

The total worldwide average effective dose from natural radiation is approximately 2.4 mSv per year; in Canada, it is 1.8 mSv. In some parts of the world, it is naturally much higher – for instance on the Kerala Coast in India, the annual effective dose is 12.5 mSv. The dose varies with the source of the radiation. For example, in northern Iran, geological characteristics result in a dose that can reach 260 mSv per year.  It is important to note that despite these high doses there have not been any adverse health effects observed among the general population [3]. By comparison, the maximum allowable annual dose (allowable by who?) for a nuclear energy worker is 50 mSv per year, and for a member of the general public it is 1 mSv/year[1] [4] . Another useful yardstick is the average radiation dose from an abdominal/pelvic CT scan, which is 10 mSv.

 

Table 2-2 Background Radiation Doses in Different Parts of the World

 

Location

mSv/a

Northern Iran

260

India

12.5

Worldwide Average

2.4

Canada

1.8

Toronto

1.6

Ottawa

1.8

Montreal

1.6

DGR (incremental dose)

0.01

 

 

 

Figure 2.2 Background Radiation Doses in Different Parts of the World[2]

Figure 2.3 Background Radiation Doses in Different Parts of the World (excluding data from Northern Iran)

 

3. Characterization (or description) of waste and long-term management

3.1  Description of Low and Intermediate Level Waste

Radioactive wastes to be accepted by the DGR are classified as solid low-level or solid intermediate-level. The classification is as described below, and is consistent with Canadian Standards Association (CSA) standard N292.3 [4].

Low level waste (LLW) consists of non-fuel waste in which the concentration or quantity of radionuclides is above the clearance levels and exemption quantities given in the Nuclear Substances and Radiation Devices Regulations [5], and which contain primarily short-lived radionuclides (i.e. half-lives shorter than or equal to 30 years).  Typically, Low-level wastes include paper, rags, tools, clothing, filters, and other materials which contain small amounts of mostly short-lived radioactivity.  LLW normally does not require significant shielding for worker protection during handling and storage. OPG LLW typically consists of: incinerator ash; compacted waste; bulk and drummed non-processible wastes; some low activity ion-exchange (IX) resins and filters from secondary side reactor process systems; and system components such as heat exchangers, feeder pipes and steam generators.

Intermediate Level Waste (ILW) consists of non-fuel waste that can contain significant quantities of long-lived radionuclides. ILW often requires shielding for worker protection during handling. OPG ILW typically consists of primary side and moderator ion exchange (IX) resins and filters; irradiated core components; and reactor fuel channel wastes from refurbishment activities.

The L&ILW are generated from a variety of activities: operational wastes (which includes all L&ILW from operation and maintenance of the reactors and their associated facilities) and refurbishment wastes (which includes component waste from major refurbishment projects, such as pressure tubes, calandria tubes, end-fittings, steam generators, and associated hardware). Decommissioning waste (which includes waste from the final dismantling of reactors and facilities) is not included in the preliminary safety report.

The DGR will not accept used fuel or recognizable fuel fragments. The DGR also excludes liquid wastes, except for small amounts of incidental liquids that are inevitably associated with the solid wastes.

Table 3-1 and  Table 3-2 provide more detailed descriptions of various L&ILW categories tracked for the DGR.

Table 3-1 LLW Categories

Waste Category

Description

Bottom Ash

Heterogeneous ash and clinker from waste incineration.

Baghouse Ash

Fine homogeneous ash from waste incineration.

Compact Bales

Generally compactible solid LLW, for example, empty waste drums, rubber hoses,

rubber area floor matting, light gauge metals, welding rods, plastic conduit, fire blankets

and fire retardant material, metal cans, insulation, ventilation filters, air hoses, metal

mop buckets and presses, electric cable (<1/4” diameter), lathe turnings, metal filings,

glass, plastic suits (Mark III/IV), rubbers, Vicraft hoods, rubber gloves.

Box Compacted

Same as compact bales.

Non-Processible

Boxed

Solid LLW that is non-compactible or has contact dose greater than 2 mSv/hr, for

example, heavy gauge metal (i.e., beams, IX vessels, angle iron, plate metal), concrete

and cement blocks, metal components (i.e., pipe, scaffolding pipes, metal planks,

motors, flanges, valves), wire cables and slings, electric cables (>1/4” diameter),

Comfo respirator filters, tools, paper, plastic, absorbent products, laboratory sealed

sources, feeder pipes.

Non-Processible

Drummed

Generally small, granular or solidified LLW, for example, floor sweepings, cleaners and

absorbents (e.g., Dust Bane, Stay Dry), metal filings, glassware, light bulbs,

bitumenized LLW.

Non-Processible

Other

Large and irregularly shaped objects such as heat exchangers, Encapsulated Tile

Holes (ETHs), shield plug containers, and other miscellaneous large objects (e.g., fume

hoods, glove boxes, processing equipment).

Low Level / ALW

Resin

Spent low level IX resin arising from light water auxiliary systems, and/or Active Liquid

Waste (ALW) treatment systems.

ALW Sludge

Sludge from Bruce two-stage ALW Treatment System.

Steam

Generators

Steam generators removed from service.

 

Table 3-2 ILW Categories

Waste Category

Description

Moderator Resin

Spent IX resin arising from moderator purification

Systems.a

Primary Heat Transport (PHT) Resin

Spent IX resin arising from PHT purification systems.a

Miscellaneous Resin

Spent IX resin and activated charcoal arising from station

auxiliary systems (e.g., heavy water upgraders).a

CANDECON Resin

Spent IX resin from chemical decontamination process for

nuclear heat transport systems.

IX Columns

Spent IX resin mainly arising from Pickering PHT

purification system, comes as package with steel

container.

Irradiated Core Components

Various replaced core components, notably flux detectors

and liquid zone control rods.

Filters and Filter Elements

Filters and filter elements from various station process

systems.

Retube - Pressure Tubes

Fuel channel waste from large scale retube.

Retube - Calandria Tubes

Fuel channel waste from large scale retube.

Retube - Calandria Tube Inserts

Fuel channel waste from large scale retube.

Retube – End-Fittings

Fuel channel waste from large scale retube.

Note: a. These ILW resins often occur mixed together in the same container in varying proportions.

 

3.1.2  Waste Containers and Packages

The combination of wastes plus container is defined as a waste package. Some waste packages currently stored at WWMF meet DGR waste acceptance criteria and are considered DGR-ready. Others will require some waste conditioning, additional decay time, and/or container overpacking or shielding.

Steam generators, and if necessary heat exchangers, would be segmented to meet the size and weight limits of the DGR main shaft cage. They would be grouted to stabilize the contents, cut into sections, and have seal plates welded on the ends. Some steam generators may be processed to recycle the inactive steel components, and the residual parts with all the radioactivity transferred to the DGR in LLW non-processible or similar containers; however, the reference forecast conservatively assumes segments.

ILW resins are stored in steel resin liners. Depending on the dose rates of the resin liners, they would be placed into disposable concrete cylindrical shield overpacks, which ensure that the resulting dose rates do not exceed the DGR waste acceptance criteria. The reference overpack options are a 250 mm thick concrete shield that holds two resin liners, a 350 mm shield that holds two resin liners, and a short 350 mm shield with a steel insert that holds one resin liner. It is expected that approximately one-third of resin liners will not require a concrete shield.

ILW irradiated core components, filters and filter elements and IX columns are presently stored in long tile-hole-equivalent (T-H-E) liners within ICs, i.e., IC-2s and IC-18s. OPG’s reference plan is that this existing ILW will be removed from the T-H-E liners and repackaged into smaller containers i.e., alternative tile-hole-equivalent liners (ATHELs) for later transfer to the DGR. These would be placed into disposable concrete cylindrical shield overpacks similar to the ones used for ILW resin liners, which ensure that the resulting dose rates do not exceed the DGR waste acceptance criteria. It is assumed that post-2018 newly arriving ILW of this type will be placed directly into smaller containers (e.g., ILW shields or ATHELs) for later transfer to the DGR.

As part of reactor lifecycle management activities, the current assumption is that replaced feeders will be cut into suitable lengths and packaged in non-processible (box) containers. Retube wastes will be emplaced using two sizes of shielded containers: RWC(PT) for pressure tubes, calandria tubes, and calandria tube inserts, and RWC(EF) for end-fittings.

3.1.3  Waste Acceptance Criteria

All LLW and ILW will be shipped or transferred to the DGR Facility in waste packages that meet the DGR waste acceptance criteria.

The DGR waste acceptance criteria have been developed to ensure that the wastes emplaced in the DGR are within the bounds of the safety assessment, design basis and regulatory requirements.

 

Table 3-3 Summary of Waste Acceptance Criteria

Criteria

Summary Description

Waste characterization

- physical, chemical, radiological characteristics of each package

Documentation

- waste packages must be tracked in OPG's waste tracking database with waste characteristics, dose rates, description of contents, etc.

- verified load statements

- supplemental info such as radiological surveys, chemical analyses, loading checklists

- notes on package design documentation, such as drawings, technical specifications, design

requirements, etc.

- transfer documents for wastes subject to additional controls

Acceptable waste

package designs

- all DGR waste package designs must be approved

Condition of waste

Container

- no significant rusting

- sound structural integrity

- no leakage

- no wobbling or tilting

Mass limits

- 35 Mg, subject to maximum design limit for each waste package type

Size limits

- must fit within internal dimensions of the DGR cage

Containment

- wastes and contamination shall be contained during handling

- all containers shall have lids

Venting

- where the potential for gas build-up exists and containers are not designed to withstand the pressure, the containers shall be vented

Identification/labelling

- containers bar-coded with OPG's waste tracking database tracking number on two adjacent vertical sides

- additional information including gross mass, dose rate, and significant non-radiological hazards to be marked on packaged with lettering at least 25 mm high

Stackability

- stable, self supporting stack of up to 6 m high

- use of standard footprints strongly encouraged

Handling

- conventional material handling equipment such as forklifts with loads of up to 35 Mg

Fire resistance

- non-combustible containers

Dose rate limits

- 2 mSv/hr on contact with external surface of waste package or shielding

- 0.1 mSv/hr at 1 m from transportation package

- exceptions approved by responsible health physicist

Radionuclide

Composition

- package amount must be reported for H-3, C-14, Cl-36, Co-60, Sr-90, Zr-93, Nb-94, Tc-99, I-129, Cs-135, Cs-137, U-235, U-238, Pu-239, Pu-240, Pu-241

Contamination limits

- removable surface contamination on package exterior to be less than 4 Bq/cm2 beta-gamma and 0.4 Bq/cm2 alpha when averaged over 300 cm2

Heat load limits

- no restriction if less than 0.01 W/m3 of waste package external dimensions

- up to 10 W/m3 by prior notification and approval for special cases

Waste form

- solids only

- sludges must have slump of less than 150 mm

Residual liquids

- generally must be less than 1% free liquid by volume

- bulk IX resins must be less than 5% free water by volume

Gas generation

- must not generate toxic gas on exposure to water

Excluded wastes

- reactive wastes, polychlorinated biphenyl (PCB) wastes, pathological wastes, ignitable wastes

- explosives, corrosives, compressed gases

- used nuclear fuel and recognizable fuel fragments

- high thermal Co-60 sources

Special notice wastes

- wastes containing significant levels of Occupational Health and Safety Act (OHSA90)

designated substances

- leachate toxic wastes

Chelating agents

- must be less than 1% by weight of package

Petroleum oils

- must be less than 1% by weight of package

 

3.1.4  Waste Characterization Program

Most L&ILW is inherently heterogeneous, with considerable variability both across waste categories, and also from package to package within a waste category. OPG has therefore supported a waste characterization and tracking program for many years.  The characteristics of various waste types have been identified, and information recorded on waste packages in an electronic database (part of waste tracking). 

The reference methodology for characterizing radioactive wastes is based on the use of gamma dose rates associated with each waste package to derive certain marker radionuclide inventories (usually Co-60 and Cs-137), and scaling factors to calculate the inventory of other radionuclides of interest. The scaling factors are derived either from experimental data on OPG waste samples, from theoretical arguments (such as activation analysis and computer modelling, comparison with incinerator stack emission data for volatile species, etc.), or from the ratio of these nuclides in used fuel. This process is consistent with international best-practice (IAEA09, ISO07).

3.1.5 Waste Tracking

Waste containers and inventories stored at WWMF are presently tracked using OPG’s Integrated Waste Tracking System electronic waste tracking database (ANDERSON05). This system, or a similar one, will be adopted for the DGR, so that waste packages will be tracked with respect to their location within the DGR. This system will contain information on the characteristics of each package, and will have the ability to produce reports on the waste inventory within the DGR at any time.

3.1.6  Waste Volumes

The amount of waste and number of packages projected over the life of OPG’s nuclear program is calculated based on the existing inventory tracked in OPG’s waste tracking database and a future waste receipt projection.

Based on the existing plus projected inventory, it is estimated that approximately 53,000 packages representing a total emplaced volume of approximately 200,000 m3 will be sent to the DGR. About 80% of the emplaced volume is LLW. Note that while refurbishment waste only makes up about 10% of the emplaced volume, it accounts for more than 60% of the radionuclide inventory at 2062.

Table 3-4 Waste Volumes in Reference Forecast (Rounded)

 

Operations LLW

Operations

ILW

Refurbishment

L&ILW

Total

Net waste volume (m3)

95,100

9.300

11,200

115,600

As-stored volume (m3)

135,000

13,500

21,700

170,200

Emplaced volume (m3)

154,700

27,600

21,700

204,000

 

Note also that the “emplaced volume” is greater than the “as-stored volume”, which is the volume of the containers in which the waste is presently stored, due to the extent of overpacking and disposable shielding used for DGR-ready packages.

Each year, approximately 5,000 to 7,000 m³ of new L&ILW is produced as a result of the operation of OPG owned or operated nuclear generating stations in Ontario, including those at Darlington, Pickering and Bruce. The waste is transported to the WWMF for processing, which may include compaction or incineration. After volume reduction, this results in 2,000 to 3,000 m³ of additional stored waste annually. To the end of 2010, the existing nuclear reactors in Ontario have produced about 84,000 m³ of waste.

The total projected emplaced waste package volume is approximately 200,000 m3with a total of about 50,000 packages.

3.1.6  Waste Package Retrieval

The materials that are placed in the DGR are considered waste and the need for retrieval is not anticipated. However, in the unlikely event that any waste package(s) would need to be retrieved from a room following emplacement, retrieval can be achieved.

A specific plan for retrieving the package(s) would be developed. First, the position of the waste package(s) to be retrieved will be identified using the waste tracking system and the number and type of packages that will have to be moved to access the identified waste package will be determined. Alternative locations, which may be temporary or permanent, for the packages will be identified. They could be relocated to another room, which is partially filled or empty. This new location could be suitable as a permanent location for these packages.

4. Long-Term Management

The DGR is the long-term management solution for the operational and refurbishment L&ILW currently stored at the WWMF, as well as the future operational and refurbishment L&ILW produced as a result of operation of OPG-owned or operated nuclear reactors.

The DGR Project is proposed because it provides a long-term management method for waste streams from OPG-owned or operated nuclear generating stations that will protect health, safety and the environment, and if necessary, will do so in the absence of institutional controls

4.1            Need for the Project

The basic need for the DGR Project derives from the fact that L&ILW consists of materials that can remain hazardous for hundreds, and is some cases, many thousands of years due the presence of long-lived radionuclides. These long timeframes require that a solution be found that protects humans and the environment, that is passive, and that does not require long-term institutional control. For shorter-lived radionuclides, near-surface disposal facilities can provide the required protection; however, for long-lived radionuclides, deep geologic disposal in suitable rock formations is the solution consistent with international guidance and practice.

The proposed site for the DGR is located on lands adjacent to the WWMF at the Bruce nuclear site within the Municipality of Kincardine.  The proposed DGR Project site was chosen because it holds two attributes that, based on international experience, are essential for the successful development of a long-term waste management facility: technical suitability, in this case geology that offers multiple natural barriers to safely isolate and contain the waste for tens of thousands of years and beyond; and an informed and willing host community.

As part of the MOU with the Municipality of Kincardine, an Independent Assessment Study [6] was undertaken concurrent with studies in support of the engineering and geotechnical feasibility of a long list of concepts for LLW management at the WWMF. The options considered were:

·         enhanced processing and storage;

·         covered above-grade concrete vaults;

·         shallow concrete vaults;

·         deep concrete vaults;

·         shallow rock cavern vaults in near surface dolostone (less than 100 m below surface);

·         deep rock cavern vaults in thick salt bed (200 to 400 m below surface);

·         deep rock cavern vaults in “tight” shale formation (400 to 600 m below surface);

·         deep rock cavern vaults in “tight” limestone formation (600 to 800 m below surface); and

·         ongoing management at the WWMF (status quo).

The results of a primary screening analysis eliminated deep concrete vaults and deep rock cavern vaults in thick salt bed from further evaluation because suitable host formations are absent. A secondary geotechnical feasibility screening showed that the shallow concrete vaults and shallow rock cavern vaults were not technically feasible at the WWMF site. The two final deep rock cavern vaults were combined and considered together as deep rock vaults (now referred to as the DGR).

The four short-listed feasible concepts were enhanced processing and storage, surface concrete vaults, deep rock vaults and status quo.

As part of the Independent Assessment Study, the four feasible concepts were compared using a number of different criteria, including:

·         engineering feasibility, which considers the preliminary designs, geotechnical feasibility, construction and operation schedule, and cost and personnel estimates;

·         safety and licensibility, which considers the routine releases, intrusion scenarios and licensibility;

·         environmental protection and feasibility, including potential effects on the physical, biological and socio-economic environments;

·         economic feasibility, including predicted employment, expenditures, municipal taxes, and population and community spending; and

·         social factors, including results of public attitude research on the options and tourism research.

These criteria were presented to the public along with descriptions of the alternatives to increase awareness and understanding of the options, and to identify issues and concerns.

To summarize, the Independent Assessment Study found that each of the four long-term management options is technically feasible and may be safely constructed at the WWMF.

There is considerable international experience using each of the options for the long-term management of L&ILW. Each option is capable of meeting stringent Canadian and international safety standards with a considerable margin for LLW. The ability of the repository concepts to accept ILW was assessed qualitatively.

 

The deep rock vault option is most preferred considering technical/safety factors and environment/social factors. The low permeability of the host rock is expected to result in deep repository concepts meeting radiological protection criteria. Surface repository concepts would require additional analysis to ascertain the degree to which ILW could be managed. An examination of the environmental protection feasibility of the options showed that all potential adverse effects could be mitigated or managed using known and proven methods. Status quo is most favourable for economic factors (i.e., lowest cost).

5.   Safety Regulations

Key concepts for long-term management are based on containment and isolation of the waste, in accordance with the CNSC Regulatory Guide G-320 (CNSC06a). The guide states that, “containment can be achieved through a robust design based on multiple barriers providing defence-in-depth. Isolation is achieved through proper site selection and, when necessary, institutional controls to limit access and land use”.

5.1            Ontario

Some regulatory requirements from the provincial jurisdiction, in particular Ontario’s Occupational Health and Safety Act (OHSA90) and conventional occupational safety standards, including those pertaining to mining aspects of the DGR, are applicable to the DGR workers.

5.2            International

The development and safety of deep geologic repositories has been the subject of international attention by the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency (NEA) for many years.

A number of technical documents are available that provide guidance on best international practices with respect both to achieving safety, and on the demonstration of safety.  Particular international documents relevant to the development and safety of the DGR are listed in Table 5-1 below.

Structured approach to safety assessment developed under the IAEA Improvement of Safety Assessment Methodologies (ISAM) program (IAEA04a) has been followed in the DGR program, as suggested in the CNSC Regulatory Guide G-320 (CNSC06a).

Specific guidance on radiation protection criteria and their application for disposal of long-lived radioactive waste has been provided by the International Commission on Radiological Protection (ICRP) in ICRP-81 (ICRP00). This guidance has been taken into account in the CNSC Regulatory Guide G-320, and has been taken into account in the DGR project.

Table 5-1 International Guidance Applicable to the DGR

Document No.

Title

IAEA SF-1

IAEA Safety Fundamentals: Fundamental Safety Principles (IAEA06a)

IAEA WS-R-4

Geological Disposal of Radioactive Waste – Safety Requirements (IAEA06b)

IAEA DS-334

Geological Disposal of Radioactive Waste (draft) (IAEA07)

IAEA DS-354

Disposal of Radioactive Waste (draft) (IAEA06c)

IAEA DS-355

The Safety Case and Safety Assessment for Radioactive Waste Disposal (draft) (IAEA08a)

IAEA SS 111-F

The Principles of Radioactive Waste Management (IAEA95)

IAEA SS 111-G-4 1

Siting of Geological Disposal Facilities (IAEA94)

IAEA-ISAM-1

Safety Assessment Methodologies for Near Surface Disposal Facilities (IAEA04a)

NEA 3679

Postclosure Safety Case for Geological Repositories (NEA04)

ICRP 81

Radiation Protection Recommendations as Applied to the Disposal of Long-Lived Solid Radioactive Waste (ICRP00)

5.3  Safety Objective

According to IAEA guidance (IAEA06b), geological disposal of radioactive waste is aimed at:

·         Containing the waste until most of the radioactivity, and especially that associated with shorter-lived radionuclides, has decayed;

·         Isolating the waste from the biosphere and to substantially reduce the likelihood of inadvertent human intrusion into the waste;

·         Delaying any significant migration of radionuclides to the biosphere until a time in the far future when much of the radioactivity will have decayed; and

·         Ensuring that any levels of radionuclides eventually reaching the biosphere are such that possible radiological impacts in the future are acceptably low.

Consistent with the above IAEA guidance and the NSCA (subparagraph 9(a) (i)), the overall safety objective of the DGR is:

To provide safe long-term management of low and intermediate level waste without posing unreasonable risk to the environment or health and safety of humans.

Conclusions on whether the overall safety objective is met by the DGR can be made by comparing the predicted performance of the DGR with performance criteria based on regulatory requirements. To allow such comparisons, specific design and safety criteria have been established for the DGR.

In addition, long-term safety of the DGR during the postclosure period is judged through how well the following safety functions are fulfilled by the repository after decommissioning:

·         Isolation of the waste away from the biosphere; and

·         Long-term containment of the waste.

The overall safety objective can be concluded to be met if it can be demonstrated that:

·         The DGR has been designed for safe construction, operation and decommissioning, incorporating good engineering practices and use of known technologies; and

·         Experience with facilities similar to the DGR demonstrates a strong operational record.

5.4            Design and Safety Criteria

Specific criteria have been established for the DGR design and safety, based on either the regulations under the NSCA or guidance from federal and provincial authorities, and international guidance. These criteria are used in DGR design and in confirming conclusions on DGR safety, reached through various assessments during both the preclosure and postclosure periods.

5.4.1  Design Criteria

Table 5-2 lists the regulations and major standards and codes applicable to the design and operation of the DGR.

Table 5-2 Regulations, Standards and Codes

Code or Standard

Applicability

Management System Requirements for Nuclear Power Plants CSA N286-05

Management systems

National Building Code of Canada

Surface buildings and structures, fire protection

National Fire Code of Canada

Fire protection systems

Occupational Health and Safety Act (Ontario) – Construction Projects Regulations (Reg. 213/91)

Surface buildings and structures

Occupational Health and Safety Act (Ontario) – Industrial Establishments Regulations R.R.O. 1990 (Reg. 851)

Surface buildings and structures

Occupational Health and Safety Act (Ontario) – Mines and Mining Plants Regulations R.R.O. 1990 (Reg. 854)

Shafts, hoists, repository, fire protection, ventilation requirements

General Requirements for Pressure-Retaining Systems and Components in CANDU Nuclear Power Plants/Material Standards for Reactor Components for CANDU Nuclear Power Plants CSA N285-08

Boiler, Pressure Vessel, and Pressure Piping Code CSA B51

Pressurized systems

Non-Rail-Bound Diesel-Powered Machines for Use in Non-Gassy Underground Mines CSA M424.2

Non-rail-bound diesel-powered Machines

Concrete Materials and Methods of Concrete Construction/Test Methods and Standard Practices for Concrete CSA-A23.1 and CSA – A23.2

Surface buildings and structures

Design of Steel Structures CSA S16

Surface buildings and structures

OPG Radiation Protection Requirements Nuclear Facilities (OPG01b)

Radiation zoning and protection

Ontario Electric Safety Code

Electrical systems

Workplace Electrical Safety CSA Z462

Electric arc flash

Use of Electricity in Mines CSA-M421

Lightning protection

Installation Code for Lightning Protection Systems CSA B72-M87

Lightning protection

5.4.2  Long Term Safety of the DGR

The postclosure safety of the repository is quantitatively assessed through considering a range of potential future scenarios. These scenarios include the expected evolution of the DGR system with time, and the potential impacts of low-probability events leading to degradation and loss of containment. Potential effects are considered for both humans and the environment.

The safety assessment has been undertaken using the following approach:

1.      the assessment context is defined, documenting the high-level assumptions and the constraints, notably regulatory requirements and assessment timeframe;

2.      the system is described, including the features relevant to postclosure safety;

3.      a range of potential future scenarios is systematically identified;

4.      conceptual and mathematical models are developed for these scenarios; and

5.      the scenarios are analyzed and the results are assessed regarding the performance of the system, and its overall robustness.

Key components of the postclosure safety assessment context are summarised in Table 5-3.

Table 5-3 Components of the Postclosure Safety Assessment

Component

Description

Regulatory Requirements and Guidance

·         Nuclear Safety and Control Act and associated regulations

·         Canadian Nuclear Safety Commission regulatory guidance document G-320 “Assessing the Long Term Safety of Radioactive Waste Management” [7]

·         Canadian Environmental Assessment Agency and Canadian Nuclear Safety Commission Guidelines for the preparation of the EIS for the DGR

Endpoints

·         Radiation dose to humans

·         Environmental concentrations of radionuclides and non-radioactive species

Criteria

Numerical criteria have been approved by the CNSC for the following [7]:

·         radiation dose limits to prevent impact on humans

·         no-effect concentration limits of radionuclides in environment to prevent impact on non-human biota

·         concentration limits for non-radioactive elements in various environmental media to prevent impact on humans and the environment

Timeframe

·         1 million year baseline

·         Encompasses the period over which the maximum impacts are expected to occur

 

6.   Socio-Economic Aspects of the DGR Project

6.1            Public Participation and Aboriginal Engagement

Many stakeholders were included the EA consultation process:

  • General public
  • Elected representative (federal, provincial, municipal,)\Governments Departments and agencies
  • Aboriginal peoples
  • Chamber  of Commerce/Business Improvement Area Groups
  • Nuclear Industry representatives/unions, etc.

 

6.2            Methods of Engagement for stakeholders consultation

Many engagement tools were used to encourage public and stakeholder participation:

·       DGR Mobile Exibit

·       DGR Open Houses

·       Websites

·       DGR Community Consultation Advisory Group (Public Attitude Research)

·       Speaking Engagements etc.

 

6.3            Engagement Strategies (Site Preparation and Construction and Operations Phases)

 

  • Provided key stakeholders and the general public with information and opportunities to
  • Discuss key DGR activities, milestones and decisions
  • Discuss results of the follow-up monitoring program and any undertakings from the regulatory approvals process.
  • Engaged those living within close proximity of the Bruce nuclear site regarding any anticipated effects on the environment and health and safety of persons, and any activities.

 

6.4            DGR Public Attitude Research Key Findings

 

  • Nuclear issues are not “top of mind” among the respondents in the Municipality of Kincardine and the surrounding municipalities.
  • A clear majority of people from the Municipality of Kincardine (90%) are confident in the management of radioactive waste of radioactive waste at the WWMF and 83% have confidence in the proposed DGR Project.
  • Respondents are satisfied living in their community (70%) and are largely committed to staying long term (69%).
  • The DGR Project is not expected to change people’s level of commitment to living in the area (92% indicated no change in the Municipality of Kincardine); level of satisfaction with living there (82% indicated no change in the Municipality of Kincardine); or feeling of personal health and safety (79% indicated no change in the Municipality of Kincardine).

 

6.5            Assessment of Economic Effects or Aboriginal

Land Leasing:

  • Likely effects on cottagers leasing lands from Aboriginal peoples and land leasing activities of First Nations people are anticipated to be similar to those economic effects expected for the Municipality of Kincardine  and surrounding municipalities.
  • A beneficial effect on business activity could be anticipated during all DGR Project phases; increased demand for leased cottage property may result due to increased population.
  • Disruption to commercial business, including land leasing activities, is not expected due to nuisance factors or traffic caused by the DGR.
  • No residual adverse effects on tourism industry are anticipated, largely because adverse effects on community character are not expected.
  • No adverse effects are anticipated that would diminish the attractiveness of the area or establish a stigma.
  • No adverse effects on residential property values are anticipated, largely because no changes in dust, noise or local traffic conditions are expected. Also, the DGR site would not likely be visible from land leased by Aboriginal peoples.
  • Increases in off-site noise during site preparation and construction and decommissioning are low in magnitude and limited in area. Effects on use and enjoyment of property are very localized.

 

6.6            Assessment of Sociological Aspects

 Addressing Social Issues:

  • OPG will continue to work with the local municipality, health and safety providers, local police, emergency medical services and other officials to mitigate any issues related to the DGR Workforce.
  • Mitigation measures to social issues related to the DGR will be community specific, and will be designed and implemented in a cooperative manner.
  • Potential measures include:
  • working with DGR contractors to minimize reliance on a transient        workforce
  • conducting orientation programs for incoming workers
  • modifying traffic management plans to address specific traffic-related issues
  • Follow-up Public Attitude Researchand ongoing OPG public affairs programs

 

6.7            OPG Community Involvement

·         The DGR is expected to strengthen OPG’s presence in the community; OPG is a positive contributor to community cohesion as seen by local residents

  • OPG’s community programs and contributions will continue to be noticeable to local residents; previous contributions include support for over:

120 local not-for-profit initiatives and 75 community events and clubs or facilities each year

Kincardine Scottish Festival, Highland Heavy Games, the Bruce County Museum and Cultural Centre, local Food Banks, minor Sports, environmental initiatives and First Lego Leagues

·         The DGR Project was developed in partnership with Kincardine and surrounding Bruce County municipalities, and the Community Partnership program was implemented to continue OPG’s positive presence.

6.8            Economic Aspects of Siting DGR

  • No significant adverse effects on the Socioeconomic Environment are expected.
  • New direct, indirect, and induced employment opportunities and labour income are expected within Kincardine and neighbouring municipalities.
  • Increased municipal revenue is anticipated through property taxes and onetime/annual payments.

6.9            Job Creation

·    Estimated on-site labour work force:

  • Site preparation and construction (~ 6 years): 80 to 200 jobs
  • Operations ( ~ 40 years): 40 jobs
  •  Decommissioning (~ 7 years): 4 to 125 jobs

·    Total employment impact is approximately 25,000 full-time equivalent person-years

  •  27% within the Municipality of Kincardine
  • 24% within the surrounding municipalities
  •  49% within Ontario and beyond

·    Number and distribution of jobs does not suggest communities will experience “boom” or “bust” effects

 

Labour Income by Phase

·         Site Prep and Construction: $612 Million (direct, indirect and induced)

  Percentage of total income by area:

§  Municipality of Kincardine: 10%

§  Surrounding municipalities: 10%

§  Province and beyond: 80%

·         Operations and Decommissioning: $830 Million (direct, indirect and induced)

 Percentage of total income by area:

§  Municipality of Kincardine: 40-55%

§  Surrounding municipalities: 35-50%

§  Province and beyond: 10%

 

6.10       Anticipated Tax Revenue

·         Bruce Nuclear Site

OPG 2009 Property Tax Payments = $5 million for lands, buildings and structures at the Bruce nuclear site (paid to Bruce County and Municipality of Kincardine)

 ~ $472,200 was for site waste management operations

·         DGR Project

Increased municipal revenue is expected to occur as a result of on- and off-site development:

§   Property taxes on- and off-site/revenue from land development

§  Revenue from land improvements (e.g., fees in lieu of building permits, development charges)

 

7.      Summary

There is a need to sequester LLW and ILW because, by definition, they exceed activity levels prescribed in regulations.  It is our opinion that the DGR for LLW and ILW is safe and does not expose the public to any risk. Any radiation levels in or near the facility would be negligible, several orders of magnitude less, compared to what exists in nature.  The word the material is classified as nuclear waste, there appears to be a general confusion amongst the public between LLW, ILW and spent nuclear fuel. This is not a depository for spent nuclear fuel.

The design meets all domestic and international requirements, and has large safety margins since it is based on the ALARA principle. 

There is a supportive host community and there are positive socioeconomic aspects associated with the facility.  In terms of industry there will be substantial employment during the construction phase as well as permanent employment for 40 years. 

In our opinion this will be one of the most benign industrial facilities in the vicinity of the Great Lakes.


8.    References

 

[1]          NWMO, DGR Preliminary Safety Report, 00216-SR-01320-00001, R000, March 2011.

[2]          Information from NuclearSafety.gc.ca

[3]          Ghiassi-nejad M, Mortazavi SM, Cameron JR, Niroomand-rad A, Karam PA., Health Physics, 2002 Jan;82(1):87-93

[4]          Government of Canada, Radiation Protection Regulations SOR/2000-203, July 2013

[5]          Canadian Standards Association (CSA). 2008. Management of Low and Intermediate Level Radioactive Waste. CSA N292.3.

[6]          Canadian Nuclear Safety Commission (CNSC). 2000. Nuclear Substances and Radiation Devices Regulations. SOR/2000-207.

[7]          Ontario Power Generation (OPG). 2004. Final Report of Independent Assessment of Long-Term Management Options for Low and Intermediate Level Wastes at OPG's Western Waste Facility.

[8]          Canadian Nuclear Safety Commission (CNSC). 2006. Regulatory Guide G-320: Assessing the Long Term Safety of Radioactive Waste Management.

[9]          Ontario Power Generation (OPG). 2011. Deep Geologic Repository for Low and Intermediate Level Waste - Preliminary Safety Report. 00216-SR-01320-00001 R000.

 

 

 

 



[1] Note that the maximum allowable dose to a member of the general public is actually less than the natural background, so this number should be interpreted to mean maximum allowable incremental dose.

[2] Note:  This is the same data that is given in Table 2.2 but it is plotted graphically.  There is not enough dynamic

range to show the difference between the DGR and, for example, Ottawa. Figure 2.3 shows this difference better.