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F-3.4.1-1 Rev 6
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UKP-GW-GL-795 Plan.doc
F-3.4.1-1 Rev 6
Westinghouse Non-Proprietary Class 3
UK AP1000TM NPP Decommissioning Plan
UKP-GW-GL-795, Revision 0
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Revision History
UK AP1000 NPP Decommissioning Plan
REVISION HISTORY
Revision
0
Description of Changes
Original Issue.
Trademark Notices
AP1000TM is a registered trademark in the United States of Westinghouse Electric Company LLC, its
subsidiaries and/or its affiliates. This mark be used and/or registered in other countries throughout the
world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of
their respective owners.
UKP-GW-GL-795
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Table of Contents
UK AP1000 NPP Decommissioning Plan
TABLE OF CONTENTS
Section
Title
Page
REVISION HISTORY.................................................................................................................................. ii LIST OF TABLES ....................................................................................................................................... vi LIST OF FIGURES .................................................................................................................................... vii LIST OF ACRONYMS ............................................................................................................................. viii 1.0 INTRODUCTION ........................................................................................................................... 1 2.0 Regulatory Background ................................................................................................................... 2 2.1 UK Decommissioning Policy ............................................................................................. 2 2.2 Nuclear Site Licence Conditions ........................................................................................ 3 2.3 Safety Assessment Principles for Nuclear Facilities .......................................................... 4 2.4 Radioactive Substances Regulation – Environmental Principles ....................................... 5 2.5 Construction (Design & Management) Regulations ........................................................... 6 2.6 International Nuclear Decommissioning Guidance .......................................................... 10 3.0 Role of Westinghouse and Utility in Decommissioning................................................................ 12 4.0 Principles Underpinning Design .................................................................................................... 13 5.0 4.1 Design Principles .............................................................................................................. 13 4.2 Operation and Maintenance Principles ............................................................................. 17 4.3 Decommissioning Principles ............................................................................................ 19 4.4 Learning from Experience Principles ............................................................................... 22 4.5 ALARP Considerations .................................................................................................... 23 4.6 Key Design Features ......................................................................................................... 24 Decommissioning Logistics ........................................................................................................... 26 5.1 Decommissioning Sequence ............................................................................................. 26 5.2 Shielding and Containment ............................................................................................... 47 5.3 Progressive Hazard Reduction .......................................................................................... 49 5.4 Preventing Early Foreclosure of Options.......................................................................... 52 UKP-GW-GL-795
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Table of Contents
UK AP1000 NPP Decommissioning Plan
TABLE OF CONTENTS (cont.)
Section
6.0 Title
5.5 Decommissioning Technology ......................................................................................... 53 5.6 Feasibility Reviews ........................................................................................................... 54 5.7 Strategy for Safety Systems .............................................................................................. 57 Timing of Decommissioning ......................................................................................................... 59 6.1 7.0 8.0 10.0 Timing of Decommissioning ............................................................................................ 59 Decommissioning Hazards and Challenges ................................................................................... 62 7.1 Experience of Decommissioning ...................................................................................... 62 7.2 Decommissioning Hazards ............................................................................................... 62 7.3 Hazard Control Measures ................................................................................................. 66 7.4 Manual and Remote Tasks................................................................................................ 66 7.5 Industrial Safety Hazards .................................................................................................. 67 Plant Status before Decommissioning ........................................................................................... 69 8.1 9.0 Page
Plant Status at End of Operational Life ............................................................................ 69 Disposability Assessment and Decommissioning ......................................................................... 72 9.1 Waste Streams .................................................................................................................. 72 9.2 Waste Routes to Interim Storage ...................................................................................... 72 9.3 Waste Stream Sensitivity to Decommissioning Processes ............................................... 76 9.4 Impact of Decontamination on Primary Waste Volumes ................................................. 77 9.5 Generation of Secondary Wastes ...................................................................................... 77 9.6 Use of BAT & Waste Management Hierarchy in Decommissioning ............................... 78 9.7 HLW and ILW Transport and Storage ............................................................................. 79 Decommissioning Plans and Programmes ..................................................................................... 83 10.1 Decommissioning Plans and Programme ......................................................................... 83 10.2 Disposability and Government Policy .............................................................................. 86 10.3 Decommissioning Lifecycle ............................................................................................. 86 UKP-GW-GL-795
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UK AP1000 NPP Decommissioning Plan
TABLE OF CONTENTS (cont.)
Section
11.0 Title
Page
References...................................................................................................................................... 88 UKP-GW-GL-795
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Table of Contents
UK AP1000 NPP Decommissioning Plan
LIST OF TABLES
None.
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Table of Contents
UK AP1000 NPP Decommissioning Plan
LIST OF FIGURES
Figure 5-1
Outline Decommissioning Sequence ................................................................................... 30 Figure 5-2
Decontamination Techniques [Reference 41] ...................................................................... 55 Figure 5-3
Decommissioning Techniques [Reference 41] .................................................................... 56 Figure 7-1
An Example of Contamination Mapping (Oskarshamn 3) .................................................. 65 Figure 9-1
AP1000 Plant Waste Facilities (No. 4, 5, 21, 22, 23, 24, 25, 26, 27, 29, 31)
and Transport Routes ........................................................................................................... 75 Figure 10-1
Outline Decommissioning Programme ................................................................................ 85 UKP-GW-GL-795
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List of Acronyms
UK AP1000 NPP Decommissioning Plan
LIST OF ACRONYMS
ALARP
BAT
BWR
CCTV
CDM
CEC
CVS
EA
ER
ESF
EU
FSC
GDA
GDF
HEPA
HHISO
HLW
HRGS
HSE
HVAC
IAEA
ILW
LC
LLW
LLWR
LFE
MPC
ND
NEA
NIA
NPP
NRC
OECD
PCCWST
PCS
POCO
PWR
QQR
R&D
REP
RO87
RPV
RWMD
SAFSTOR
SAP
SCV
SGS
SPHSE
SQEP
As Low As Reasonably Practicable
Best Available Techniques
Boiling Water Reactor
Closed Circuit Television
Construction (Design and Management) Regulations 2007
Cavity Enclosure Container
Chemical and Volume Control System
Environment Agency
UK AP1000 Environment Report [Reference 15]
Engineered Safety Features
European Union
Final Site Clearance
Generic Design Assessment
Geological Disposal Facility
High Efficiency Particulate Air
Half-height ISO containers
High Level Waste
High Resolution Gamma Spectroscope (a waste package assay instrument)
Health and Safety Executive
Heating, Ventilation, and Air Conditioning
International Atomic Energy Agency
Intermediate Level Waste
Licence Conditions
Low Level Waste
Low Level Waste Repository
Learning from Experience
Multi-Purpose Canister
Nuclear Directorate
Nuclear Energy Agency
Nuclear Installations Act 1965
Nuclear Power Plant
United States Nuclear Regulatory Commission
Organisation for Economic Co-operation and Development
Passive Containment Cooling Water Storage Tank
Passive Containment Cooling System
Post Operational Clean Out
Pressurised Water Reactor
Quinquennial Review
Research and Development
Radioactive Substances Regulation Environmental Principle
Regulatory Observation RO-AP1000-087 on Decommissioning
Reactor Pressure Vessel
Radioactive Waste Management Directorate
Safe Storage (nuclear plant in retirement)
Safety Assessment Principle
Secondary Containment Vessel
Steam Generator System
Self-Priming High Solids Epoxy
Suitably Qualified and Experienced Personnel
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List of Acronyms
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LIST OF ACRONYMS (cont.)
SSC
VVM
WENRA
WSS
Systems, Structures, and Components
Vertical Ventilated Module
Western European Nuclear Regulators Association
Solid Radwaste System
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1.0 Introduction
1.0
UK AP1000 NPP Decommissioning Plan
INTRODUCTION
UK nuclear regulators reviewed Westinghouse decommissioning plans as part of Step 4 of
the Westinghouse AP1000TM nuclear power plant (NPP) Generic Design Assessment (GDA).
They identified many decommissioning aspects that the GDA submission needed to address.
These were originally captured in the revised response to regulatory observation
RO-AP1000-087 (hereafter abbreviated as RO87), which was submitted on the 28th
December 2010.
The information provided in RO87 was used in this document (UKP-GW-GL-795) to
demonstrate that:
1. It is feasible to decommission an AP1000 plant using current technology
2. Decommissioning issues were appropriately considered in the overall design
The utility operator is expected to prepare their own decommissioning plan and update it
from time to time.
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2.0 Regulatory Background
2.0
REGULATORY BACKGROUND
2.1
UK Decommissioning Policy
UK AP1000 NPP Decommissioning Plan
The UK government policy on decommissioning [Reference 1] describes decommissioning
as follows:
“Decommissioning is a staged process through which a nuclear facility, at which normal
operations have finally ceased, is taken out of service, including full or partial dismantling of
buildings and their contents. It may include other operations such as the decontamination of
buildings which are not to be dismantled and the remedial treatment or restoration of the
land under and around the facility. The objective of decommissioning is to remove the hazard
the facility poses progressively, giving due regard to security considerations, the safety of
workers and the general public and protecting the environment, while in the longer term
reducing the number of sites and acreage of land which remain under regulatory control.”
The policy outlines the government’s expectations of nuclear facilities operators regarding
decommissioning. The policy specifies that operators must:

Produce and maintain a decommissioning strategy and plan for their site

Consider all relevant factors in the decommissioning strategy, and transparently assess
and present them, supported by objective information and arguments

Address the future use of sites in good time and make decisions which take both local
factors and the wishes of the local community into account

Avoid creating radioactive wastes in forms that may foreclose options for safe and
effective long-term waste management options

Consider the benefits of delaying operations to allow radioactive decay to occur

Minimise the volumes of radioactive wastes that are created

Avoid creating wastes until long-term waste management solutions have been identified

Make Progressive and substantial reductions in radioactive discharges to the marine
environment through the rigorous application of As Low As Reasonably Practicable
(ALARP) and Best Practical Means. (Since publication of the UK decommissioning
policy, Best Practical Means has been superseded by BAT.)

Review their strategies when changes in circumstances, including relevant Government
policies, make this necessary

Subject strategies to regular periodic reviews, at least every five years, by the Health and
Safety Executive (HSE) in consultation with the Environment Agency (EA)

Ensure that their decommissioning work is adequately funded

Maintain the knowledge base, records and skills necessary to their decommissioning
operations and management of associated wastes

Identify, implement and share best practice
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UK AP1000 NPP Decommissioning Plan
The decommissioning strategy should consider relevant developments in UK radioactive
waste management policy. The policy currently assumes that:
1. Spent fuel reprocessing will not be available to new nuclear power stations
[Reference 36]
2. National repositories will be developed to accept intermediate level waste (ILW) in 2040
and high level waste (HLW) waste in 2075 [Reference 23].
To ensure that operators’ decommissioning strategies remain soundly based as circumstances
change, they will be reviewed quinquennially by the HSE, who will consult with the EA.
2.2
Nuclear Site Licence Conditions
Under the Nuclear Installations Act 1965 (NIA), as amended [Reference 2], a nuclear power
plant requires a site licence granted by the HSE. This licensing function is administered on
the HSE’s behalf by its Nuclear Directorate (ND). The HSE has developed a standard suite of
licence conditions (LCs) that are attached to all nuclear site licences. Many of these
conditions apply throughout the whole lifecycle of the facility from initial design and
installation through operation to final decommissioning.
However, licence condition LC35 is specifically concerned with decommissioning and states:
35
Decommissioning
(1)
The Licensee shall make and implement adequate arrangements for the
decommissioning of any plant or process which may affect safety.
(2)
The Licensee shall make arrangements for the production and implementation of
decommissioning programmes for each plant.
(3)
The Licensee shall submit to the Executive for approval such part or parts of the
aforesaid arrangements or programmes as the Executive may specify.
(4)
The Licensee shall ensure that once approved no alteration or amendment is made to
the approved arrangements or programmes unless the Executive has approved such
alteration or amendment.
(5)
The aforesaid arrangements shall where appropriate divide the decommissioning
into stages. Where the Executive so specifies the Licensee shall not commence nor
thereafter proceed from one stage to the next of the decommissioning without the
consent of the Executive. The arrangements shall include a requirement for the
provision of adequate documentation to justify the safety of the proposed
decommissioning and shall where appropriate provide for the submission of this
documentation to the Executive.
(6)
The Licensee shall, if so directed by the Executive where it appears to them to be in
the interests of safety, commence decommissioning in accordance with the aforesaid
arrangements and decommissioning programmes.
(7)
The Licensee shall, if so directed by the Executive, halt the decommissioning of a
plant and the Licensee shall not recommence such decommissioning without the
consent of the Executive.
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2.0 Regulatory Background
UK AP1000 NPP Decommissioning Plan
It is important that when a nuclear facility reaches the end of its operational life it is
decommissioned in a safe and controlled manner and not left to pose a hazard for current
and future generations. The purpose of this Condition is therefore to require the Licensee to
have adequate arrangements for the safe decommissioning of its facilities. It also gives the
HSE the power to direct the Licensee to commence decommissioning of any plant or facility
to prevent it being left in a dangerous condition or to ensure decommissioning takes place in
accordance with any national strategy. The Condition also gives the HSE the power to halt
any decommissioning activity if HSE has concerns about its safety.
2.3
Safety Assessment Principles for Nuclear Facilities
The HSE is responsible for regulating nuclear safety, including safely managing, conditioning
and storing radioactive waste on nuclear licensed sites.
The HSE have developed Safety Assessment Principles (SAPs) [Reference 3] that the ND
uses to ensure that a consistent and uniform approach is applied when assessing nuclear
facility safety cases. The HSE requires Licensees to provide safety cases demonstrating that
their facilities operate safely at all stages of their life, including decommissioning. A safety
case for operation must be prepared before starting operation; and a safety case for
decommissioning must be prepared before conducting decommissioning.
The SAPs are used for regulatory assessment throughout the lifecycle of an activity on a
nuclear licensed site. In the assessment process dutyholder submissions are examined to
determine that:


Risks are ALARP
Appropriate attention has been given to both:
–
–
Aspects important to safety
Radioactive waste management and decommissioning
Paragraphs 684 to 739 of the SAPs [Reference 3] are devoted to decommissioning. These
paragraphs contain eight principles related to decommissioning:
DC.1
Facilities should be designed and operated so that they can be safely
decommissioned.
DC.2
A decommissioning strategy should be prepared and maintained for each site and
should be integrated with other relevant strategies.
DC.3
Decommissioning should be carried out as soon as is reasonably practicable taking
relevant factors into account.
DC.4
A decommissioning plan and programme should be prepared and maintained for
each nuclear facility throughout its life-cycle to demonstrate that it can be safely
decommissioned.
DC.5
The facility should be made passively safe before entering a care and maintenance
phase.
DC.6
Throughout the whole life-cycle of a facility the documents and records that might be
required for decommissioning purposes should be identified, prepared, updated and
retained.
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2.0 Regulatory Background
2.4
UK AP1000 NPP Decommissioning Plan
DC.7
Organisational arrangements should be established and maintained to ensure safe
and effective decommissioning of facilities.
DC.8
The safety management system should be periodically reviewed and modified as
necessary prior to and during decommissioning.
Radioactive Substances Regulation – Environmental Principles
The EA is responsible in England and Wales for regulating both:


Discharges to the environment
Disposal of radioactive waste on or from nuclear licensed sites
The EA has produced a set of Radioactive Substances Regulation Environmental Principles
(REPs) [Reference 4]. These REPs form a consistent and standardised framework for the
technical assessments and judgements that the EA make when regulating organisations under
the Radioactive Substances Act 1993 [Reference 5].
Many of these REPs apply throughout the lifecycle of the facility from initial design and
installation, through operation to final decommissioning, and many refer to applying BAT.
Paragraphs 148 to 159 of the REPs address decommissioning [Reference 4] and contain the
following principles.
Principle DEDP1 – Decommissioning Strategy
150
Each site should have a decommissioning strategy that is updated and refined at
appropriate intervals.
Principle DEDP2 – Decommissioning Plan
152
There should be a decommissioning plan for each facility and this should be updated
and refined throughout its operating life and during decommissioning.
Principle DEDP3 – Considering Decommissioning during Design and Operation
154
Facilities should be designed, built and operated using the best available techniques
to minimise the impacts on people and the environment of decommissioning
operations and the management of decommissioning wastes.
Principle DEDP4 – Discharges during Decommissioning
156
Aerial or liquid radioactive discharges to the environment during decommissioning
should be kept to the minimum consistent with the decommissioning strategy for the
site.
Principle DEDP5 – Legacy Wastes
158
UKP-GW-GL-795
Decommissioning strategies and plans should provide for the timely
characterisation, retrieval, conditioning and packaging of legacy radioactive wastes.
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2.0 Regulatory Background
2.5
UK AP1000 NPP Decommissioning Plan
Construction (Design & Management) Regulations
The Construction (Design and Management) Regulations 2007 (CDM) [Reference 6]
integrate health and safety into the management of the project and encourage everyone
involved to work together to:

Improve planning and managing projects from the beginning

Identify hazards early on so that they can be eliminated or reduced at either the design or
planning stage; and so that the remaining risks can be properly managed

Target effort where it can do the most good in terms of health and safety

Discourage unnecessary bureaucracy
These regulations focus attention on planning and management throughout the entire lifecycle
of the plant. They aim to ensure that health and safety considerations are treated as an
essential, but normal, part of a project’s development; starting with its design, then through
construction and commissioning to operation and to eventual decommissioning
[Reference 7].
CDM applies to most common building, civil engineering and engineering construction work,
if the construction work is expected to either:


Last longer than 30 days
Involve more than 500 person-days of construction work
Regulation 2, Interpretation – provides the following clarification of the term “construction
work”:
“Construction work” means the carrying out of any building, civil engineering or
engineering construction work and includes:
a)
the construction, alteration, conversion, fitting out, commissioning, renovation,
repair, upkeep, redecoration or other maintenance (including cleaning which
involves the use of water or an abrasive at high pressure or the use of corrosive or
toxic substances), de-commissioning, demolition or dismantling of a structure;
b)
the preparation for an intended structure, including site clearance, exploration,
investigation (but not site survey) and excavation, and the clearance or preparation
of the site or structure for use or occupation at its conclusion;
c)
the assembly on site of prefabricated elements to form a structure or the disassembly
on site of prefabricated elements which, immediately before such disassembly,
formed a structure;
d)
the removal of a structure or of any product or waste resulting from demolition or
dismantling of a structure or from disassembly of prefabricated elements which
immediately before such disassembly formed such a structure; and
e)
the installation, commissioning, maintenance, repair or removal of mechanical,
electrical, gas, compressed air, hydraulic, telecommunications, computer or similar
services which are normally fixed within or to a structure,
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2.0 Regulatory Background
UK AP1000 NPP Decommissioning Plan
Regulation 29 refers specifically to demolition or dismantling, and states:
29(1)
The demolition or dismantling of a structure, or part of a structure, shall be planned
and carried out in such a manner as to prevent danger or, where it is not practicable
to prevent it, to reduce danger to as low a level as is reasonably practicable.
29(2)
The arrangements for carrying out such demolition or dismantling shall be recorded
in writing before the demolition or dismantling work begins.
Nuclear decommissioning operations will comprise a number of separate projects or work
packages that fall within the remit of CDM. CDM management practices and risk prevention
principles will be applied to each decommissioning project, and this will become the primary
mechanism to ensure risks associated with each activity are identified, prevented or
controlled.
CDM defines roles for the various organisations involved in construction projects
[Reference 7]:

Client
The client is an organisation or individual for whom the project is carried out. The client
has certain responsibilities under CDM. The client may opt to employ a competent
advisor to help ensure that their duties are carried out. The client is responsible for
confirming the competency of such an advisor should they opt to employ one.
The client should avoid specifying how the work should be undertaken, as this would
make them responsible as designers under the regulations. Where a CDM co-ordinator is
appointed, the client must ensure that they are allowed sufficient time and resources
following their appointment to carry out their duties.
The client will be the utility that owns and operates the AP1000 NPP. In
decommissioning an AP1000 plant, it is expected that the client will employ a competent
CDM co-ordinator and a specialist nuclear decommissioning contractor as the Designer
of decommissioning works, the Principal Contractor or the Contractor.

CDM Co-ordinator
The CDM co-ordinator will act on behalf of the client and advise on health and safety
management matters. They will help the client to prepare the health and safety file.
Whereas during the construction phase, the Health & Safety file would be prepared using
design documentation, the Decommissioning Health & Safety File should include:
–
The construction Health & Safety File
–
Documentation pertaining to plant modifications
–
Documentation pertaining to incidents during the operating lifetime of the plant
which may be considered to cause a hazard during the decommissioning process.
CDM co-ordinators can be appointed independently of any other role on the project team,
or they may combine this work with another role, for example, project manager, designer
or principal contractor. Where the role is combined, it is crucial that the CDM
co-ordinator has sufficient independence to carry out their tasks effectively.
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2.0 Regulatory Background

UK AP1000 NPP Decommissioning Plan
Designers
The designer is a person responsible for specifying the design of the facility. These
specifications may include drawings, or schedules of equipment. The designer’s decisions
will influence the way that the facility is operated and maintained. In addition, the
designer is responsible for ensuring that the facility can be safely decommissioned with
the avoidance of unresolved legacy issues. To do this, the designer has a duty to eliminate
hazards and reduce risks during the design process, and to provide information about
remaining risks. In carrying out their work, designers are required to avoid foreseeable
risks to the health and safety of those involved or affected by the construction, use,
maintenance and demolition of the structure. In doing so, they must eliminate hazards
which may give rise to risk and reduce the remaining risks from any hazards. Both these
elements must be done “so far as is reasonably practicable, taking due account of other
relevant design considerations.” Design risk assessments should be carried out to identify
where improvements to the design can be made to minimise risk in the future.
The designer is also required to take all reasonable steps to provide with his design
sufficient information about aspects of the design of the structure or its construction or
maintenance as will adequately assist the CDM co-ordinator to comply with his duties
under CDM, including his duties in relation to the health and safety file.
As a designer of the AP1000 plant, Westinghouse will make provision for the transfer of
knowledge relating to the construction, operation and maintenance of the AP1000 plant
[Reference 15, Ch 1.4]. The knowledge transfer and management arrangements
developed to support operation of the AP1000 plant will provide the foundations of the
arrangements required for decommissioning.

Principal Contractors
The key duty of the principal contractors is to properly plan, manage and co-ordinate
work such that risks are properly controlled. The principal contractors will also ensure
that there is safe working, co-ordination and communication between contractors. They
will ensure that should a Contractor sub-let their awarded work, that the sub-contractor is
suitably qualified and experienced to undertake the awarded work before being allowed
to commence work.

Contractors
Contractors employed during the work will be assessed for their suitability to undertake
the work to be awarded to them. The principal contractor will perform this assessment,
and declare the contractor suitably qualified and experienced, or otherwise. When the
contractor is awarded the work, it is their responsibility to ensure that their personnel are
fully trained such that they are able to safely carry out the work awarded to them. The
Contractors will be required to inform the Principal Contractors should they wish to
sub-let their awarded work.
Because the regulations apply to the entire lifecycle of the plant, ongoing knowledge transfer
is key to complying with the regulations. The Health and Safety File facilitates this
knowledge transfer. With respect to the Health and Safety File clients, designers, principal
contractors, other contractors, and CDM co-ordinators all have the following legal duties:

CDM co-ordinators must prepare, review, amend or add to the file as the project
progresses, and give it to the client at the end of project
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
Clients, designers, principal contractors and other contractors must supply the
information necessary for compiling or updating the file

Clients must keep the file to assist with future construction work

Everyone providing information should ensure that it is both accurate and provided
promptly
The Health and Safety File should contain the information needed to allow future
construction work including cleaning, maintaining, altering, refurbishing and demolishing to
be carried out safely. Information in the file should alert those carrying out such work to
risks, and should help them to decide how to work safely. The file should be readily
accessible to all parties, and should be kept up to date after any relevant work or surveys.
To comply with CDM, each decommissioning project or work package will require a Health
and Safety File to be developed. The Health and Safety File should compliment the detailed
decommissioning plan and decommissioning safety case. The client and CDM co-ordinator
should agree to the file’s scope, structure and format at the start of a project. To allow
personnel carrying out the work to identify and address likely risks, the Health and Safety
File will include sufficiently detailed information about the following health and safety
issues:

A brief description of the work to be performed

Any residual hazards that remain and how they have been dealt with (for example
radiological surveys or other information concerning radioactive contamination,
contaminated land, or buried services)

Key structural principles (for example, bracing, sources of substantial stored energy –
including pre- or post-tensioned members) and safe working loads for floors and roofs,
particularly where these may preclude placing scaffolding or heavy machinery

Hazardous materials used (for example residual chemicals, decontamination chemicals,
or special coatings which should not be burnt off)

Information about removing or dismantling the installed plant and equipment (for
example any special arrangements for lifting, order or other special dismantling
instructions)

Health and safety information about equipment supplied for cleaning or dismantling the
structure

The nature, location and markings of significant services, including underground cables,
radiation monitoring equipment or fire-fighting services

Information and as-built drawings of the structure, its plant and equipment (for example
radiological area classification, means of safe access to and from service voids, fire
doors, and compartmentalisation)
Implementing CDM will ensure that the health and safety issues associated with each
decommissioning project or work package will be addressed before the work commences.
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2.6
International Nuclear Decommissioning Guidance
2.6.1
International Atomic Energy Agency (IAEA)
The IAEA’s Statute authorises the Agency to establish safety standards to protect health and
minimise danger to life and property. The IAEA must use these standards in its own
operations, and a State can apply them through its own regulatory provisions for nuclear and
radiation safety.
IAEA safety standards establish an essential basis for safety of all nuclear installations and
also cover decommissioning. The IAEA provides many documents related to nuclear safety
and technical guidance that addresses decommissioning. In later sections, this
decommissioning plan refers to many of the IAEA documents [Reference 16 – 22, 27].
2.6.2
Western European Nuclear Regulators Association (WENRA)
WENRA is a network of Chief Regulators of European Union (EU) countries with nuclear
power plants, Switzerland and other interested European countries that were granted
Observer status.
WENRA’s main objectives are to:

Develop a common approach to nuclear safety

Provide an Independent capability to examine nuclear safety in applicant countries

Be a network of chief nuclear safety regulators in Europe to exchange experience and
discussing significant safety issues.
WENRA has produced many safety reference levels for decommissioned facilities that are
intended to provide a good basis for future harmonisation on a European level
[Reference 29]. These safety reference levels are addressed in chapters on:




Safety Management
Decommissioning Strategy and Planning
Conduct of Decommissioning
Safety Verification
Decommissioning safety reference levels mainly address the radiological hazards resulting
from activities associated with facility decommissioning primarily decommissioning after a
planned shutdown. However, WENRA advises that non-radiological hazards arising during
decommissioning activities should be given due consideration during the planning process
and in the risk analyses as far as they may influence the radiological hazards or risks.
2.6.3
Nuclear Energy Agency (NEA)
The Nuclear Energy Agency (NEA) is a specialised agency within the Organisation for
Economic Co-operation and Development (OECD). The mission of the NEA is to assist its
member countries (including the UK) in maintaining and further developing, through
international co-operation, the scientific, technological and legal bases required for the safe,
environmentally friendly and economical use of nuclear energy for peaceful purposes. The
NEA works closely with the IAEA.
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In 2010 the NEA produced a report offering an overview of approaches and views in
applying experience from decommissioning to the design and licensing of third generation
reactor systems [Reference 30]. The report uses information obtained from a survey of
reactor design organisations, electricity producers and regulatory authorities concerned with
the development and implementation of new reactor systems.
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Utility in Decommissioning
3.0
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ROLE OF WESTINGHOUSE AND UTILITY IN DECOMMISSIONING
In its role as designer, Westinghouse recognises that it must offer a plant design that is
amenable to decommissioning. Utilities that become the owner/operators of the plant are
responsible for the initial decommissioning plans. As a vendor, Westinghouse does not write
a decommissioning plan and, as agreed to earlier with ND, there will not be a formal
decommissioning plan or a decommissioning safety case for GDA.
However, as indicated in the response to technical query TQ AP1000-330 [Reference 8] high
level reference only to management arrangements for decommissioning will be made in GDA
documentation. Knowledge transfer and the management arrangements developed to support
operation are expected to provide the foundations of the arrangements required for the
decommissioning process. However, to ensure that this is so, it will be reviewed as the
operational management arrangements are created, noting that there are 60 operating years
available before the decommissioning phase.
Westinghouse will support the utility in providing decommissioning-related information to
the utility, allowing it to finalise its initial decommissioning plan. Examples of the type of
information that could be provided are:





Construction
Structural integrity
Dismantling procedures
Decommissioning waste volumes
Plant life variables that the utility will need to monitor
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4.0
PRINCIPLES UNDERPINNING DESIGN
4.1
Design Principles
4.1.1
Regulatory Guidance
The AP1000 NPP design both facilitates and minimises decommissioning as well as the
associated waste management operations and costs. The principles underpinning the design
are consistent with



The UK government policy on decommissioning (see Section 2.1)
The HSE Safety Assessment Principles for Nuclear Facilities (see Section 2.2)
The EA Radioactive Substances Regulation – Environmental Principles (see Section 2.3)
HSE Safety Assessment Principle DC.1 states “…facilities should be designed and operated
so that they can be safely decommissioned” [Reference 3]. Adopting this principle ensures
that decommissioning and waste retrieval requirements are incorporated into plant design.
Measures that demonstrate adopting this principle include:

Design measures to minimise activation and contamination, for example selecting
materials around the reactor that prevent large quantities of long-lived radionuclides from
forming (such as using low-cobalt steel)

Physical and procedural methods to prevent the spread of contamination

Control of activation (for example, by controlling primary circuit water chemistry)

Design features that facilitate decommissioning and reduce dose uptake by
decommissioning workers

Consideration of the implications for decommissioning when proposing modifications to
and experiments on the facility

Identification of reasonably practicable changes to the facility to either facilitate or
accelerate decommissioning

Minimising the generation of radioactive waste
Additionally, Safety Assessment Principle RW.2 states “…the generation of radioactive
waste should be prevented or, where this is not reasonably practicable, minimised in terms of
quantity and activity.” This principle may be adopted through specific design provisions,
construction methods, and commissioning, operational and decommissioning arrangements
that either avoid creating radioactive waste or reduce generated radioactive waste to the
minimum level throughout the lifetime of the facility.
In addition to the HSE Safety Assessment Principles, the EA Radioactive Substances
Regulation – Environmental Principles also provide guidance on how decommissioning and
waste minimisation may be incorporated into plant design. Principle DEDP3 “Considering
Decommissioning during Design and Operation” states “…facilities should be designed, built
and operated using the best available techniques to minimise the impacts on people and the
environment of decommissioning operations and the management of decommissioning
wastes”
[Reference 4].
Furthermore,
principle
DEDP4
“Discharges
during
Decommissioning” states “…aerial or liquid radioactive discharges to the environment during
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decommissioning should be kept to the minimum consistent with the decommissioning
strategy for the site.”
The following sections describe how the AP1000 plant design addresses these principles.
4.1.2
Minimising the Generation of Radioactive Waste
The current design of the AP1000 plant is the result of a design philosophy and design
process that emphasised safety and simplicity. The design process used throughout the
development of the AP1000 plant is to create a safe nuclear power plant with costs, radiation
exposures and radioactive discharges as low as reasonably practicable.
The AP1000 European Design Control Document [Reference 9] outlines the principles
underpinning the AP1000 plant design that minimise the creation of radioactive waste during
operations and decommissioning. The AP1000 plant was designed with fewer valves, pipes,
and other components; so, compared to plants of a similar size, less waste will be generated
during both maintenance activities (repair and replacement) and decommissioning. Compared
to similar nuclear power plants, the AP1000 plant has roughly:




50 percent fewer valves
35 percent fewer pumps
80 percent less piping
80 percent fewer heating, ventilation, and air-conditioning (HVAC) systems
Considering these characteristics, the decommissioning phase of the AP1000 plant may be
shorter, and decommissioning activities may produce smaller activated/contaminated material
quantities that should be treated and conditioned for interim storage and final disposal as
radioactive waste. This lends itself to:



Simpler strategies
Shorter decommissioning timescales
Lower requirements for decommissioning phase funding
In addition to the reduced building volumes and plant inventory, many features have been
incorporated into the AP1000 plant design that reduce the production of activated corrosion
products. Such measures include the selection of materials used in the design and controls
imposed on construction. A consequence is reduced operational radiation exposure during the
normal operational life. This results in both reduced radioactive inventory and reduced
residual active material mass during decommissioning.
The only effective way to reduce activation of fixed structures is to control material
composition and attempt to reduce major element concentrations. Each of the materials in and
around the core has a major element concentration that cannot be removed. These include
iron and nickel for steels, and calcium for concrete. Efforts to reduce trace elements are of
little benefit because their contribution to the total is minimal compared to the radioactive
inventory contribution from the major elements.
One case where controlling a trace element is appropriate is steel, where reducing cobalt has
significant benefits. Activation Co-59 produces Co-60, which is the largest contributor to
occupational radiation exposure in pressurised water reactors (PWRs). The reactor internals
will be the most highly activated structures in the plant. The cobalt level in these structures
has been restricted to below 0.05 percent by weight.
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The UK AP1000 Environment Report (ER) offers detailed descriptions of how the AP1000
plant has been designed to minimise radioactive waste generated during operation
[Reference 15]. In particular:

ER Section 2.6 offers further examples of AP1000 plant design decisions related to waste
minimisation, waste generation, and waste disposal that reinforce the concept of safety
through simplicity, ALARP, and BAT.

ER Section 3.2 describes several ways in which the release of radioactive emissions from
the AP1000 plant is reduced at the source.

ER Section 3.3 describes the handling and treatment of gaseous radioactive waste and
offers details of the BAT assessments carried out for the treatment of gaseous radwaste.

ER Section 3.4 describes the handling and treatment of liquid radioactive waste and gives
details of the BAT assessments carried out for the treatment of liquid radwaste.

ER Section 3.5.4 describes how waste minimisation, storage and disposal principles are
incorporated into all AP1000 plant life cycle stages, including design, construction,
operation and decommissioning.

ER Table 3.1-1 “Nuclear BAT Management Factors and AP1000 Plant Features” outlines
the ways in which:
–
–
–
–
4.1.3
Radioactive waste generation is minimised
Resources are used efficiently
Emissions are reduced
Waste is stored in a passively safe and retrievable form
Preventing the Spread of Contamination
The AP1000 plant design provides features that protect against the occurrence of, spread of,
and thus potential personnel exposure to radioactive contamination. Activated corrosion
products are the major contributor to surface contamination. In the AP1000 plant design, the
following steps were taken to limit corrosion products generation.

Injecting zinc acetate into the primary system inhibits general corrosion and primary
water stress corrosion cracking. When adding 10 parts per billion (+/– 5ppb) of zinc,
material corrosion rates can be reduced by a factor of three or more.

To mitigate the effects of activated corrosion products, primary circuit water chemistry
will be maintained at pH levels of 6.9 to 7.4.

Primary and auxiliary materials selected for use in the AP1000 plant have a cobalt
content of less than 0.2 percent by weight. High cobalt stellite alloys are not used in
primary circuit components.
The spread of contamination is also limited by a structural design that incorporates a number
of structural modules. These are composite structures made of concrete-filled steel plate
where the steel plate face forms the walls of rooms. They are easily decontaminated and
prevent water borne contaminants from possibly leaching into the concrete. In the limited
areas where concrete walls are exposed to potential contamination, these walls are coated
with a coating that can be decontaminated. All steel surfaces exposed to potential
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contamination will also have surface finishes that will both prevent penetration and facilitate
decontamination [Reference 9].
Incorporating thicker and larger plates in the spent fuel pond also limits the spread of
contamination. Where possible, this plate becomes the formwork. This eliminates potential
voids between the plates and the concrete, which in turn, eliminates potential unsupported
plate areas and thus, potential for high stress areas. The larger plates significantly reduce the
amount of welds, particularly in situ welding. Additionally, reducing the weld surfaces
reduces plate surface finish damage, making decontamination easier. Leaks in the spent fuel
pond are primarily associated with welds that, in this design, have been significantly reduced
in quantity. Furthermore, leak chases are included. These indicate evidence of leakage and
direct any contaminated leakage flow to the waste handling systems. They also prevent active
fluid from leaching into the concrete if a leak occurs. The same advantages also apply to the
refueling water storage tank and the reactor cavity.
HVAC in the secondary containment areas also limit the spread of contamination.
ER Section 2.9.5 [Reference 15] gives further details of how the AP1000 plant minimises the
potential for both leakage and spreading of radioactive contamination using structure, system,
and component designs and operational procedures.
UK AP1000 Radioactive Waste Arisings, Management & Disposal report [Reference 12]
outlines the principles considered in the Radwaste and ILW Store Buildings design with
respect to decommissioning. These include:

Wherever practical, surfaces exposed to contamination will be minimised; and their
surface coatings will be impermeable and readily decontaminated. For example, inside
containment [Reference 9, Ch 6.1.2.1]:
–
Carbon steel and structural modules within containment are coated with either
inorganic zinc, with an epoxy top coat or self-priming high-solids epoxy (SPHSE).
–
Where practical, miscellaneous carbon steel items (such as stairs, ceilings, gratings,
ladders, railings, conduit, duct, and cable tray) are hot-dip galvanized.
–
Steel surfaces subject to immersion during normal plant operation (such as sumps and
gutters) are stainless steel or are coated with SPHSE applied directly to the carbon
steel without an inorganic zinc coating.
–
Exposed concrete surfaces inside containment are coated with an epoxy sealer to help
bind the concrete surface together and reduce dust that can become contaminated and
airborne.
–
Floors subject to heavy traffic or contaminated liquid spills are coated with
self-leveling epoxy or SPHSE floor coating.

The plant will be designed for the containment and recovery of possible spillage of
contaminated material.

To prevent the spread of airborne contamination during operations, the ventilation system
will be designed using depression gradients, such that the air flows from of
low-contamination areas to areas of potentially high contamination.
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
Cabinets (for example, sample cabinets) will be robustly constructed and made as small
as practicable where used, whilst still fulfilling their design function.

Both internal and external cabinet surfaces will be smooth and free from cracks, sharp
edge projections and recesses where contamination could collect.

To prevent potentially spreading airborne contamination to the operating areas, all
cabinets will be vented via the radioactive ventilation system and maintained at a
depression.
Design Features that Facilitate Decommissioning
In addition to minimising decommissioning waste and contamination, the AP1000 plant
design incorporates features that facilitate decommissioning. These features are described
later in Section 4.6.
In addition to the design features that minimise waste, reduce contamination and facilitate
decommissioning the plant containment and structures are designed to retain their integrity
for the expected operational life of the plant and the subsequent decommissioning period. The
life of the radwaste building will be extended beyond the assumed 60-year operating period
to allow the building to support decommissioning. The operational period of ILW store
building is 100 years.
4.2
Operation and Maintenance Principles
To operate and maintain the plant with respect to future decommissioning, Westinghouse
recommends that the Licensee adopt the following principles:
4.2.1
Regulatory Compliance
The Licensee will be ultimately responsible for plant operation and maintenance. When
operating the plant, the Licensee must follow regulatory requirements that are informed by
the policy and regulatory guidance provided by:




4.2.2
The UK government
The HSE
The EA
International bodies, such as the IAEA (see Section 2.0).
ALARP
The Licensee should ensure that the ALARP principle is considered in all operational and
maintenance tasks that are important to safety, radioactive waste management and
decommissioning.
4.2.3
BAT
To optimally protect human health, wildlife, organisms, and the wider environment, the
Licensee should ensure that the BAT principle is adopted when managing radioactive
substances and radioactive waste. The implementation of BAT during operations should
reduce the amounts of waste and radioactive contamination that must be dealt with during
decommissioning.
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Operation and Maintenance
The Licensee will need to establish preventive maintenance programmes to ensure that the
AP1000 plant continues to operate safely. These programmes include surveillance, inspection
and maintenance of structures, systems and components (SSCs) in accordance with
appropriate safety margins, engineering practices and quality levels.
The way that a plant is operated and maintained will also significantly impact both:


The decommissioning strategies that can be implemented
The nature and amounts of waste that will have to be dealt with.
For example, operational procedures and good housekeeping must ensure that radioactive
contamination does not spread throughout the plant and that provision is made to carry out
maintenance activities that minimally spread contamination can be carried out. It is also
important for the plant to be decontaminated throughout its operating lifetime, because this
will minimise the potential operator dose uptake during decommissioning operations.
To ensure a high level of plant safety as the plant ages, the Licensee should develop an
inspection and maintenance programme that effectively and proactively manages SSC aging.
As the plant ages, there is an increased probability of both single component failures and
common cause failures. Operating experience [Reference 19] has shown that many SSC
failures occur as a result of aging mechanisms, such as:







General and local corrosion
Erosion–corrosion
Radiation and thermal embrittlement
Fatigue
Creep
Vibration
Wear
These failures can affect plant safety through abnormalities of process systems. The
programme should cover the entire plant lifetime including the decommissioning period,
which may extend over many years. The objective is to provide for the timely detection and
mitigation of significant aging effects in nuclear power plant systems, structures and
components important to plant safety and reliability during operation and decommissioning
[Reference 20].
4.2.5
Waste Minimisation
The basic AP1000 plant design principles that minimise creating radwaste during operations
and decommissioning are well established [References 12 and 15]. The Licensee should
follow these principles, which include:




Good housekeeping
Optimisation of plant shutdowns
Waste segregation
Volume reduction
To prevent additional secondary waste from arising during plant operation, the operational
procedures that the Licensee uses should take into account waste minimisation and best
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practices. This will ensure that all wastes can be disposed of through already proven or
planned routes when the plant is eventually decommissioned.
The on-site waste inventory should be minimised before decommissioning by promptly
transferring waste to the available national waste disposal repositories, if these repositories
are available and have capacity to accept new build waste. Some ILW and HLW may have to
be transferred to on-site waste stores before this transfer is possible.
4.2.6
Plant Modification
If the licensee should modify the plant during the operational life, then the effect on
decommissioning, such as future access, should be considered.
4.2.7
Data Collection, Record Management and Corporate Memory
HSE Safety Assessment Principle DC.3 states “…throughout the whole life-cycle of a facility
the documents and records that might be required for decommissioning purposes should be
identified, prepared, updated and retained” [Reference 3] (see Section 2.3).
Adopting this principle requires the Licensee to develop robust data collection and record
management procedures to handle the large amount of monitoring, process equipment and
maintenance information that must be retained to facilitate decommissioning. This includes
records of any design or materials changes, and any extensive maintenance or refurbishment
work carried out. These records are important to both:


Document the plant condition before transitioning from operation to decommissioning
Ensure minimal loss of corporate memory before the decommissioning phase begins
4.3
Decommissioning Principles
4.3.1
Decommissioning Strategy
The Licensee will develop a decommissioning strategy that is updated and refined at
appropriate intervals throughout the AP1000 plant operation and decommissioning phases.
The decommissioning strategy will provide an overview of a Licensee’s approach to
decommissioning its nuclear liabilities. The strategy should be integrated with other relevant
strategies (for example, waste strategy), and take into account government policy and
regulatory guidance.
A decommissioning strategy should take into account the following principles [Reference 1]:

Ensuring worker and public safety

Maintaining site security

Minimising waste generation and providing for effective and safe management of wastes
which are created.

Minimising environmental impacts, including reusing or recycling materials whenever
possible

Maintaining adequate site stewardship

Using resources effectively, efficiently and economically
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
Providing adequate funding

Maintaining access to an adequate and relevant skills and knowledge base

Using existing best practice wherever possible

Conducting research and development (R&D) to develop necessary skills or best practice

Consulting appropriate public and stakeholder groups on the options considered and the
contents of the strategy
These principles should be applied throughout each decommissioning programme to ensure
that programmes are optimised.
4.3.2
Initial Decommissioning Plan
The Licensee will develop an initial decommissioning plan for each AP1000 plant. After final
shutdown the initial decommissioning plan will be refined to a final decommissioning plan
(including radiological characterisation, for example). The initial decommissioning plan will
be updated and refined throughout its operating life and converted to the final
decommissioning plan.
Section 10.1.3 and Figure 10-1 outline a preliminary decommissioning programme.
4.3.3
Decommissioning Safety
As part of the initial decommissioning plan, the Licensee will take into account basic safety
issues and demonstrate that decommissioning can be safely conducted using proven
techniques or ones being developed.
Toward the end of the facility’s operational life, the decommissioning safety assessment will
be developed in increasing detail in conjunction with the decommissioning strategy and final
decommissioning plan.
The decommissioning safety assessment will address the changing hazards that may occur at
different decommissioning stages. During decommissioning, the safety assessment will be
updated to:


4.3.4
Identify the impact of the changes
Demonstrate that the risks remain at an acceptable level whilst the work is undertaken
CDM for Decommissioning Work Packages
The Licensee will implement CDM to integrate health and safety into the managing of each
decommissioning work package (see Section 2.5).
4.3.5
Skills and Training
The Licensee should maintain the knowledge base, records and skills necessary to their
decommissioning operations and management of associated wastes. This should include staff
retention, recruitment and training; and preserving documentation necessary fully to underpin
the operations.
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UK AP1000 NPP Decommissioning Plan
Radiation Risk Reduction
A staged decommissioning approach is used to systematically and progressively reduce
radiation risk [Reference 28]. During the first stage fuel will be removed from the reactor and
checks are performed to ensure that containment is maintained. After fuel is removed, the
primary circuit is decontaminated using existing primary circuit systems.
During the second stage preparatory work will be performed to enable the remaining
decommissioning activities. This includes conversion of the fuel handling area for processing
decommissioning waste. This will follow an appropriate cooling period for the final charge of
spent fuel in the pond.
In the third stage the remaining radioactive materials will be systematically removed to
reduce risk until there are no radiological restrictions on the site.
4.3.7
Dose Minimisation
Dose minimisation techniques will be established in response to radiation and contamination
survey results. These surveys will be conducted to determine the radionuclides, maximum
and average dose rates, and contamination levels of inner and outer surfaces of structures or
components throughout the reactor installation.
Dose to decommissioning operators will be reduced to ALARP levels (see Section 4.5).
Methods to achieve ALARP [Reference 28] include:
4.3.8

Design features that minimise dose

Assessing potential options and choosing arrangements that are ALARP

Reducing the dose rate, for example through decontamination, shielding or remote
working

Reducing the time spent close to the source.
Minimisation of Discharges during Decommissioning
During decommissioning, aerial or liquid radioactive discharges to the environment should be
minimised consistent with the site decommissioning strategy.
4.3.9
Waste Minimisation
The waste management hierarchy will be integrated into the design, operation and
decommissioning plans for the AP1000 plant. To maximise reuse and recycling, the
integrated waste strategy [Reference 11] will be reviewed during the operational phase and
before and during the decommissioning phase.
Where practicable, decontamination will be used to recover materials and to reduce the
radiological waste classification so that the wastes can be disposed at very low level waste or
conventional non-radioactive disposal facilities.
Wastes generated during decommissioning will be segregated into different waste types to
allow optimal use of the waste routes available at the time of decommissioning.
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UK AP1000 NPP Decommissioning Plan
Passively Safe Storage of ILW and HLW
Interim storage will be used for HLW and ILW generated at the end of plant life and during
decommissioning [Reference 11]. The on-site HLW and ILW stores will be designed to
contain the wastes in a passively safe condition (that is one that will not require active safety
systems during the interim storage period).
4.3.11
Decommissioning End-Point
An important principle of decommissioning is to establish the potential future uses of the site.
One decommissioning objective should always be to remove hazards so that the site does not
present an unacceptable level of risk to either human health or the wider environment. The
end-point includes the decommissioning of all facilities including the interim storage
facilities.
A Licensee may decide that the end-point for decommissioning on a nuclear site is at
delicensing. Delicensing involves:


the Release of the land from regulation under the Nuclear Installations Act 1965 (as
amended)
the Release of the facilities operator from his period of responsibility for any nuclear
liability
Delicensing may not always be the end-point of a decommissioning project. For example, the
Licensee may either plan to operate another nuclear facility on the same site or to maintain
institutional control.
4.4
Learning from Experience Principles
4.4.1
National and International Experience
International experience is an important source of information on best practice. International
organisations (for example, IAEA, WENRA, NEA, EU), regulatory bodies (for example,
HSE, US Nuclear Regulatory Commission [NRC]), and scientific and technical publications
and conference proceedings all publish learning from experience (LFE). As part of their
normal business practice, both Westinghouse and the Licensee will monitor these resources.
Relevant information will be appropriately recorded.
The IAEA has established a decommissioning network, which will assemble organisations
with specific experience and competence in decommissioning, and that are willing to share
their experience with other organisations.
4.4.2
Licensee Experience
The Licensee will likely have accumulated its own operating and decommissioning
experience from operating and then decommissioning other nuclear power plant facilities.
The Licensee will use this experience, combined with experience gained from the
decommissioning specialist contractors it employs, to inform AP1000 plant
decommissioning.
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Westinghouse Experience
Westinghouse will share the LFE processes it has in place with the Licensee (for example the
Corrective Action Process and the Lessons Learned data base). To assist the Licensee with
their decommissioning plans, Westinghouse will provide the Licensee with LFE resources
about:

Early Westinghouse PWR plants decommissioning (see Section 7.1.1)

The first AP1000 plants (for example, in China and the USA) that are likely to be
decommissioned ahead of any UK plants
The Westinghouse management arrangements contain the requirements to alert and involve
Licensees in the discussion and resolution of safety or environment related learning events.
4.5
ALARP Considerations
4.5.1
Plant Design
The AP1000 plant has been designed to keep residual risks As Low As Reasonably
Practicable. In addition to the design features described in Section 4.6, examples of
decommissioning ALARP considerations in the plant design include:
4.5.2

The modular design of the plant allows multiple components to be removed together
from the reactor and transferred to a purpose-built waste facility so that they can be
dismantled in a lower dose area. Waste processing and volume reduction activities will be
conducted in the reactor building only where necessary. This will also provide controlled
areas where measures to reduce personnel exposure to ALARP can be implemented (for
example, by maximising the opportunities for remote applications for cutting and
handling in a dedicated facility).

Equipment maintenance access requirements will be incorporated into the plant design.
For example, access for remote working will similarly support decommissioning (for
example Sections 5.1 and 5.2.3).

The use of embedded pipes was minimised consistent with maintaining radiation doses
ALARP. To the greatest extent possible, pipes were routed in accessible areas such as
dedicated pipe routing tunnels or pipe trenches that will offer good conditions for
decommissioning [Reference 15].

Equipment carrying radioactive fluid is generally installed in separate areas and rooms
for maintenance and ALARP considerations.
Decontamination
Adopting ALARP considerations when planning and executing decommissioning activities is
essential to protect workers, the public and the environment. Decontamination is important to
ensure that exposures are kept ALARP. ALARP decontamination considerations should be
assessed in the final decommissioning plan and should consider the following [Reference 32]:

The relative benefit of decontamination versus no decontamination

The target decontamination level
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
The estimated doses to workers

Possible aerosol generation

The likelihood that available techniques will achieve the target level on particular
components

An ability to demonstrate by measurement that the target level was reached

The availability of facilities required for decontamination and their eventual
decommissioning

The cost of the application compared with the expected benefit (for example the cost of
decontamination versus the cost of disposal of original material)

An estimate of the volume, nature, category and activity of any primary and secondary
wastes

The compatibility of these wastes with existing treatment, conditioning, storage and
disposal systems.

Any possible deleterious effect of decontamination on equipment and system integrity

Any possible on-site and off-site consequences resulting from decommissioning activities

Non-radiological hazards (for example the toxicity of solvents used)
Shielding and Ventilation
ALARP methods will also define the need to use shielding and ventilation based on radiation
and contamination surveys to determine the following for inner and outer surfaces of items to
be removed from the facility:



4.5.4
Radionuclides
Maximum and average dose rates
Contamination levels
Working Area and Personnel Monitoring Programmes
ALARP methods also require implementing regular working area and personnel monitoring
programmes. Personnel monitoring programmes should be conducted using management
systems to determine the workers individual and cumulative dose. If the dose deviates from
an acceptable level, immediate corrections can be made to protect individuals.
4.6
Key Design Features
4.6.1
Design Features Supporting Decommissioning
AP1000 plant design features that facilitate decommissioning include the following:

Separation of radioactive and non-radioactive equipment, and minimisation of radioactive
systems.
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
The modular design and construction means that the modules can be removed in a
manner similar to the way they were installed. However, some segmentation of the
modules may be required to facilitate lifting operations due to the additional weight of the
concrete added during construction.

Adding a larger capacity hoist module gives the polar crane structure sufficient capacity
to handle heavy equipment. Additionally, the polar crane can accommodate the upper
assembly of the steam generators between the girders.

Where practicable, floor slabs have been designed to support the weight of equipment
during the decommissioning process.

As far as possible, the plant layout ensures that there is space to remove whole plant
items and modules. Laydown areas have also been provided to protect and wrap
potentially contaminated equipment before transporting it to the site decontamination and
sorting facility(s).

Access routes for equipment have been considered in the design. Routes are available
through equipment hatches, and temporary access through the steel containment is easily
provided and controlled.

The design facilitates sampling during the decommissioning process.

Specific equipment at the lower levels includes removable shielded hatch covers (for
example, chemical and volume control system [CVS] demineralisers).

Removable gratings have been used for floors to facilitate the transporting and handling
of equipment.

Where appropriate, the design allows sections of plant to be isolated and decommissioned
independently of others. In particular, service systems will include isolation valves to
allow areas to be decommissioned and taken out of service without affecting ongoing
service supplies needed elsewhere.

The plant electrical power distribution has been arranged with decommissioning in mind.
Prior to decommissioning, a new electrical supply to the ILW store will be installed
which will also supply the radwaste building. This will allow the main nuclear island to
be dismantled whilst retaining the use of the Radwaste building.

All sumps are plate-lined, welding is minimised, and surfaces are treated to limit crud
retention.
Section 4.1 outlines some of the ways in which the AP1000 plant design facilitates
decommissioning through limiting the spread of contamination and minimising
decommissioning waste.
In addition to the features incorporated into the plant design, decommissioning will also be
facilitated through:

3D and 4D modelling to optimise dismantling tasks

Incorporating lessons learned and best available techniques as described in Section 4.4.
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5.0
DECOMMISSIONING LOGISTICS
5.1
Decommissioning Sequence
5.1.1
Overview of the Outline Decommissioning Sequence
There are three well defined decommissioning strategies [References 9 and 16]:
1. Immediate dismantling
2. Deferred dismantling (also known as “safe enclosure”)
3. Entombment
The AP1000 NPP design is amenable to each of these decommissioning strategies, but lends
itself most easily to the immediate dismantling option (see Section 6.1). This section outlines
a method by which the AP1000 plant can be safely decommissioned in three stages.
Before beginning decommissioning, a temporary decommissioning facility suitable for use as
an intermediate-level waste/low-level waste (ILW/LLW) solid waste processing area will be
constructed within the site boundaries. The facility will be large enough to store at least two
steam generators, one reactor vessel, and sundry other equipment; and it will include a remote
handling and waste reduction process area. This is not shown in Stage 1 because it is assumed
that this will be designed and installed prior to decommissioning. Figure 9-1, Item 27
identifies the area currently allocated for the GDA site’s temporary decommissioning facility,
and it is considered close enough to perhaps tie into the plant’s radioactive drain system.
Activities to be completed during each decommissioning stage are as follows:


Stage 1
–
Maintenance of buildings and systems until dismantling
–
Remove fuel from the reactor to the spent fuel pool
–
Remove radioactive inventory from the reactor system and non-decommissioning
service systems
–
Contamination mapping
–
Post-operational clean out and in situ decontamination (for example chemical
decontamination of contaminated circuits) of non-decommissioning service systems
and reactor system
–
Establishment of new radiation control areas based on the above actions as work
progresses
–
Review service systems configuration (for example electrical, piping, HVAC,
radwaste, drainage, monitoring) and plan for their use/modification in
decommissioning activities
–
Remove fuel from the spent fuel pool to the spent fuel store
Stage 2
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–
Modify existing service systems and install temporary service systems to meet
decommissioning requirements (for example, shielding, monitoring, ventilation
systems)
–
Remove radioactive inventory from the spent fuel pool system and the fuel handling
system
–
Dismantle non-radioactive, non-decommissioning service systems from the turbine
building and dispose/recycle inert waste
–
Post-operational clean out and in situ decontamination of the spent fuel pool system
and the fuel handling system
–
Conversion of fuel handling building into an interim waste storage, decontamination,
waste reduction, packaging, and processing area for ILW
–
Contamination mapping of potentially radioactive, non-decommissioning service
systems
–
Dismantle radioactive, non-decommissioning service systems
–
Ex situ decontamination of dismantled radioactive, non-decommissioning service
systems (for example, in decommissioning facility, spent fuel pond, fuel handling
area, hot machine shop, radwaste building)
–
Treat, package and dispose of
non-decommissioning service systems
–
Contamination mapping of potentially radioactive systems in containment
waste
from
potentially
radioactive,
Stage 3
–
Dismantle reactor system. Assess radioactive contamination levels of dismantled
equipment and securely protect for transportation to waste handling areas
–
Remove radioactive inventory from residual radioactive systems in containment
–
Contamination mapping of other potentially radioactive systems and buildings
–
Dismantle and remove non-radioactive, decommissioning service systems from
turbine building yard. Replace with temporary decommissioning service systems, as
required
–
Decontaminate dismantled reactor system ex situ (for example, in the
decommissioning facility, spent fuel pond, fuel handling area)
–
Post-operational clean out and in situ decontamination of radioactive systems in
containment, the containment building (including cutting, processing, and removal of
contaminated concrete in the containment building) and the shield building
–
Remove radioactive inventory from residual radioactive systems in auxiliary building
–
Dispose/recycle inert waste from non-radioactive systems from turbine building and
yard
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–
Package and dispose of radioactive waste from reactor system
–
Post-operational clean out and in situ decontamination of residual radioactive
systems in auxiliary building
–
Dismantle and remove residual radioactive systems from containment
–
Dismantle and remove containment building, shield building and turbine building
–
Ex situ decontamination of dismantled radioactive systems from containment, the
dismantled radioactive components of the containment building and shield building
–
Dispose of inert waste from dismantled turbine building
–
Dismantle and remove remaining radioactive and non-radioactive decommissioning
service systems in auxiliary building
–
Decontaminate, dismantle, and remove the auxiliary building, radwaste building,
decommissioning building, and LLW store. Replace with temporary systems as
required
–
Ex situ decontamination of contaminated decommissioning service systems and the
auxiliary building, radwaste building, decommissioning facility and LLW store
–
Treat, package and dispose of radioactive waste from dismantled radioactive systems
from containment, the dismantled contaminated components of the containment
building and shield building
–
Dispose/recycle inert waste from dismantling non-radioactive decommissioning
service systems and residual radioactive systems in auxiliary building
–
Dismantle and remove temporary decommissioning service systems, wastewater
system, annex building and diesel generator building
–
Treat, package and dispose of radioactive waste from potentially radioactive
decommissioning service systems, including temporary service systems
–
Dispose/recycle inert waste from temporary decommissioning service systems,
wastewater system, annex building and diesel generator building
–
Maintain ILW and HLW stores until transfer of all waste to repositories
The decommissioning strategy will be reviewed and updated throughout the plant lifecycle to
take into account the current status of the plant, availability of new technologies and changes
in regulations. It may be possible for activities identified in one stage to be carried out during
other stages, if appropriate (for example, removal and dismantling of the reactor pressure
vessel may be carried out earlier, depending on activity levels).
At this stage it is envisaged that, before decommissioning commences, a temporary building
would be erected within the site boundaries for use as a solid waste processing area for ILW
and LLW. Deferring the construction of the decommissioning facilities to the end of
operations allows for the latest techniques and technologies to be used, and would allow for
the prevailing regulatory requirements to be implemented. The building should be large
enough to facilitate storage of at least two steam generators, and one reactor vessel and other
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sundry equipment. It would include an HVAC system, remote handling and waste reduction
areas; and will tie into the plant active drain system. Waste would be reduced to appropriate
sizes and processed in accordance with applicable radiological disposal requirements. ILW
waste would also be stored and processed in the modified spent fuel building, as described in
the Stage 2 decommissioning activities. The Licensee may consider other opportunities for
dismantling and materials treatment (for example, use of off-site facilities)
The outline decommissioning sequence is shown diagrammatically in Figure 5-1.
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Figure 5-1 Outline Decommissioning Sequence
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5.1.2
UK AP1000 NPP Decommissioning Plan
Stage 1 Decommissioning

Maintenance of buildings and systems until dismantling
After ceasing power generating operations, all buildings and systems will be maintained
as required until they are ready to be dismantled.

Remove fuel from the reactor to the spent fuel pool
Removal and transport of spent fuel from the reactor to the spent fuel pool, per the
processes established at that time. Spent fuel will remain in the spent fuel pool until the
heat output from the fuel has reduced sufficiently to meet the current Radioactive Waste
Management Directive (RWMD) criterion for acceptance.

Remove radioactive inventory from the reactor system and non-decommissioning service
systems
The reactor system and the non-decommissioning service systems radioactive inventory
will be removed and treated by the existing plant radwaste systems. The term
“radioactive inventory” refers to the radioactive contents of vessels, tanks, pipes such as
water, chemicals, settled solids, ion exchange resins and filter cartridges. At this stage,
fixed contaminated equipment will not be removed.

Contamination mapping
Contamination mapping of all radioactive and potentially radioactive systems will be
carried out to ensure that

–
The health and safety risks associated with maintaining and dismantling radioactive
systems are understood
–
The decontamination and radioactive waste treatment options can be assessed
Post-operational clean out and in situ decontamination (for example, chemical
decontamination of contaminated circuits) of non-decommissioning service systems and
reactor system
The reactor cavity and radioactive, mechanical and fluid systems not required for
decommissioning will be decontaminated and cleaned in-line with the decontamination
strategy described in the UK AP1000 NPP Decontamination Considerations
[Reference 42]. Contaminated piping circuits will likely be decontaminated by an
oxidative decontamination process, although other processes (for example,
oxidative/reductive processes) may be considered. The oxidative process involves ion
exchange. Anion and cation resins will be used, become contaminated, and must be
treated as radioactive waste. The AP1000 plant circuits are anticipated to be significantly
less contaminated due to the chemistry and the addition of zinc acetate. This
decontamination method will require the provision of dedicated reagent tanks, pump skid,
and the like. It may be possible to utilise existing plant equipment or alternatively explore
the option of using mobile equipment while developing the final decommissioning plan.
All waste, radioactive or not, will be treated in accordance with the requirements in force
at the time of decommissioning. The system will be designed to minimise exposure and
limit the potential for contaminated water to leak and contaminate surfaces.
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UK AP1000 NPP Decommissioning Plan
Establishment of new radiation control areas
When the plant enters the decommissioning phase, the existing radiological control areas
will remain in place. As work progresses and radioactivity levels permit, the boundaries
will be successively reduced.

Review requirements of decommissioning service systems
The decommissioning requirements for all service systems will be reviewed. Service
systems may need to be modified according to decommissioning work package
requirements. The need for temporary service systems will be identified (for example,
electrical, piping, ventilation, monitoring and shielding systems). During
decommissioning, the existing plant radwaste systems will remain in use for as long as
possible during decommissioning. The Auxiliary Building ion exchange facility may be
used to treat decontamination and radioactive drain reagents. The treatment route will be
assessed to determine if further treatment units (for example, neutralising, filtering) will
be required to achieve required discharge consents. Final effluent disposal will either be
by routing to the site effluent treatment facility, discharge to outfall, or by collecting and
transferring to a designated waste effluent treatment facility operated by a contractor. A
mobile encapsulation plant will be used as far as practicable for the decommissioning and
disposal of solid waste (for example, high efficiency particulate air [HEPA] filters, spent
resins). When the auxiliary building and, in particular, the liquid radwaste facilities are to
be decommissioned, it will be possible to re-locate the encapsulation plant to either the
radwaste building, or to the temporary decommissioning structure for further use. The
liquid radwaste equipment could then be cleaned, size reduced and encapsulated, if
needed. The point at which the existing plant radwaste systems will be decommissioned
will be defined in the final decommissioning plan. However, it is assumed it will be
during Stage 3.

Remove fuel from the spent fuel pool to spent fuel store
Once the remaining spent fuel in the spent fuel pool has sufficiently cooled, the spent fuel
assemblies will be moved by the fuel transfer system from the spent fuel pool to the cask
loading pit. Here they will be placed into dry storage canisters that will be filled with
inert gas and sealed. The sealed canisters will then be cleaned and decontaminated before
being transferred to the spent fuel store using an appropriate transport vehicle.
5.1.3
Stage 2 Decommissioning

Modify existing service systems and install temporary service systems to meet
decommissioning requirements (for example, shielding, monitoring, ventilation systems)
Modify existing service systems and supply temporary service systems to meet
decommissioning requirements in line with the plans developed during the Stage 1
review.

Remove radioactive inventory from the spent fuel pool system and the fuel handling
system
The radioactive inventory of the spent fuel pool system and the fuel handling system will
be removed and treated by the existing plant radwaste systems. Contamination mapping
will be performed to assess the decontamination and cleaning requirements, and the
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health and safety risks associated with maintenance and dismantling of the radioactive
systems.

Dismantle non-radioactive, non-decommissioning service systems from the turbine
building and dispose/recycle inert waste
Non-radioactive service systems (for example, circulating water cooling system) that are
no longer needed to support decommissioning activities will be dismantled and removed
per the schedule established at that time. Temporary lifting and handling equipment to
support plant dismantling will be brought in and installed. The radioactive drain system
will remain in operation. All waste is expected to be non-radioactive and will be disposed
off/recycled by the procedures in place at that time.

Post-operational clean out and in situ decontamination of spent fuel pool system and fuel
handling system
The spent fuel pool, the fuel handling system, and other associated areas will be
decontaminated and cleaned as required in-line with the decontamination strategy
described in the UK AP1000 NPP Decontamination Considerations [Reference 42]. As
described in Stage 1, the existing ion exchange facility may be used to treat
decontamination and radioactive drain reagents.

Fuel handling building conversion
All fuel from the reactor and the spent fuel pool will have been removed during Stage 1
of decommissioning. The spent fuel racks will be dismantled, and the pond will be
cleaned and decontaminated. The HVAC systems will remain in operation. The fuel
handling building will then be used as a decontamination and waste reduction area. This
building, together with the temporary facility described earlier, will be sufficient to store
all of the dismantled equipment removed from the plant during decommissioning. The
basic premise is that there is significantly less building volume and equipment and that
less of this reduced volume will be contaminated. As items are confirmed as clean, they
will be removed from the site and disposed. This facility will be used for processing ILW
waste. To accommodate their disposal, items will be reduced in size to fit in the standard
waste casks (for example, RWMD boxes and drums [Reference 35]) in use for that waste
level at that time. Equipment for cutting and volume reduction activities will be installed
in this building. The reduction process will be designed to minimise personnel exposure,
and minimise and contain the generation of dust.

Contamination mapping of potentially radioactive non-decommissioning service systems
Contamination mapping of all radioactive and potentially radioactive systems will be
performed to ensure that:

–
The health and safety risks associated with maintaining and dismantling the
radioactive systems are understood
–
The decontamination and radioactive waste treatment options can be assessed
Dismantle radioactive non-decommissioning service systems
Detailed plans for dismantling all major equipment will be developed in the final
decommissioning plan (see Section 4.3.2 and Section 10.1). Such plans will include
provisions to limit personnel hazards (see Section 7.2). Installed in-cell handlers will be
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used to remove large plant items. Where no man access exists, remote in-cell handling
equipment will be provided to facilitate dismantling the plant.
To remove large plant items and modules, plans cognisant of the technical guidance, such
as IAEA Technical Report “Heavy Component Replacement in Nuclear Power Plants:
Experience and Guidelines” [Reference 34], will be developed. This report describes the
organisational requirements necessary for large and complex removal projects, and
describes suitable strategies and sequences for removing specific heavy components.
Section 5.1.7 presents examples of outline sequences for removing large items,
contaminated items and difficult-to-access items.
Before being transported to the LLW or ILW areas, dismantled equipment will be
packaged and examined to determine radioactivity levels and to confirm packages are
securely protected for transportation. Stations will be set up in the auxiliary and
containment buildings to provide this function. Items determined to be clean will be
routed to an interim clean storage area, where they will await dispatch.

Ex situ decontamination of dismantled radioactive non-decommissioning service systems
Decontamination of dismantled radioactive components will take place ex situ in the
various facilities provided (for example, the converted spent fuel pool/fuel handling area,
the decommissioning facility, the health physics and hot machine shop, the radwaste
building).

Treat, package and dispose of waste from radioactive non-decommissioning service
systems
Radioactive waste will be classified, treated and disposed following established
procedures for operational LLW and ILW, as described in the UK AP1000 Radioactive
Waste Arisings, Management and Disposal [Reference 12].

Contamination mapping of potentially radioactive systems in containment
Contamination mapping of potentially radioactive systems inside the containment
building will be carried out to ensure that both the health and safety risks associated with
maintaining and dismantling the radioactive systems are understood, and that the
decontamination and radioactive waste treatment options can be assessed.
5.1.4
Stage 3 Decommissioning

Dismantle reactor system, assess radioactivity levels of dismantled equipment, and
securely protect for transportation to waste handling areas
The reactor system will be dismantled and components will be removed and transported
to the waste processing facilities. The radioactivity levels of the reactor pressure vessel
should have dropped sufficiently to allow the dismantled components to be transported
within the regulations applicable at that time. The hot and cold legs will be remotely cut
and removed using techniques that minimise personnel exposure. The reactor pressure
vessel will be disconnected from its supports, and all piping will have been previously cut
and removed. Details of this operation will be included in the final decommissioning
plan. A temporary shielding structure that will cover the reactor pressure vessel and
internals during lifting, upending, and transporting to the processing facility will be
produced. The polar crane will be used for lifting and upending components to and on the
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135 ft (41.15m) level. The reactor pressure vessel and its transport shield can then be
passed through the new opening to the jacking tower, lowered to the transporter, and then
taken to the processing facility. These activities can be largely completed remotely.
An option for the reactor pressure vessel will be to dismember it in situ, package it in the
approved containers, and then remove the individual containers. This will require the
provision of temporary shielding and cutting processes, which will allow for flooding the
reactor vessel compartments to provide additional shielding. Alternatively, reactor
vessels in the United States have been disposed intact. The internals will have been
removed, decontaminated, wrapped, and taken to the processing area. The transport route
will be through the opening provided at the 135 ft (41.15m) elevation.

Remove radioactive inventory from residual radioactive systems in containment
Some systems in the containment building are expected to be radioactive (this will have
been confirmed by the contamination mapping performed at the end of Stage 2). The
existing plant radwaste systems will remove and treat the radioactive inventory of these
systems.

Contamination mapping of other potentially radioactive systems and buildings
Contamination mapping of the remaining radioactive decommissioning service systems
will be performed before cleaning, decontaminating, and dismantling these systems.
Before dismantling, contamination mapping of the containment building and shield
building, auxiliary building, radwaste building, and the temporary decommissioning
facility will be carried out to assess if any of the structures or concrete are activated or
contaminated.

Dismantle and remove non-radioactive decommissioning service systems
The non-radioactive decommissioning service systems in the turbine building will be
dismantled and removed per the schedule established at that time. Temporary lifting and
handling equipment will be brought in and installed to support plant dismantling.
Temporary service systems will be provided to facilitate ongoing decommissioning
activities, as required.

Decontaminate dismantled reactor system ex situ (for example, in the decommissioning
building, spent fuel pool, fuel handling area)
Decontamination of radioactive dismantled reactor components will take place ex situ in
the various facilities provided (for example, converted spent fuel pool/fuel handling area,
the decommissioning building, the health physics and hot machine shop, the radwaste
building).

Post-operational clean out and in situ decontamination of radioactive systems in
containment, the containment building (including cutting, processing, and removal of
radioactive concrete in the containment building) and the shield building
The radioactive systems in the containment building will be cleaned and decontaminated
in-line with the decontamination strategy described in the UK AP1000 Decontamination
NPP Considerations [Reference 42]. Contaminated concrete in the containment building
will be cut, processed and removed. Concrete is considered to be contaminated no more
than 3 ft (0.9m) from the exposed outer edges. The cutting program may use a diamond
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wire sawing technique, which will cut through the steel facing. Any active concrete will
be removed under controlled conditions and processed separately. Diamond wire sawing
is relatively clean, because the dust can be controlled and the limited quantities of water
needed can be controlled and maintained within a closed circuit. Typically, blocks of
concrete are cut, and the contaminated pieces are removed. The removed pieces are sized
for packaging in the approved disposal containers. Once the wire is passed around the
area to be cut, the machine can be located away from the radioactive area and shielded to
minimise personnel exposure. The only anticipated active concrete will be local to the
reactor pressure vessel or possibly the spent fuel pond area. The same techniques will be
used in both areas. After removing the contaminated concrete, the containment and shield
buildings will be cleaned and decontaminated as required.

Remove radioactive inventory from residual radioactive systems in auxiliary building
The existing plant radwaste systems in the auxiliary building will be the last radioactive
decommissioning service systems to be decommissioned. The radioactive inventory of
these systems will be removed and treated by temporary service systems.

Dispose of/recycle of active and inert waste
As decommissioning progresses, wastes from dismantled non-radioactive systems and
non-radioactive wastes segregated from dismantled decontaminated radioactive systems
will be disposed/recycled/recovered by the procedures in place at that time. Radioactive
wastes will be classified, treated and disposed following established procedures for
operational LLW and ILW described in the UK AP1000 Radioactive Waste Arisings,
Management and Disposal [Reference 12]. The order in which waste from specific
systems will be disposed can be seen in Section 5.1 and in Figure 5-1.

Post-operational clean out and in situ decontamination of radioactive decommissioning
service systems in auxiliary building
The radioactive decommissioning service systems in the auxiliary building will be
cleaned and decontaminated in-line using the decontamination strategy described in the
UK AP1000 Decontamination NPP Considerations [Reference 42].

Dismantle and remove residual radioactive systems from containment
The residual radioactive systems in the containment building will be dismantled.
Dismantled components will be examined to determine radioactivity levels, securely
packaged, and transported to the appropriate waste processing facilities.

Dismantle and remove containment building, shield building and turbine building
At this point, the dismantling of buildings will begin. Removing the containment
building, shield building and turbine building is a bulk material removal undertaking.
However, it is simplified because the structures are freestanding and can be
systematically removed from the top. The containment building steel pressure vessel will
be stable throughout the dismantling process and can support significant weight in terms
of access attachments for personnel, and cutting equipment. During the removal process,
individual sections of plate will need to be stabilised. Waste will be monitored prior to
transportation but, because none of it should be radioactive, it is expected to go directly
to the clean waste areas.
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Ex situ decontamination of dismantled radioactive systems from containment and the
dismantled contaminated building components
Ex situ decontamination of dismantled radioactive decommissioning systems and the
dismantled active components of the containment building and shield building will take
place in the various waste handling facilities (for example, in the decommissioning
building, radwaste building).

Dismantle and remove remaining radioactive and non-radioactive decommissioning
service systems and residual radioactive systems in auxiliary building
The radioactive and non-radioactive decommissioning systems in the auxiliary building
will be dismantled and removed per the schedule established at that time. Temporary
lifting and handling equipment will be brought in and installed as required to support
plant dismantling. Dismantled equipment will be packaged and examined to determine
activity levels and to confirm that packages are securely protected for transportation to
the waste processing areas.

Decontaminate, dismantle and remove the auxiliary building
The auxiliary building will be systematically removed by cutting through the composite
steel/concrete structure. Temporary HVAC together with temporary sealing for openings
will ensure that airflow is controlled and inward only. None of this concrete will be
activated; however, some of it may be contaminated. Significant areas of the walls are
steel faced and can be easily decontaminated. A mobile decontamination unit will be used
for this purpose. It is recognised that the tanks in the lower auxiliary building due to
access considerations may need to be removed to a local decontamination area first, and
then taken to the processing facility. Such items, due to access considerations, may need
to be disassembled before removal. This then allows work to decontaminate the walls in
the lower auxiliary building to proceed. Explosive removal of concrete has not been
considered in the Nuclear Island. Floor slabs will be supported, cut, and systematically
removed to expose the equipment in the rooms. Much of this equipment will be
radioactive and will be removed, packaged for local transport, and taken to the waste
processing area. Work to disconnect this equipment, most of which are modules, from
their systems will have been previously completed. Dismantled components will be
packaged and examined to:

–
Determine radioactivity levels
–
Confirm that packages are securely protected for transportation to the temporary
decommissioning facility.
Decontaminate, dismantle and remove the radwaste building
Before it is dismantled, the radwaste building will be manually cleaned and
decontaminated, as required. The building HVAC system will be maintained to minimise
dust released, and equipment will be manually dismantled and size reduced. The HVAC
system will then be similarly decommissioned, and the building cladding will be
removed, followed by demolition of the structure itself. Health Physics will survey all
wastes before disposal, but this waste is anticipated to be free release discharge.

Decontaminate, dismantle and remove the LLW store
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The LLW store will be manually cleaned as required. It is not expected to be
contaminated, but decontamination may be carried out if required. A temporary HVAC
system will be provided to minimise dust released, and equipment will be manually
dismantled and size-reduced. All wastes will be surveyed by Health Physics prior to
disposal, but this waste is anticipated to be free release discharge.

Ex situ decontamination of dismantled radioactive decommissioning service systems and
auxiliary building components
Radioactive components from the decommissioning service systems and the dismantled
auxiliary building will be ex situ decontaminated in the temporary decommissioning
facility.

Treat, package and dispose of radioactive waste from dismantled radioactive
decommissioning service systems and buildings
All remaining radioactive waste from dismantled radioactive decommissioning service
systems and buildings will be treated and packaged in the temporary decommissioning
facility. Waste packages will be examined to:

–
Determine radioactivity levels
–
Confirm that packages are securely protected for transportation to interim
stores/disposal facilities/free-release, as appropriate
In situ decontamination and dismantling of the decommissioning building
After processing radioactive waste from dismantled radioactive decommissioning service
systems and buildings, the temporary decommissioning facility will be decontaminated
and decommissioned. The HVAC system will be operated while the remaining equipment
is cleaned and dismantled using the mobile decontamination unit. The HVAC system will
then be shut down, cleaned, surveyed, with the spent filter cartridges disposed of as ILW
waste. All the equipment and structures will be surveyed prior to disposal, and it is
anticipated it will be LLW or free discharge. Temporary waste handling and processing
facilities will be provided.

Dismantle and remove temporary decommissioning service systems
The temporary decommissioning service systems provided to facilitate decommissioning
of the existing systems and buildings will be decontaminated and decommissioned.
Dismantled components will be examined to determine radioactivity levels and packaged
to ensure that the contents are securely protected for transportation to the temporary
waste handling and processing facilities. Waste packages will be examined to:

–
Determine radioactivity levels and to
–
Confirm that packages are securely protected for transportation to interim
stores/disposal facilities/free-release, as appropriate
Dismantle and remove waste water system, diesel generator building and annex building
At this stage, the waste water system, the diesel generator building, and the annex
building will be dismantled. All waste is expected to be non-radioactive and will be
disposed according to the procedures in place at that time.
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Dispose of radwaste and non-radioactive waste
All radioactive waste generated during the final building and systems dismantling will be
processed, packaged and transported to interim storage as described in Section 9.0.
Non-radioactive waste will be disposed according to the procedures in place at that time.

Maintain ILW and spent fuel stores
The interim ILW and spent fuel stores will be required until the respective national
repositories are available to receive waste packages for final disposal. Additionally, spent
fuel will need to remain in the spent fuel store until the heat output from the fuel has
reduced sufficiently to meet the current RWMD criterion for acceptance. Consequently,
the ILW and spent fuel stores will need to be maintained until all waste packages have
been transported off the site for final disposal.
5.1.5
Decommissioning of Waste Stores

ILW store dismantling
The store design life is substantially longer than the power generation facility, and as
such, it will be a standalone structure supported by its own services and utilities. If the
national ILW repository is available before AP1000 plant operations end, ILW
decommissioning waste could be shipped directly to the repository without interim onsite storage. Having the national ILW repository available before AP1000 plant
operations end will also allow the on-site ILW store to be decommissioned and
dismantled using the decommissioning facilities created for the AP1000 plant.
Once all the stored waste packages have been removed from the ILW store to off-site
storage (either the national ILW repository or another appropriate storage facility), the
building can be decommissioned. To minimise dust and effluent releases, the building
HVAC and effluent collection systems will continue to operate. The vault crane will be
moved to the maintenance bay for decontamination. It will then be used to lift and
remove equipment from the building. Equipment will be decontaminated, and the
reagents used will be collected in the effluent collection system. The effluents will be
routed to a temporary mobile treatment facility (for example, filtration, ion exchange, pH
neutralisation) before discharge to site drainage. Most of the equipment will be located
outside of the vault and is expected to be clean. A Health Physics survey will be carried
out, and the waste will be characterised as either LLW or free discharge. Dismantling and
cutting will be performed manually. Once the equipment has been removed, the effluent
collection system will be similarly drained, flushed and decommissioned. Any
contamination will be removed by scabbling or other suitable techniques (for example,
peelable coatings), and disposed of as LLW or ILW, as required. The HVAC system will
then be decommissioned, decontaminated and surveyed, with the spent cartridges being
disposed as ILW using a temporary mobile encapsulation facility. The remaining building
structure will be demolished and monitored, and is anticipated to either be free discharge
to landfill or to be used on site as backfill.

Dismantling of Spent fuel store
The on-site spent fuel store will be decommissioned after transferring stored spent fuel to
the national HLW repository (once available). On site storage up to 160 years after plant
start-up may be required to allow all the fuel to decay before shipment to the national
HLW repository (see Section 10.1.2). The decommissioning facilities created to handle
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the AP1000 plant and the ILW store arisings will need to be dismantled and disposed of
following a period of decontamination.
When the HLW repository is available, the spent fuel will be shipped to it, and the spent
fuel store can be decommissioned. This will require using temporary facilities to handle
wastes arising as a result of the dismantlement of the spent fuel store. Based on the
Holtec dry storage system, these wastes will consist of the steel closure lid, steel
container and divider shells, the concrete support foundation and the top surface pad
[Reference 11]. Holtec’s dry storage system materials were selected to ensure that there
were no chemical, galvanic, or other reactions among the materials of the HI-STORM
100U System, their contents, and the operating environment. The HLW store comprises a
number of subterranean cylindrical steel and concrete cavities (see Section 9.7.3) which
are expected to be free from contamination. This will be confirmed by a survey of surface
radioactivity. If necessary, the cavities will be decontaminated, and any radioactive waste
arising from decontamination will be handled and disposed of as either LLW or ILW,
using a temporary mobile encapsulation facility. Following decontamination, the
remaining waste materials will be characterised as either LLW or free discharge.
5.1.6
Decommissioning Work Package Management
To facilitate organization and management of decommissioning activities, they will be
broken down into a number of manageable work packages directly related to the various
AP1000 plant systems being decommissioned. This approach is consistent with IAEA
guidance based on lessons learned from global large nuclear facility decommissioning
experience [Reference 8].
During detailed decommissioning design the site licensee may decide to rationalise these
work packages or group packages together in a different way, which may in turn change the
proposed sequence. The exact sequence will be determined during detailed decommissioning
design.
Management of each work package will require developing the following:

The purpose and description of the task

Applicable criteria including:
–
–
–
Engineering and technical requirements
Health, safety and environmental protection requirements
Applicable standards

The activity sequence of events

The basic safety considerations of the activity, including descriptions of the nature and
source of any hazards with potentially significant risks

The quantities, characteristics and disposition of wastes arising
The work packages will refer to other documents, such as the Radiological Health and Safety
Manual, the Waste Management Manual, the Security Plan, the Quality Assurance
Programme, the Fire Protection Programme
CDM management practices and risk prevention principles (see Section 2.5 of Reference 1)
will be applied to each decommissioning work package, and this will become the primary
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mechanism to ensure that risks associated with each activity are identified, prevented or
controlled.
During decommissioning, the RWMD letter of compliance process will be followed to ensure
that regulatory approval to dispose of radioactive waste is obtained.
Detailed work procedures identify:



The step-by-step instructions to perform each task
The required equipment and associated operating parameters
Safety precautions and disposal methods, as applicable
Detailed work procedures may be general or specific.
When developing and executing work packages, human factors will be considered to ensure
that work is carried out as safely, reliably and efficiently as possible and that dose to
operators is minimised. This will be achieved by considering the ergonomics of specific work
procedures, including the design of tools and equipment to be used, the design of the work
location and the organizational structure put in place to facilitate and control the work.
Interdependencies between work packages may exist. Wherever the execution of a work
package may affect or depend on other work packages, this will be incorporated into the work
package in all applicable places (for example, establishing success criteria, identifying
boundaries and interfaces).
A “gated” management process will be implemented for the creation and management of
work packages. The management process consists of a series of steps. The work that needs to
be completed and the criteria that need to be satisfied (via external review) are specified in
each step. Work can only progress to the next step when all the criteria have been satisfied.
The flow chart below illustrates the decommissioning gated management process steps:
Step 1
Step 2
Step 3
Step 4
Step 5
Appraisal
Selection
Definition
Execution
Feedback
The following subsections specify the work that needs to be completed in each step.
5.1.6.1
Step 1 – Appraisal

Confirm licensee specifications are valid and viable

Establish success criteria

Identify and notify stakeholders

Assess if options are technically feasible and deliverable

Appraise costs, timescales and risk for each option

Identify boundaries, interfaces and waste streams
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
Assess if options meet all environmental, health & safety, and legal policies

Identify licensing strategy (for example, requirements for safety case [ND],
environmental permitting [EA], waste package [LLWR, RWMD], and funding liability
[DECC])

Carry out studies to identify all significant risks and hazards (for example, contamination
mapping) and develop risk mitigation plans

Fully define objectives and strategies for Step 2

Produce a programme and estimates for Step 2

Agree to next gate deliverables

Submit Pre-conceptual Letter of Compliance to RWMD
Work will progress to the next step when the options have been endorsed for further study.
5.1.6.2
Step 2 – Selection










Develop work package layout
Develop boundaries, interfaces and waste streams
Optioneering to select ALARP approach
Develop hazard management and environmental, health & safety controls
Develop operational strategies for the work package
Engineering & process definition and underpinning of selected technology
Define products, inventories and wastes
Submit conceptual Letter of Compliance
Identify risks and develop risk mitigation plans
Produce programme and estimates for Steps 3 and 4
Work will progress to the next step when the preferred option and delivery strategy has been
endorsed.
5.1.6.3
Step 3 – Define

Finalise boundaries and interfaces

Define plant space management arrangements and waste routes

Define hazard management arrangements – operating controls and features/licensing
arrangements

Confirm the probability and impact of all remaining risks and establish how risks will be
mitigated

Define management controls and operational arrangements

Define service requirements for each work package

Consider reconfiguration of existing services/installation of temporary services
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
Specify the product and wastes quality standards

Submit interim Letter of Compliance

Technology underpinning confirmed and incorporated into the final design

Engineering underpinning confirmed and incorporated into the final design

Cost of work defined supported by quotes and historical norms
Work will progress to the next step when proceeding with the selected option to full
implementation has been sanctioned.
5.1.6.4
5.1.6.5
Step 4 – Execution

Detailed design

Testing mock up equipment, if necessary

Training workers, if necessary

Submit final Letter of Compliance (decommissioning can only commence when this has
been received)

Equipment procurement, if necessary

Equipment manufacture, if necessary

Equipment installation, if necessary

Reconfiguration of existing services/installation of temporary services

Execution of the work package, following procedures developed during detailed design

Confirmation that the work package has been completed, and that all isolations and
restrictions related to the work package have been removed
Step 5 – Feedback

Decommissioning activities to be recorded

Assess outcomes against established success criteria

Lessons learned to be identified within organisation and disseminated to other interested
parties (for example, stakeholders, regulators and international community)

Compliance of the work with nuclear site license conditions

Programme and budget for the executed work package

Post-job brief to be prepared
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If the work package includes provision of structures, systems and components for use by
others (for example, provision of temporary decommissioning facility), then plant and
equipment will undergo plant and equipment operational trials, radioactive commissioning
and handover to the decommissioning site team.
BAT and ALARP assessments will form integral parts of each work package, and reviews
will be conducted at regular intervals, as appropriate.
5.1.7
Examples of Decommissioning Work Package Activities
The following subsections present three examples of decommissioning work packages,
offering further insight into a typical approach that may be followed. A detailed work
package management plan (see Section 3.0), which will develop engineering designs, risk
assessments, methodologies and management practices for each work package, will support
each example. The licensee will develop the actual decommissioning approach and work
package management plan during the detailed decommissioning design.
5.1.7.1
Large Item Removal – Steam Generators
An example of large item removal is decommissioning the two steam generators. The steam
generators form part of the Steam Generator System (SGS) and are located inside
containment. The strategy for decommissioning the steam generators involves in situ
decontamination followed by removing the entire steam generator to a laydown area through
an opening cut into the containment/containment shield building. The steam generator may
then either be further decontaminated ex situ before being wrapped for transport to the
designated disposal facility, or it may be cut up into smaller pieces to aid decontamination
and disposal.
One possible steam generator decommissioning work package sequence is as follows:

Stage 1 contamination mapping

Post Operational Clean Out (POCO)
–
–
–
–
In situ decontamination by circulating chemical cleaning agents
Pressure pulse cleaning and water slap methods
Sludge lancing
Water flush and drain system

Stage 2 pre-dismantling contamination mapping

System dismantling
–
Insulation is removed. During this process the bottom part of the steam generator is
filled with water to reduce dose rate.
–
Dismantle minor connecting systems, small equipment, and instrumentation systems,
and remove for ex situ decontamination, as required.
–
Provide temporary shielding cover to reduce exposure from openings, as required.
–
Provide access through the containment at the 135 ft (41.15m) operating deck level.
The temporary opening will be equipped with an airlock that will remain closed
except for the period when the steam generators are passing through it.
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–
Remove steel platforms on top of the steam generator housing at approximately 10 ft
(3.04m) of concrete, one side
–
Remotely cut and remove the steam dome and internal moisture separators. These are
not expected to be contaminated and will significantly reduce the steam generator
weight and height.
–
Provide a temporary shielding cover to reduce exposure from the steam generator
tubes, and remotely affix it in place.
–
Remotely cut the hot and cold legs. The legs will be filled with water up and into the
steam generator channel head to provide the maximum personnel protection.
–
Provide temporary shielding cover to reduce exposure from hot and cold leg
openings.
–
Lift the lower assembly with the polar crane, which will be equipped with a
construction trolley sized to lift the steam generator lower assembly. The polar crane
main structure is sized to hoist the complete steam generators.
–
Transfer the lower assembly to a horizontal position using an “upender” designed for
this purpose. Take the lower assembly through the opening to a jacking tower to
lower it to a transporter to transfer it to the laydown area.
Decontaminate and/or dispose of the steam generator according to the detailed
decommissioning plan. Options include:
–
Transporting the steam generator(s) whole to the LLW repository
–
On-site size reduction for decontamination in the decontamination building before
transferring it to either the LLW repository or free release/recycling, as appropriate.
Section 7.2 identifies the hazards that may be encountered when decommissioning the steam
generators. These include radiation hazards, chemical hazards associated with
decontamination chemicals and industrial hazards associated with construction, working at
height and heavy lifting/dropped loads.
5.1.7.2
Contaminated Item Removal – Spent Resin Tank
The decommissioning of the two spent resin tanks is an example of decommissioning a
contaminated item. The spent resin tanks are stainless steel tanks with a volume of 8.5m3
each. They are located in the Spent Resin Tank Room at ground level in the southwest corner
of the Auxiliary Building and form part of the Solid Radwaste System (WSS). As such, they
will be one of the last radioactive systems to be decommissioned.
The strategy for decommissioning the spent resin tanks involves transferring the radioactive
ion exchange resin to the ILW encapsulation facility. The tanks will then be decontaminated
in situ and either removed whole or in pieces for further ex situ decontamination and
recycling/free release or disposal as LLW.
One possible sequence for decommissioning the spent resin tanks is as follows:

Stage 1 contamination mapping/identifying radioactive inventory
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POCO
–
–
–
–
Fluidisation and pumping out the spent resin to the ILW encapsulation system
Water flushing the system
In situ chemical decontamination
Water flush and drain system

Stage 2 pre-dismantling contamination mapping of tank surfaces and connecting pipe
systems

System dismantling
–
Dismantle connecting systems, small equipment, and instrumentation systems, and
remove them for ex situ decontamination, as required.
–
Provide temporary shielding cover to reduce exposure from openings, as required.
–
Remove obstructions and interferences.
–
Create opening in Spent Resin Tank Room (wall or ceiling) large enough to extract
the spent resin tanks. Consideration must be given to the timing of this activity to
ensure continued structural integrity and that any necessary access to adjacent areas
and levels above is maintained.
–
Remove tank retention bolts.
–
Lift tank and place on transport vehicle located in Rail Car Storage Bay.
–
Transfer tanks to Decommissioning Building.

Monitor residual radioactivity of tanks and determine whether they can be reused. If not,
cut up tanks and decontaminate pieces ex situ using Decontamination Building facilities.

Dispose to the LLW repository or free release/recycle, as appropriate.
Hazards associated with decommissioning the spent resin tanks include radiation hazards,
chemical hazards associated with decontamination chemicals and industrial hazards
associated with creating room openings and lifting/cutting of tanks.
5.1.7.3
Difficult to Access Item – Passive Containment Cooling System Tank
An example of a system that is logistically difficult to access during decommissioning is the
Passive Containment Cooling System (PCS). Despite the fact that some of this system is
located within containment, it is not expected to be radioactively contaminated.
The strategy for decommissioning the PCS combines in situ dismantling with removal of
small components from within containment, and removal of the large Passive Containment
Cooling Water Storage Tank (PCCWST) externally by cutting out concrete sections in the
shield building roof.
The PCCWST is incorporated into the shield building structure above the containment vessel.
The interior wetted walls of the tank are lined with stainless steel plate. The conical shield
building roof supports the PCS tank. The conical roof is constructed as a structural steel
module and is lifted into place during construction. Steel beams provide permanent steel liner
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and concrete structural support. The concrete is cast in place. Shear studs connect the
concrete and steel liner.
To prevent freezing, the storage tank isolation valves along with the PCCWST discharge
piping and associated instrumentation between the PCCWST and the downstream side of the
isolation valves, are contained within a temperature-controlled valve room. This valve room
is located inside containment at an elevation of 160 ft 6 in (48.92m) above ground level.
One possible sequence for decommissioning the PCS is as follows:

Stage 1 contamination mapping/surface swabbing of PCS in containment

POCO is unlikely to be required and would be limited to external surface cleaning

Stage 2 contamination mapping/surface swabbing of PCS in containment

System dismantling
–
Dismantle connecting systems, small equipment and instrumentation systems, and
remove for ex situ decontamination, as required.
–
Provide rigging systems to lower tank isolation valves from valve room to the
operating deck staging area at elevation 135 ft 3 in (41.22m), and subsequently
remove it from containment.
–
Removal of the PCCWST and the water distribution bucket located inside the
containment shield building will take place externally by cutting out concrete
sections in the shield building roof, and accessing the steel tank liner and water
distribution bucket from above. Cranes, external scaffolding and selected cutting and
demolition equipment will be required.
–
Removed sections will be lowered to ground level by the crane, and they will be
transferred to the laydown area for further size reduction

Equipment removed from containment will be transferred to the Decommissioning
Building for radioactive contamination monitoring and ex situ decontamination, if
required.

PCS components will be disposed as free release, recyclable or to the LLW repository, as
appropriate.
Hazards associated with decommissioning the PCS include radiation hazards that may be
associated with working inside containment and industrial hazards associated with
construction, working at height, and heavy lifting/dropped loads.
5.2
Shielding and Containment
5.2.1
Existing Shielding and Containment
The AP1000 plant design incorporates significant shielding and containment designed to be
effective under all conditions anticipated in the plant life cycle.
The containment building is an integral part of the overall containment system. It prevents
releasing airborne radioactivity following postulated design-basis accidents, and shields the
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reactor core and reactor coolant system during normal operations. Personnel airlocks and
equipment and maintenance hatches permit access to containment. The health physics office
in the annex building can control access to containment. The control room is shielded against
radiation so that it is possible for personnel to occupy it under accident conditions. The fuel
handling and storage facility is designed to prevent inadvertent criticality and to maintain
shielding and cooling of spent fuel.
The shield building provides shielding for both the containment vessel and the radioactive
systems and components located in the containment building. During accident conditions, the
shield building provides the required shielding:


Radioactive airborne materials that may be dispersed in the containment
Radioactive particles in the water distributed throughout the containment
The cylindrical section of the shield building also protects the containment building from
external events and is a key component of the PCS.
The auxiliary building provides protection and separation for the seismic Category I
mechanical and electrical equipment located outside the containment building. It provides
protection for the safety-related equipment against internal or external events, and it also
provides shielding for the radioactive equipment and piping housed in the building. In the
auxiliary building, the fuel handling area is designed to prevent inadvertent criticality, to
maintain shielding and spent fuel cooling; and to provide protection for the spent fuel
assemblies, the new fuel assemblies, and the associated radioactive systems from external
events. Also, the fuel handling area is constructed to prevent release of airborne radiation
after any postulated design basis accident that could result in damage to the fuel assemblies or
associated radioactive systems and cause unacceptable site boundary radiation levels. Also,
the control room inside the auxiliary building is shielded against radiation so that personnel
can continue to occupy it under accident conditions.
Certain areas of the annex building, such as the hot machine shop and the control support
area, have shielding to protect against low level radiation from either internal sources or
external sources under accident conditions. This is accomplished by using either reinforced
concrete walls or reinforced masonry walls.
As described in the outline decommissioning strategy (Section 5.1.1), wherever possible, the
existing shielding and containment in each plant section remains in place until all the
radioactive equipment has been removed and all required decontamination has been
completed.
To allow decommissioning equipment access, some shielding modifications are expected. To
ensure that any shielding changes meet ALARP considerations, revised shielding assessments
will be performed as part of the decommissioning safety assessment.
5.2.2
Temporary Shielding and Containment
Whilst the shielding and containment incorporated into the AP1000 plant design will be
retained for as long as possible during decommissioning, there will be some requirements for
temporary shielding and containment to exist, including:

Temporary decommissioning facility shielding and containment. To reduce personnel
exposure to radioactivity, this facility will be designed to provide maximum remote
working opportunities during decommissioning waste processing and handling.
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
A temporary HVAC system and a mobile decontamination unit will be needed for
auxiliary building decommissioning.

A temporary HVAC system and a mobile decontamination unit will be needed for
temporary decommissioning facility decommissioning.

Temporary covers and shielding will be needed for open pipe and vessel ends exposed
during cutting operations and dismantling.

A temporary shielding structure will be produced to cover the reactor pressure vessel and
internals during lifting, upending, and transporting to the processing facility.

Temporary enclosures and tents with HVAC facilities may be used around dismantling
works where there is a risk of generating airborne radioactive material.
Remote Operations
Where possible, dismantling operations for highly radioactive components will be performed
remotely to reduce occupational radiation exposure (see Section 7.4.1). Remote dismantling
operations may take place behind shielding barriers, underwater, in fume hoods and in
inaccessible areas.
5.3
Progressive Hazard Reduction
5.3.1
Regulatory Strategy
As part of the ND’s regulatory strategy, the decommissioning programme proposed by a
licensee must demonstrate that it systematically and progressively reduces the hazards
presented by the nuclear facility [Reference 33].
Decommissioning may proceed as a continuous activity, or if there are safety benefits, as a
series of sequential stages, the end result of each stage being a significant reduction in hazard.
The hazard level that a particular plant on a site poses influences each stage’s order, timing
and extent. Actions should be prioritised based on the need to reduce the largest hazards or
those with high risk. The ND will require licensees to justify the order and timescales on
which they address each hazard.
Generally, the ND expects the most radioactive or mobile material to be removed or
immobilised (or both) in the shortest time possible. Further actions should follow on
timescales appropriate for the remaining hazards they address. Eliminating the highest
hazards in a nuclear facility should not diminish consideration of the residual hazard(s). In
some cases, actions that temporarily increase risk may be required to reduce the overall
hazard. Risks at each activity stage must be substantiated and demonstrated to be acceptable
and ALARP.
These principles are inherent in the AP1000 plant decommissioning sequence, as described in
Section 5.1. The sequence is a logical one that progressively reduces hazards. The following
subsections demonstrate this for each stage of decommissioning.
5.3.2
Stage 1 Decommissioning
The fuel rods are the plant components with the highest residual radioactivity. As can be seen
in Section 5.1.2, the first step of Stage 1 decommissioning occurs when the fuel rods are
removed from the reactor to the spent fuel pool. The reactor cavity and other areas that have
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been directly exposed to the fuel will have the next highest residual radioactivity.
Consequently, the next step of the decommissioning sequence is decontaminating these areas.
Residual radioactivity levels will be reduced by removing the radioactive inventory of service
systems no longer needed for decommissioning.
Contamination mapping of potentially radioactive systems and structures will be carried out
to ensure that:

The health and safety risks associated with maintaining and dismantling the radioactive
systems are understood

The decontamination and radioactive waste treatment options can be assessed
The results of these surveys will inform the selection of cleaning and decontamination
processes of the non-decommissioning service systems. Decontamination of contaminated
circuits will further reduce residual radioactivity levels before the commencement of
radioactive component removal.
As decommissioning progresses, the radiological control boundaries will be successively
reduced along with the progressive hazard reduction.
ILW and LLW will continue to be generated throughout the decommissioning process.
Therefore, the existing plant radwaste systems must be retained for as long as possible. The
precise time that the radwaste systems will be decommissioned cannot be specified at this
stage. However, it is assumed that it will be during Stage 3.
As soon as it has cooled sufficiently, the remaining spent fuel will be placed into dry storage
canisters and transported to the spent fuel store.
5.3.3
Stage 2 Decommissioning
To meet the requirements of decommissioning activities, existing service systems will be
modified and temporary service systems will be installed.
As soon as possible, after removing the final spent fuel, the radioactive inventory of the spent
fuel pool system and fuel handling system will be removed, and post-operational clean out
and in situ decontamination will be carried out. This will ensure that residual radioactivity
levels are ALARP before converting the fuel handling building to an ILW waste processing
facility. This conversion is in line with the principle of converting existing parts of the plant
to facilitate decommissioning. The spent fuel racks will be dismantled, and the pond will be
cleaned and decontaminated allowing the area to be used as a decontamination and waste
reduction area. Potential hazards associated with these decontamination and waste reduction
processes will also be reduced. For example, the volume reduction process will be designed
to minimise personnel exposure, and minimise and contain the generation of dust.
The first systems dismantled will be easily removable non-radioactive systems and service
systems that are not required for decommissioning. This reduces hazards associated with
subsequent decommissioning activities by removing potential obstructions and creating
space.
Further contamination mapping will be performed prior to the dismantling of radioactive nondecommissioning service systems. At this point, the reactor pressure vessel and internals are
the components with the highest residual radioactivity. However, these components can only
be removed when the radioactivity levels have dropped to the point where the disassembled
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pieces can be transported within the regulations applicable at that time. This may be a
significant period of time and will result in the plant being maintained in safe storage
(SAFSTOR) for this period. For this reason, the reactor pressure vessel and internals will not
be removed until Stage 3 of decommissioning. The sequence for removing radioactive
equipment will continue to follow the principle of removing components with the highest
residual radioactivity. Section 5.1.7 outlines the sequence for removing the steam generators.
Radiation and security controls will be implemented to both determine the radioactive
contamination levels of dismantled equipment and to confirm that it is securely protected for
transportation. This will allow clean equipment to be routed to an interim clean storage area.
This will reduce the volume of waste sent to the temporary decommissioning facility, to the
radwaste building, and to the converted fuel handling building. Radioactive components will
be transported to the appropriate waste handling area for decontamination, processing and
packaging.
5.3.4
Stage 3 Decommissioning
Contamination mapping of the remaining potentially radioactive systems and buildings will
be performed at the end of Stage 2 and at the beginning of Stage 3.
By this point the reactor pressure vessel (RPV) and the internals radioactivity levels should
have dropped sufficiently to allow them to be removed. The internals will be removed,
wrapped, and taken to the processing area for decontamination and processing. Section 5.1.4
outlines the procedure under which the reactor pressure vessel will be removed.
After removing the RPV, non-radioactive decommissioning service systems will be
dismantled and removed and replaced by temporary service systems, as required. The
non-radioactive systems will be removed first because this work will be easier than removing
the radioactive systems. The radioactive decommissioning systems will be systematically
decommissioned as specified in Section 5.1.4. Residual hazards for each system will be
progressively reduced by removing the radioactive inventory, post-operational clean out, in
situ decontamination, dismantling and removal of components to the waste processing areas.
In the waste processing areas, in situ decontamination will be performed, as appropriate, and
the waste will be processed, packaged and transported to the appropriate interim store or final
disposal repository, if available.
At this time, the only potential radioactive contamination remaining will be in structures and
concrete in the following:





Containment building
Shield building
Auxiliary building
Radwaste building
Temporary decommissioning facility
The contamination mapping carried out at the beginning of Stage 3 will establish this. Any
activated or contaminated concrete will be removed under controlled conditions and
processed separately, as described in Section 5.1.4.
After removing the contaminated concrete, the containment building, shield building and
turbine building will be decontaminated in situ and dismantled. Radioactively contaminated
components from the containment and shield buildings will undergo ex situ decontamination,
processing, packaging and disposal.
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Note that the turbine building could be dismantled during Stage 2 because the systems in the
building are not expected to be contaminated, and will be dismantled and disposed of during
the first 2 years of Stage 2 decommissioning. However, other plant areas that present a
greater residual risk are expected to be decommissioned first; and so dismantling the turbine
building will be delayed until the containment and shield buildings are ready to be
dismantled.
The last potentially radioactively contaminated buildings to be dismantled will be the
auxiliary building, the radwaste building and the temporary decommissioning facility. This
will allow the waste processing and handling facilities in these buildings to remain in service
for as long as possible. The auxiliary and radwaste buildings will be dismantled first, and
radioactive waste will be transported to the decommissioning facility for decontamination,
processing and packaging. To facilitate this work, temporary decommissioning service
systems and a mobile decontamination unit will be provided. The temporary
decommissioning facility will be the last potentially radioactively contaminated building to
be dismantled. It will be decontaminated in situ prior to dismantling, and waste will be
classified and disposed of via appropriate routes.
The remaining buildings and systems will all be non-active and will be disposed of according
to the procedures in place at the time.
5.4
Preventing Early Foreclosure of Options
5.4.1
Design
The modular design of the AP1000 plant means that modules can be removed in a similar
manner to the way they were installed, albeit with some segmentation to facilitate lifting
operations (see Section 4.6.1). Likewise, the design facilitates potentially large and complex
decommissioning activities (for example, removal of the largest, heaviest and most highly
radioactive components, as described in Section 5.1), thus avoiding irreversible construction
processes.
5.4.2
Operation and Maintenance
To avoid early foreclosure of decommissioning options, the Licensee will need to ensure that
the plant is appropriately operated and maintained. Section 4.2 outlines proper operation and
maintenance principles, including:

Preventive maintenance programmes will be established, including SSC surveillance,
inspection and maintenance, in accordance with appropriate safety margins, engineering
practices and quality levels.

Operational procedures ensuring that radioactive contamination does not spread
throughout the plant will be established.

Decontamination will be carried out as required throughout the plant operating lifetime;
this will minimise the potential operator dose uptake during decommissioning operations.

An appropriate inspection and maintenance programme will be developed and
implemented to reduce the probability of structure, system and component failures
resulting from aging (see Section 4.2.4).
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
Robust data collection and record management procedures will be implemented to handle
the large amount of monitoring, process equipment and maintenance information that
must be retained to facilitate decommissioning.

Wherever possible, the Licensee will also implement lessons learned from experience,
industry good practice and guidance to improve plant operation and maintenance (see
Section 4.4).
5.5
Decommissioning Technology
5.5.1
Currently Available Technology
Decontamination and dismantling techniques and tools for nuclear facilities are widely
available, as a large number of successfully completed decommissioning projects have
demonstrated [Reference 30, 37, 38, 40]. All decommissioning tasks, especially ones
concerning reactors, can be accomplished using existing techniques for which expertise is
already available.
The AP1000 plant can be decommissioned using currently available technology; novel
technologies or pieces of equipment are not needed. Existing technology used in the
decommissioning methodology includes:

Temporary lifting and handling equipment to clean and decontaminate the plant.

Cranes to move heavy and large components, such as steam generators.

Remote in-cell handlers to facilitate dismantling of the plant.

Dedicated reagent tanks and pump skid for the chemical decontamination process.

A mobile encapsulation plant for disposing of solid waste, as required.

Scabbling to reduce/remove contaminated wall surfaces.

Diamond wire cutting technique to remove contaminated concrete.

Radioactive monitoring equipment.
When the time comes for the plant to be decommissioned, the techniques to be adopted will
be assessed to ensure that BAT and ALARP processes are used throughout the
decommissioning process. However, currently recognised decontamination techniques are
summarised in Figure 5-2, and currently recognised decommissioning techniques are
summarised in Figure 5-3 [Reference 41].
5.5.2
Implementation of New Technology
As outlined in the previous section, the decommissioning of the AP1000 plant can be
completed using current technology. However, the decommissioning strategy will be
reviewed and updated throughout the plant lifecycle to take into account the current status of
the plant, availability of new or enhanced technologies, and changes in regulations.
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Deferring decommissioning facility construction until the end of operations allows the latest
techniques and technologies to be used, and the prevailing regulatory requirements to be
implemented.
5.6
Feasibility Reviews
5.6.1
Licensee Reviews
The Licensee will perform feasibility reviews to validate methodology/techniques used to
decommission the AP1000 plant. The initial decommissioning plan will be reviewed and
updated throughout the operational phase; the final decommissioning plan will be reviewed
and updated during decommissioning.
The arrangements for managing decommissioning activities will also be reviewed regularly
as part of the Licensee’s compliance with the CDM Regulations [Reference 6]. These reviews
will ensure that:

Construction, dismantling and demolition work is performed without risk to the health
and safety of any person so far as is reasonably practicable.

Throughout decommissioning, the standard of welfare facilities is adequate for all
persons.

Any structure designed to be used as a workplace has been designed considering the
provisions of the Workplace (Health, Safety and Welfare) Regulations 1992.
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phase
5. Chemical
pastes
4. Chemical gels
removal
Surface
Mechanical
Liquefaction
Degradation
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55
Figure 5-2. Decontamination Techniques [Reference 41]
9. Wet abrasives
8. Ultrasonic
cleaning
2. Swabbing
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1. Drill and
1. Melting
1. Thermal
spall/expansive
degradation
Sponge
2. Detergent &
2.
Ligh
grout
blasting
surfactants
ablation
2. Microbial
3. Organic
2. Jackhammer
solvents
CO² blasting
3. Foam
3. Laser
degradation
decontaminat 3. Scarifying/
Ablation
High pressure 4. Exothermic
ion
scabbling
metallised
liquid nitrogen
4. Microwave
powders
blasting
4. Water
4. Grinding/ shaving scabbling
(thermite)
flushing
/ planning
5. Freon/Arklone
5.
Strippable
jetting
5. Steam
5. Metal / concrete
coatings
cleaning
6. Wet ice
milling
blasting
6. Low/high
pressure
7. Vibrating
water jets
abrasive media
Cleaning
Wet surface
Blast cleaning 1.Vacuuming 1.Bleaching
cleaning
Dry surface
1.Strong mineral 1.Electropolishing/1.
acids
Electroetching 2.
2. Bases &
2. Electro kinetics
alkaline salts
3.
3. Complexing
4.
agents
Abrasive
methods
Electrochemical
Decontamination
Techniques
Dissolution
Chemical
Figure 1: Decontamination
techniques grouping
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Foam Filling
Painting
Passive aerosol generation
(fogging)
2.
3.
4.
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1.
Fixing
contamination
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Laser cutting
High pressure water jet cutting
12.
13.
56
Figure 5-3. Decommissioning Techniques [Reference 41]
Electrical cutting
11.
Power nibblers
7.
Thermal cutting
Shears (crimp & shear tool)
6.
10.
Sawing (slitting saws, reciproating saws)
5.
Orbital cutters
Oxyacetylene torches
4.
Abrasive cutting wheels
Linear shaped charges
3.
9.
Arc Saw
2.
8.
Plasma arc
1.
Cutting
Decommissioning
techniques
Decommissioning Techniques Grouping
Ball & Chain
‘Spider’
Diamond sawing or
coring
4.
5.
Drilling and rock
splitting
2.
3.
Blasting
1.
Demolition
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UK AP1000 NPP Decommissioning Plan
HSE reviews
The HSE, in consultation with the environment agencies, will perform quinquennial reviews
(QQR) of the Licensee’s decommissioning strategies to ensure that they remain soundly
based as circumstances change. Guidance on the QQR process is provided in Appendix 7 of
the HSE Technical Assessment Guide T/AST/026 “Decommissioning on Nuclear Licensed
Sites” [Reference 33].
5.7
Strategy for Safety Systems
5.7.1
Safety Systems Incorporated in the Design
The AP1000 plant design incorporates a range of Engineered Safety Features (ESFs) that
protect the public if radioactive fission products are accidentally released from the reactor
coolant system [Reference 9, Ch 6 and Reference 28 Section 6.4]. The ESFs include:







Passive Containment Cooling System (PCS)
Passive Core Cooling System
Containment Isolation System
Containment Hydrogen Control System
Containment Leak Rate Test System
Main Control Room Emergency Habitability Systems
Fission Product Removal and Control Systems
Each of these systems will remain functional whilst fuel remains in the reactor. The main
control room emergency habitability systems will be maintained whilst the control room
function is required.
In addition to the ESFs, the AP1000 plant has a number of auxiliary systems [Reference 9,
Ch 9] that provide safety and support functions during decommissioning:



Fuel storage and handling
Water systems
Process auxiliaries:
–
–
–
–
–





Compressed air
Plant gas
Sampling systems
Drainage
Chemical volume control system
Air-conditioning, heating, cooling, and ventilation systems
Fire protection system
Communication system
Plant lighting system
Standby diesel generators and diesel fuel oil system
The fuel storage and handling systems will remain operational until all fuel is removed to dry
storage. The other systems will remain functional during decommissioning both while their
services are required, and while the residual risks they are designed to mitigate are still
present. Where the above systems cannot be decommissioned progressively, it may be
necessary to bring in temporary equipment to provide these services. These requirements will
be assessed in detail in the final decommissioning plan.
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The AP1000 plant has Emergency Response Facilities [Reference 28, Section 13.3] which
include the Main Control Room, the Operational Support Centre and the Technical Support
Centre. During an emergency, these areas provide staff communication and assembly points.
Furthermore, the Technical Support Centre also provides plant management and technical
support to operating staff during an emergency. These facilities will be retained as long as
possible during decommissioning. If necessary, to ensure that effective emergency response
facilities are available throughout decommissioning, appropriate temporary facilities will
replace them.
5.7.2
Plant Management Safety Systems
In addition to the safety systems incorporated into the design, the Licensee will need to
implement appropriate plant management safety systems to cover decommissioning
activities. The Licensee will be responsible for plant safety and environmental management
throughout both the operating life and eventual decommissioning of the plant. Westinghouse
will liaise with the site Licensee regarding safety and environmental management throughout
all phases of the plant life cycle [Reference 15].
Westinghouse can support the Licensee by providing technical and design information that is:


Relevant to decommissioning
Assists the Licensee in:
–
–
–
Choosing the decommissioning strategy (Section 4.3.1)
Preparing the initial decommissioning plan (Section 4.3.2)
Creating the CDM file for decommissioning work packages (Section 4.3.4)
Knowledge transfer and the management arrangements developed to support AP1000 plant
operation are anticipated to provide the foundations of the required decommissioning process
arrangements. However, a review of this will be performed while the operational
management arrangements are created to ensure that this is true. Decommissioning
arrangements are expected to be reviewed every five years.
5.7.3
Other Safety Systems Required During Decommissioning
In addition to the plant safety systems incorporated into the design (Section 5.7.1) and the
plant management safety systems (Section 5.7.2); the decommissioning activities may
require:


Safety systems to be modified as dismantling progresses
Other safety systems for temporary decommissioning facilities to be added
These requirements will be identified in the final decommissioning plan.
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6.0
TIMING OF DECOMMISSIONING
6.1
Timing of Decommissioning
6.1.1
IAEA Decommissioning Strategies
There are three main options that the IAEA recognise as viable strategies for
decommissioning nuclear power plants [Reference 16]:

Immediate dismantling is the strategy in which the equipment, structures, components
and parts of a facility containing radioactive material are removed or decontaminated to
a level that permits the facility to be released for unrestricted use as soon as possible
after permanent shutdown. In some cases, where unrestricted release is not feasible, the
facility may be released from regulatory control with restrictions imposed by the
regulatory body. The implementation of the decommissioning strategy begins shortly after
permanent termination of operational activities for which the facility was intended,
normally within two years.

Deferred dismantling is the strategy in which the final dismantling of the facility is
delayed and the facility is placed into long term storage where it is maintained in a safe
condition. This strategy may involve some initial decontamination or dismantling, but a
major part of the facility will remain for a certain time period in a caretaker mode. This
time period might range from a few years to over 50 years, after which time the
decommissioning process will be completed and the facility can be released from
regulatory control. The deferred dismantling option is often used at multi-facility sites
when one or more of the facilities are shut down while others continue to operate. This is
especially true of facilities that share some common systems.

Entombment is the strategy in which the radioactive contaminants are encased in a
structurally long lasting material until the radioactivity decays to a level that permits
release of the facility from regulatory control. The fact that radioactive material will
remain on the site means that the facility will eventually become designated as a near
surface waste disposal site and criteria for such a facility will need to be met.
Although the assumed strategies tend to be classified as either immediate dismantling or
deferred dismantling, quite a few variations exist within these two categories [Reference 9].
For example, some utilities propose what could be considered to be a “rapid” immediate
dismantling, with all work being completed in about 10 years. Meanwhile, others consider a
more prolonged dismantling period of between 20 to 40 years as immediate dismantling.
Under the deferred dismantling option, a variety of deferral or dormancy periods are
considered. This results in dismantling being completed in periods ranging from about 40 to
around 100 years. The parts of the plant for which dismantling is deferred also varies. On
some sites, it is effectively the dismantling of significant radioactive plant and structural
parts, such as the reactor that is being deferred. Meanwhile, all other parts of the plant and
buildings are dismantled more “immediately”. Additionally, in a deferred dismantling
strategy, in some cases, the work is assumed to require 24-hour onsite staffing, while in other
cases, remote surveillance measures are considered acceptable. Following a safe enclosure
period, some utilities consider that radiation levels will have sufficiently reduced to allow
simpler reactor dismantling technologies to be used (for example, that fully remote operations
will not be required).
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The IAEA identifies immediate dismantling as the preferred decommissioning strategy
[Reference 17]. However, they recognise that when all relevant factors are considered there
may be situations where immediate dismantling is not practical. These factors may include:

Availability of:
–
–
–



6.1.2
Waste disposal or long term storage capacity for decommissioning waste
A trained workforce
Funds
Co-location of other facilities on the same site requiring decommissioning
Technical feasibility
Optimisation of the radiation protection of workers, the public and the environment
UK Government Policy on Timing of Decommissioning
The UK government policy on decommissioning [Reference 1] states:
“Decommissioning operations should be carried out as soon as reasonably practicable,
taking all relevant factors into account as provided for in the relevant operator’s strategy
and plan. The Government recognises that decommissioning operations may, however,
involve two or more separate stages spanning a number of decades. It may also be more
appropriate to delay particular operations to benefit from new or developing technologies or
from further development of existing best practice, or to take advantage of radioactive decay.
The Government confirms that, as with regulatory approval, the relevant factors, and their
respective importance, can only be determined on a case-by-case basis.”
6.1.3
Westinghouse Assumption
The nature of the AP1000 NPP design is amenable to all the decommissioning strategies
described in Section 6.1.1. However, for the purposes of GDA, Westinghouse assumes that
the decommissioning strategy will be immediate dismantling. This is consistent with both the
preferred IAEA decommissioning option (see Section 6.1.1) and the current UK Government
policy (see Section 6.1.2).
This assumption presumes that suitable waste disposal sites or interim waste storage facilities
are available for both low level and other intermediate and high level radioactive wastes.
Advantages of the immediate dismantling strategy include:

All radioactivity above specified levels is removed and properly disposed of or stored at
an interim facility as soon as feasible, thereby minimizing risk to population and the
environment.

The site may be reused in a timelier manner than under alternative strategies and is
appropriate for any suitable desired end state of the site.

The operating workforce, which is highly knowledgeable about the facility, is available to
support (and possibly plan and perform) the decommissioning activities.

The social impact of shutdown on the local community is potentially limited.
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
Utilising currently available waste disposal facilities removes any uncertainty with
respect to their future availability.

Potential cost savings resulting from future price escalation (because most activities
undertaken during immediate dismantling would also be performed during deferred
dismantling).

This approach does not require the significant civil engineering and maintenance effort
associated with deferred dismantling, where the facility must be placed into long term
storage to prevent deterioration of structures and systems and to maintain a safe condition
for an extended period.

Reduced maintenance and security costs.
The disadvantages of the immediate dismantling strategy include:

The potential for higher worker exposure (because there will be less time for radioactivity
to decay).

A larger initial commitment of financial resources.

A larger immediate commitment for waste disposal or storage space.
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7.0
DECOMMISSIONING HAZARDS AND CHALLENGES
7.1
Experience of Decommissioning
7.1.1
Decommissioning Experience
Westinghouse has a wealth of nuclear decommissioning experience, including
decommissioning PWRs and other reactor types. Westinghouse will draw upon this
experience to assist the Licensee in developing a safe and cost-effective decommissioning
strategy for the AP1000 NPP.
A non-exhaustive list of Westinghouse’s global decommissioning experience includes:

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Fort St Vrain (US)
Vandellos 1 (Spain)
Zorita (Spain) (PWR)
CIEMAT Research Facilities (Spain)
Three Mile Island (TMI) 2 (US) (PWR)
Yankee Rowe (US) (PWR)
Trojan (US) (PWR)
Qinshan (China) (PWR)
Connecticut Yankee (US) (PWR)
San Onofre (US) (PWR)
Shoreham (US)
Fukushima 2 (Japan)
Forsmark 1/2/3 (Sweden)
Oskarshamn 1/2 (Sweden)
TVO Olkiluoto 1/2 (Finland)
Chooz A (France) (PWR)
KNK (Germany)
Past experience indicates that record keeping and plant knowledge are important to facilitate
the decommissioning process. The decommissioning plan should reflect that retaining this
knowledge base would be a major contributor to a safe, efficient and cost-effective
decommissioning process.
Westinghouse believes the decommissioning experience accumulated thus far indicates that
the most appropriate philosophy for decommissioning is the same as that embodied in the
AP1000 plant construction plan: Remove components/modules as complete units as far as
possible [Reference 30].
The first AP1000 plants are being constructed in both China and the USA. These plants will
probably reach the end of their operational life before any plant built in the UK. If these
plants are subject to immediate dismantling, then the lessons learned from their
decommissioning may benefit the final decommissioning plan required for any UK AP1000
plant.
7.2
Decommissioning Hazards
7.2.1
Significant Hazards
The major hazards anticipated during the decommissioning process are industrial and
radiological. From past decommissioning experience, these two hazards can interact and
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complicate each other. For example, inadequately considering decommissioning activities
during structural design has caused difficulties in retrieving radioactive waste, and in some
cases, has hampered sampling to determine the hazardous radiological characteristics.
The major anticipated hazards are as follows:

Radiation Hazards
–
–
–
–

Industrial Hazards
–
–
–
–
–
–
–
–
–
–
–

Heavy/dropped loads
Working at height
Tripping/falling
Fire
Noise
Vibration
Dust
Electric shock
High temperatures
High pressures
Enclosed spaces
Toxic Materials
–
–
–
7.2.2
Direct radiation
Radionuclide inhalation/ingestion
Loss of containment
Radioactive waste
Residual chemicals
Decontamination chemicals
Non-radioactive waste
Identification of Hazards
Referring to the final decommissioning plan, a decommissioning safety case will be
developed by the operator. During decommissioning, the safety case will be updated when
necessary to reflect the impact of facility modifications and to address the changing nature of
the hazard.
The Pre-Construction Safety Report uses a number of checklists to identify hazards to
consider in the AP1000 plant Design Basis Assessment, Probabilistic Risk Assessment,
Severe Accident Analysis, and internal and external hazard assessment [Reference 28].
Collectively, these checklists provide comprehensive and systematic hazards identification.
These checklists will be reviewed, and hazards will be assessed for their applicability to the
decommissioning operations.
The final decommissioning plan will identify a number of individual decommissioning work
packages. The CDM will be implemented through either the operator’s or decommissioning
contractor’s management system will to ensure that the hazards associated with each
individual decommissioning work package will be fully evaluated before the project is
initiated. A key aim of CDM is to identify hazards early on so that they can be eliminated or
reduced at the design or planning stage, and so that the remaining risks can be properly
managed (see Section 2.5).
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Radiological and non-radiological hazards will be surveyed as an important input to the
decommissioning safety assessment and for implementing a safe approach during the work.
The survey should be conducted to identify the inventory and location of radioactive
materials and other hazardous materials. Special attention will be paid while characterising
any fissile material that may be left in the plant. Uncertainty about the fissile material
amounts could have severe consequences if criticality assessments are incomplete or wrong
[Reference 27].
The radiation and contamination surveys determine the:



Radionuclides
Maximum and average dose rates
Contamination levels for inner and outer surfaces throughout the facility
Contamination mapping may then be used to plot the location of the hazards to facilitate the
development of appropriate safety precautions and to help prioritise the decommissioning
works. Figure 7-1 illustrates an example of contamination mapping for the Oskarshamn 3
boiling water reactor (BWR) plant in Sweden.
Particular decommissioning challenges are associated with removing heavy components, for
example:





Steam generators
Reactor vessel head
Reactor vessel internals
Pressuriser
Reactor coolant piping/recirculation piping
Radiation hazards associated with decommissioning these heavy items may arise during
decontamination operations and when the internal contaminated surfaces are exposed during
opening or cutting operations. Other radiation exposures may occur during the preparation of
containment openings and when large items are handled in the designated laydown areas. The
safe hoisting and handling of these large items presents a significant industrial hazard. To
minimise the risks associated with such operations, lifting engineers will prepare lifting plans.
The polar crane used during site operation will be retained and ready for use during
decommissioning of heavy items. The polar crane structure has sufficient capacity to handle
heavy equipment with the addition of a larger capacity hoist module. Additionally, the polar
crane can accommodate the steam generators upper assembly between the girders.
7.2.3
Protection Measures
The measures to protect against significant hazards are similar to the control measures
identified in subsection 7.3.1.
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Figure 7-1. An Example of Contamination Mapping (Oskarshamn 3)
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7.3
Hazard Control Measures
7.3.1
Control Measures
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The control measures required for each individual decommissioning work package will be
developed as part of the final decommissioning plan and related decommissioning safety case
and project-specific CDM implementation. For each activity, consideration will be given to:

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
ALARP/BAT
Skills/Training
Written procedures
Area classification
Decontamination (chemical/mechanical)
Access/egress
Ventilation
Shielding
Remote operations/robotics
Scaffolding/safe lifting/cranes
Temporary facilities
Personal protective equipment
Dose rate/occupancy factors
Fire safety
Noise reduction
Monitoring
Laydown areas
Waste treatment/packaging
During Stage 1 decommissioning, the first contamination barrier is kept as it was during
operation, but the mechanical opening systems are permanently blocked and sealed (for
example, with valves or plugs) [Reference 9]. This allows ancillary systems to be
decommissioned and removed to reduce hazards from minor systems, whilst using
radioactive decay in the main hazards to naturally reduce the potential dose uptake from
activities.
Equipment will be removed floor by floor so that activities in a particular plant area are
controlled and can be easily supervised. Where practicable, work will be performed away
from access/egress points. Where such work is unavoidable, consideration will be given to
scheduling this work last, again, if practicable. Remaining contaminated items will then be
surveyed, and as before, appropriate work systems will be put in place to enable their safe
removal.
7.4
Manual and Remote Tasks
7.4.1
Remote Operations
The outcome of the contamination mapping exercise will largely determine whether human
or robotic decommissioning methods are selected (see Section 7.2.2). This information will
be used to assess the predicted cumulative dose for the task with the available shielding
layout. Other important factors in selecting human or robotic tasks are:


Robotic equipment design and availability
Working area accessibility
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Where possible, to reduce the occupational radiation exposure, dismantling operations for
highly radioactive components will be performed remotely. For example, Westinghouse has
removed the reactor pressure vessel and internals remotely at several nuclear sites (for
example, Fukushima 2, San Onofre, Yankee Rowe).
Examples of operations that could also be done manually in low radioactive contamination
areas (but are typically undertaken in high radioactive contamination areas) include:
7.4.2

Mechanical, thermal or hydraulic cutting (for example, diamond wire cutting, band saw,
abrasive water jet cutting, plasma arc cutting)

Grinding, blasting, welding, hammering

Lifting, jacking

Small component disassembly

Sludge recovery

Decontamination

Monitoring (for example, TV cameras, radiation)

Radioactive waste handling (ILW and HLW)
Manual Operations
To decommission the facility, many manual tasks need to be undertaken. Skilled personnel
should be engaged to assess, plan, and safely execute the required tasks under controlled
systems of work. This may include removing smaller items such as pumps and pipe work and
some decontamination and size reduction operations carried out in the Radwaste Building, for
example.
At certain stages, manual dismantling may be aided by partially or completely
decontaminating the structures or systems to be dismantled to reduce the level of required
radiological controls.
Low activity or clean plant and equipment are assumed to be removed manually by suitably
qualified and experienced personnel (SQEP).
Removal of Steelwork and concrete structures may involve using both SQEP and remotely
operated equipment.
7.5
Industrial Safety Hazards
7.5.1
Industrial Safety Hazards
Industrial safety hazards are identified in subsection 7.2.1, and control methods are shown in
subsection 7.3.1.
The anticipated significant non-radioactive hazards may include:

Access – Provisions for items such as access routes/equipment hatches/removable
gratings/lifting equipment have been included in the design, and would be subject to
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routine monitoring and controlled under a written system of work to ensure that
unnecessary dose is not accrued by transient workers, for example.

Building integrity –Buildings will have stood for more than 60 years, and therefore need
to be regularly inspected to ensure that their integrity remains, and that no unacceptable
hazards arise during decommissioning work. Where new access points need to be
provided, assessments will be made to ensure that the building structural integrity is not
compromised.

The environment – Environmental conditions (for example, flood risk) may change over
time. Therefore, periodic reviews of external risks to site integrity that the environment
will be undertaken to assess the impact (if any) on proposed decommissioning activities.

Lifting activities – Lifts will be assessed to determine their potential impact on
safety-critical systems, if a failure under load occurs. During decommissioning activities,
an increased number of lifts is expected compared to operational activities.

Manual operations – Skilled trades such as scaffolding/welding/fitting will be required to
do much of the decommissioning work. Controlling the tasks using a safe system of
work, pursuant to CDM, will ensure that risks to personnel are reduced to tolerable
levels. Working at height can pose a particular risk from falls, or dropped objects. A high
standard of safety principles and clear expectations will supplement the safe approach to
decommissioning.
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8.0
PLANT STATUS BEFORE DECOMMISSIONING
8.1
Plant Status at End of Operational Life
8.1.1
Status of Plant
For GDA purposes, after completing its 60-year plant design objective, the AP1000 NPP is
assumed to be decommissioned by immediate dismantling (see Section 6.1.1). Furthermore, it
is assumed that the plant systems, structures and components (SCC) will have been well
maintained during its operational life through the utility implementation of an effective
inspection and maintenance programme following principles similar to those laid out by the
IAEA [References 19 and 20].
Detailed planning for transition from operation to decommissioning should begin during
operation. Preparatory actions to implement the decommissioning plan should be taken
immediately after permanent shutdown to ensure a controlled transition and the best use of
resources [References 21 and 22].
Stage 1 decommissioning is assumed to commence following normal end-of-life electrical
power generation cessation and safe plant shutdown to stable conditions (see Section 10.1.1).
Immediately upon end-of-life cessation, the plant will be in an idle condition with all
necessary support facilities (for example, power, utilities, cooling, shielding, ventilation
systems, effluent treatment, drainage, security) remaining operational and maintained.
Nuclear fuel will be present in the reactor and the spent fuel pond. There will be some LLW,
ILW, and spent fuel waste (HLW) generated during normal plant operations present in their
dedicated on-site storage locations. On-site chemical inventories will be at low levels, and
tanks will contain residual unused chemicals. Some accumulation of potentially radioactive
corrosion products is to be expected in the radioactive water systems.
8.1.2
Pre-decommissioning Plant Characterisation
Following the decision to cease operations, a number of surveys may be needed to
characterise the plant condition in terms of chemical, physical and radiological condition
before decommissioning can commence. Thorough characterisation of the plant condition
will:


Provide advance guidance and direction to the required decontamination efforts
Identify areas that meet unrestricted release criteria
The surveys also provide a baseline that allows remediation progress to be monitored.
Many facilities will have radioactive, hazardous, or mixed wastes. A robust effort to
characterise the condition of the plant will yield important insight into both the handling of
these wastes and the actions that will need to be considered to ensure efficient and timely
disposal before decommissioning.
8.1.3
Pre-decommissioning Planning Management and Operational Issues
There are a number of issues related to decommissioning planning that the utility
management will need to address some years before the plant is finally shut down. Several of
these issues, listed below, were selected based on international experience [Reference 18].

Stakeholder issues including staff and public relations
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Regulatory and licensing issues including Environmental Impact Assessment
Organisational restructuring
Decommissioning activities and technology
Training and retraining
Defueling and fuel management
Waste management and disposal
Funding and finance
Project strategy, planning and contracting
Records and documentation
Rigorously identifying and storing the appropriate design and operational records needed to
support decommissioning before shut down is an important planning element. Such data
management and record keeping are essential to avoid losing nuclear facility operating
experience when it is shut down. Site Licence Condition LC6 requires documentation related
to licence conditions, including design and construction information relevant to
decommissioning, to be retained for a period of 30 years after the plant is decommissioned
and decontaminated. Arrangements to fulfil these requirements will be made with the
Licensee [Reference 14].
it is also fundamental to achieving safe and effective decommissioning that all waste streams
and routes for storing, transporting, and disposing of these wastes are fully identified and
agreed upon before decommissioning work commences. This includes identifying any
necessary waste transport containers, vehicles and associated infrastructure (for example,
facilities for scrap/material release, waste treatment and conditioning, waste characterisation,
interim storage buildings and final disposal). Westinghouse has submitted its integrated waste
management strategy [Reference 11] to identify its proposals and expectations in this respect.
The Licensee will be requested to provide an inventory of radioactive wastes on-site every
3 years to support the UK Radioactive Waste Inventory collated by the NDA on behalf of the
Government [Reference 39].
The AP1000 plant layout identifies areas within the site boundaries that are suitable to use as
LLW/ILW solid waste processing areas. These areas must be established before dismantling
work begins. The facilities will be large enough to allow at least two steam generators, one
reactor vessel, and sundry other equipment to be stored; and will include a remote handling
and waste reduction process area. The identified area will be close enough to tie into the
plant’s radioactive drain system and will be designed in detail and installed prior to
decommissioning.
8.1.4
Assumed Target Decommissioning End Point
The main objectives of decommissioning any industrial site, including a nuclear power plant,
are:
1. To ensure that the decommissioned site does not present unacceptable risks to either
human health or the wider environment
2. To allow the site to be released for future use (for example, for redevelopment or use as
an open space)
For the purposes of the plan, the decommissioning operations are assumed to include at least
the following:

Removal of all hazardous materials (including wastes)
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Dismantling of plant/machinery
Demolition of structures to 1 m below ground level
To achieve these objectives, the overriding requirement would be to safely remove
radioactive materials (to include spent fuel, radioactive wastes, and contaminated plant
components).
If appropriate, the utility operator will address the intended decommissioning end point and
subsequent management of contaminated land as part of their update of the Decommissioning
Strategy document. The assumed target decommissioning end point is described below:

Plant and equipment will be decontaminated and radioactive materials will be removed
and disposed of via the appropriate route.

Structures will be de-planted and demolished during final site clearance (FSC).

Buildings will be removed to a depth of 1 m below ground level.

Roads, car parks, underground services, and similar structures will be removed.

The radioactive drains and outfall will be removed, and foul and surface water drains will
be removed if they are less than 1 m below ground level.

Basements, if present, will be demolished to 1 m below ground level and any remaining
subsurface structures will be punctured to assist drainage.

Land requiring remediation will be identified and treated appropriately.

Ground will be appropriately landscaped, and land drains installed if required.

After FSC, the site will be released from its Nuclear Site Licence if it is considered
appropriate.
Note that the eventual end use will be subject to consultation and rationalisation with UK
government policy at the time, and may also be subject to conditions attached to the original
planning consent.
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9.0
DISPOSABILITY ASSESSMENT AND DECOMMISSIONING
9.1
Waste Streams
9.1.1
Decommissioning Waste
Table 3.5.10 outlines the estimated radwaste arising from decommissioning and dismantling,
and Appendices A3, A4, A5 and A6 of the AP1000 Environment Report offer additional
details [Reference 15].
The tables identify wastes generated from decommissioning large-volume components
(Appendix A3), small-volume components (Appendix A4), and from demolishing plant
modules (Appendix A6). In addition to specifying the expected waste classification level
(ILW or LLW) and the volume and mass for each type of waste, the tables outline how the
wastes will be pre-conditioned and disposed.
Pre-conditioning methods that will be used for decommissioning waste include:

Immobilisation in cementitious grout within a 3m3 RWMD approved box

Compaction (possible super compaction) into a 200 l RWMD approved drum and placed
into a half-height International Standards Organization (HHISO) container

Placement in “baskets” in an RWMD approved box (possibly grouted)

Monitoring and swabbing (over a period of time) with potential cleaning/decontamination
or size reduction

Size reduction and placement in HHISO

Collection in 200l drum
Disposal methods for decommissioning waste include:



Storage in site ILW store until UK repository becomes available
Transportation to LLW repository
Recycle or free issue (release/clearance from regulatory control)
Any assumptions made when decommissioning and dismantling radwaste was estimated are
identified in the notes accompanying the tables.
In addition to the components and modules identified as decommissioning radwaste above,
there may be additional radwastes associated with decontamination activities. These
secondary wastes are discussed in Section 9.5.
9.2
Waste Routes to Interim Storage
Waste routes out of the building to interim storage are discussed below.
9.2.1
High Level Waste
HLW (spent fuel) to be removed during Stage 1 decommissioning will be handled,
transported and stored similarly to the way spent fuel generated during the AP1000 NPP
operational phase is [Reference 12].
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Spent fuel will be handled, transported and stored using the Holtec HI-STORM 100U
underground storage system [References 12 and 15] or an equivalent system. Up to 32 spent
fuel assemblies will be loaded into a multi-purpose canister (MPC) positioned in a HI-TRAC
transfer cask in the cask loading pit inside the auxiliary building’s fuel handling area. The
MPC and transfer cask will then be transferred from the cask washdown pit to the railcar bay,
and loaded onto the low profile transporter. The MPC and transfer cask will be transported
out of the auxiliary building on the low profile transporter via the access hatch on the east
side of the building. Here, a HI-TRAC crawler will pick them up, and they will be transported
to the HI-STORM 100U spent fuel store via the route shown in Figure 9-1. Inside the spent
fuel store, the MPC will be stored in a vacant HI-STORM 100U cavity.
9.2.2
Intermediate Level Waste
Much of ILW that will be generated during decommissioning is related to decommissioning
of reactor components. Decommissioning these components occurs in the fuel handling
building after it is converted to an interim waste storage, decontamination, waste reduction,
processing and packaging area in Stage 2 decommissioning (see Section 5.3.3). The cranes
previously used for spent fuel handling will transfer the generated ILW to the railcar bay. The
mobile ILW encapsulation plant located in the railcar bay may be used to treat and package
operational ILW. This equipment is designed to encapsulate ILW RWMD waste packages
(drums or boxes).
Prior to being transported to the ILW store, the waste package will be placed in an overpack
that provides shielding, limiting exposure to operators or the public. A self-propelled trailer
will be used to move the waste packages to the ILW store. The trailer will leave the auxiliary
building via the access hatches on the east side of the building and will travel to the ILW
store via the route shown in Figure 9-1. Inside the ILW store, the packages will be placed and
recovered using an overhead crane. Shipment from the ILW store will only occur when a
national repository is available.
9.2.3
Low Level Waste
LLW will be bagged, collected manually, and transported to areas of the waste accumulation
room inside the radwaste building. The waste will be packaged in HHISO containers located
in the radwaste building loading bay. Either a fork truck or the overhead crane will offload
the HHISO containers. A fork truck specifically designed for container handling will
transport the HHISO containers from the radwaste building. Usually, the HHISO containers
will be transported by road from the radwaste building to the low level waste repository
(LLWR) at Drigg for final disposal. However, if the LLWR is unavailable for any reason, the
HHISO will be transported to LLW buffer store. The fork truck will leave the radwaste
building via the access hatches on the east side of the building and will travel to the buffer
store via the access road to the south of the radwaste building, as shown in Figure 9-1.
Large decommissioning items may be transferred to the laydown area located to the west of
the nuclear island (see Figure 9-1). If the large equipment is to be decontaminated, size
reduced and packaged, this will take place in the adjacent temporary decommissioning
building. Details of equipment and activities to take place in the decommissioning building
will be subject to detailed decommissioning planning. However, moving decommissioning
radwaste between the nuclear island, laydown area, decommissioning building and interim
radwaste stores will take place along the common access roads to the south and west of the
nuclear island (see Figure 9-1).
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Waste Transportation
Waste will be transported using appropriate transportation packages and vehicles. During
transportation site traffic will be controlled, and waste transfer will be marshalled to minimise
risking transportation accidents.
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Figure 9-1. AP1000 Plant Waste Facilities (No. 4, 5, 21, 22, 23, 24, 25, 26, 27, 29, 31) and Transport Routes
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9.3
Waste Stream Sensitivity to Decommissioning Processes
9.3.1
Effectiveness of Decontamination
Decontamination processes will be adopted to change the classification of the
decommissioning waste to the lowest level possible. Where possible, the aim will be to
reduce ILW to LLW and LLW to free release material. Effective monitoring of radioactivity
before and after decontamination is important in confirming the waste classification and
reducing quantities of ILW and LLW produced during decommissioning.
In each ILW and LLW waste classification, decommissioning waste mass is sensitive to the
effectiveness of the selected decontamination techniques (see Section 9.4).
9.3.2
Waste Form
The waste streams produced during decommissioning are sensitive to different
decommissioning processes because different processes may result in different forms of
waste being produced. For example:

Different size reduction processes may generate different amounts of radioactive fines,
which may produce:
–
–
9.3.3
Wastes that require stabilisation
Dusts that require filtration

Different decontamination processes may produce different solid or liquid secondary
wastes (see Section 9.5).

For ILW the stabilisation techniques and grout addition rates may vary according to the
form of the waste, and this will affect the total number of waste packages generated.
Waste Volume
The waste volume generated during decommissioning is sensitive to the decommissioning
processes adopted.
For example:

Adopting refined cutting techniques (for example, diamond wire sawing) can remove
contaminated concrete up to a specific contamination depth, and can minimise the
amount of concrete classified as radioactive waste.

Disposal of large items without size reduction may be possible, but would result in
disposal of a much larger waste volume.

Selecting compaction or super compaction will affect the volume of LLW produced.
The disposability assessment [Reference 15] identifies the mass of the radioactive
components requiring decommissioning, which is unlikely to vary significantly as a function
of different decommissioning methods. However, the waste radioactive components volume
is sensitive to the size reduction and packaging techniques that may be employed.
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9.4
Impact of Decontamination on Primary Waste Volumes
9.4.1
Decontamination
Decontamination will be used to reduce the decommissioned material contamination levels
where BAT and ALARP analyses indicate that it is beneficial. For example, where
practicable, decontamination will be used to recover wastes so that they can be reused or
recycled. Decontamination may also allow the radiological waste classification of primary
wastes to be reduced from ILW to LLW, or from LLW to free release for disposal through
conventional non-radioactive routes. Wastes generated during decommissioning will be
segregated into different waste types to allow optimum use of the various decontamination
routes available at the time of decommissioning.
The majority of waste arisings from decommissioning small and large volume components,
as identified in appendices A3 and A4 of the AP1000 UK Environment Report
[Reference 15], will undergo swabbing/monitoring to determine their radioactivity levels and,
if necessary, will be subject to decontamination. The aim is to reduce the radioactivity of the
waste so that it may be reclassified. The wastes will be monitored to assess if this has been
achieved, and further decontamination will be carried out if necessary. This process will
result in reductions in the volumes of ILW and LLW requiring processing and disposal.
Also, before demolition, plant modules will be monitored to assess the contamination levels
of the steel and concrete. Decontamination of the modules will be carried out to reduce the
radioactivity levels from LLW to free release levels where possible. Decontamination will be
aided by using:

Concrete walls with decontaminable coating

Steel surfaces with surface finishes that will prevent penetration of contamination (see
subsection 4.1.3).
Credit for the effect of decontamination on the estimated ILW and LLW arisings was not
taken in the disposability assessment [Reference 15]. At this stage no estimate of the benefit
that decontamination processes may provide in reducing ILW and LLW decommissioning
waste quantities was made.
9.5
Generation of Secondary Wastes
9.5.1
Liquid Wastes
Some decontamination operations will produce secondary liquid wastes. These may include:

Detergent wastes
Wet surface cleaning operations may produce detergent wastes. These will be collected
and monitored to assess radioactivity levels. When the radioactivity levels are above
acceptable limits, the wastes will be treated through the existing plant radwaste systems.
However, if this is not possible, wastes will be processed using mobile equipment.
Processing will typically involve a concentration step (for example, evaporation or
reverse osmosis) to reduce the volume of waste to the extent necessary to allow an
encapsulation plant to immobilise the concentrate in a cementitious grout. The liquids
remaining after the concentration step (for example, condensed evaporator distillate) will
be monitored to assess if further treatment is required before discharge.
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Chemical wastes
Decontamination processes, such as chemical dissolution, will produce chemical wastes.
These will be collected in the chemical waste tank, and pH and other chemical
adjustments will occur as required. The wastes will be monitored, and if the radioactivity
levels are deemed to be above acceptable limits, then the waste will be processed using
mobile equipment or combined with detergent wastes and processed as detergent waste.
9.5.2
Solid Wastes
Some decontamination operations will produce secondary solid wastes. These may include:

Filters
Some decontamination and decommissioning operations will produce airborne
particulates. These operations include cutting, abrasion, mechanical surface removal (for
example, scabbling), and demolition. These operations will be performed using the BAT
and ALARP techniques to minimise the amount of dust produced. However, some dust
capture by air filtration will be necessary, and this will generate spent filters as a
secondary waste product. Spent filters will be monitored to assess radioactivity levels so
that they can be processed and disposed of appropriately.

Swabs
Dry surface cleaning will generate contaminated swabs as secondary waste. These will be
monitored, classified and disposed appropriately.

Radioactive resins
An oxidative decontamination process will likely be used to decontaminate contaminated
piping circuits. This involves an ion exchange process using anion and cation resins to
absorb radioactive ions released by the process. These resins will become radioactive and
will be encapsulated in a cementitious grout in a similar way to operational ion exchange
resins.
9.5.3
Quantity of Secondary Wastes
The quantity of secondary wastes that will be generated from the decontamination processes
will be dependent on the methods employed and the extent of their use. As the initial
decommissioning plan is developed in detail, this will need to be defined, but as of this
revision, no estimates have been made.
9.6
Use of BAT & Waste Management Hierarchy in Decommissioning
9.6.1
Waste Hierarchy
The licensee will implement the waste hierarchy principles (avoid, minimise, recycle/reuse,
abate) during decommissioning operations (subsection 4.3.9). Throughout the plant
operational life and during decommissioning, the Licensee will review the AP1000 Integrated
Waste Strategy [Reference 11] to maximise waste reuse and recycling wherever possible.
The AP1000 plant design includes features that avoid and minimise waste production
(Section 4.1.2). Decontamination techniques will be selected to minimise the waste produced
and to reduce the radiological waste classification during decommissioning (Section 9.4).
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Where waste is produced it will be treated by appropriate techniques (foe example,
segregation, size reduction, encapsulation, packaging) to ensure that it is in the most
appropriate form for disposal.
9.6.2
BAT Assessment and Implementation
The plant has been designed and will be operated and maintained to avoid early foreclosure
of decommissioning options, which may restrict the implementation of BAT (Section 5.4).
The Licensee will regularly review the methodology/techniques used to decommission the
AP1000 plant throughout the life cycle (Section 5.6). As part of these reviews, the
methodology and techniques that are considered BAT will be regularly assessed. By deferring
decommissioning facility design and construction until near the end of operations, it will be
possible to ensure that the techniques associated with the most up-to-date BAT assessment
are built into the constructed decommissioning facilities.
9.7
HLW and ILW Transport and Storage
9.7.1
Transportation of ILW and Spent Fuel
Section 9.2 describes how to transport processed and packaged spent fuel and ILW from the
auxiliary building to the respective on-site stores.
9.7.2
ILW Store
Once decommissioning ILW has been processed and packaged in the auxiliary building, it
will be transported to the ILW store for storage until a national ILW repository becomes
available. Subsection 3.5.8.2 of the AP1000 Environment Report [Reference 15] describes
the ILW store proposed for the generic site.
The store will be a reinforced concrete structure with 1m thick walls that can be extended at
appropriate intervals to suit new ILW waste arisings. The ILW store incorporates a receipt
area with waste package assay equipment and a shielded vault serviced by a certified nuclear
crane. Office and administration space, and an equipment room housing HVAC and electrical
and mechanical equipment are included in an annex to the main store building.
Packages will be transferred into the receipt area through a shielded door, and then
transported by the crane to a position in the store vault determined using the tag information.
Package position will be recorded in the control log for ease of future retrieval. Packages will
be placed in the store, layer by layer, to limit the potential topple height. Layers will be
constructed from the furthest point of the store working backward to the receipt area. The
chosen transfer path for placing/retrieving a package will minimise the effective drop height.
The store design and operation will allow individual packages to be retrieved and visually
examined. Close circuit television (CCTV) in the import/inspection area will facilitate this.
The first phase of construction will provide an ILW store suitable for 20 years of ILW
production. Extensions to the store will be sized to accommodate future waste arisings and
are expected to be added in 20-year increments. The ILW store will be designed for a total
inventory of 60 years of operational waste arisings from one AP1000 plant. The ILW store
has a 100-year design life and could be used to both retain ILW after an AP1000 plant is
decommissioned and until the national ILW repository becomes available.
Every ILW waste package will be “finger printed” in the ILW store using a High Resolution
Gamma Spectroscope (HRGS) to monitor its radioactivity level before it is transferred to the
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ILW Store Vault. Waste package inventory records will be completed according to the
required regulations to maintain an inventory record of each waste package and its location
inside the ILW Store Vault. Because all waste packages sent to the store will be ILW and are
expected to remain ILW, no segregation will be required within the store vault.
When an ILW waste repository becomes available in the UK, the ILW waste packages will
be removed from the store and monitored again with the HRGS before being sent to the
repository, along with its associated waste package inventory record. The same facilities used
during placement of the ILW packages into the store will also be used to ship the ILW
packages (that is, the area at one end of the store building and the store building crane).
If the HRGS result of a package indicates the radionuclides in the package have decayed so
much that the package could be LLW, the package will be temporarily placed in an LLW
storage area. At that time, the LLW disposal facility will be contacted to ensure the
appropriate records are prepared for LLW disposal at that time.
All ILW packages will be visually inspected during handling, and if defects or external
damage are found, the package will be flagged as “rogue” and placed/sealed within a
secondary containment vessel (SCV) prior to storage. A “rogue” package might arise from:

Overfilling a package during encapsulation, causing spillage and contamination on the
outer surface.

Malfunction during lidding, causing an unsealed package.

Corrosion/damage to the package, resulting in a containment failure.
Any “rogue” package will be transferred to an SCV. The SCV is a container designed
similarly to the RWMD packages that is sized to fit over the RWMD 3m3 box/drum. If due to
quality assurance inspections an ILW package is found to be “rogue”, it will be inserted into
an SCV, lidded, and positioned in the store as normal. A small batch of empty SCVs will be
stored in the radwaste building until required.
9.7.3
Spent Fuel Store
Once all the spent fuel has been processed and packaged in the auxiliary building, it will be
transported to the spent fuel store for storage until a national HLW repository becomes
available. Subsection 3.5.8.3 of the AP1000 Environment Report [Reference 15] describes
the spent fuel store proposed for the generic site.
The proposed spent fuel store is Holtec International’s underground dry spent fuel dry storage
system, the HI-STORM 100U System, although alternative systems may be considered.
The HI-STORM 100U System is a vertical, ventilated dry spent fuel storage system. Holtec
and Westinghouse have confirmed that the Holtec equipment can fit in the AP1000 plant
areas that must be travelled to transfer spent fuel from the spent fuel pool to the underground
storage area. The system consists of three primary components:
1. HI-STORM 100U underground vertical ventilated module (VVM)
The VVM provides for storage of a multi-purpose canister (MPC) in a vertical
configuration inside a subterranean cylindrical cavity entirely below the top-of-grade.
The principal function of the VVM structure is to provide the biological shield and
cooling facility.
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The MPC storage cavity is defined by the cavity enclosure container (CEC), consisting of
the container shell integrally welded to the bottom plate. In the installed configuration,
the CEC interfaces with the surrounding subgrade for most of its height except for the top
region, where it is girdled by the top surface pad. The CEC is a closed-bottom, open-top,
thick-walled cylindrical vessel with no penetrations or openings. Thus, groundwater has
no path to intrude into the interior space of the MPC storage cavity.
Corrosion mitigation measures commensurate with site-specific conditions are
implemented on below-grade external surfaces of the CEC. All external and internal
surfaces of the VVM are coated with an appropriate surface preservative. An optional
concrete encasement around the coated external surface of the CEC may be added to
control pH at the CEC-to-subgrade interface. Nevertheless, in the structural evaluations, a
corrosion allowance equal to 3mm on the external surfaces of the VVM in contact with
the subgrade is assumed in the structural evaluations. The closure lid is a steel structure
filled with shielding concrete, and incorporates a specially designed air ventilation
system.
2. MPCs, each containing 32 spent fuel assemblies
The MPC and HI-TRAC in the HI-STORM 100U System are identical to those in the
Holtec above-ground system that has been used for several years. The MPC is a single
package, equally suitable for onsite storage, transport, and permanent disposal in a future
repository. The MPC is constructed entirely of stainless steel alloy materials with the
exception of the Metamic, a fixed neutron absorber, which is contained within the
canister for criticality control. The fuel assembly basket contained in the MPC is a
honeycomb multi-flanged plate weldment that forms the square fuel cells in the basket.
Complete edge-to-edge continuity exists between the continuous cells that provides an
uninterrupted heat transmission path, making the MPC an effective heat rejection device.

The top end of the Holtec MPC uses a closure system that includes a lid equipped
with vent and drain ports used to remove air and water and to backfill the canister
with inert gas (helium)

A closure ring used to provide a redundant confinement boundary for the MPC lid
The vent and drain ports are covered, helium leak-checked, and seal-welded before the
closure ring is installed. The closure ring is a circular ring that is edge welded to the
canister outer shell and lid. The MPC lid provides sufficient structural capability to
permit the loaded MPC to be lifted by threaded holes in the MPC lid.
The heat from the fuel stored in the core region of the basket is removed by the
thermosiphon (circulatory) action. As a result, high heat rate fuel (gamma radiation
emitted is proportional to the heat emission rate from the fuel) can be placed in the core
region, surrounded by cooler (and older) fuel in the periphery. This approach, known as
“regionalised” storage, is extremely effective in mitigating the dose that is laterally
emitted from a basket. The effectiveness of regionalised storage in reducing dose derives
from the fact that almost 95 percent of the dose from the basket comes from the
peripheral fuel; the inner region fuel is almost entirely noncontributory to the dose.
3. HI-TRAC transfer cask, which holds the MPC during loading operations
HI-TRAC is the acronym for Holtec International transfer cask – or “shuttle cask” – for
the HI-STORM 100U. HI-TRAC is a slim cylindrical cask with removable bottom and
top lids. HI-TRAC can be mounted on top of a HI-STORM 100U overpack to deliver or
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retrieve an MPC. HI-TRAC is a heavy-walled steel and lead cylinder with a water jacket
attached to the vessel exterior. The main structural function of HI-TRAC is provided by
carbon steel. Water provides the main neutron shielding, and lead provides the main
gamma shielding.
Spent fuel will remain within the spent fuel store for a determined period of time to
enable the heat-generating capacity of the spent fuel assemblies to reduce enough to meet
the required standards for the national Geological Disposal Facility (GDF). At the
proposed high burn-up rates, RWMD has estimated that dry cask storage for up to
100 years may be necessary to allow it to cool sufficiently before it can be transferred to
an approved RWMD disposal canister for final disposal. However, Westinghouse expects
that the repository design may be reconsidered on the basis of current worldwide
expectations from spent fuel characteristics, which would allow for shorter dry cask
storage periods.
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DECOMMISSIONING PLANS AND PROGRAMMES
10.1
Decommissioning Plans and Programme
10.1.1
Decommissioning Nuclear Island – Plans and Programme
Final decommissioning plans for the nuclear island will be developed in detail by the utility
before conducting decommissioning.
After 60 years of operation, both The normal end-of-life cessation of electrical power
generation and the safe shutdown of the plant to stable conditions occur. The outline
decommissioning plan assumes that the immediate dismantling strategy begins soon after
shutdown of the plant, usually within 5 years.
Stage 1 decommissioning may begin as soon as the plant stops generating power; however,
there is a need to allow the core barrel radioactivity levels to decay to a level where the barrel
can be safely dismantled and packaged in accordance with the requirements in force at that
time. It is anticipated Stage 1 of decommissioning will take approximately 10 years to ensure
that the final fuel removed has sufficient time to cool in the spent fuel pond.
Decommissioning Stages 2 and 3 will be approximately 6 years each.
This results in an overall schedule for immediate dismantling of the nuclear island of between
22 and 27 years.
10.1.2
Decommissioning Interim Waste Stores – Plans and Programme
detailed decommissioning plans for the interim waste stores will be developed by the utility
during the period at which waste and spent fuel are transferred to the repository.
LLW Store
Currently, LLW is disposed at the Low Level Waste Repository (LLWR) in Cumbria.
However, this facility is expected to close in 2050 [Reference 23]. Consequently, any LLW
generated while decommissioning a future AP1000 NPP will need to be disposed at a future
LLWR that is not yet specified. The LLW from decommissioning will be transported from
the site to the repository immediately after on-site packaging.
The LLW buffer store is a covered area comprising a concrete hard standing area with a steel
framed canopy designed to store half-height ISO containers (HHISO) that have been filled in
the radwaste building. The plant will be able to store up to 2 years generation of operational
LLW in the LLW buffer store. The decommissioning LLW volume may require a larger
LLW buffer store to accommodate 2 years of generation. If this is the case, then the storage
area and canopy may need to be extended. The LLW buffer store will only be dismantled
after completing all radioactive decommissioning activities at the end of Stage 3
decommissioning.
ILW Store
It is assumed that the national ILW repository will be available in 2040 [Reference 23], well
before plant operations cease and decommissioning operations begin. Transferring ILW from
the on-site interim ILW store will begin as soon as the ILW repository is available to receive
new build ILW. The ILW from decommissioning will be transported from the site to the
repository immediately after on-site packaging. However, the interim ILW store will have a
100-year design life. Thereafter, all ILW must be transferred to the repository. Before
reaching the ILW store 100-year design life, it must be decided either to immediately
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dismantle the ILW store or defer the ILW store dismantling until spent fuel stores are
dismantled 161 years after plant start-up. If safety analysis permits, it may be beneficial to
extend the ILW store life until completing all radioactive decommissioning activities. This
would allow the ILW store to temporarily house any ILW that may be generated during
dismantling of the spent fuel stores.
Spent Fuel Stores
The source of HLW in the AP1000 plant is spent fuel. The GDA AP1000 Disposability
Assessment [Reference 24] assumes that spent fuel will be overpacked for disposal. Under
this concept, spent fuel would be sealed inside durable, corrosion-resistant disposal canisters
manufactured from suitable materials, which would provide long-term containment of the
radionuclide inventory. Current Radioactive Waste Management Directorate (RWMD)
generic spent fuel disposal studies define a temperature criterion for the acceptable heat
output from a disposal canister. To ensure that excessive temperatures do not damage the
bentonite buffer material placed around the canister in the disposal environment, a 100°C
temperature limit is applied to the inner bentonite buffer surface [Reference 24]. Based on a
canister containing four AP1000 plant fuel assemblies, each with the maximum burn-up of
65 GWd/tU and adopting the canister spacing used in existing concept designs, it would
require on the order of 100 years for the radioactivity, and hence heat output, of the AP1000
fuel to decay sufficiently to meet this temperature criterion [Reference 24]. This cooling
period is built into the decommissioning programme for the spent fuel store described below.
However, RWMD proposes to explore how this cooling period can be reduced, and any
reduction or changes to the cooling requirements will affect the spent fuel store
decommissioning programme.
In the AP1000 plant, spent fuel is generated at the end of each fuel cycle at intervals of about
18 months, and each refuelling offload into the spent fuel storage pool is 68 fuel assemblies.
The spent fuel storage pool will be operated to retain capacity for one full core offload
(157 fuel assemblies). The spent fuel pool design allows 889 fuel assemblies to be stored.
Therefore, the spent fuel storage pool can hold ten refuelling offloads, representing
approximately 18 years of operation. When the spent fuel pool storage period ends, spent fuel
must be transferred to an interim spent fuel store to complete the 100-year cooling period.
The first HLW will be cool enough to be transferred to the HLW repository after this
100-year cooling period. The national HLW repository is assumed to be available in 2075
[Reference 23].
Each interim spent fuel store will have a 100-year design life. Unless the design life of the
HLW interim stores can be safely extended, the final phase of HLW interim store
construction will need to occur 60 years after plant start-up to ensure that the spent fuel
removed in Stage 1 decommissioning can be safely stored for 100 years before being
transferred to the HLW repository. All decommissioning HLW will be transferred to the
repository 160 years after plant start-up.
An assessment will need to be made when the spent fuel stores reach the end of their 100 year
design life, to determine if they should be immediately dismantled or if dismantling can be
deferred until all the HLW interim stores have been emptied 160 years after plant start-up.
The baseline programme assumes that the LLW, ILW and spent fuel stores will be dismantled
161 years after plant start-up, with completion of this decommissioning 163 years after plant
start-up.
10.1.3
Overall Decommissioning Programme
The overall decommissioning programme is shown in Figure 10-1.
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Figure 10-1. Outline Decommissioning Programme
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10.2
Disposability and Government Policy
10.2.1
Disposal of AP1000 Plant Decommissioning Wastes
Westinghouse has submitted a document that addresses disposal of operational and
decommissioning wastes originating from an AP1000 plant [Reference 12]. The
decommissioning wastes fall into the following categories:




Non-radioactive waste
LLW
ILW
Spent fuel
These wastes are similar to operational and legacy wastes from existing nuclear facilities, and
the disposability challenges are well known.
Existing recovery, recycling and disposal facilities are widely available in the UK to deal
with non-radioactive wastes.
Currently, LLW is disposed at the Low Level Waste Repository (LLWR) in Cumbria. This
facility has limited capacity, and although there are plans to increase capacity by adding new
vaults, the facility is still expected to close in 2050 [Reference 25]. Any LLW
decommissioning waste generated after operation of the first AP1000 plant has ceased after
60 years will need to be disposed of at a future LLWR that is not yet specified.
Currently, there are no existing ILW or HLW radioactive waste disposal facilities operating
in the UK. It is assumed that the national ILW repository will be available in 2040, and the
national HLW repository will be available in 2075 [Reference 23]. At that time, the ILW and
HLW decommissioning waste will be transferred to these facilities for final disposal.
10.2.2
RWMD Disposability Assessment
RWMD has concluded that ILW and spent fuel from both operation and decommissioning an
AP1000 plant should be compatible with plans for transport and geological disposal of higher
activity wastes and spent fuel [Reference 24].
RWMD expect that future refinements to the assumed radionuclide inventories of the higher
activity wastes and spent fuel will eventually support and substantiate these conclusions,
complemented by developing more detailed proposals for packaging the wastes and spent
fuel and a better understanding of the waste packages expected performance [Reference 24].
This information would be developed as part of the detailed site specific decommissioning
plan that the utility would prepare during the AP1000 plant operational phase.
Based on the GDA Disposability Assessment for the AP1000 plant, RWMD has concluded
that, compared with legacy wastes and existing spent fuel, no new issues that challenge the
fundamental disposability of the wastes and spent fuel are expected to arise from operation of
such a reactor [Reference 24].
10.3
Decommissioning Lifecycle
The decommissioning programme shown in subsection 10.1.3 identifies the key
decommissioning activities associated with the AP1000 plant. Based on this programme the
AP1000 plant operational lifecycle is 60 years, with final decommissioning of the reactor and
all associated facilities completed 22 years later, and final decommissioning of the interim
HLW and ILW radwaste stores being completed 103 years later.
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The decommissioning timeline highly depends on the requirement to allow the HLW spent
fuel to cool for 100 years before it can be accepted in the HLW repository (see subsection
10.1.2). If this cooling period can be reduced, then the decommissioning programme can be
similarly reduced.
The lifecycle of the interim waste stores is also strongly dependent on final waste repository
availability. The HLW and ILW stores have a 100-year design life. However, HLW and ILW
are envisioned to be transferred to waste repositories at the earliest opportunity.
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REFERENCES
1. The Decommissioning of the UK Nuclear Industry’s Facilities. Policy statement by the
UK Government and Devolved Administrations. DTI, 2004.
2. Nuclear Installations Act 1965, 1965 CHAPTER 57.
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January 2008.
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6. The Construction (Design and Management) Regulations 2007, SI 320, 2007.
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Regulations 2007 Approved Code of Practice, HSE 2007.
8. GDA Technical Query TQ AP1000-330 “Expectations of Operating Utility Management
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9. Westinghouse Report EPS-GW-GL-700, Rev. 1, “AP1000 European Design Control
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11. UKP-GW-GL-054, Revision 1, “UK AP1000 Integrated Waste Strategy,” Westinghouse
Electric Company LLC, 2011.
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and Disposal,” Westinghouse Electric Company LLC, 2011.
13. UKP-GW-GL-055, Revision 2, “UK AP1000 Radioactive Waste Management Case
Evidence Report for Intermediate Level Waste,” Westinghouse Electric Company LLC,
2011.
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Company LLC, 2011.
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No. 62, IAEA, 2009.
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20. Implementation and Review of a Nuclear Power Plant Ageing Management Programme,
IAEA Safety Report Series No. 15, IAEA, 1999.
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Report Series 420, IAEA, April 2004.
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23. UK Radioactive Higher Activity Waste Storage Review, Issue 1, NDA, March 2009.
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for Wastes and Spent Fuel Arising from Operation of the Westinghouse AP1000, NDA
Technical Note No. 11339711, NDA, October 2009.
25. Future LLW Operations 2007 – 2048, LLWR Ltd.
26. LLW Repository Strategic Review, NLWS/LLWR/01 – Issue, NDA, January 2009.
27. Decommissioning of Nuclear Fuel Cycle Facilities, IAEA Safety Guide No. WS-G-2.4,
IAEA, 2001.
28. UKP-GW-GL-793, Revision 0, AP1000 Pre-Construction Safety Report, Westinghouse
Electric Company, LLC, 2011.
29. Decommissioning Safety Reference Levels Report, Version 1, WENRA, March 2007.
30. Applying Decommissioning Experience to the Design and Operation of New Nuclear
Power Plants, Nuclear Energy Agency, OECD, 2010.
31. “Decommission of AP1000,” Presentation to EA/ND, Westinghouse Electric Company
LLC, Pittsburgh, 25 February 2010.
32. Decommissioning of Nuclear Power Plants and Research Reactors, IAEA Safety Guide
No. WS-G-2.1, IAEA, October 1999.
33. Technical Assessment Guide T/AST/026 “Decommissioning on Nuclear Licensed Sites,”
Rev. 2, HSE, March 2001.
34. Heavy Component Replacement in Nuclear Power Plants: Experience and Guidelines,
IAEA Nuclear Energy Series No. NP-T-3.2, IAEA, 2008.
35. Generic Repository Studies, Generic Waste Package Specification, Vol. 1, Nirex Report
N/104 2005.
36. “A White Paper on Nuclear Power,” Department for Business, Enterprise & Regulatory
Reform, January 2008.
37. Decommissioning of Nuclear Facilities – It can be done, NEA#06829, NEA, 2010.
38. Decontamination and Decommissioning of Radiologically Contaminated Facilities,
Interstate Technology & Regulatory Council, January 2008.
39. UK Radioactive Waste Inventory, NDA.
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40. State of the Art Technology for Decontamination and Dismantling of Nuclear Facilities,
Technical Report Series 395, IAEA, 1999.
41. Review of Clean Up and Techniques applicable to discharges from Decommissioning,
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