Varying definitions and boundaries

The most important factor influencing the lifetime cost of a nuclear reactor is capital or construction cost. Investment in electricity generation is a multi-year affair and can take a decade or more from an early planning stage, conducting environmental impact assessments, obtaining construc­tion permits, actual construction and plant commissioning before the plant produces the first kWh. Necessarily the actual financial outlays are spread over this period.

Despite its crucial role in determining the economics of nuclear power, investment costs of nuclear power plants remain a mysterious affair. The fact that the investment in a nuclear power project encompasses numerous, often site — or project-specific components ranging from site acquisition and preparation to bid evaluation, construction, licensing and grid integration, mandates unambiguously defined project boundaries, i. e., what is included in an investment cost quotation and what is not. It also requires clarity about the cost of finance during construction, currency exchange rates used, inflation over the construction periods, taxes or subsidies. Otherwise cost comparisons are meaningless.

At the most aggregate level, total investment costs equal ‘overnight costs’ (OC) plus interest during construction (IDC). The term OC is often used to express what the investment of a project would cost if it were built ‘over­night’, i. e., as if money had no time value.

IDC are the financing costs for plant construction until the plant is con­nected to the grid and generates revenues. Because it can take as much as 10 years or more to bring a nuclear power plant from planning to comple­tion, IDC alone can tilt the balance between an economically viable or unviable project. Their long construction periods and high up-front invest­ment requirements make nuclear power projects very sensitive to IDC, and thus to construction delays.

The EIA Directive requires the production of an Environmental Statement as the primary output of the EIA process, the minimum contents of which are prescribed by the EIA Directive and are closely aligned to that of the Environmental Report in the SEA process. Annex III of the EIA Directive dictates the minimum information that is to be provided as part of the Environmental Statement including, among others, a description of the physical characteristics of the project including its land-use requirements, an estimate (by type and quantity) of expected residues and emissions associated with the activity and a description of the likely significant effects of the proposed activity on the environment. A non-technical summary of the information is also to be included in the Environmental Statement so as to ensure that the implications of the scientific informa­tion are readily accessible by the general public. As with the SEA process, a crucial requirement of the EIA Directive is that the Environmental Statement identifies any measures envisaged to ‘prevent, reduce and where possible offset any significant adverse effects on the environment’ (EIA Directive, Annex 3, paragraph 5). The Environmental Statement is to be made available to the relevant members of the public, along with the application for development consent, and the public is to be given the opportunity to express its opinion on the project before any decision to initiate the project is taken.

However burdensome the EIA process may appear, one fundamental factor (and some may argue flaw) in the process is that an Environmental Statement which suggests significant harm to the environment does not actually prevent an authority from granting its consent for the activity in question. While the EIA Directive expressly requires the decision-maker to take the findings of the Environmental Statement into consideration, there is no overt obligation on the decision-maker to withhold consent for development where the negative environmental effects appear dispropor­tionately greater than the benefits that the activity would bring. Equally, neither is there an obligation on authorities to afford particular weight to the views of the public — although the public has the right to be consulted during the process, the practical value of that right is merely procedural.

Nonetheless, the authority must inform the public of its ultimate decision as well as the reasons and considerations upon which it is based.

The requirements consider three types of external events: aircraft crashes, chemical explosions, and other important human-induced events. The requirements consider only accidental events of that type; terrorist attacks by the same means with the only purpose of producing harm are excluded.

The currently high level of air traffic and the expected increase in the future may represent a substantial hazard to NPPs, despite the high level of safety in current and future air flights. The hazards due to the crash of a large passenger airplane include the impact itself, the explosions and large fires that may be produced. The risk can be reduced by avoiding having flight corridors close to NPPs.

The proximity of chemical installations and transportation routes also creates risks of explosions, deflagrations and large fires with the correspond­ing release of toxic substances that may affect the safety of the operating NPP; compensatory measures should be applied or the site discarded. Other important human-induced events include large fires, for instance forest fires, collisions of ships with water intake structures, and the presence of electro­magnetic waves with the potential of affecting the plant information, control and instrumentation systems.

An IAEA safety guide describes the many external human-induced events to be considered in the evaluation of a site for a NPP (IAEA, 2002a). The guide considers each one of the events, defines the associated hazards and determines how to obtain the main parameters to be used in the design basis of the plant to cope with such hazards.

Requirements regarding the supplier’s engineering organisation, engineer­ing and design process, design interfaces management, use of computer- assisted design (CAD) software and tools, project design manual, review and approval of design documentation by the owner and other aspects related to the management of the project engineering process will be speci­fied in this section.

19.9.2 Procurement and supply chain management

This section should specify the requirements for the procurement process, formation of the supply chain for the project, use of subcontractors (i. e. from a list of owner-approved subcontractors), inspection plans, approval of manufacturing procedures, and project procurement procedures manual.

19.9.3 Project risk evaluation

This section should specify the requirements concerning the economic and financial aspects, as well as project scheduling and development, and other relevant information required by the vendor for the preparation and peri­odic updating of a project risk evaluation report, enabling the owner to assess the risk status of the project throughout its duration.

The operator of a nuclear plant is responsible for its safety. An important operating discipline is a robust performance improvement programme. The programme should have several elements in an overall interactive model. Elements of the model could include self-assessments, operating experience feedback, conduct of operations, performance assessments, oversight stan­dards, engineering programmes, and processes for dealing with any gaps that are identified. These elements would then fit into an overall perfor­mance model that has the following steps: (1) obtain the results for a per­formance monitoring/assessment element; (2) identify the gaps; (3) analyze and identify solutions; (4) implement the solutions; and (5) continue moni­toring. The objectives are to identify and correct problems, to identify and correct any negative trends before they become an issue, and to raise the sensitivity of management and staff to the importance of constant diligence and questioning attitudes. While a comprehensive discussion of all the ele­ments of a performance improvement model is beyond the scope of this chapter, two of the elements can be mentioned for illustration: self-assess­ment and operating experience feedback.

Self-assessment is a general process that can encompass both plant com­mercial performance and safety. Since performance and safety are bound together, self-assessments are a particularly important part of the overall safety structure. The IAEA Safety Requirements (IAEA, 2006d) for a nuclear facility management system mandate that: ‘Senior management and management at all other levels in the organization shall carry out self­assessment to evaluate the performance of work and the improvement of the safety culture’. Individuals and workgroups must assess their perform­ance against the licensee’s safety goals including the operating license requirements and other nuclear industry safety standards. One of the values of self-assessments is that they also recognize strengths and good practices that exceed the current requirements, and these might be used to enhance performance in other areas.

A policy should be developed that lays out the objectives and procedures for performing the self-assessments. Such a policy could include the scope for the assessments, the frequency, the process roadmap, the reporting and review mechanisms, quality assurance requirements, and what benchmarks should be employed. In general, the process roadmap would involve the preparation of annual plans indicating the areas that will be assessed and the schedules, the formation of self-assessment teams to carry out the reviews, conducting and documenting the assessments, analyzing the results, taking corrective actions, and communicating the status. This process would then be followed up by an evaluation of the effectiveness and quality of the review, as well as of the lessons learned for improvement.

Two types of gaps can be identified in this way, the first being where current safety requirements, regulatory or otherwise, are not being met and corrective actions must be taken immediately, and the second being where requirements are being met but there is opportunity for improvement. In this case, although the requirements are still being met, the performance may be trending away from acceptable standards. Ideally, it is the second type of gap that would eventually come to dominate the self-assessment process. This would demonstrate that the licensee was proactively determin­ing the precursors to any potential diminishing of safety and addressing them immediately. Therefore, self-assessments are a key activity for pre­venting operational complacency.

Analysis and feedback of operating experience is recognized as a valid tool to enhance safety in the IAEA Fundamental Safety Principles. By any measure, a nuclear power plant is a complex technology. The plant contains about 100 major systems that fall into four groups: nuclear systems, fuel and refuelling systems, secondary plant systems, and electrical systems. Due to ageing, configuration changes, and equipment upgrades, each of these systems requires verification on a continuous basis to ensure that the systems continue to meet safety and operational requirements. Operating experience, as part of the overall plant performance model, is a valuable tool for helping to ensure this, since it enables the licensee to apply previous information to anticipate and address issues before they occur. INSAG-21 has pointed out the importance of operating experience feedback for life cycle management and backfitting of nuclear facilities, as well as for improv­ing operating and regulatory practices, to enhance the global nuclear safety regime (INSAG, 2006).

Operating experience information covers all aspects of the NPP’s opera­tion and has implications for both plant performance and safety. With respect to plant performance, particular attention is paid to outages, planned and otherwise. Outages can be classified as planned (under operator control), unplanned (causes under operator control), and external (not under operator control). Planned outages include refuelling, inspection, maintenance, testing, and upgrades. Unplanned outages include those due to human error, equipment failure, operating margins, and regulatory/ licensing issues. Externally driven outages include grid failure following electricity demand, and environmental conditions.

The IAEA has established guidelines to enhance operating experience feedback (IAEA, 2006f). According to this Safety Guide, an effective system for the feedback of operational experience relating to safety should have the following elements:

• Reporting of events at plants

• Screening of events — primarily on the basis of safety significance

• Investigation of events

• In-depth analysis, including causal analysis, of safety-significant events

• Recommended actions resulting from the assessment, including approval, implementation, tracking and evaluation

• Wider consideration of trends

• Dissemination and exchange of information, including by the use of international systems

• Continuous monitoring and improvement of programmes for the feed­back of safety-related operational experience

• A storage, retrieval and documentation system for information on events.

The licensee should develop a comprehensive operating experience pro­gramme with input from a variety of internal and external sources. One international tool for operating experience is the Incident Reporting System (IRS) jointly developed by the IAEA and OECD/NEA (IAEA, 2008). The IRS reports contain information on NPP events that are of significance to safety and the safety lessons that can be learned to assist in reducing recur­rence of events at other plants.

More information on specific topics is also available. This includes an Information System on Occupational Exposure, which was started by the OECD/NEA and is now jointly maintained with the IAEA. Other projects at the OECD/NEA addressing specialized areas include the International Common-Cause Failure Data Exchange, the Fire Incident Records

Exchange, the Piping Failure Data Exchange, the Exchange of Operating Experience Concerning Computer Based Safety, and Stress Corrosion Cracking and Cable Ageing.

Another tool is the World Association of Nuclear Operators (WANO) series of documents on Operating Experience Report and Significant Operating Experience Report processes. However, WANO information is generally restricted to its members. There are also specific reactor tech­nology groups, such as the CANDU, Westinghouse, General Electric and KWU Owners’ Groups, which deal with design-specific operational feed­back, although some information may be restricted to group members. There are also national and regional institutions which interchange operat­ing experience.

In response to restrictions on some operating information that could impact safety, INSAG has pointed out in both INSAG-21 (INSAG, 2006) and INSAG-23 (INSAG, 2008b) that there is considerable room for improve­ment in the transparent sharing of safety information, both nationally and internationally. INSAG-23 also notes that:

It is widely observed in all fields of human activity that serious accidents are nearly always preceded by less serious precursor events. If lessons can be learned from the precursors and these lessons put into practice, the probability of a serious accident occurring can be significantly reduced. . . While the con­tinued strong safety performance by operators is encouraging, safety significant events continue to recur in nuclear installations. This indicates that operators are not learning and applying the lessons that experience can teach us.

As a result of their assessment, INSAG has proposed several recommen­dations to improve international operational feedback. However, imple­mentation of the feedback still rests with the licensee. It is important to ensure that operating experience is being used effectively throughout the licensee’s organization at all levels, and for both safety and operational performance. This includes the various processes for information gathering and analysis, experience application, auditing, and training.

According to the remarks made by the Honourable Peter B. Lyons, Commissioner from the US Nuclear Regulatory Commission at the 2008 International Congress on Advances in Nuclear Power (ICAPP’08), ‘Creating, sustaining, and growing a population of educated, trained, and experienced personnel from which the nuclear industry need to recruit in order to accomplish their goals is a challenge among government, industry, and academia’.

It is widely recognized that national development in the nuclear industry requires a scientific and technological infrastructure. Such an infrastructure is mainly found in:

• National and private research and development (R&D) institutes

• Institutes and laboratories for standardization and calibration

• Higher education institutions

• Vocational schools for practitioners and professional training centres

• Scientific academies and professional associations

• National industry.

All these organizations create knowledge in one way or another but three of them are particularly important: R&D institutes, higher educational insti­tutions and vocational schools.

Finally, it is necessary to design the facilities and equipment at the training centre. The description of the training resources will include buildings, class­rooms, laboratories, simulators, mock-ups and other training delivery set­tings and equipment.

Among the training tools the simulator deserves special attention. According to IAEA (2009b) a key lesson learned regarding commissioning of a nuclear facility is the importance of having a plant-referenced, full — scope control room simulator available well in advance of nuclear facility operation. This simulator not only provides a unique tool for training nuclear facility control room personnel, but also is important for tasks such as normal, abnormal and emergency operating procedure development and validation, development and validation of commissioning tests, validation of digital control systems, and training of other plant personnel.

For many new nuclear facility projects, a full-scope simulator is provided as part of supplying the nuclear facility package. Integrating the simulator development and training schedule with the overall commissioning sched­ule is very important. According to the US Nuclear Regulatory Commission, the simulator should be ready for training three years before the fuel load.

Important considerations regarding simulator information are a simula­tor configuration control process description; a plan for acquiring, validat­ing and using a plant reference simulator (or if a plant reference simulator is not yet available, a description of how and when a part-scope or non­plant-referenced simulator will be used during the training and how and

when that simulator will become plant-reference); a list of unresolved simu­lator deficiencies and recent simulator fidelity data; and simulator perfor­mance indicators or process description.

The operating organization as well as the regulatory body would require extensive technical support in a number of areas for efficient operation and effective safety regulation of the NPP. Such support would be needed to tackle problems that may arise during operation as also to obtain a proper understanding and assessment of the ageing-related degradation of systems, structures and components and to find appropriate solutions for their longer-term management. Also further analysis and experimental work may become necessary in the light of new information from research or operat­ing experience. In addition, over a period of time the safety standards may get revised, leading to the need for implementation of safety upgrades that might need substantial engineering development. To cater to these needs requisite laboratories and engineering development facilities should be established and expertise generated for their effective functioning. Some examples of the facilities required are metallurgical laboratories for carry­ing out failure analysis of radioactive as well as non-radioactive compo­nents, assessment of the extent of irradiation-induced embrittlement in materials and post-irradiation examination of reactor fuel. Examples of facilities for engineering development are those required for testing of tools and procedures for complex repair and inspection jobs, environmental qualification and endurance testing of components and development of remotely operated tools.

Capabilities are also needed to carry out various studies and analyses such as on atmospheric dispersion of radioactivity under different weather conditions, analysis of ageing structures to check on their continued capac­ity to withstand design loads, and periodic updating of the probabilistic safety analysis for a quantitative assessment of the current safety status of the NPP. There are other areas also in reactor physics, reactor chemistry, control and instrumentation and computer-based systems where the techni­cal support organizations will have to play a strong role in support of the NPP operation and regulatory effort.

Experience shows that total dependence on the reactor vendor for tech­nical support over an extended time is neither feasible nor desirable. It is therefore necessary to establish technical support organizations in the country well before the start of NPP construction and to staff them with personnel who have received advanced training in specific fields. Facilities for analysis and engineering development should be set up and progres­sively augmented for effective functioning of the technical support organi­zations. Expertise available in the various academic and professional institutions in the country should also be utilized such as by awarding research projects to these institutions for specific development jobs and by inducting their experts in advisory committees and in development of national safety standards.

The Enhanced CANDU 6 (EC6) is a 740 MWe pressure tube reactor designed by Atomic Energy of Canada Limited (AECL). The EC6 design benefits from the proven principles and characteristics of the CANDU 6 design, which is currently in operation in several countries in the world, such as natural uranium fuel, two independent safety shutdown systems, a separate low-temperature, low-pressure moderator (which provides an inherently passive heat sink by permitting heat to be removed from the reactor core under abnormal conditions), a reactor vault that is filled with cool light water (which surrounds the reactor core, providing another passive heat sink), on-power refueling, and a modular, horizontal fuel channel core. The EC6 design includes a more robust containment with thicker walls and a steel liner, enhanced severe accident management, addi­tion of the emergency heat removal system as a safety system, improved shutdown performance for larger loss of coolant accident margins, and a plant life of 60 years with one life extension of critical equipment such as fuel channels and feeders at mid-life. The Canadian Nuclear Safety Commission (CNSC) is currently conducting the design review of the EC6.

The concept of defence in depth was first utilized in military practice. Scattered defensive outposts are, of course, vulnerable to defeat through local attack. To overcome this weakness, the defence in depth idea estab­lished methodologies through which the outposts were linked together via communication channels and response doctrine that specified assistance from one outpost to others nearby, along with an established deep configu­ration of outposts that together formed a strong network of defence.

Defence in depth

This useful concept has been adopted by the world nuclear industry. A good summary of the application of this concept can be found in the IAEA report titled Defence in Depth in Nuclear Safety (INSAG-10, 1996). This document is based on the original description of this concept published in an earlier
document (INSAG-3, 1988). Similar descriptions have appeared in design and safety-related documents published over the past few decades. Figure 10.3 illustrates the overall concepts of defence in depth. In this view, the processes are separated into two parts — prevention and mitigation, respec­tively. Some reactor designs may have different specific elements in some of these positions; however, the principle remains the same — there are multiple levels of defence against transfer of radioactive materials from their normal positions in the reactor to the public or environment.