The selection of any new site for a large industrial installation requires a systematic approach, as described in Fig. 18.1. First of all, a whole country or a region of it is selected based on economic considerations, proximity to an electricity market and social demands, such as the convenience of boos — tering the development of a particular region.

Within the region, selection of one or more zones will have to be deter­mined by a general analysis of some basic parameters. At this first stage, the availability of cooling water is the most restrictive technological require­ment; generally the areas of interest are limited to rivers, lakes or coastal sites. Artificial lakes could also be built on smaller tributaries and the use of cooling towers may open more possibilities, although the proximity to a large body of water is always recommended.

An analysis of the geology of the region originally selected will determine which areas must be discarded because of high seismicity or for other reasons; the meteorology of the region will determine the hazards associ­ated with extreme meteorological events; zones that are densely populated or near large population centres (over 25,000 residents) will also be dis­carded as it would be difficult to establish efficient emergency procedures; and sites close to large industrialized areas or areas with high agricultural

or ecological value should also be dismissed. The systematic approach of all these criteria will divide the region in question into zones or areas in which NPPs may be situated. At this time, a gradation of the areas found could also be established. Countries have evaluated the maximum nuclear capac­ity which could be installed along a given river, large lake or coastal region.

The zone or zones of interest are then studied in depth to determine the optimum sites where the plant or plants could be built. Two types of studies are conducted; on the one hand, detailed economic, geological, hydrological, meteorological and social and demographic studies will determine some basic plant design parameters, while on the other hand the plant or plants to be constructed will determine some basic requirements from the site, mainly related to the ultimate heat sink, redundant electrical power supply, the need to have an efficient emergency management system, the release of radionuclides during normal operation and accident conditions, the man­agement of radioactive waste, and the size and geometries of the buildings to be erected.

The initially separated studies described above are later subjected to a compatibility study between the plant and the site, covering all types of parameters which constitute the basis for the selection. The site information gathered and the compatibility of the site and the selected technology or technologies will constitute the basis for requesting the site approval from the Regulatory Body in accordance with the regulations of the country.

In the past, governmental institutions and large utilities, under their areas of influence, have conducted studies to determine the best locations for building nuclear power plants and fuel cycle installations. For new entrants the development of such a bank of potential sites is highly recommended. The current social opposition to nuclear power makes it difficult to find new sites for nuclear power plants and related facilities.

A section of the SS document should be dedicated to specifying the scope of supply for nuclear fuel and associated services. Alternatively, this portion of the scope could be included in the nuclear fuel (NF) document of the BIS, which would be a comprehensive, self-supporting document dedicated entirely to the scope of supply and services, technical requirements and commercial conditions for nuclear fuel.

Standard practice is to request the following from the complete plant supplier (turnkey approach), from the nuclear island supplier (split-pack­age approach) or from the NSSS supplier (multi-package approach): [105]

draulic design documents; provision to the owner of all necessary fuel data for him to achieve fuel procurement from third parties, if he should so decide in the future; and supply of all quality assurance and quality control manuals, procedures and records related to the nuclear fuel supply.

• Investment for reload batches. As an option, the bidder is usually requested to submit a proposal for the provision of a limited number of reload batches (usually two or three, sometimes more), sufficient for the owner’s fuel specialists to familiarise themselves with the nuclear fuel and reactor core design and gain sufficient knowledge to decide whether to continue with the original fuel supplier or purchase it on the market from third parties. An alternative to requesting a specific number of reload batches consists in requesting the supply of reload batches neces­sary for a given period of operation (e. g. 4 or 5 years), after which familiarisation is expected to be achieved.

• Together with each subsequent reload, the supplier is normally requested to provide the associated fuel management services (e. g. core design, safety analysis, reload licensing).

As already described, an application for a site license, whether part of a more general license or a stand-alone application, needs to identify the precise site on which the applicant proposes to build a nuclear power station, and the characteristics of the site need to be described, as do the mutual interactions between the plant and the site. The submitted docu­mentation should be analyzed by the RB against established safety prin­ciples, such as: [110]

of implementing emergency countermeasures (including possible evac­uation of areas around the site). As already mentioned, the potential difficulties of a nuclear emergency within a more general emergency should also be considered by the RB.

• The social, economic and environmental effects of the NPP. An NPP will have effects on the surrounding population and the environment. The RB should consider the population density and the proximity to large and medium cities, technological parks, recreational areas, national parks and heritage locations which may become heavily affected by radiation releases of a certain magnitude.

The analysis by the RB experts requires knowledge and experience in earth sciences to determine the magnitude of the maximum possible natural events, as well as experts on man-made events, and people with experience of emergency planning. The best help to the RB will probably come from national institutions dealing with such phenomena and activities. A site license will typically contain requirements and limitations on the site’s preparation activities that may be conducted before construction begins.

Many RBs require that a local environmental impact statement (EIS), assessing the impact caused by the future power plant, is also submitted by the applicant and analyzed by the RB at the time of site evaluation. It is also customary at this stage to inform stakeholders of the project and to allow them to formally make representations detailing any reservations they may have about the project. Some countries, for example the UK, require that a Public Inquiry should be called at which the applicant is invited to present their case and to hear and consider the contributions and concerns raised by participating stakeholders. The involvement of stake­holders in nuclear issues before deciding the construction of a new nuclear power plant is recommended in INSAG-22 (INSAG, 2008a).

At a very early stage (around 12 years before the first fuel loading) it is critical to plan the human resources requirements with a long-term vision. The following items must be considered within the human resources planning:

1. Job profiles and selection criteria. Once the decision about the project organization and the future plant organization is made, the next step will be to identify the job positions needed for the tasks to be accom­plished during the different stages of the project, including plant opera­tion and maintenance. With the detailed matrix (job position — number of candidates — year of hiring) already made, it will be useful to group the future job positions into families, such as managers, engineers, instructors or regulators, then afterwards select from them the best candidates for the different jobs.

The number of new employees to be contracted in a specific year will depend on the capability to train them and the predicted allocation of manpower for the different tasks within the project.

2. Training programmes. It will be necessary to design different training programmes as the project progresses. During the first years it will be enough to provide some generic nuclear training with different content for each group to become familiar with the specific areas in the nuclear field. Table 6.1 includes typical training modules to be implemented for the different stakeholders during the first steps of a nuclear programme.

Depending on the activity involved, more specific training will be delivered, complementing the generic nuclear part, in areas such as project management, quality assurance and quality control, licensing, norms and regulations, among others.

E&P C&C EM RB O&M

NPP Fundamentals Applied Thermodynamics •

and Fluid Mechanics Mechanical & Electrical •

Components

Control & Instrumentation •

Strength of Materials •

Nuclear Physics •

Thermohydraulics •

Nuclear Reactor Chemistry •

Design and NPP Design Criteria •

Engineering Safety Assessment •

Technical Specifications •

Design Engineering •

Nuclear Safety and Safety •

Culture

Nuclear Technology Nuclear Steam Supply System

Reactor Auxiliaries •

Systems

Plant Services •

Safeguard Systems •

Water-Steam Cycle •

Reactor Control, Limitation •

and Protection System Effluent treatment systems •

E&P: Engineering and Procurement. C&C: Construction and Commissioning. EM: Electrical, Mechanical and Instrumentation Equipment Manufacturers. RB: Regulatory Body. O&M: Plant Operation and Maintenance.

Some in-plant training, either tutoring or ‘shadowing’, is highly rec­ommended, in order to become familiar with the allocation of systems and main equipment, the functioning of the operation departments and the internal norms and procedures. This experience can be gained in reference plants that the vendor has previously built or, if appropriate, in other plants belonging to the owner/operating organization.

3. Task assignment within the project and professional development. As soon as the training is finished the professionals and technicians will be assigned to a specific task within the project. The project manager should perform technical competency-based assessments, leadership development, and succession planning for future high-responsibility assignments.

According to the IAEA (2008a), the influencing factors that can reduce

the human resource requirements are as follows.

• Those NPP operating organizations considering adding new nuclear units have to assess the extent to which the current workforce can be effectively utilized for the commissioning and operation of the additional units and in this way provide an opportunity to evaluate the possibility of sharing common services for the whole fleet (e. g. the Quality Assurance department or even Maintenance). It is possible to achieve as much as a 30% reduction in manpower requirements for the next reactor when maintaining an efficient organizational structure.

• Where an owner/operator owns or operates units at more than one location, a different organizational structure may be used to improve efficiency. Many functions can be centralized in the parent organization. It is common to find fleet nuclear companies that have an average of 20% fewer personnel due to the economies of scale.

• Some new advanced reactors have a more simplified design and fewer systems and components, therefore the staffing reductions for a passive light water reactor plant compared to a current nuclear plant could be about 40%.

• Finally, another factor that can affect the number of resources needed is the possibility of contracting specialized services externally. Although operating organizations tend to conduct maintenance activities them­selves, rather than contracting with a vendor, there are some exceptions for outage-related work, where most operating organizations continue to rely on external support, particularly for specialized maintenance and inspections of major equipment. Engineering and technical support are other services susceptible to be contracted out. In those cases the licensee retains the primary responsibility for the safety of such operations.

152 Infrastructure and methodologies for justification of NPPs

FP7 is the short name for the Seventh Framework Programme for Research and Technological Development. This is the EU’s main instrument for funding research in Europe and it will run from 2007 to 2013. FP7 is also designed to respond to Europe’s employment needs, competitiveness and quality of life.

The framework programme for nuclear research and training activities will comprise Community research, technological development, interna­tional cooperation, dissemination of technical information and exploitation activities as well as training.

Sustainable Nuclear Energy Technological Platform, SNETP (2007)

The SNETP was officially launched in 2007. Today, SNETP gathers about 70 European stakeholders from industry, research and academia, technical safety organizations, non-governmental organizations and national repre­sentatives. SNETP aims to support fully through R&D programmes the role of nuclear energy in Europe’s energy mix, and its contributions to the security and competitiveness of energy supply, as well as to the reduction of greenhouse gas emissions. To achieve this objective, SNETP has elabo­rated a Strategic Research Agenda (SRA) that identifies and prioritizes the research topics.

SNETP has set up a specific Working Group dedicated to Education, Training and Knowledge Management (ETKM) issues, with the support provided by the European Nuclear Education Network (ENEN) Association. This workforce will in part also provide qualified staff to Europe’s nuclear industrial sector to accompany the development of the sector in the next decades.

Reactor core management starts with working out the scheme for initial fuel loading in the core with control rods in appropriate positions and the required concentration of neutron poison in the reactor moderator to ensure that the prescribed level of sub-criticality is maintained all the time. The next step is to compute the core configuration for its first criticality and to prepare the procedure for achieving first criticality. Special instrumenta­tion is installed in the core for measuring the very low neutron flux in the core during the approach to first criticality. It may also be necessary to install a neutron source to ensure that the reactor startup instrumentation is comfortably on-scale before starting to reduce the boron concentration in the moderator water or withdrawing control rods for making the reactor critical.

After the reactor is made critical, the predicted and the actual core con­figuration for criticality are compared and any significant differences or anomalies are resolved. Thereafter various tests are conducted with the reactor at low power, such as measurement of reactivity worth of control rods and various coefficients of reactivity. Other checks such as measuring neutron flux at different axial and radial locations in the core, establishing the relation between reactor thermal power and neutron power, effective­ness of radiation shielding and response of the NPP control system during situations like partial or total load throw-off are carried out at different power levels during the reactor power ascension stages.

Thermal hydraulics computations are done to assess the power delivered from fuel to coolant and various important thermal parameters such as the fuel rod linear power, the fuel centre temperature, the fuel clad surface temperature and the temperature gradient across the fuel thickness, both for steady-state condition and under transients such as tripping of main coolant pumps and movement of control rods. Computations using coupled neutronic and thermal hydraulic codes facilitate obtaining a better and more holistic assessment of the reactor core.

After fuel burn-up proceeds to a level at which there will not be sufficient reactivity available to operate the reactor, the core will have to be refuelled. Refuelling is generally done by removing high burn-up fuel from the central zone of the core, moving the low burn-up fuel from the outer zone to the central zone and loading fresh fuel in the outer zone. As refuelling is done in the reactor shutdown state with the reactor vessel head open, care needs to be taken to prevent open vessel criticality of the core, especially when control rods are removed for maintenance work.

For performing the above activities a strong reactor physics group has to be developed at the NPP with proficiency in use of neutronic and thermal hydraulic computational codes and a good understanding of the reactor core behaviour. Based on the results of various measurements and operat­ing experience, the computational codes will have to be fine-tuned or upgraded. This group will advise the plant management on the refuelling of the core and will work out the detailed refuelling scheme. For reactor designs with on-power refuelling, this task has to be performed on a day — to-day basis. This group will also analyse reactivity anomalies and other reactivity-related events as and when they occur and advise the plant man­agement on the corrective actions. As the neutronic and thermal hydraulic behaviour of the core is one of the most important areas of reactor safety, the staff of the regulatory body also need to have proper understanding and appreciation of effective safety regulation. Many regulatory bodies have standing expert groups to advise them in these areas. Such advisory groups may comprise experts from within the regulatory body, the technical support organization and academic institutions and even personnel from the utility headquarters who are not directly involved in managing the reactor core. While assistance from the reactor vendor may be obtained during the initial few years for managing the core, it is absolutely essential that a high level of expertise in this field be progressively developed in the operating organization, the technical support organization and the regula­tory body for managing the reactor in the long term as also for future expansion of the nuclear power programme.

As with any other large industrial installation, a nuclear power plant, fuel cycle facilities and associated activities produce substantial non-radiological physical and chemical impacts on their surroundings during pre-construc­tion, construction, operation and dismantling. These impacts are generally considered in an environmental report, a licensing requirement in most countries. The USA NRC requirements are among the most developed regulations, and they also include economic and radiological impacts (NRC, 1984).

Nuclear regulatory authorities are not the only regulators intervening in this process; other local, regional and state authorities also participate in the review, generally in a coordinated manner, to verify compliance with other regulations on environmental protection, such as those related to air and water quality and on land use.

For non-radiological impacts, any environmental impact study starts by describing the affected territory and its current use, industrial and recrea­tional development; the surface and ground water hydrology and the use given to such waters; the meteorology and air quality, and the terrestrial and aquatic ecology, amongst the major aspects. With this knowledge, the impacts during pre-construction and construction activities are analysed, generally divided into three levels of significance: small, moderate, or large. For all these impacts, mitigation measures are also considered, though, in such an analysis, unavoidable adverse environmental impacts can be found for which no practical means of mitigation are available.

During pre-construction and construction activities, the major unavoid­able environmental impact would be the land to be occupied by the plant buildings and the land used temporarily for construction purposes; addi­tional land will also be needed to build new or widen existing roads and electrical energy corridors. The high energy intensity of nuclear power does not need additional land for storing new and used fuel, as is the case for fossil fuels (mainly coal). Considerations should also be given, among other things, to the use of water and construction materials; the effects that exca­vation and dewatering will produce on groundwater aquifers; the ecological impact on terrestrial and aquatic losses; and the increase of traffic and health effects due to fugitive dust, noise and transportation. The purpose of these considerations is to evaluate them and to define mitigation and con­trols aimed at lessening the adverse impacts.

The land used for buildings will be considerably improved by trees, gardens and lawns. When the plant ends its useful lifetime, decommissioning will restore the site and make it useful for other purposes (decommissioning is considered in detail in Chapter 24 of this book). The operation of nuclear power plants is very clean; negligible amounts of conventional pollutants are released and solid conventional waste is very limited. Non-radiological health impacts to members of the public, including etiological agents, noise, electromagnetic fields, occupational health, and transportation of materials and personnel are minimal and well controlled to verify compliance with applicable regulations.

During operation, the major non-radiological impacts are the use of water to cool the condenser and the effects that heat rejection produces in the affected water bodies. Nuclear power plants have a low thermal effi­ciency of about one-third; therefore two-thirds of the heat produced in the reactor core by fission is waste heat that has to be rejected to water bodies and eventually to the atmosphere. Some 5% of this heat is released within the plant itself; therefore some 62% of the generated heat has to be removed by the condenser water. Large quantities of water have to pass through the condenser to remove such heat, and the flow of water depends on the limit put into the outlet temperature; for a 10°C increase, in 1 GWe plant, some 45 m3/s is needed.

To achieve heat rejection, two types of systems have been developed. Once-through systems take water from a large water body and discharge it to the same water body at a higher temperature and at a different point. Only plants built in coastal locations and in the proximity of large rivers or lakes can use such systems. In closed circuit systems, the warmed water is cooled in a cooling tower or spray pond and recirculated through the con­denser. In this way, the waste heat is finally deposited into the air. Combinations of once-through and closed circuits are frequently found, the once-through systems being used when temperature limitations in the receiving water body can be complied with (mainly during the winter), and closed circuits otherwise.

Once-through systems may cause damage to living organisms in the water body as a result of changes in temperature, the impingement of larger organisms in the water-intake screens and the entrainment of smaller organisms that pass through the condenser. Deleterious effects may also be produced by the use of chlorine and biocides, by changes in the water quality, mainly oxygen content and increases in salinity. All these effects have the potential of introducing changes into the aquatic ecosystem; they should be known to verify compliance with environmental regulations.

A large variety of chemicals are added to wet cooling towers to control bacteria and prevent corrosion. Such chemicals might eventually be dis­charged to an adjacent water body when recirculation water retaining impu­rities is taken out of the tower; this is called a blowdown. Blowdowns have to be controlled and should not be discharged to public waters without treat­ment and control in accordance with regulations. Such chemicals may also escape to the atmosphere with small water droplets in the drift, i. e. steam released from the cooling towers; these chemicals will be deposited and will accumulate on surfaces near the plant. The effects of this have to be control­led; in modern cooling towers, drift is limited and the effects minimal.

Although non-radiological effects are well recognized, their study and quantification constitute the basis of their mitigation, until compliance with existing regulations. In general, nuclear power plants create a clean environ­ment with limited physical and ecological impacts.

2.7 Conclusions

The justification principle has not been systematically developed for appli­cation to nuclear power plants, fuel cycle facilities and related relevant activities, despite the fact that it is a key requirement in international and supranational regulatory activities. Only the UK has so far developed regu­lations and guidance for justification of nuclear energy, now being applied in the justification of some new nuclear designs. Other countries have devel­oped detailed regulations regarding environmental analysis which also include social and economic aspects and are close to justification exercises. Most frequently, economic advantages and benefits are the only basis for decisions.

The application of the justification principle, as defined in the IAEA Fundamental Safety Principles, and within a well-defined and complete process, will serve to present to society a valid account of the benefits derived from nuclear energy and the risks and detriments associated with it. These studies will facilitate public understanding and help in decision processes.

There are many examples regarding nuclear installations and relevant related activities where justification could provide valid insights to high — level decision-making processes. The elements to be taken into account, the justification process itself, the definition of a justification authority, and the value and limitations of the justification decision, all need to be defined formally. Valid tools are already available to define and quantify some of the key elements which are part of the justification equation; others need further research and development.

Normally, the plant owner must justify sufficient nuclear safety provisions to satisfy the requirements of labour regulatory bodies as well as those deemed essential by union and non-union staff. This requirement becomes an integral part of the regulations related to safety in the workplace, also known as industrial safety. The owner must train and maintain the vigilance of staff exposed to ionizing radiation in the course of their duties.

It is a truism that actual plant malfunctions never go ‘by the book’; that is, they are always unique and do not conform to the exact sequence defined in the deterministic or the probabilistic safety analyses. Furthermore, there is never full assurance that all possible failure modes and combinations have been investigated. The most likely cause of this diversity of cause and effect is the known complexity of the plant systems, combined with the much larger complexity that arises from innate human diversity at the operating staff level. Human behaviour, both as individuals and in groups, can exert very large positive as well as negative effects on calculated fre­quencies and consequences. Put in another way, a highly competent operat­ing crew can safely operate even a seriously flawed plant design; at the same time an incompetent operating crew is capable of doing great damage to even an extremely well-designed plant. Lastly, considering the long time span of plant operation (50 to 100 years), all of the important variables can range from fully satisfactory at one point in time to unsatisfactory at a later time. Human managers are always responsible for sustaining high perfor­mance (and hence low risk) at all times — and even they are never perfect. Hence, there is a basic need for audits by an independent regulatory agency.

Taking into account the above described UNSCEAR estimates for the effects that can be attributed to the normal operation of NPPs, and its own findings, the ICRP recommended the use of ‘detriment-adjusted nominal risk coefficients’ for the only purpose of radiological protection at low doses. These coefficients are numerals expressed in % per unit dose, which — mul­tiplied by dose — aim at quantifying the plausibility or ‘degree of belief’ of latent effects as a result of radiation exposure. The coefficients are nominal, in the sense that they do not necessarily correspond to a real value, since they relate to hypothetical (not real) people who are averaged over age and sex. Since the different possible effects may cause distinct detriment to people, the coefficients are multidimensional, quantifying the plausible expectation of harm, and including among other factors the weighted plau­sibility of fatal and non-fatal harm, and life lost should the harm actually occur.

The detriment-adjusted nominal risk coefficients recommended by ICRP are:

• for malignancies,

— 5.5% Sv-1 for a whole population

— 4.1 % Sv-1 for an adult population

• for hereditable effects,

— 0.2% Sv-1 for a whole population

— 0.1 % Sv-1 for an adult population

which result in a combined value of

• 5.7% Sv-1 for a whole population

• 4.2% Sv-1 for an adult population.

The risk coefficients imply a central assumption of a linear dose-response relationship for the induction of cancer and heritable effects, according to which an increment in dose would induce a proportional increment in risk even at low doses. This assumption is essential for the practical implementa­tion of the system of radiation protection (see hereinafter) in order to provide the basis for the summation of doses of various levels, from differ­ent sources, and from external exposure and from intakes of radionuclides (Beninson, 1996).

It is again emphasized that the risk coefficients are nominal, i. e. artificially constructed using average phantoms. Figure 11.4 indicates how the effective dose is constructed using these phantoms (ICRP, 2007b).

International radiation safety standards have taken the UNSCEAR esti­mates and ICRP recommendations into account, rounding an overall nominal risk coefficients to ~5% Sv-1, as the basis of the requirements for limiting radiation risks. This is because, while it is not demonstrable, it is considered plausible that risks be attributable to radiation exposures, even at low doses, and therefore for reasons of social duty, responsibility, utility, prudence and precaution, it is ethically required that regulatory bodies do ascribe such nominal radiation risks to prospective exposure situations. However, both UNSCEAR and ICRP had made clear that while nominal risk coefficients can be used for attributing risk for purposes of prospective planning, they cannot be used for attributing factual health effects retrospectively.