Category Archives: Infrastructure and methodologies for the. justification of nuclear power programmes

Responsibilities of regional and local authorities

The preparedness for nuclear emergency requires many arrangements in emergency intervention zones established in the surroundings of nuclear facilities according to regulatory criteria. Regional and local authorities play an important role in the implementation of such arrangements, since they have a detailed knowledge of the geographical, economic and social conditions of these zones. The responsibilities and functions of regional and local authorities are usually targeted to address logistic and operational issues of the nuclear emergency plans. In discharging these responsibilities, regional and local authorities usually assume the following logistic and operational functions:

• Establishing local action plans and procedures to ensure that emergency countermeasures can be implemented in such a way that every poten­tially affected person will be adequately protected in case of emergency

• Providing adequate facilities to implement emergency countermeasures, including adequate centres to concentrate, monitor, decontaminate and take social care of victims, including relocation if needed

• Providing medical facilities and resources adequately equipped for first aid involving medical care of victims potentially irradiated or contaminated

• Providing adequate facilities to store emergency equipment in suitable conditions to be used in case of emergency

• Establishing and implementing training programmes for intervention personnel, and promoting their participation at all levels of responsibility

• Preparing, conducting and evaluating periodic drills and exercises orga­nized to train responders and verify plan effectiveness

• Develop and put in practice adequate public information programmes aimed at teaching people how to protect themselves in case of emer­gency, and efficiently transmit information needed to manage the emer­gency in the most efficient way.

Sources of radioactivity

In a nuclear power plant there are two sources for the production of radio­active substances, the fission by neutrons and absorption of neutrons (trans­mutation by neutron absorption) taking place in the fuel itself, and the irradiation of material in the reactor that is exposed to the neutrons from the fission process (activation). The radioactive substances produced from the first source are fission products and transuranic elements (elements heavier than uranium). The fission products are the lighter elements (e. g. cesium, strontium and iodine) that are created when the heavier atoms (e. g. uranium or plutonium) are split (fissioned) and energy is released. The transuranic elements (e. g. plutonium, americium and curium) are generated by the absorption of neutrons in uranium and the successively created transuranic elements. The amount of fission products and transuranic ele­ments is directly coupled to the energy that has been generated. The spent fuel is highly radioactive and will need shielding and cooling for the subse­quent handling.

A typical composition of spent nuclear fuel is shown in Table 14.1. The fission products and transuranic elements are kept in the fuel and contained by the fuel cladding. They will only be released to other parts of a nuclear power plant if the fuel cladding is damaged. Minor amounts could also

Table 14.1 Composition of spent nuclear fuel (PWR, 60 MWd/kg U, 15 years out of the reactor): the most important radionuclides











Fission products

Transuranic elements



























































































emanate from fuel contamination on the outside of the fuel cladding that remains after the fuel fabrication.

The second source of radioactive substances in a reactor, activation prod­ucts, is the result of irradiation of material in the reactor by neutrons from the fission process. Only material inside the reactor pressure vessel and in the concrete that immediately surrounds it will be exposed to sufficient neutron fields for activation. The highest activity will be generated in the core components holding the fuel and in other internal parts in the pressure vessel. Also material contained in the coolant or coolant-moderator water, which passes through the reactor core, could become activated.[80] This could be metal ions or particles from corrosion in the primary circuit of the reactor or other trace elements contained in the coolant or coolant-moder­ator. Radioactive substances thus created could then be transported through the primary system of the reactor and contaminate surfaces and filters, thus creating a radiation field around these components and in the end a radio­active waste.

A list of typical activation products is given in Table 14.2. To minimize the creation of activation products, one strives to keep the primary circuit water very clean through ion exchange and mechanical filtering as well as

Table 14.2 Activation products in fuel cladding and mechanical components (PWR, 60 MWd/kg U, 15 years out of the reactor): the most important radionuclides


Half life (years)

Activity (Bq/tU)

Activation products C-14





























to reduce the corrosion by adjusting the chemical environment, e. g. by adding lithium hydroxide or hydrazine to the coolant. Also gaseous radioac­tive fission and activation products are formed and transported by the coolant and coolant-moderator to a degasification system.

Overnight costs

The principal components of OC are engineering-procurement- construction (EPC) costs, owner’s costs and contingency costs. EPC repre­sent the bare costs of plant construction comprising direct (equipment, material, labour) and indirect (engineering/construction services) compo­nents. Under an EPC contract, the contractor — usually the vendor — is responsible for the engineering design, including adaptation to match site and other location-specific conditions, production or procurement of the necessary plant components as well as materials, and plant construction. Given the complexity of a nuclear power plant and its financial dimensions (economic risks), the contractor usually subcontracts parts of the work or shares parts with the plant owner or both. In essence, subcontracting and owner involvement are a measure of risk management (plant completion risk).

Owner’s costs are additional investment expenditures borne by the plant owner and usually relate to costs associated with property (land) acquisi­tion, site selection and preparation, bid evaluation, cooling infrastructure, administration and associated buildings, site works, project management, permits, legal services, licences, local taxes, staff and operator training, and possibly also expenditures for connecting the plant to the grid, i. e., switch­yards and transmission infrastructure.

Contingencies are provisions for any unforeseen or unplanned expendi­tures associated with the project. They are generally estimated as a specified percentage of EPC but also depend on the type of contract arrangement (turnkey contract or several contracts managed by the plant owner or cost — plus contracts).

The absolute and relative values of these OC components depend on location and plant design and therefore can vary considerably even within a country and for plants of similar design and size. The major factors in variability across countries include domestic labour and material costs, site-specific conditions and readily available infrastructures, finance arrangements and interest rates, institutional and regulatory framework, standardization and multiple-plant versus single-plant construction (econo­mies of scale). They are also a function of the localization rate, i. e. the ratio between imported and locally manufactured or procured components and participation in the civil works. The availability of nuclear-specific skilled tradespeople and engineering capability is another factor affecting OCs. Site- and geography-specific conditions may add costs to an otherwise standard design, such as additional design and engineering costs for meas­ures to protect a plant in an earthquake-prone location.

Unit size and plant design are other factors that explain OC cost differ­ences. Typically, smaller plants have higher specific investment costs (i. e. dollars per kW(e)) than larger plants, since certain cost components are relatively independent of size. For example, Westinghouse’s AP-1000 design is 80% more powerful than its AP-600 design, but the AP-1000’s overnight costs are only 15% to 18% higher than the AP-600’s (RWE Nukem, 2002).

Regulatory intervention can add to overnight costs, especially if it requires design modifications once the project is well underway. The exact impact of regulatory changes on cost is elusive because the regulatory process varies across regions. A number of studies have tried to quantify the impact of regulation on nuclear power investment costs but have not generated broadly applicable quantitative results beyond the straightforward reality that construction delays increase investment costs (Mooz, 1979; Paik and Schriver, 1979; Komanoff, 1981; Zimmerman, 1982; Cantor and Hewlett, 1988; McCabe, 1996; Canterbery et al., 1996).

For new designs, or for construction in new environments, OC may include first-of-a-kind (FOAK) costs. FOAK costs include a particularly high share of contingency costs to cover unforeseen events given the lack of experience with the design, the environment or the country. They can add as much as 35% to OC (UoC, 2004). Costs are lower for subsequent units, but some (decreasing) additional costs will persist until experience has been accumulated on several (about five to eight) essentially identical designs. For example, Progress Energy recently announced overnight costs of $3376/kW(e) for a second AP-1000 at its Levy County site, substantially lower than the first unit’s $5144/kW(e). And the Russian Federation’s Kaliningrad-2 cost $1667/kW(e), half the cost of Kaliningrad-1. In these examples the cost reductions also reflect the facts that some site preparation costs incurred for the first unit are not reincurred for the second unit and the vendors’ allocations of costs among the two units are to some extent arbitrary.

The specific components of FOAK costs are uncertain and prone to escalation. For example, the OC cost estimate for Olkiluto-3, a FOAK third- generation European Pressurized Reactor (EPR), has reportedly risen from €3.0 billion to €5.3 billion due to construction delays caused by FOAK — related quality issues, design revisions and approvals, and logistic challenges not experienced for a long time (NW, 2010a; KPMG, 2010).

International comparisons of investment costs are also often obscured by the unavailability of information about the exchange rates[93] that are used and, if escalation costs are included, about the components that are affected and the escalation rates assumed. Finally, OC may include the initial core load of nuclear fuel.

The percentage of each OC cost component varies according to several studies that include both data for plants that have been built and estimates for future plants (Kozlov, 2004; UoC, 2004; Scoggs, 2007). For example, EPC ranges from 73% to 97%, owner’s cost between 2% and 15% and contin­gencies between 1% and 13% of total OC. The different percentages reflect various cost-shaping factors such as plant design, whether it is built on an existing or greenfield site, economies of scale (in terms of both unit size and the number of units previously built), contractual arrangements and the cost of labour. For example, low owner’s costs may indicate a project built on an existing site or a local government subsidy for site development and preparation. Low contingency costs might indicate a turnkey contract,[94] while high contingency costs might indicate a cost-plus EPC contract.

Practical considerations

A key practical consideration in relation to EIA is who should be respon­sible for carrying out the EIA assessment and producing the Environmental Statement, and thus bear the costs of doing so. The EIA Directive stipulates that the developer is to carry out the EIA and provide the requisite infor­mation to the authority. For the purposes of the EIA directive, the ‘devel­oper’ is defined as either the person making an application for authorisation of a private project, or the public authority which initiates a project. Accordingly, the obligation to carry out the EIA lies firmly with the party initiating the project, and in the case where this is a private party, the obli­gation on the authority is to ensure that the EIA has been properly formu­lated. This will necessarily mean that early engagement with the process is essential, for both the developer and the authority, particularly so that detailed arrangements for public consultation can be coordinated.

However, the EIA process does not (and should not) end with the deci­sion of the authority giving consent for the activity to proceed. Where an activity is deemed to be justified in light of its environmental effects, the activity and its effects on the environment should be subject to appropriate supervision. This process is known as ‘monitoring’ and can be distinguished from the main EIA process and preparation of the Environmental Statement on the basis that it should continue throughout the life of the activity in question. The purpose of monitoring is essentially to ensure that the environmental effects which were identified in the EIA Environmental Statement were correct, but also to provide authorities with sufficient infor­mation to enable them to decide whether enhanced measures are required to mitigate the environmental damage that will occur. This additional facet of EIA is not necessarily present in all Member States’ domestic legislation. In the UK, for example, a development consent granted in reliance on EIA will usually have conditions attached where these are seen as necessary to ensure that environmental impacts are no greater than predicted. However, the Environmental Statement does not, of itself, create an enforceable set of standards to be applied to the development. In a nuclear context, moni­toring will necessarily extend beyond the life of a nuclear power plant, and continue throughout the decommissioning phase, with the primary purpose being to ensure that any hazardous or radioactive substances remaining on the nuclear-licensed site do not cause material harm to the natural environ­ment. In the United Kingdom, a separate, comprehensive EIA procedure must be complied with before the process of dismantling or decommission­ing a nuclear reactor can commence.

Under the EIA Directive, EU Member States are also required to engage in dialogue with the European Commission for the purposes of exchanging information on the experience gained in applying the EIA process. The reasoning behind this obligation may have had something to do with the discretion that Member States are afforded in setting the threshold of ‘sig­nificance’ in determining whether a proposed Annex II activity requires an EIA. This would seem to be supported by the EIA Directive, which further requires that Member States inform the European Commission of any criteria and/or thresholds adopted for Annex II projects, so as to ensure relative harmonisation of EIA standards across Member States. This is no doubt a fundamental, but secondary, requirement of the EIA Directive, since it does not actually establish any obligatory environmental standards that must be adhered to — quality control is largely a matter for individual Member States. Nonetheless, the European experience of EIA has shown that it is a valuable and successful tool in ensuring that a national planning system adequately addresses and adapts to environmental concerns.

Other important considerations

The requirements also consider a rather long list of local phenomena that may produce harm to the safety of the NPP, such as volcanism, sandstorms and subsurface freezing of sub-cooled water. There are also phenomena which may have an impact on the long-term removal of decay heat; in this respect consideration should be given to air temperature and humidity, water temperature, available flow of water and natural and human-induced phenomena which could impair the loss of the heat removal function, such as insufficient river flow, loss of the available water reservoir, water intake blockage by marine organisms or freezing of cooling towers, among others, All these aspects have to be considered to include appropriate preventative and mitigation devices, equipment and procedures.

The IAEA has not yet developed any safety guide to measure the hazards associated with these varied concerns. Only a safety guide on volcanic hazards in site evaluation for nuclear installations is in preparation (IAEA, 2009). The guide, based on a previous 1997 document, includes the knowl­edge gained in the science of volcanology and associated risks mainly due to the enormous amount of volcanic ashes that are injected into the upper atmospheric levels, which is already evident in the effects produced on air traffic. The fallout of a large amount of ashes on the plant premises, water intakes and roads could impair the safe operation of a nuclear power plant. As in the past, there could also be mega-volcanic eruptions with the poten­tial of blocking sunlight, which may have a serious impact on all types of installations and on life on earth. These extreme effects are generally outside the scope of the design.

Project documentation

This section of the PI document includes requirements on the documenta­tion to be submitted with the bid by the vendor; the list of project document types to be submitted to the owner’s review and/or approval; the require­ments for the preparation of the list of project documents, grouping docu­ments by project phase (e. g. design, procurement, construction, testing and commissioning, plant operation and maintenance); a description of the owner’s review and approval process for project documentation; document formatting and submittal requirements (e. g. electronic and/or hardcopy, format); specific requirements for the vendor’s documentation concerning package plants; and final project documentation to be handed over to the owner by the vendor for project records and plant operation and maintenance.

19.9.4 Information management system

The information management system (IMS) describes the owner’s minimum requirements to be used by the vendor during the design, procurement, construction, testing and commissioning stages of the project. This may include specific requirements, such as the use of certain software applica­tions and databases to ensure compatibility with the owner’s system and eventual transfer of project information for use during plant operation and maintenance, once construction is completed by the vendor. Special care should be taken in identifying which information in the vendor’s databases is to be made accessible to the owner at all stages of the project, so as to facilitate project control and monitoring by the owner.

Another significant aspect of the vendor’s IMS that requires close atten­tion in this section of the PI document is the software tools that the vendor intends to use at the design stage for the performance of engineering and design activities such as 3D modelling of structures, preparation of HVAC ductwork and cable raceway layouts, production of piping and instrumenta­tion diagrams (P&IDs), schematic and wiring diagrams, piping isometrics, engineering of component databases, and computer codes for engineering calculations, to ensure smooth transferral to the owner at the end of the project and future use in plant modifications and upgrades during the operation phase.

Periodic safety reviews

A periodic safety review (PSR) is a comprehensive assessment of safety that is normally carried out at defined intervals as prescribed in the license. Plant ageing, configuration changes, modifications to procedures, significant events, operating experience, and other safety reviews occur over the life­time of a plant and the PSR is a systematic way of assessing the cumulative effects of any changes to plant safety. In addition, a PSR takes into account advances in safety standards since the time of construction or the previous review. The safety of future operation of the plant can be evaluated from the PSR.

The scope of the PSR includes an assessment of plant design and opera­tion against the current safety standards and practices. Therefore, the PSR is a tool for securing a high level of safety throughout the NPP’s operating lifetime, taking into account changes in the plant and the evolution of safety knowledge. The PSR does not replace the routine safety reviews of nuclear power plant operation, which are the primary means of safety verification throughout the plant operating cycle. The IAEA provides recommenda­tions and guidance on how to conduct the PSR (IAEA, 2003e).

From experience, the IAEA recommends that a PSR should be first undertaken about 10 years after the start of plant operation and that sub­sequent PSRs should be done every 10 years. The 10-year period is based on the expected developments in safety standards from both experience and ongoing R&D, and from the expected rate of the changes that could affect the plant. The PSR covers all aspects of operations, including manage­ment structures, reporting systems, staff experience and competence, plant configuration, safety culture, knowledge management, ageing effects on the SSCs, radiological protection, emergency planning, and operating experi­ence, to mention just a few. Owing to its comprehensive scope, the PSR provides reassurance to both the licensee and the regulator that the licens­ing basis for the NPP is still valid.

The licensee has prime responsibility for performing the PSR. The start­ing point is agreement between the licensee and RB on the scope, schedule, and requirements for the review. Owing to the broad scope of the assess­ment, a PSR is a complex task that could take up to a maximum of three years to complete. Therefore, the IAEA has broken down the review into five subject areas with 14 safety factors to ensure that the review is com­prehensive (IAEA, 2003e). These cover the plant (plant design, actual con­ditions of the SSCs, equipment qualification, ageing), safety analysis (deterministic safety analysis, probabilistic safety analysis, hazards analysis), performance and feedback from experience (safety performance, use of experience from other plants and research findings), management (organi­zation and administration, procedures, the human factor, emergency plan­ning) and environment (radiological impact on the environment). Each of these factors is reviewed and assessed against current safety standards and practices. In addition, IAEA (2003e) recommends a global assessment to integrate the results of the review of the safety factors.

Where necessary, corrective actions are determined and implementation plans are enacted. Since these actions lead to safety improvements, an objective is to complete as many of the actions as possible within the time frame of the PSR. The end point of the PSR is regulatory approval of the integrated programme to address any outstanding safety issues. Any safety gaps that cannot be reasonably addressed would require further assessment of the risk and justification to allow the plant to continue operation.

The PSR is a major undertaking that involves considerable planning and preparation. To initiate the review, the licensee establishes a dedicated project management team, develops guidance documentation laying out the scope and methodologies, defines the documents to be produced and their formats, develops a QA plan, prepares the review plan and budget, and secures approvals. The plan is then executed with many activities carried out in parallel, including, to a reasonable extent, the amelioration of issues as they are identified. This is followed by execution of an integrated plan to implement corrective actions and/or safety improvements. Further details are contained in IAEA (2003e) and recent experience in IAEA Member States is detailed in IAEA (2010b).

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

The development of a national academic programme for the education of the necessary scientists, engineers and other technicians to support techni­cal research would also be expected to be in place as part of the commit­ment to the development of the required national capabilities.

To build new nuclear knowledge it is particularly interesting to partici­pate in R&D projects related to nuclear disciplines, such as nuclear fuel, nuclear materials, management of radioactive waste, nuclear safety or radia­tion protection.

Examples of international R&D platforms are:

• The IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), established in 2001

• The European Commission within the Seventh Framework Programme (2007-2013) for nuclear research and training activities

• The European Union Sustainable Nuclear Energy Technological Platform (SNTP)

• The ‘Global Nuclear Energy Partnership’ (GNEP), founded in September 2007

• Generation IV International Forum (GIF), since January 2000.

Training for human performance improvement

Appropriate attitudes on the part of nuclear facility personnel have to be ensured. Due attention should be paid to the fact that the required attitudes cannot be achieved only through education and training. Attitudes also depend on individual characteristics and organizational culture. The behav­iour of nuclear facility managers and their ability to be everyday role models for their personnel are crucial factors.

Managers of nuclear facilities should embrace their roles in evaluating training to improve its effectiveness and to improve performance, in the same way as they embrace performance improvement. They are responsible for the behaviour of their employees and for the consequences of that behaviour. Following training, managers should observe the performance of their recently trained employees, provide timely, behaviour-specific feed­back to those employees, evaluate the impact on organizational perform­ance, and provide feedback to the trainers so that they can improve the quality of the training.

As indicated by the National Academy for Nuclear Training (2002) in their document ACAD 02-004, the training organization can support line managers and encourage professionalism through activities such as the following:

• Provide training that improves station and personnel performance.

• Ensure that the training staff serve as a role model for other personnel by exhibiting a high level of professionalism while conducting training in every setting.

• Ensure that training personnel model the standards and expectations held by the line managers; conduct training according to clear standards of performance and behaviour; let personnel know what is expected of them and hold them accountable; emphasize pride of ownership and accountability; and clarify the method for performing tasks correctly.

• Provide input to managers to encourage them to recognize exceptional personnel performance during training. Likewise, recognize superior instructor performance. Course completion certificates and awards can foster a sense of training and qualification accomplishment in personnel and their instructors.

Training should be considered a strategic tool to foster human performance excellence and therefore the improvement of safety and reliable and effi­cient plant operation.

6.6 Sources of further information and advice

This section introduces some relevant training issues which enlarge the information supplied in the chapter.

Development of national safety standards

Based on experience in design, operation and regulation of NPPs, several countries have developed their national safety standards for siting, design, construction, operation, decommissioning and quality assurance aspects of NPPs. Some international organizations have also developed safety stan­dards for NPPs that codify the good practices followed globally. In the beginning, a country starting its nuclear power programme can adopt or utilize these available international safety standards as appropriate. However, after gaining some experience it is advisable that the country’s regulatory body develops its own safety standards. To start with, the emphasis should be on developing those standards where the internation­ally available safety standards are found to be not directly applicable. This could be due to the specificities of the NPP design adopted or on account of local conditions such as climate, soil characteristics and expected frequency or magnitude of natural phenomena like precipitation, earth­quakes, etc., that may be significantly different from those in other countries.

For developing safety standards the regulatory body can engage experts from its own staff and from the operating organization, the technical support organization and academic and professional institutions in the country. For ensuring good quality, formal mechanisms should be in place for thorough review of the draft documents before their publication. In addition to the safety standards that specify the safety requirements, supplementary docu­ments like safety guides and safety manuals that provide details on the means to fulfil the safety requirements also need to be developed. The exercise of developing national safety standards is by itself a good means for enhancing the national technical competence.

216 Infrastructure and methodologies for justification of NPPs