Before fuel loading, the operators must be trained and qualified to operate the fuel handling equipment. Detailed procedures and operating instruc­tions must be prepared and exercised during the training period with dummy fuel assemblies. Strict attention to criticality such as boron concen­tration levels in pressurized water reactors (PWR) is essential at this stage. Once fuel is loaded, for light water reactors (LWRs) the upper vessel inter­nals and the pressure vessel head are installed. At this point, the operator carries out additional mechanical and electrical tests to verify that the reactivity control systems are functioning properly and reliably. The initial core monitoring system data will familiarize the operator with some practi­cal reactor core experience.

Some additional tests are normally performed just before initial critical­ity to provide further assurance that the plant systems and components required for plant operation perform as expected. The plant is then brought from cold shutdown to hot shutdown to initial criticality for the start of low-power physics testing. A variety of tests are performed to confirm the core design values as used in the FSAR and other technical analyses. Reactor power is then raised through steps with test programmes at each step. The tests include physics measurements, plant shutdown and heat removal capabilities, power transients, loss of site power tests, and instru­mentation and control checks.

After full power is reached and maintained for a period of time, the plant should be shut down and thoroughly inspected, and the commissioning data assessed. Any changes to the plant would be evaluated thoroughly to ensure that safety margins meet the design specifications and that the plant can perform reliably. Finally, plant acceptance testing is performed to ensure that the plant meets the contractual output. The plant operating staff typi­cally become proficient in the operation and maintenance of components and systems during the commissioning activity.

During operation, the licensee is required to maintain detailed records concerning operations, and the nature of these records is stated in the operating license. These can include the results of effluent and environmen­tal monitoring programmes, operating and maintenance procedures, results of the commissioning programme, results of inspection and maintenance programmes, and the nature and amount of radiation, nuclear substances and hazardous substances within the nuclear facility.

The operator must also manage plant configuration changes and the status of the SSCs over the life of the plant. A key aspect of this is the management of ageing, including both degradation and obsolescence, par­ticularly for those SSCs important for safety. It is likely that the licensee will have to demonstrate to the regulator that it has a comprehensive and systematic management programme to address SSC ageing. The IAEA has published recommendations for the establishment, implementation, and improvement of ageing management programmes that can be used to develop an effective strategy (IAEA, 2009d). According to this guide: ‘Evaluation of the cumulative effects of both physical ageing and obsoles­cence on the safety of nuclear power plants is a continuous process and is assessed in a periodic safety review or an equivalent systematic safety reas­sessment programme.’ The science, technology and regulatory aspects of ageing in nuclear power plants are considered in detail in Tipping (2010).

The major part of NPP commissioning covers a period of about 2-3 years from the finished erection of the first systems (electrical energization) up to the start of the commercial operation of the station. About 40-50 profes­sionals are usually assigned to perform the tasks for this activity. Major support will be provided by manufacturers with the participation of plant operation and maintenance personnel.

The overall manpower requirements for the stage of plant operation are not so much directed by the plant output capacity as by the policies regard­ing the uses of external contractors. For guidance, the manpower require­ments can be defined as an average of one worker for 1 megawatt electrical of gross capacity of the plant (680 MWe gross capacity would require approximately 680 workers). This average value of workers per MWe is not linear and tends to be as low as 0.7 as the plant gross capacity increases, especially for multiunit stations.

A comprehensive training system should include full and accurate descrip­tions of the following elements.

Training regulatory requirements

It is necessary to identify the national and international standards for train­ing that are applicable and the training ‘certification’ model according to the training regulatory requirements. It is essential to decide the roles played by the plant management, the training organization, the regulator or, if appropriate, an external independent organization in the ‘certification’, ‘accreditation’ or any other concept that will assure the quality of the train­ing programmes.

A good example of an external certification model is the training accredi­tation process conducted by the National Academy for Nuclear Training of INPO.

The need for human resources in nuclear power programmes 183 Training organization, management and staffing

An important consideration when organizing a training system is to clearly define the role of the line managers, as they are the owners of the training programme of their people, the role of the training manager, as the training consultant and administrator, and the responsibilities for training and quali­fication of the line managers, training managers and plant personnel.

The training organization should be designed so as to include the func­tions and responsibilities of instructors and other training staff and the description of the training committees (expected attendees, meeting fre­quency and proposed agenda). The training organization manual has to include a detailed description of the administration of training and qualifi­cation activities. Finally, it is recommended to identify some indicators of training effectiveness.

There are several areas in which national competence needs to be devel­oped to be able to service the NPP for its proper upkeep during the longer — term operation and for future expansion of the nuclear power programme in the country. The longer-term operation would include possible extension of operation beyond the initially licensed operating period. In addition to building national technical capabilities in the major areas described in Section 7.7, other important areas for managing the nuclear power pro­gramme in the longer term are building human resources, developing tech­nical support organizations, developing national safety standards and engaging in international cooperation.

The Advanced Power Reactor 1400 (APR1400), with a rated power of 1400 MWe, is the largest two-loop PWR currently available. The APR1400 is an evolutionary reactor developed in the Republic of Korea, based on the accumulated experience from the design and operation of the 1000 MWe Korean Standard Nuclear Power Plant and from the EPRI URD (EPRI, 1995, 1999). The APR1400 incorporates a number of improvements to meet operators’ needs for enhanced safety, performance and economics and to address new licensing requirements such as the mitigation of severe acci­dents. The APR1400 has a very characteristic configuration, with two large steam generators and four reactor coolant pumps in a ‘two hot legs and four cold legs’ arrangement. The APR1400 also features fully digital instrumen­tation and control (I&C), and a main control room designed with full consideration of human factors. The APR1400 incorporates safety systems with both active and passive characteristics, and has also been designed to take advantage of modularization and prefabrication construction tech­niques to ensure a predictable construction budget and schedule. Two

APR1400 units are currently under construction in the Republic of Korea (Shin-Kori 3 and 4), and they are expected to enter commercial operation in 2013-14. The APR1400 has also been selected for the first four units that will be built in the United Arab Emirates.

Senior management of any organization whose operation has been autho­rized by a national regulatory body must be knowledgeable of its obliga­tions under the operating licence and must be directly involved in setting and sustaining safety policy of the organization (IAEA, 2006). It has been shown that effective leadership is required to sustain a high level of performance, because senior management sets the ‘tone’ for the whole organization.

Early in the process of starting a nuclear program, governmental authori­ties of the country must establish a stable relationship between several aspects of the program. Figure 10.1 presents one model of a stable safety management relationship.

The Challenger space orbiter failure in 1997 and the Columbia failure in 2003 each displayed several contributing elements, but the root cause in both cases was human error. Risks were taken without full understanding of the probabilities and without proper balance in senior management deci­sion making — as judged post facto. It is interesting that in none of these cases was the future risk of the event recognized before the fact, even

Table 10.1 A typical accident — Toronto Airport

though more junior staff gave clear warnings in all three cases. (The same pattern existed before the recent oil drilling disaster in the Gulf of Mexico.)

There is a direct correlation between the basic principles of radiological

protection recommended by ICRP and basic universal ethical doctrines.

This correlation can be described as follows (Gonzalez, 2010): [7]

• The principle of inter generational prudence is based on aretaic ethics (namely virtue ethics) which is expressed with the aphorism ‘Do good that will not be returned’ and is the basis for complying with the UN’s Precautionary Principle (UNESCO, 2005).

Teleological and utilitarian ethics belong to a family of ‘social-oriented’ ethics; deontological and aretaic ethics belong to a family of ‘individual — oriented’ ethics. In relation to radiation protection, namely for keeping humans safe from radiation harm or injury, teleological and utilitarian ethics would aim at the principles for protecting society as a whole, while deontological and virtue ethics are more focused on individual protection.

The principles and their ethical foundations are interrelated and appli­cable to all exposure to radiation risk, namely to exposures to ‘certain’ doses and to exposures to the ‘potential’ of doses. Moreover, distinctly from con­ventional ethical approaches, the ICRP principles harmonize all the prevail­ing ethical doctrines and use all of them in conjunction, as illustrated in Fig. 11.5 (Gonzalez, 2010).

Building bridges among the ethical doctrines and applying them to radio­logical protection has historically been at the roots of the ethic accomplish­ment of the ICRP recommendations.

The occurrence of a nuclear emergency will lead to a sequence of response actions focused on managing the incident and mitigating its effects (the responsibility of the site operator), and protecting the public against actual or potential effects of the incident (the responsibility of the site operator, and governments through the respective emergency planning and pre­paredness authorities). Many activities will be undertaken by the operator and respective orders of government (local, regional, national and, where appropriate, international or neighbour countries) for responding to the emergency in a timely and adequate way.

Rapid and effective coordination among all organizations involved is a crucial issue for a successful response. Coordination among these organiza­tions requires implementation of a well-structured action plan based on an efficient network of command and control centres. Usually, every response organization has its coordination centre to command its tasks and coordi­nate them with the rest of the response organizations. Every command and control centre should have clearly established its role and be endowed with sufficient human and technical resources to fulfil its mission.

12.7.1 Coordination centres operated by licensees

The operator usually has two emergency centres closely interconnected. The first centre is located in the facility and the second in a location outside the areas likely to be affected by the emergency. In addition to these two emergency centres, the licensees also operate an emergency centre in the company headquarters, mainly to be informed, but also to help the reactor operators in the implementation of severe accident guidelines and proce­dures. The functions of the operator’s centres are:

• To provide technical support to the personnel operating to bring the plant to safe conditions as soon as possible and minimize the impact of the emergency on the facility and its workers as well as reduce the uncontrolled release of radioactive material to the environment

• To identify and request external aid required for the plant, according to the evolution of the emergency

• To provide public authorities that manage the external emergency with available information on the emergency at the facility, to facilitate the implementation of measures to protect the population

• To direct urgent off-site emergency activities until the public authorities assume direction of operations.

To adequately fulfil this mission the operator’s operational centres have access to all available information on plant design and operation; adequate procedures to operate the plant in degraded safety conditions; simulators of the behaviour of the plant that can predict the evolution of any event and anticipate the most appropriate mitigation measures; detailed informa­tion on the geography and on-line meteorological data for the site to evalu­ate atmospheric dispersion of potential uncontrolled release of radioactive material by air; redundant connections to the emergency coordination centres used by public authorities to take decisions, to inform them on the development of on-site emergency and request their help if necessary; direct connections with suppliers of equipment and services; connections with the nuclear facilities or similar technology and electricity generation of other countries; and connection with centres of water resources manage­ment, national meteorological services and other relevant coordination centres that operate networks or systems relevant for emergency response.

The fuel for current water reactors is in the form of pellets of uranium dioxide or a mixture of uranium and plutonium dioxide (MOX fuel). The uranium enrichment (content of uranium-235) is typically 3-5% in light water reactors. The pellets are very stable ceramic cylinders about 1 cm in diameter and 1 cm high. The pellets are placed in sealed thin metal tubes (e. g. of stainless steel or zirconium alloy), which are kept together as bundles to form a fuel element. The fuel element, which typically contains between 60 and 300 fuel pins, can be handled as an entity (Fig. 14.2). Fresh nuclear fuel elements need to be handled with care to avoid contamination and mechanical failures, but do not require radiation shielding. After the fuel has been used in the reactor it can still be removed and handled as an intact

fuel element. It is, however, highly radioactive due to the formation of fission products and transuranic elements in the fuel and activation prod­ucts in the fuel element structure. The typical composition of spent fuel (excluding the fuel element structure) is:

• 95% uranium (remaining enrichment about 0.8%)

• 1% plutonium

• 4% fission products and transuranic elements other than plutonium.

A more detailed composition of a typical LWR fuel element is given in Table 14.1. Some of the fission products are very short-lived with half-lives of a year or less, while others have half-lives ranging from 30 years to mil­lions of years.

The spent fuel element has a high concentration of different radionu­clides that decay by emitting a-, P — or y-radiation or undergo spontaneous fission that emits neutrons. The a — and P-radiation is mainly absorbed in the fuel itself and is the energy dissipated as heat (decay heat), while the y — and neutron radiation is more penetrating so that the spent fuel will require shielding. Some neutrons also generate additional fission in the fuel, which will require control of the spent fuel configuration to avoid criticality. The spent fuel thus needs shielding and cooling during the subsequent handling. After removal from the reactor the fuel is stored under water for several years to allow cooling. During the first year the decay heat goes down rapidly as the short-lived fission products decay. After about five years the decay heat is dominated by cesium-137 and strontium-90, which both have a half-life of about 30 years.

Spent fuel remains radioactive for very long times, hundreds of thousands of years, and will eventually need final geological disposal to ensure long­term isolation from humans and the environment. In Fig. 14.3 the radioac­tive decay for spent fuel is shown. The curve shows the toxicity index, which takes into account not only the activity but also the harm the radioactive substance would give if incorporated into the body (essentially eaten). After the first few years the toxicity is dominated by cesium and strontium. After a few hundred years the toxicity will be dominated by the transuranic ele­ments, such as plutonium and americium. By removing plutonium and pos­sibly also some other transuranic elements the long-term toxicity and heat release can be reduced, but it is generally considered that long-term geo­logical isolation will still be needed.