Results of ageing studies generally points out that rotating machines (motors, motor generator sets, turbo-Generators may become more maintenance intensive. A systematic study of such rotating machines has thus become part of PLiM. Long term operation of plant may require replacement of some of these to assure long term safety and reliable operation.

1.8.1. CONTROL, INSTRUMENTATION EQUIPMENT OBSOLESCENCE AND COMPUTER SYSTEM UPGRADES

It is not possible to follow a uniform policy on ageing management of control & instrumentation for a nuclear power plant. Hence individualized approaches are to be determined for different cases.

Early HWRs were designed with a mixture of computers for direct digital control of the major plant systems as well as analogue electronic instrumentation and control equipment, most of which is not expected to last the 30-60 year range of operating life. Moreover, the issue of C&I equipment obsolescence is considered of high importance due to lack of original equipment vendors, rapid electronic technology development, and replacement of process control analogue instruments with digital electronics.

For HWR LTO, a systematic approach to identify/deal with obsolete equipment and a long range plan to address instrument obsolescence is recommended. Normally these issues are being dealt with for attending to PLiM for life attainment as well.

It was not possible to follow a uniform policy on ageing management of control & instrumentation for a nuclear power plant. Hence individualized approaches are used for different commodities. The first task was to identify the applied major areas, which require replacement/up gradation. This was mainly based on the maintenance feedback over the last two decades. The major areas which required life management/replacement are:

• Indicating alarm meters

• Channel temperature monitoring system

• Certain electronic transmitters & controllers

• Fire alarm & beetle monitoring system

• Protection system

From experience with RAPS it has been seen that a proper type of instrument, selected during the construction stages of a plant can give good performance over a number of years of operation. As revealed from the several case studies, the problem with most of the instruments is mainly due to unavailability of spares. However it is recognized that high temperature and radiation environment does affect the operation of some of the instruments as seen in the case of terminal blocks, located in CTM junction boxes in fuelling machine vaults. Thus normal operating temperature and humidity conditions for each location, especially inside reactor building is to be assessed carefully since this has bearing on the long term performance.

Inside reactor building instrument location is also to be selected carefully so as to avoid high temperature and high radiation environment. Thus it is better to install all field instruments, especially electronic instruments in accessible areas. During up gradation of RAPS-2, we had relocated all electronic instruments to accessible areas.

As discussed above, effective PLiM programmes should integrate both safety and operational performance concerns. Regulatory attention focuses on the former, with the primary interest being on obtaining assurances in effective ageing management of SSCs important to safety. In practice, the specific aspects of regulatory requirements and expectations differ considerably between member states. In general, however, they follow the guidance outlined in the IAEA safety standards, safety guides and technical report series [1-20] described in Section 1.4.

The operating NPPs have been licensed on the basis of the requirements established during the design process. To demonstrate compliance with regulatory requirements, the day-to-day operating envelope must be maintained within the bounds of the assumptions of the plant safety analysis. These operating envelopes include:

• Special / critical safety system setpoint limits and system availability.

• Acceptable range of process parameters.

• Allowable equipment configuration and operating states.

Some examples of regulatory approaches and strategies for ensuring the implementation of effective and systematic ageing management practices are summarized in the following sections.

Fig. 5. Process for renewal of operating permit beyond design life time in Spain.

The initial dry fuel loading and the on-power refuelling of the horizontally oriented pressure tubes has caused minor scratching of the lower quadrant of the tubes by the fuel bundle bearing pads. The use of stainless steel shims during initial fuel loading, in recent years, has eliminated the scratching at this stage. Examination of removed tubes has shown that the scratches are rounded and shallow and tests have shown that they are unlikely to cause Delayed Hydride Cracking (DHC) initiation under reactor operating conditions.

The high flow rate of the coolant through the fuel bundles causes bundle vibration, which results in minor fretting of the tube wall by the bearing pads. In reactors with a 12-bundle fuel string (i. e. all CANDU 6 and Pickering units), experience from the examination of removed tubes and from the many periodic and in-service inspections performed to date, has shown that these fret marks are shallow and are not likely to initiate DHC. In reactors with a 13- bundle fuel string, fuel bundle bearing pad fretting in the inlet rolled joint area, particularly at the burnish mark, has resulted in deeper fret marks.

Debris can possibly come from material left in the Primary Heat Transport System during construction/installation, from in-service degradation of components, or from use of unfiltered make-up water to the PHTS. Debris, which becomes entrained in the coolant and then trapped in the fuel bundles or between the bundles and the tubes, can result in debris fretting damage of both the fuel sheaths and the pressure tubes. The fret marks in the pressure tubes can be deep and may require tube removal, although cracking or tube failure have not been observed. The occurrence of severe debris fretting in pressure tubes is of a low frequency and random thus it is not seen as a generic ageing mechanism.

Inspection and monitoring activities are designed to detect and characterize significant SSC degradation before safety and production margins are compromised. Together with an understanding of ageing degradation, the results of inspections and monitoring and the subsequent trending of degradation parameters, provide a basis for decisions regarding the type and timing of maintenance actions and decisions regarding operational changes and design modifications to manage detected ageing effects. A risk informed methodology can contribute to minimizing ageing by providing the basis for targeted and more effective inspection and assessment.

A proactive monitoring, inspection and trending programme can be used to detect steam/feedwater piping wall thinning, due to flow-assisted corrosion (FAC), to characterize degradation rates and locations and, if necessary, predict when repair/replacements need to be implemented.

Complementary to Maintenance Enhancement Project, CNE-Prod has started in 2003 the development of a Plant Life Management (PLiM) Programme strategy for Cernavoda Unit 1. The programme use approaches that build on PLiM experience at other advanced CANDU and/or PWR Stations from western countries but implementation strategy is specific to Cernavoda Unit 1. The objective is to start early in plant life on systematic ageing management and hence reap benefits during the on-going plant operation, which is expected to be longer than that for the early CANDU plants.

The overall approach adopted is to consider all the issues relating to physical plant ageing as well as “institutional” ageing through an integrated approach. The physical plant assessment focuses on the continuing ability of the structure, system or component to meet the specified performance standards over its design life. If, in spite of maintenance and or refurbishment, a critical structure or component is deemed not likely to be able to meet its performance requirements, then a timely replacement must be undertaken with due consideration to the associated cost-benefits justification.

Figure A. III.1 below, shows the key elements of the life management programme for physical plant, as follows:

• Based on a screening methodology, a set of critical systems, structures and components is identified consistent with meeting the station goals and objectives.

• Life assessment studies are performed for critical structures and components that are generally passive in nature and typically designed not to be replaced as part of normal maintenance programme. The objective is to assess relevant degradation mechanisms and develop an appropriate ageing management plan to achieve design life with the option for life extension.

• Maintenance optimization studies are performed for critical systems with emphasis on active components (that are generally designed to be replaced as part of the normal maintenance programme), in order to preserve the defined systems functions. [8]

• Obsolescence studies are performed on generic plant components that cannot be maintained or refurbished in a cost effective manner due to several factors (such as availability of spares and new developments in technology that make replacement a viable option). Obsolescence relates primarily to instrumentation and control equipment and computer systems.

• Integrated safety and performance assessment is performed with AECL[9] support in order to demonstrate continued compliance with the safety and licensing basis requirements as the plant ages.

•

OPEX programme addresses emerging key issues that may adversely affect plant safety and reliability and may not be addressed by the assessment process described above. This programme relies on monitoring of the operating experience feedback, recent R&D activities, and new developments in regulatory requirements and industry practices.

Fig. A. III.1. Plant life management — Physical plant assessment.

Figure A. III.2 below shows a corresponding block diagram covering the impact of “institutional” ageing. If the management infrastructure is allowed to degrade beyond what are normally accepted industry standards, then the plant will be forced to shutdown until the underlying issues can be resolved in a satisfactory manner. The major initiatives for improvement dealing with institutional ageing included in the station technical programmes schedule are:

• Enhance day-by-day engineering & technical support to plant emerging issues

• Increase technical unit personnel efficiency

• Enhance engineering analysis expertise

• Technical support to Unit 2 commissioning

• Optimize design bases

• Improve design engineering processes

• Improve station governing documents (station policies and procedures)

• Complete implementation of self assessment process

• Optimize data collection & electronic performance indicators reporting

• Upgrade OPEX process

• Optimize corrective actions process

• Improve first line supervisors management skills

Plant Life Management

Fig. A. III.2. Plant life management — Institutional assessment.

Legal limitations are in place concerning radiological doses that NPP personnel may accumulate during a given period of time. Legally allowed doses are laid down conservatively such that they are not expected to cause any damage to health. An individual annual dose of 20mSv is such a value. Considering that AMP/PLiM activities may involve special circumstances that may lead to increased personnel exposure to ionizing radiation, it is essential that all such tasks are planned in advance with a view to limiting the dose. The concept of as low as reasonably acceptable (ALARA) must be rigorously applied. Shielding, distance and time spent on the activity must be optimized to create those conditions amenable to ALARA principles. Practice on mock-ups will facilitate a rapid and technically sound work programme. Therefore, radiological protection measures, where applicable, become an integral part of PLiM.

In recent years, there have been considerable advancements in tubing eddy current testing technology and also much better knowledge of tubing degradation mechanisms and which EC inspection techniques can be best used for detection. Also, improvements in analysis and interpretation of eddy current data and use of these data for predicting early signs of tubing degradation have been developed. A proactive SG ageing management programme uses the results of the life assessment work, couples it with these advanced inspection and interpretation techniques, and then develops an enhanced SG tubing inspection programme for plausible tubing ageing degradation. The objective is to have as-early-as-possible identification of any possible tubing degradation by focusing inspection effort on the “age — sensitive” regions of the tube bundle with appropriate techniques capable of detecting the plausible degradation. Typical examples of information not previously available from eddy current inspection, but now available, is quantification of the depth and extent of deposits on the tubing primary side, and detection of tube-to-support gaps for use in vibration and fretting wear assessments.

A. I.2.1. LIFE MANAGEMENT PROGRAMME FOR THE PRESSURE TUBES

The objective of formulating the life management programme for the PT has been the understanding of the various degradation mechanisms and their mitigation through research and development in the fields of design, manufacture, operation, in-service inspection and life extension. Some of the aspects of life management are listed below:

• Design modification and improvement in manufacturing procedures.

• Inspection and monitoring of the degradation related parameters.

• Development of methodologies for assessment of fitness for service with due importance to safety and not being unduly conservative.

• Development of analytical codes for degradation modelling and residual life estimation of PT.

• Development of life extension tools and technologies.

• Post irradiation examination studies.

• Replacement of degraded PTs.

A. I.2.1.1. Design modifications and improvements in manufacturing procedures

Problem due to deuterium pick-up in pressure tube has been mitigated to a large extent by reducing the maximum initial hydrogen in the as-manufactured condition to 5 ppm from the earlier value of 25 ppm and changing the material to cold worked Zr-2.5Nb from cold worked Zircaloy-2. The specification has been modified to limit the content of chlorine, phosphorus and carbon achieved through quadruple melting. Such tubes are being used in all reactor currently under retubing as well as new ones under construction.

The effect of axial creep-growth elongation of the PT on the service life of the coolant channel has been resolved by providing sufficient length for the bearing sleeve. Through suitable design of the coolant channel it has now been ensured that diametral creep-growth and bending creep-sag of the coolant channel do not limit the life of Zr-2.5Nb PT.

The objectives of this pilot project can be summarized as follows:

• Develop a general procedure to apply PLiM methodology, including guidelines for life assessment reports (LA’s), condition assessment reports (CA’s) and systematic assessment of maintenance reports (SAM’s).

• Test these procedures and get feedback from the field application in order to improve them.

• Convince all plant staff about the importance of the PLiM programme and the role they should play as active participants.

• Gain experience in the application of the methodology to start the next stage of the programme.

The first approach to a life assessment study was performed by CNE personnel for the steam generators during the last two years. This study, together with the knowledge acquired during training, has set the basis to develop the life assessment procedures and guidelines.

Besides steam generators, other four major components are being analyzed: Main feedwater pumps, moderator heat exchanger and pressure and inventory control system feed pumps. The selection of these components resulted from a screening stage in which safety and economical issues are considered, in order to determine the most important structures, systems & components (SSC’s) to be analyzed. During the screening it is also possible to determine for which components a LA analysis is needed, and for which a CA would be sufficient. It is worth noticing that a residual risk is associated with any screening process since there are components for which a lower depth or no analysis is indicated.

Life assessment: analysis is a deep study that is intended to gain as much knowledge as possible about the component. To achieve this, three main tools are used, documentation review, interviews and walk downs.

(a) Documentation review: the extent and usefulness of this review strongly depends upon the information management policy followed by the plant. If the information was properly managed, details from construction, major maintenance and design modifications, historical operation parameters, failures and inspection plans applied can be obtained.

At CNE, most of this information is correctly storaged and classified. However, it was sometimes needed to check few reports from early operation years with responsible personnel. In these cases, cooperation among different plant areas resulted essential.

Beside plant information, it is necessary to review international literature in order to get updated information about what is being done in similar plants around the world. As this regard, international organizations such as International Atomic Energy Agency (IAEA), Electric Power Research Institute (EPRI), World Association of Nuclear Operators (WANO), and CANDU Owner Group (COG) are important sources of information.

(b) Interviews: as it was previously stated, interviews were found to be a powerful tool for those cases where information was not either clear or available. However, even having enough information about the component, the operational experience from operation and maintenance staff is always valuable and should not be underestimated.

(c) Walk downs: walk downs should be aimed to detect anomalies that cannot be appreciated using the other tools described above. Inefficiency of supports and insulators are classical defects that can be detected during walk downs.

As well as it is done for interviews, walk downs are carefully prepared trying to do specific question in order to get specific answers.

Once the information is collected, reports are made consisting in an introduction to the component features, component historical behavior, review of the maintenance, inspection and monitoring practice used in plant, and finally the analysis of the Ageing Related Degradation Mechanisms (ARDMs). Conclusions from these reports will allow us to do recommendations in order to reach design life satisfactorily, and to consider the possibility of extending component life.

Condition assessment: analysis methodology is similar to that used for Life Assessments, but in a lower depth. CA’s are mostly performed for active components whose performance can be easily followed by operation parameters, or for those components that were found to be not so critical during screening stage.

Systematic assessment of maintenance: is a detailed analysis of the maintenance strategies that are being used for a system. As a result of the application of a SAM study, the following is expectable:

• Identify the optimum preventive maintenance programme

• Obtain the right balance between preventive and corrective maintenance

• Provide means to monitor maintenance efficiency and effectiveness

CONTRIBUTORS TO DRAFTING AND REVIEW

1st Consultants Meeting

Vienna, Austria: 10-13 January 2005

2nd Consultants Meeting

Mississauga, Canada: 23-25 November 2005

[1] Systematic ageing management process is the application of the Plan-Do-Check-Act cycle to operations, maintenance, and engineering actions aimed at achieving effective ageing management, which is based on the understanding of ageing.

[2] Definition of scope — Typically, the process starts with a complete listing of all the SSCs that constitute the “plant” (such as the subject index (SI) listing, typically consisting of over 1000 items). These are reviewed to eliminate SIs that are not relevant to safety or power production.

[3] Work breakdown — The SIs selected for review are grouped into common subject areas. These common subjects are then divided up into subcomponents corresponding to the disciplines involved and the responsible organizations.

[4] Inspection results are documented to allow assessment of degradation rate for each plausible degradation mechanism. Where inspection results are not available, they are scheduled and focussed to assess the degradation mechanism of concern.

• Once degradation rates are known, a life assessment for the equipment can be made. If the life assessment reveals a life that is less than the target life, programmes are required to address the gap.

[5] Components with * met the criteria for NPLA components but were covered by other programs.

[6] EPRI — Electrical Power Research Institute.

[7] NMAC — Nuclear Maintenance Application Center.

[8] EPRI Preventive Maintenance Basis Database Client/Server, Versions 6.0/1.0, Product # 1009584

[9] AECL- Atomic Energy of Canada Limited.