Inspection programmes

In the present pressure tube inspection programme, the following parameters are being monitored:

• Irradiation induced dimensional changes, which include axial elongation.

• PT wall thickness and PT-CT gap at 6’O clock and presence of PT-CT contact if any.

• Location of GS.

• Service induced flaws.

• Deuterium concentration profile.

A number of tools and techniques have been developed for carrying out the aforementioned inspections. The diameter measurement and the sag measurement have not been part of the regular inspection programme. However, tools have been designed for carrying out such measurements for limited use in dry channel. The description of these tools and techniques are given in the subsequent paragraphs.

A tool for dry channel visual inspection called DRYVIS (dry channel visual inspection system) has been developed and is based on a pneumatically operated tube walker, which can be remotely made to crawl in the desired direction within a tubular component. The device can carry any transducer for carrying out inspection. In DRYVIS, the system comprises tube walker, radiation resistance video camera, a separate illumination head and grating for sizing of indications. This system has been used to carry out visual inspection within drained and defuelled pressure tubes, as well as calandria tubes.

The present report references a variety of other IAEA publications that have been developed in the IAEA project on Safety Aspects of NPP Ageing. Some of these deal exclusively with HWR components; others include HWR components in a report that also includes light water reactor (LWR) components. For background, the project and its main products, including the previously issued programmatic guidelines, are summarized in this section.

The IAEA initiated information exchange on safety aspects of NPP ageing in 1985 to increase awareness of the emerging safety issue relating to physical ageing of plant SSCs. In 1989 a systematic project was begun aimed at assisting Member States in understanding ageing of SSCs important to safety and in effective ageing management of these SSCs in order to ensure their integrity and functional capability throughout their service life. This project integrated information on the evaluation and management of safety aspects of NPP ageing generated by Member States into a common knowledge base, derives guidance, and assists Member States in the application of this guidance. Main publications of the project [1-14, 20, 21] fall into five groups. Figure 2 shows the stature of IAEA publications related with ageing management and PLiM.

Awareness

Following the first International Conference on Safety Aspects of Ageing and Maintenance of Nuclear Power Plants [1] which was organized by the IAEA in 1987, increased awareness of physical ageing of SSCs and its potential safety impact was achieved by the development and wide dissemination in 1990 of IAEA-TECDOC-540 on Safety Aspect of Nuclear Power Plant Ageing [2]. In the 1980s most people believed that classical maintenance programmes were adequate for dealing with the ageing of NPPs. However, in the 1990s the need for ageing and life management of NPPs became widely recognized.

Programmatic guidelines

The following programmatic guidance reports have been developed using experience of Member States.

• Data Collection and Record Keeping for the Management of Nuclear Power Plant Ageing [3] provides information on the baseline, operating and maintenance data needed and a system for data collection and record keeping.

• Methodology for the Management of Ageing of Nuclear Power Plant Components Important to Safety [4] gives guidance on screening SSCs to make effective use of limited resources and on performing ageing management studies to identify or develop effective ageing management actions for the selected components.

• Implementation and Review of Nuclear Power Plant Ageing Management Programmes [5] provides information on the systematic ageing management process and an organizational model for its implementation.

• Equipment Qualification in Operational Nuclear Power Plants [6] provides the current methods and practices relating to upgrading and preserving equipment qualification in operational NPPs and reviewing the effectiveness of plant equipment qualification programmes.

Component specific guidelines

Based on current experience, practices in Member States and methodology for the management of ageing of NPP components important to safety [4], guidelines to assess and manage ageing of major NPP components important to safety have been developed through technical meetings and coordinated research projects (CRPs). Main objectives for each specific component guideline are:

• To identify significant ageing mechanisms,

• To document the current practices for the assessment and management of ageing,

• To assist Member States in establishing a systematic ageing management programme.

The comprehensive technical guidelines have been issued with common contents as shown below [11-21].

• Component description

• Component design basis

• Potential ageing mechanisms

• Inspection, monitoring and maintenance requirements, techniques and practices

• Methods for the assessment of degradation

• Methods for the mitigation of degradation

• Systematic component specific ageing management programme (AMP)

Ageing Management Review Guidelines [15] is a reference report for Ageing Management Assessment Teams (AMAT) and for utility self-assessments; these reviews can be programmatic or problem oriented. The focus of the project work has progressively shifted from developing awareness, to preparing programmatic, and then component specific guidelines. In
future, the focus will be on providing services to assist Member States in the application of the guidelines. A reduced effort will be maintained to facilitate information exchange through the preparation of additional guidelines and updating of existing guidelines.

Safety Guide on Ageing Management is being developed for NPP and research reactor. This SG (DS-382) will provide guidelines for operators on how to improve the existing AMPs and implement the effective AMPs for new NPPs. Regulators will use it to set up requirements and verify NPP’s compliance with requirements.

I""dS 382 I

under

developing

I

J

Fig. 2. Structure of IAEA publications related with PLiM and ageing management.

Once progress has been made on the ageing assessments, it is important to implement the disposition of recommendations into plant programmes. Effective plant practices in monitoring, surveillance, maintenance, and operations are the primary means of managing ageing. From general experience with HWR PLiM programmes, it can usually be expected that the ageing assessment programme will lead to modifications and enhancements, but not necessarily replacements, of the pre-existing plant ageing programmes. However, a successful PLiM programme is measured by the plant performance indicators. This requires a structured and managed approach to the implementation process. The overall objective is to optimize plant programmes for ageing management, both for the remaining design life period and for the plant long term operation.

First, the recommendations for changes to current plant programmes that result from the ageing assessments are systematically dispositioned. Following this, PLiM implementation strategy may be developed on Table 4. In this step, plant staff considers the implementation options (implementation procedures, cost-benefit analysis, schedule & outage plans, affected departments, task responsibilities, work priority, etc). PLiM changes are then input to outage and operational planning process.

Detailed life assessments cover both the pressure retaining components (e. g. casing, case cover, closure bolting, and stuffing box gland) and also the rotating element of the pump. The shaft and other components of the rotating element of HTS pumps are critical sub-components that must meet all the operational conditions for design life. Fatigue is the degradation mechanism of concern, particularly given that it is very difficult to get any degradation indicator prior to shaft failure.

To perform fatigue life assessment of operating pump shafts, a number of techniques have been used, specifically tailored to reflect actual plant conditions (as distinct from the more conservative design-basis assumptions). The methodology provides a more accurate estimate of shaft fatigue life for the actual plant-operating environment.

The fatigue life of the shaft is very dependent upon the radius provided at each of the shaft notches or grooves, which are the primary sources of stress concentration. The notch radii are not always specified on the supplier’s drawing or often not verified after manufacturing, even if specified. Hence it is important that the actual notch radii be known for accurate determination of the fatigue life of the pump shaft. To do this, replicas are taken of actual groove profiles on a plant pump shaft (usually a spare one). Then the actual shaft notch radii are determined from the replicas in a metrology laboratory. The stress concentration factors based on these actual groove radii are then used in subsequent fatigue evaluation.

Fatigue life assessment for operating mechanical loadings further involves a number of other special techniques to accurately represent actual plant conditions. Bending stresses in the shaft are a particularly important loading. Specialized analysis techniques have been developed to represent the pump and motor bearing stiffness in the structural model and to enable use of actual operational vibration data on the pump/motor set in the shaft bending loading assessment.

Thermal fatigue life assessment of pump internals is a relatively complicated and expensive undertaking. However, a detailed thermal and stress analysis approach has also been developed to assess thermal conditions associated with cold injection flows during normal and abnormal operation.

While HTS pump shafts and other important internals are non-pressure boundary components, the above techniques have been used in fatigue assessment of these components for the original plant life. These methodologies have proven to reflect the excellent plant performance to date of HWR HTS pumps and are useful approaches to assess fatigue life extension capability of these important sub-components.

An in depth review of existing fire alarm was carried out after the fire incident at Narora Atomic Power Station. This review had revealed that the coverage of the system is not adequate and does not meet the current standard. Hence it was replaced by a state of art addressable type of system, supplied by an Indian firm. The number of detectors was also increased to cover all the areas adequately and also different type of detectors like optical etc were also used.

As part of this up gradation the following additional features were also provided.

• Digital type linear heat sensor cable for fire detection of cable trays of cable bridge

• Aspirating type smoke detectors for fire detection in moderator room which is a high radiation area.

• Beam detector for boiler room for wide area coverage in high temperature areas

• Flame detectors for fire detection in oil tanks.

Different samples of cables were collected in RAPS and were subjected to ageing test to estimate their residual life. For most of the cables residual life was found to be about 8 to 10 years.

Collecting samples of trunk cables is found impracticable due unavailability of adequate length of samples, without disturbing the cable terminations. Hence it is recommended that wherever feasible extra length shall be provided. Alternatively dummy samples may be kept at appropriate locations for testing purpose.

Emergency core cooling system

Emergency core cooling system, existing in RAPS and MAPS was a low-pressure injection system, using moderator pumps. As part of safety up gradation it was decided to retrofit high pressure injection system, in line with the current practices followed.

The existing instrumentation and controls was using pressure switches for detecting the LOCA condition and pneumatic DP transmitters for sensing direction of injection. As part of up gradation the entire instrumentation was modified, using electronic pressure transmitters for detecting LOCA condition and DP transmitters for sensing direction of injection. A test facility was also provided for online testing of all the motorized valves of the system along with their logic.

The CNSC has not issued explicit regulatory requirements on ageing management. However, a number of age-related regulatory requirements are included in several regulatory publications, including:

• Class I Nuclear Facilities Regulations (requiring licensees to describe “the proposed measures, policies, methods and procedures for operating and maintaining the nuclear facility”);

• Requirements for Containment Systems for CANDU Nuclear Power Plants (R-7);

• Requirements for Shutdown Systems for CANDU Nuclear Power Plants (R-8);

• Requirements for Emergency Core Cooling Systems for CANDU Nuclear Power Plants (requiring that safety systems are available to operate when called upon, R-9);

• Reliability Programmes for Nuclear Power Plants (requiring development of system availability limits and minimum functional requirements, and description of the inspection, monitoring, and testing activities designed to ensure system availability, Regulatory standard S-98); and

• Other specific conditions of an NPP operating license.

In order to address ageing, Canadian CANDU NPP utilities are required to inspect and perform material surveillance according to the technical requirements of CSA standards:

• Periodic inspection of CANDU nuclear power plant components (N285.4),

• Periodic inspection of CANDU nuclear power plant containment components, (N285.5),

• In-service examination and testing requirements for concrete containment structures for CANDU nuclear power plants (N287.7), and

• Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU nuclear power plants (N285.8).

These requirements include inspection techniques, procedures, frequency of inspection, and evaluation of inspection results, disposition, and repair. Maintenance programmes are required for the purpose of limiting the risks related to the failure or unavailability of any significant SSC (See Table 3).

In addition to the above, CNSC adopted industry fitness-for-service guidelines as regulatory means to address ageing management of the special components such as SG tubes and feeders.

Ageing management of the pressure tubes requires a strategy that effectively addresses all ageing mechanisms so that the core remains fit-for-service. The two key aspects of the strategy are: 1) appropriate inspections involving measurement of axial elongation, radial expansion, and sag of the tubes as well as volumetric inspection for flaws and monitoring of the deuterium concentration; and 2), material surveillance to confirm the acceptability of the fracture toughness and delayed hydride cracking characteristics of the material as it ages.

Dimensional changes

During reactor operation, the conditions of temperature, stress and neutron flux change the dimensions of the pressure tubes. Irradiation-induced and thermally induced deformation of fuel channel components will, in the absence of other mechanisms, eventually establish fuel channel life. The following inter-related dimensional changes occur in pressure tubes during normal reactor operation:

• Axial elongation

• Diametral expansion

• Wall thinning

• Sag

Pressure tube axial elongation due to irradiation can require remedial action, and, in the extreme, become a tube life limiting factor if the bearing length provided by the design is not sufficient to accommodate the projected axial elongation for the design life. The difference in axial elongation rates between neighbouring channels is also monitored to ensure that interference between feeders or problems with fuelling machine access does not occur.

Elongation of all fuel channels can either be measured using the fuelling machine or measured periodically during planned outages using specialized gauging tools. These inspections provide information on the elongation rate of each individual channel as well as providing data to determine the variability in the creep and growth properties of the tubes. Current understanding of irradiation induced axial elongation indicates that elongation rates may be slightly non-linear. Therefore continued frequent monitoring is required to determine when the channels will go off bearing and to identify which channels are affected and will require remedial action. If the channels are predicted to go off bearing before the design life is reached, the following actions can be implemented:

• Shifting the channels to recover any available bearing travel on the current fixed end

• Defuelling a small number of channels

• Replacing a small number of channels

• Demonstrating that off bearing operation is acceptable

This section addresses PLiM at Pickering, Darlington and Bruce Power (prior to decontrol in 1999). Descriptions of Bruce Power PLiM programmes post decontrol will be added at a later time. The table below lists the in-service dates of these multi-unit CANDU plants.

Over the past ten to fifteen years, all of the Canadian CANDU owner/operators and AECL have been involved in Plant Life Management (PLiM) programmes and activities. The PLiM programmes have used various terminology (such as Life Cycle Management and Integrated Age Management) and there have been some variations in detailed activities, as reflected by various ageing issues and utility practices. Overall the effort has been primarily focused to ensure that CANDU plants will operate successfully and reliably through their design life and to preserve the option to extend plant life. Comprehensive CANDU Plant Life Management (PLIM) programmes have been developed from the operations knowledge gained from the Pickering A, and Bruce A stations, and from the four original CANDU 6 plants. In addition relevant information from the CANDU industry research and development programmes, and other national and international sources have been used.

Processes to systematically identify and evaluate the critical systems, structures, and components (CSSC’s) in Canadian HWRs have been implemented. Ageing assessments involving evaluation of degradation mechanisms that could affect fitness for service for their planned life, have been performed for many CSSCs. Plans have been developed to ensure that plant surveillance, inspection and maintenance programmes monitor and mitigate component degradation important for plant life attainment. There is also effort to incorporate ongoing improvement in equipment reliability programmes such suggested in the INPO AP-913 guideline on Equipment Reliability.

As CANDU plants in Canada continue to age, new challenges are being identified and fed — back into improvements in both the Ageing Assessments themselves and to the important link between PLiM programmes and business planning. PLiM is giving new capability and insight into O&M challenges, that is being used to help improve current plant performance. For instance, some CANDU owner/operators are considering extending duration between outages and PLiM considerations have been a significant element in evaluating the changes. The extension of PLiM techniques to facilitate planning of major plant modifications is becoming more important particularly now that several utilities are embarking on major life extension projects, via refurbishment programmes.

The following is a brief summary of recent PLiM activities in Canada.

3.1. SCREENING SSCS

Those SSCs that are to be included in a PLiM programme are usually identified by a systematic screening process, which prioritizes the systems based on their importance to achievement of plant goals, such as nuclear safety, environmental safety, and production reliability. In addition, structures and components whose failure would result in a major replacement cost or in a significant loss of production capability are also typically considered.

The first steps in screening the SSCs are to customize the generic PLiM screening methodology for use at a specific plant. Standard criteria are safety, production, environmental impacts, worker safety and cost. A risk-based approach can be applied to develop the plant specific criteria, developing quantitative weighting measures, and then applying the plant specific screening process to the plant SCC list, to rank their importance. The resulting list of prioritised SSCs (sometimes known as the CSSCs — the Critical Systems, Structures and Components) will allow them to be included in the PLiM programme. Also the implied “residual” risk of not including other SSCs will be identified. The rigour of the ageing assessment, in terms of resources required and depth of evaluation will be determined by the SSC priority.

In situations where parts of a PLiM programme are already underway without undertaking the up-front SSC screening, a systematic procedure for SSC screening should be used to verify that there are no gaps in identification of the remaining SSC assessments to be performed. The SSC prioritisation also assists utilities in identifying less critical SSCs. Figure 8 shows a typical level of assessment versus SSC criticality as determined by screening.

For the most critical components and structures that provide mainly passive functions and are subject to long term ageing degradation mechanisms, life assessments are performed. Some of these key steps in the life assessment process and interfaces between the various groups are shown diagrammatically in Fig. 9.

Typical CSSC life assessment methodology is based upon IAEA methodology, as detailed in [4]. A particularly important part of the process is to understand and assess the importance of all the ageing degradation mechanisms that can impact on the functions of the SSC. Another important aspect is to tailor the generic methodology, including the diagnostic and assessment methods and techniques, to the specific technology and characteristics of the SCC under consideration. Some typical examples and component specific considerations are given in the Appendices.

Many components in the early HWR plants had a very good service record with little or no significant degradation history to date. However, this excellent in-service experience (and hence lack of degradation data) provides a unique challenge for the CSSC life assessments within the PLiM programme. In performing the systematic and detailed assessments, a key activity is diagnosis of the operational history for ageing indicators, as well as a thorough understanding of applicable degradation behaviour.

With little degradation data from the plant, the challenge is to provide a reasonably comprehensive and detailed assessment of ageing effects for the next 20 to 30 years of operation. To meet this challenge, a thorough understanding of the applicable degradation mechanisms and the associated —“stressors”— is used. This understanding derives from research and development programmes, integrated with knowledge from relevant field data of other plants.

An in-depth understanding of the plant operational history and the current plant programmes related to ageing are both key inputs to the life assessment process. Involvement of utility staff in the process is encouraged. Developing an efficient and effective team selected among key utility staff (such as ageing management experts, system engineers, component engineers, reliability engineers and maintenance personnel) contributes significantly to success and value of the PLiM assessment programme.

Fig. 9. Life assessment process and interfaces.