A root cause of premature ageing degradation can be traced often to a lack of communication, documentation, and coordination between design, operation, and maintenance organizations. This results from the general situation that, except for major SSCs (such as fuel channels), there is not explicit responsibility on one plant person to achieve a specific SSC lifetime.

The effectiveness of SSC ageing management can be significantly improved and its service life significantly extended by the coordination of all relevant programmes and activities. This coordination generally requires a modest engineering investment in order to understand and take into account the design assumptions/ limitations, significant ageing mechanisms and the impact of operation and maintenance activities on these mechanisms and the rate of SSC degradation.

After assessing remaining life of a component, recommendations are to be developed for more detailed life evaluations and in-depth analyses of fuel channel for identified degradation areas, then the recommendations to be performed as a part of Phase II PLiM programme. The followings can be issued for the recommendations of life management of fuel channels: stress analysis for nip-up condition and feeder coupling load, bearing chamfer operation for channel shift work, preparation for large scale fuel channel replacement, revision of in-service inspection programme, and etc.

A. II.5. CONCLUSIONS

The methodology for ageing assessment of CANDU6 plant was introduced by showing the experience of Wolsong-1 plant lifetime management Phase I project. The life evaluation of major critical systems, structures, and components (CSSCs), and embedded commodities has been performed. Those CSSCs were selected by screening and prioritized methodology developed by KEPRI. General approach of ageing assessment, relation of PLiM and periodic safety review (PSR), detail screening method used were explained with an actual experience in technical evaluations.

1.1.2. Availability of qualified NPP personnel

Availability of qualified NPP personnel is an important issue for PLiM; this includes validated staff selection methods and organizational issues. Even if NPPs are technologically at a high level they still require to be operated in a manner that conforms to safety prescriptions, technical specifications (TS), and good practices, which includes avoidance of transients, for example. However, due to the slowdown in new NPP construction, particularly in the West, over the last 20 years, there has been little incentive for young engineers to embark on a career in the nuclear power sector. Universities and other seats of learning have thus stopped or drastically reduced nuclear technology courses as a response to the falling demand.

Extensive analysis and studies of HWR steam generators have already been completed, including an IAEA TECDOC that covers CANDU steam generators [I.2]. Typically, a comprehensive PLiM Life Assessment or a Life Cycle Management plan specific to the individual plant’s steam generators is completed and factored into the in-service inspection and maintenance to ensure plant life attainment. These plans are updated periodically as part of the plant life management programme for this component.

A detailed and comprehensive life assessment of the steam generating equipment will include the pressure boundary, the external support structure, the tubing, and all the key internal sub­components. Tubing is a key sub-component. For CANDU-6 NPPs (and also for Indian PHWRs built after Madulas power station (MAPS)), the SGs tube are with Alloy 800M (M means “modified”) and have experienced relatively little SG tube corrosion to date. For instance, at the Wolsong NPP Unit 1 plant (that has 21 years of in-service experience), there are only 9 SG tubes plugged, none as a consequence of corrosion, 7 of these before in-service operation, out of the total population of over 14,000 tubes. For other CANDU-6 SGs and Indian PHWRs the situation is similar. Elsewhere, the record with Alloy 800 SG tubing is similar after more than 25 years’ in-service experience.

This excellent service record requires a rather novel approach to predicting future performance, such as the potential for tubing corrosion degradation.

The assessments involve a very thorough review of tubing corrosion mechanisms that can occur in nuclear steam generators. The knowledge from R&D studies of SG tubing corrosion behaviour in various chemistry environments has been a key element of this methodology.

• First, a detailed assessment is made of worldwide experience with Alloy 800M, and other steam generator tubing alloys. From this review, all the specific types of corrosion mechanisms, and the chemistry environments that have been instrumental in causing tubing corrosion degradation, are systematically identified and the key stressors assessed.

• Second, each of these tubing corrosion situations is evaluated for relevance to the particular plant’s steam generator design and operation. The tolerance of the Alloy 800 tubing to the presence of plausible aggressive secondary side impurities (such as lead, sulphides & chlorides) in various plausible ranges of chemistry conditions that might exist in steam generator crevices (such as in the tubesheet sludge pile or in tube-to — support gaps that have become blocked with deposits), is assessed. (note: it seems obvious that experts would perform this assessment).

• Additionally, other degradation mechanisms are reviewed for their impact on SG condition, operation, and future life. These include mechanisms related to thermal — hydraulics such as vibration and fretting, particularly with respect to the potential for fretting of the tubing against the U-bend support structures. Fouling, both of the primary and secondary sides of the SG can also significantly reduce operating efficiency, and the efficiency of the station output.

While SG tubing degradation is considered the largest single potential source of SG problems, this alone is not the only important factor in determining the prognosis for achieving 50-year life. Steam generator non-tube components also present a challenge in estimating current condition as well as future life. Hence, the HWR steam generator life assessment also considers the other components in a SG that could compromise life. As there are a very large number of individual components in a nuclear SG, these are grouped into the following categories:

• Primary side pressure boundary

• Secondary side pressure boundary

• External supports

• Primary side internals

• Secondary side internals

As with the tubing, detailed consideration is given to all potential degradation mechanisms, from world wide and other HWR experience. Next the potential for plausible degradation is assessed, given the HWR NPP’s design and operation.

Based on the detailed plant SG studies performed to date, the overall SG condition at several CANDU 6 plants appears to be good with no obvious compromise to attaining the design life. However, there is sufficient uncertainty over the condition of SG secondary side internals that the life extension assessment requires additional inspection and analysis. The conclusions and recommendations are focused on chemistry control, proactive inspections/monitoring programme, and periodic cleaning on both the primary and secondary sides of the steam generators.

Despite limited information on the condition of the secondary side internals, the overall current condition of steam generators is sufficiently good to attain design life subject to a continued programme of inspections, cleaning and chemistry control. Similarly, the prognosis for life extension is also good, provided a proactive age management strategy is adopted and implemented.

For these components preventive maintenance, ISI, and condition monitoring/assessment are possible to mitigate Ageing and are replaced in a planned basis during operating phase. For examples: end shield cooling system equipments, Calandria vault cooling system components, PHT feed pumps, turbine generator system, process water systems piping and equipments, feedwater system piping and equipments, heat exchangers, diesel generators, UPSs, batteries.

A. I.1.1.4. Category 4 — Other SSCs

These are safety related support systems managed by planned preventive maintenance, ISI & conditioning monitoring/assessment and are routinely replaceable. For examples: air compressors & instrument air systems, heavy water recovery dryers, main exhaust fans, transformers, power and control cables.

There are 13 units of PHWRs in operation (refer to Table in Chapter 1 in country report) and another 5 under construction at present. The design details of plants including material specifications and quality control techniques used have seen improvements from plant to plant. Also the local environment control, operating practices and chemistry control have been upgrading from time to time. NPCIL has developed plant specific life management programme(s) to effectively monitor the condition of SSCs and take corrective action in time to maintaining safety margins. To effectively manage the ageing of SSCs the plant needs to have a programme that provides timely detection and mitigation of ageing degradation in order to ensure that the required integrity and functional capability of SSCs are maintained through out the service life and for long term operation.

This activity has as purpose to arrive at the end of the design life with a safe operation. The SECs that are not in the pilot plan, they are includes because at the end of the design life they will be replaced, they are the pressure tubes, calandria tubes, feeders. Secondary side piping is not included also. For these components we must assume future degradation behaviour, for this reason an especial inspection and maintenance to be implemented to achieve the design life.

A. V.2.1.1. Pressure tubes

• Deuterium pick-up in the body of the pressure tubes: we carry out scrape sampling programme to provide deuterium data so that the station-specific deuterium ingress model can be updated. It will also be used to give indications of changes in the deuterium uptake rate.

1.1. BACKGROUND

The design life of a nuclear power plant (NPP) does not necessarily equate with the physical or technological end-of-life (EOL) in terms of its ability to fulfill safety and electricity production requirements. Operating equipment, generically called systems, structures and components (SSCs) in a NPP is subjected to a variety of chemical, mechanical and physical conditions during operation. Such stressors lead to changes with time in the SSC material properties, which are caused and driven by the effects of corrosion, varying loads, flow conditions, temperature and neutron irradiation, for example.

Even allowing for significant ageing effects in SSCs, it is quite feasible that many NPPs will be able to operate for times in excess of their nominal design lives, provided appropriate and proven ageing management measures are implemented in a timely manner. This aspect has been recognized by operators and regulators alike, as seen in the number of license renewal applications and approvals, respectively, in the USA, and, elsewhere, by extending licensing procedures, primarily based on periodic evaluation of safety, i. e. periodic safety reviews (PSR).

In general, heavy water reactor (HWR) NPP owners would like to keep their NPPs in service as long as they can be operated safely and economically. Their decisions depend upon the business model. They involve the consideration of a number of factors, such as the material condition of the plant, comparison with current safety standards, the socio-political climate and asset management/ business planning considerations.

Historically, most NPP owners (including owners of HWRs) felt that their routine maintenance, surveillance and inspection programmes would be adequate in dealing with the ageing processes that would occur at their plants. However, starting in the early 1990s and following, it has become widely recognized that a more systematic and comprehensive approach generally known as integrated plant life management (PLiM) or life cycle management (LCM) is needed.

Fig. 1 shows the number of HWR reactors by age. 18 reactors have been operated more than 20 years and 5 reactors have been operated more than 30 years.

Fig. 1. Number of HWR reactors by age as of October 2005.

Plant ageing if uncontrolled may increase the probability of failures, possibly leading to accidents. Examples of safety implications of unchecked ageing are hydrogen pickup during normal operating conditions (NOC), the formation of hydride blisters on the pressure tube

leading to pressure tube rupture, PT sag leading to PT/CT contact, flow assisted corrosion (FAC) in piping and outlet feeders, leading to leaks and breaks.

There are failures that are not covered in reliability studies. These are failures that typically go undetected and therefore unpredicted by system testing such as in the case of the MOV torque switch settings. As friction increases settings are increased and may be increased to the point where the motor seizes or other system failures occur. Another similar instance is instrument loop reliability and the effects of the instrument error on performance and safety. Signal drifts, as the instrument ages, may lead to out of specification signals and failures, for example neutronic instrument dynamics and response.

Safety system settings can also be affected by ageing due to changes in process conditions (trip margins). Similarly, during an accident, safety system settings can be affected by the ageing state of the equipment and by changes due to the progression of the accident itself such as the case of the pressure tube (PT) blisters, stress corrosion cracking (SCC) of boiler tube and subsequent leaks, cable deterioration due to radiation exposure, material properties (affecting their ability to withstand failures), PT diametral creep creating flow redistribution in the core, bundle bypass, etc. whereby the critical channel power (CCP) effect is initially positive and then goes strongly negative.

The main components which constitute reactor structure include: Calandria vessel, end shields, Calandria supports, end-shield ring, dump tank (RAPS, MAPS, Pickering A), ring thermal shield and ion chamber mountings and if applicable inaccessible piping like that of moderator.

Extensive analysis and studies of HWR reactor structures have already been completed, including an IAEA TECDOC that specifically covers CANDU reactor assemblies [1.5]. Typically, a comprehensive PLiM life assessment or a life cycle management plan specific to the individual plant’s reactor structure is completed and factored into the in-service inspection and maintenance to ensure plant life attainment. These plans are updated periodically as part of the plant life management programme for this component.

No known degradation mechanisms have been identified (Exeption is RAPS 1 End Shield crack at location suspected to have high residual stress due to a local repair in the carbon steel calandria side tube sheet. The design and material of construction has been changed in all units from MAPS2 for PHWRs in India) that will limit the life of the critical (non-replaceable) calandria and end shield assembly to less than 60 years. No problems specific to CANDU 6 operating units requiring repairs or replacements have been identified. Problems at the older CANDU units in Ontario are not likely to occur at CANDU 6 or at current PHWR design employed in Indian units because of design changes incorporated.

There is no significant concerns are seen for life attainment. Additional inspections/assessments may be required for long term operation.

CANDU Nuclear Plant Life Assurance Programme (NPLA) reports were prepared for reactor structures at the Bruce and Pickering sites in the early 1990s, that identified plausible ARDMs and some areas of uncertainty, which required further investigation (e. g. Pickering A shell — shield supports).

I.4. OTHERS

Results of ageing studies in RAPS and MAPS indicated that the power MG set being rotating machines equipment had become more maintenance intensive. In case of Power MG set of Madras Atomic Power Station, problems were also experienced with respect to following: —

• Overloading of the MG set during 2-moderator pumps operation on class-II bus.

• Poor quality of rewound machines

• Poor quality of backup 250 VDC batteries resulting in low voltage DC input to DC motor of M. G. set

• Problems in MG set control circuit resulting from drifting in control cards.

• Spurious operation of over speed switch.

The above had caused (i) reduced availability of class-II supply (ii) need for frequent rewinding of machine and (iii) increased burden on maintenance staff.

In order to take care of above, the rotating power MG sets have been replaced by static UPS system. This system is rated at 600 KVA as against 400 KVA rating of MG set. This enhancement in KVA rating is needed to take care of (i) Overloading problem under two moderator operation (ii) to enable this UPS to soft start big moderator pump motors (135 KW) without any trip problem and (iii) To clear the fuses up to 100 A in the circuits feeding downstream loads.