Category Archives: Nuclear power plant life management processes: Guidelines and practices for heavy water reactors
Effective PLiM of HWR NPPs implies more effective use of plant data trends using instrumentation and monitors that often already exist at the plants. Also the integrity of ageing related plant data is a significant concern in the nuclear industry. With the passage of time, retention, integrity, and accessibility of data histories within the plant culture and organization can become issues. Plants that start early on organizing data systematically will be better able to optimize their maintenance strategies.
Hence, effective PLiM programmes also deal with plant data management issues to ensure high quality data for both current and future ageing/life assessments. For example, the ageing assessments on important fluid retaining SSCs show that the effects of chemistry are among the most important factors affecting degradation. To track these, an advanced system chemistry monitoring and diagnostic system can be used to advantage. For instance, to maximize steam generator tubing life, it is necessary to identify the effects of impurities in the secondary side water on local steam generator crevice chemistry and fouling. On-line access by the operators to current and past chemistry conditions (such as available from such advanced monitoring and diagnostic chemistry systems) enables appropriate responses and facilitates planning of shutdown maintenance actions (such as cleaning of specific areas). This is a significant means to successful management of health and long life of this critical plant component.
As maintenance strategies move to more condition-based decision making, effective use of age-related information at the plant and the timely flow of this information to key decision makers becomes a greater challenge to manage. Each monitored parameter of importance to ageing will be used to determine when to take an appropriate action. Therefore, it becomes important to ensure that the appropriate personnel see the requisite information, and to track their response as follow up action is taken. It also means that personnel will be expected to deal with a more significant number of potential actions, so care must be taken to not simply overload them with information and requests. To facilitate these changes to condition-based decision making advanced maintenance information, monitoring, and control systems can be used to benefit. These types of systems provide an interface for users to access specific health monitor information and an electronic portal to the maintenance review and to the work management system.
The cost and duration of a LSFCR outage can vary substantially. Shorter LSFCR outages can be achieved but with increased up-front development cost and labour premium costs during execution. Typical pre-planning activities for an LSFCR have found that the facilities provided in the existing HWR designs for undertaking LSFCR may be limited and access constrained. These limitations should be addressed up-front, as they could have a major impact on the duration and cost of retubing.
To achieve the optimum balance between lost production during the outage and overall cost, an economic model is typically developed to assess various alternatives. Alternative methods of retubing HWR reactors have been developed to address these issues.
The scope of work related to feeder pipe replacement during the LSFCR activities will depend on the success of the mitigation programme being implemented. A number of options are under consideration to mitigate the flow assisted corrosion of outlet feeder pipes to ensure the design life is achieved, see component section below.
This section covers plant life management in general at Indian PHWRs. Plant life management of PHWRs in India generally follows the practices identical to CANDU reactors as given in Chapter 3. Nuclear Power Corporation of India Ltd (NPCIL) conducts PLiM exercise at its plants as per requirements specified in NPCIL HQI-7005 and Atomic Energy Regulatory Board safety guide SG-O-14. These guidelines cover the requirements that are essentially based on the information available from national and international experiences, IAEA (For example IAEA Technical Reports Series No.338, IAEA-TECDOC-1188, IAEA-TECDOC-1037, IAEA — TECDOC-540). Within the operating license, the Indian Regulatory Body — Atomic Energy Regulatory Board (AERB) grants initial authorization for a specified period which may range from five to nine years and renewal of authorization for further specified periods after assessment of safety.
For renewal of authorization, comprehensive safety review of plants is required considering the cumulative effects of plant ageing and irradiation damage, results of in-service inspection, system modifications, operational feedback & status of performance of safety system etc. AERB Safety Guide on Renewal of Authorization for operation of Nuclear Power Plants SG-O — 12 covers the requirements.
This process of safety review for renewal of authorization is carried out several times periodically during the design life of the NPPs. This comprehensive safety review is termed as periodic safety review (PSR). For this purpose, standard categorization of systems, structures and components SSCs into following four categories has been made. The health of all the systems is reported as part of PSR based on the ageing management applicable to each SSC.
Major critical SSCs Limiting plant life. Critical SSCs Important SSCs Other SSCs
A. I.1.1. SCREENING OF SSCs
The major critical components are those of which integrity & functional capabilities have to be ensured during plant operation & shutdown conditions. These have the highest safety significance. These components are non replaceable and control the plant life. For Examples: Calandria, Endshield components, Calandria tubes, Incore components for reactivity mechanisms, Moderator system piping (inside Calandria Vault).
These components are required for plant operation & shutdown condition. They have high
safety significance as major critical components. Usually they are difficult to replace due to
radiation exposure, long shut down period and high cost. Examples: PHT system piping and equipments, pressure tubes, steam generators, primary coolant pumps, PHT feeders, ECCS system piping and equipments, Shut down cooling, moderator cooling heat exchanges and pumps.
The design and application of a plant life management (PLiM) programme has been a concern for both Comision Nacional de Energia Atomica (CNEA) and Central Nuclear Embalse (CNE).
Recently, the utility (CNE) and the research and development institution (CNEA) have signed an agreement in order to enface together the problem as a whole. Plant operation experience provided by CNE and research resources of CNEA are sinergycally combined to achieve the objectives of this challenging task.
The first step of the project was an intensive training course taken by engineers from the utility and from the research institute. Trained staff from both institutions is leading the pilot project and building the PLiM team up, incorporating and training new personnel.
This activity this organized one in two ways, one to assure the safe operation until the end of the design life and the preparation for the operation for long time, for this we are working in a pilot project.
The time is right to address nuclear power plant life management and ageing management issues in terms of processes and refurbishments for long term operation and license renewal aspects of heavy water reactors (HWRs) because some HWRs are close to the design life. In general, HWR nuclear power plant (NPP) owners would like to keep their NPPs in service as long as they can be operated safely and economically. This involves 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.
This TECDOC deals with organizational and managerial means to implement effective plan life management (PLiM) into existing plant in operating HWR NPPs. This TECDOC discusses the current trend of PLiM observed in NPPs to date and an overview of PLiM programmes and considerations. This includes key objectives of such programs, regulatory considerations, an overall integrated approach, organizational and technology infrastructure considerations, importance of effective plant data management and finally, human issues related to ageing and finally integration of PLiM with economic planning.
Also general approach to HWR PLiM, including the key PLiM processes, life assessment for critical structures and components, conditions assessment of structures and components and obsolescence is mentioned. Technical aspects are described on component specific technology considerations for condition assessment, example of a proactive ageing management programme, and Ontario power generation experiences in appendices. Also country reports from Argentina, Canada, India, the Republic of Korea and Romania are attached in the annex to share practices and experiences to PLiM programme.
This TECDOC is primarily addressed to both the management (decision makers) and technical staff (engineers and scientists) of NPP owners/operators and technical support organizations, and will be also of interest to NPP regulators and designers. It is intended to be a living publication and will be periodically updated and supplemented as new knowledge is gained. The guidance provided is applicable also to future HWR NPPs.
This TECDOC has been prepared by a group of experts from five Member States namely: Argentina, Canada, India, the Republic of Korea and Romania. The work of all contributors to the drafting and final review of this report, identified at the end of this TECDOC, is greatly appreciated.
In particular, the IAEA acknowledges the contributions of, R. Versaci (Argentina), A. Blahoianu and C. Moses (Canada), S. A. Bhardwaj (India), Kyung Soo Lee, Ill Seok Jeong (Republic of Korea) and P. Barbulescu (Romania). Special thanks are due to F. Nuzzo (Canada) and J. Nickerson (Canada), who also chaired the technical meetings. The IAEA officers responsible for this publication were Ki-Sig Kang and J. Cleveland of the Division of Nuclear Power.
Sharing of experience and learning from operating experience (including errors), from R&D programmes and the assessment of performance trends should be used as input to an effective PLiM programme. However, relying solely on a “learning by mistake” strategy is not advisable, because it involves reliance on “unanticipated ageing” and unanticipated ageing is costly. It generally requires more extensive equipment refurbishment and costs replacement power, if it causes a forced maintenance shutdown. Examples of unanticipated ageing are:
• EWS spray header erosion in CANDU 6
• Feeder thinning rates and higher flow assisted corrosion of carbon steel
• Pressure tube Spacer shifting and repositioning
• Pressure tube creep rates higher than anticipated.
• Steam generator (higher than anticipated deposition rates, divider plate accelerated leakage rate)
If unanticipated ageing (learn by mistake attitude) remains unchecked, it has the potential to also adversely affect safety, due to undetected changes in equipment condition and it tends to produce “institutional or safety culture degradation” as well.
The integrity of the reactor building for the additional operational life period is an important PLiM consideration for LTO decisions. Extensive analysis and studies of CANDU6 concrete containment structures have already been completed, including an IAEA TECDOC that covers some CANDU containment considerations [I.4]. Typically, a comprehensive PLiM Life Assessment (LA) specific to the individual plant is completed and then a detailed Ageing Management Plan (AMP) is prepared. These are then factored into the in-service inspection and maintenance to ensure plant life attainment. These plans are updated periodically as part of the overall plant life management programme.
The plausible degradation mechanisms for the containment structure have been identified; the most important being minor concrete cracking and a slight increase in permeability of containment, and changes in construction joints and cold joints. The main ageing mechanisms causing the degradation were freeze/thaw cycles, concrete shrinkage and creep and the repeated containment leak rate test. In addition the Alkali Aggregate Reaction specific to one plant and the chloride penetration at another contributed to the degradation. Corrosion of reinforced steel or corrosion and loss of pre-stressing force in pre-stressed containments where ever used are in general to be assessed for PLiM and LTO.
For LTO a detailed ageing management plan has been developed to better understand the impact of degradation mechanisms on the long term performance of the containment structure. The work includes a thorough review of site documentation, of world experience of the various ageing degradation mechanisms that could affect containment performance with time, and those applicable to the plants under consideration. Current knowledge is supplemented by an enhanced inspection and monitoring programme. At one plant, this includes design, data acquisition and installation of a system of specialized instrumentation to obtain detailed information about the behaviour of the reactor building prior to, during and after the containment building is pressurized for an in-service leak-rate test. Indian PHWRS recently constructed have provided instrumentation to monitor pre-stressing and health of cables. Typically, the Reactor Building leak-rate test is performed once every 3 to 5 years.
In developing ageing management strategies for CANDU 6 concrete containment buildings, concrete ageing experience gained at other facilities is being used. For instance, at the Gentilly site in Quebec, there are two reactor buildings. In addition to the Gentilly-2 plant, owned by Hydro Quebec, there is an earlier prototype CANDU system including a containment building (Gentilly-1) that is owned by AECL. The reactor at Gentilly-1 has been decommissioned and most of the radioactive materials removed to other sites.
The containment structure at Gentilly-1 is now over 30 years old. Importantly, there are many design and construction similarities to CANDU 6 reactor buildings. Observations made on the effects of ageing of the Gentilly-1 structure have been used to guide ageing assessments on other CANDU concrete containment structures. In addition, because there are no concerns for adverse effects arising from an accidental event, Gentilly-1 has been used to safely test and develop technologies that are appropriate for the assessment of ageing effects on structures at operating plants. For instance, conventional techniques including coring and testing the concrete and visual observations on the condition of the structure were supplemented with special methods developed to measure the state of stress in the structure. The techniques for these latter measurements were developed in connection with its fuel-waste management studies.
Subsequently, the technologies developed at Gentilly-1 have been applied at other HWR NPPs. For instance, at one HWR NPP, new instrumentation was applied to the containment structure alongside older mechanisms and a PC-based data acquisition system was installed, so that instruments could be monitored simultaneously and in real time. These instruments were placed to measure the stresses and deformations that occurred in the structure during the pressurized leak-rate test. Equally important, the measurements provided information on the effects of the environment on the structure. Figure 11 shows the mechanisms affecting the long performance of containment concrete building.
Acids and Bases
Elevated Temp./Thermal Cycling
Mild Prestressing Liner Steel Systems Reinforcing
Corrosion Elevated Temp. Irradiation Fatigue
Loss of Prestressing Force
xxxx xxxxx X xxxx
Fig. 11. Mechanisms affecting the long term performance of containment concrete buildings.
During 1989, failure of inlet manifolds in both the units MAPS was noticed. The failure was to an extent that the zircaloy inlet manifold got ruptured and its broken pieces were found in the moderator circuit. On internal inspection of calandria, extensive damage to the manifolds was discovered. Under these conditions, it was not possible to operate the reactors. Hence, the reactors were shut down. Short term immediate/intermediate measure were carried out to bring the units to power again. As part of permanent rehabilitation these units are now fitted with three numbers of perforated tubes, called spargers, installed after removal of calandria tubes and pressure tubes in three lattice locations at the bottom of Calandria as an alternate moderator inlet path providing full moderator flow in required flow pattern.
• Take moderator outlet from the original calandria outlet, while maintaining design flows through calandria sprays and other auxiliary cooling circuits.
Fig. A. I.2. Moderator Inlets and Outlets 94 After Permanent Rehabilitation
Drilling of pipe penetrations through fuelling machine vault floors for routing the piping.
All the above activities required extensive theoretical and experimental work before its actual implementation in the plant. As the en masse coolant channel replacement (EMCCR) for MAPS units was due, it was decided to carry out this job during the period when the units will have to be shut down for longer time. The installation is successfully completed in MAPS-2 which is now operating at 100% FP.
In many countries, the safety performance of the NPPs is periodically followed and characterized via the periodic safety review (PSR) approach . The regulatory review and acceptance of the PSR gives the licensee the permission to operate the plant for up to the end of the next PSR cycle (usually 10 years). The regulatory system does not limit the number of PSR cycles, even if the new cycle is going beyond the original design lifetime of the plant. The only condition is to demonstrate the safety of the plant operation for the next PSR cycle while maintaining safety and operational margins.
The PSR is a tool that may be used by regulators for the identification and resolution of safety issues in NPPs. In this framework, continued operation may be strived for by applying the results of the PSR, by identification and resolution of the safety issues as a condition of operation for the new PSR cycle. The PSR is not an adequate tool to control changes and tendencies with an evolution time shorter than 10 years. It is also not a suitable system in case the licensee needs a technological guarantee for a long term operation longer than 10 years; in many cases economical considerations suggest an extension of 20 years, or more, of the original design life. Figure 4 shows the flowchart of an overall process for periodic safety review of NPP.
However, it must be noted here that the concept of PSR was developed to be part of the normal regulatory or safety monitoring process, and not specifically to justify beyond design life operation of a plant. The PSR was originally used primarily to assess the safety status of the plants designed to early standards. In these cases, the PSR gives an overall review of all aspects of plant operation that may be relevant to safety. This review includes subjects such as emergency arrangements, organization and administration, procedures, research findings and feedback of experience. All of them are mainly relevant to current operation, and not directly related to the justification for continued operation.
A PSR implemented beyond the original NPP’s design life may require a deeper safety review, addressing the following:
• Evaluation of the plant safety against current standards;
• A new evaluation and/or qualification for items affected by time-dependent phenomena;
• The AMP, which has to be extended over the extended operating life; and
• A new safety assessment, to show that the as-designed conservatism (not the safety margin) may be reduced, based on improved plant operation practices and better understanding of the degradation mechanisms. The overall safety margin must be kept consistent with current safety requirements.
In conclusion, a full scope PSR applied with a view beyond design life operation is fully not different in principle than a usual PSR applied during the design life at ten-yearly intervals, but the emphasis has to be oriented to the ageing of SSCs limiting the total plant operational life and always on the related safety issues. Table 2 shows the list of PSR implementation in Member States.
Table 2. List of MS of PSR implementation
Design life for the plant and major components is targeted to be a certain value at the design stage (say 30 Years). Service life, during which components operate safely and reliably, may exceed the design life. This is because the actual operation in terms of say fatigue cycles, corrosion etc., for most of the components may be considerably lower than design assumptions. Thus selected components may be replaced at end of design life (say 30 years) to provide a longer service life for the plant as a whole. Therefore, unless there are critical components that cannot be physically replaced or refurbished at their end of life, the optimum life for the plant may be based on economics rather than technical issues.
The most expensive component that requires replacement in order to extend plant life is the pressure tubes. Pressure tube replacement or large scale fuel channel replacement (LSFCR) is required at about 30 years, due to ageing degradation similar to what was described in section 3. The second most expensive component to replace for LTO is the steam generators. The current generation steam generators using improved materials (alloy 800) and operational practices is to obtain high life and may not require replacement along with pressure tubes. The CANDU/ PHWR plants have all implemented steam generator life cycle management plans to inspect, monitor and mitigate steam generator degradation to achieve the design life. However, older plants may need to time Steam Generator replacement along with pressure tube replacement for LTO, so that replacement of steam generators and pressure tubes could be accomplished in the same outage.
Bulk feeder replacement is another significant activity that may be required for life extension, depending on the effectiveness of current and planned activities to mitigate feeder wall thinning due to Flow accelerated corrosion (FAC). Replacing feeders during the LSFCR may actually reduce the duration of the LSFCR by improving access to the fuel channels.
The long shut down period at LSFCR stage could also be used to take up PLiM exercise for LTO of other major SSCs like Containment, electrical systems and control and instrumentation. The PliM considerations and status with regard to such major component specific technology in CANDU/PHWRs is covered in this Chapter.
In case of India PHWRs, the following components were designated to be managed by Ageing management programme (AMP).
• Pressure tube Calandria tube & core internals
• Thermal shields, end shields
• Hanger rods
• Containment structure & Calandria vaults
• PHT headers & feeders
• Steam generator
• PHT pump body
• Bleed condenser / pressuriser / accumulators
• Shut off rod, adjuster rod & drive mechanism
• Instrumentation & control cables and connectors
• Diesel generators, motor generator sets, ACVRS
• Feed & Bleed lines & fm supply pump lines including dump & control valves, dousing system