Handbook of Ageing Management for Nuclear Power Plants

03 April 2014

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FOREWORD

One of IAEA’s statutory objectives is to “seek to accelerate and expand the contribution of atomic energy to the development of peace, health and prosperity throughout the world”. One way this objective is achieved is through the publication of the IAEA document series that include the IAEA Safety Standards and the IAEA Nuclear Energy Series (NES). The IAEA Safety Standards establish “standards of safety for protection of health, and minimization of danger to life and property” as stated in Statute Article III, A.6. The safety standards consist of Safety Fundamentals, Safety Requirements, and Safety Guides. They represent the international reference for nuclear safety for the benefit of the Member States, their regulatory bodies and other national authorities. They are written in a regulatory style and are binding on the IAEA for its own operation. Beside the safety standards, the IAEA produces a Nuclear Energy Series of documents designed to encourage and assist Member States in the development and application of nuclear energy for peaceful purposes. The NE series contains the best technologies and practices available as well as future advancements. They are written as guidelines, reports, handbooks and their content is exclusively based on input from leading subject matter experts for the benefit of Member States, government officials, nuclear power plant owners and operators, support organizations, academia and others. Within the nuclear energy series, one prominent collection is the engineering documentation dealing with plant ageing management, with degradation mechanisms, failure prevention and mitigation programmes. They contain elements of material science and techniques as applied to ageing components, systems and structures. The same collection includes documentation on Plant Life Management (PLiM) that considers ageing management in the context of the NPP entire lifecycle even long term operation beyond the originally assumed operating period taking into consideration management objectives including economic aspects and the market in which the plant operates. PLiM techniques allow at the same time the optimization of monitoring, surveillance and maintenance programmes. It can achieve all this by relying on advanced deterministic and probabilistic techniques, the use of an integrated ageing knowledge base, backed by R&D and operation feedback. Beyond the plant ageing documentation, the agency also organizes international conferences as well as working groups, coordinated research projects, technical and consultative meetings on structures, systems and component ageing in nuclear power plants as well as plant life management. Out of these initiatives the division of nuclear power of the IAEA nuclear energy department has recently prepared a generic model on Plant Life Management for Long Term Operation. Building on this vision, this handbook on ageing management in nuclear power plants is intended as an easily accessible compendium of the ageing management knowledge base, including the insights from modern Plant Life Management programmes, as they are being implemented by an increasing number of IAEA Member States. The work contributed by the subject matter experts involved in the drafting and in the review of this handbook, is greatly appreciated. They are acknowledged at the end of this document. The IAEA officer responsible for this publication is K.S. Kang of the Division of Nuclear Power.

CONTENTS 1. 

INTRODUCTION .........................................................................................................................6  1.1.  1.2.  1.3.  1.4.  1.5. 

2. 

OVERVIEW OF AGEING MANAGEMENT ............................................................................11  2.1.  2.2. 

3. 

Basic Concepts .............................................................................................................11  General features of ageing programmes .......................................................................12  2.2.1.  Data collection and record keeping..................................................................12  2.2.2.  Continuing equipment reliability improvement programmes ..........................13  2.2.3.  Relation between ageing management and maintenance programmes of safety critical components ...............................................................................14  2.2.4.  Ageing Management Assessment and SCreening of SSCs .............................16  2.2.5.  Components Life evaluation ............................................................................18 

PROACTIVE AGEING MANAGEMENT IN OPERATING NPPS ..........................................24  3.1.  3.2.  3.3.  3.4.  3.5.  3.6. 

4. 

Background .....................................................................................................................6  Terminology and definitions...........................................................................................6  Objective.........................................................................................................................8  Scope ..............................................................................................................................8  Document Structure (to be updated after all sections are complete) ..............................9 

Ageing Considerations in design ..................................................................................25  Ageing Consideration during fabrications & construction. ..........................................27  Ageing considerations during commissioning ..............................................................27  Ageing Management in Operation................................................................................28  3.4.1.  Refurbishments, Modernizations and Power Uprating ....................................33  3.4.2.  Long Term Operation: .....................................................................................34  Ageing Management Pre-requisites for Decommissioning [More input required] ......37  Effective ageing management programmes..................................................................37 

AGEING MECHANISMS ...........................................................................................................40  4.1.  4.2. 

Radiation DAMAGE ...................................................................................................43  Fatigue ..........................................................................................................................46  4.2.1. Mechanical fatigue..............................................................................................46  4.2.2. Thermal fatigue...................................................................................................47  4.2.3. Vibration fatigue .................................................................................................48  4.2.4. Fatigue by geometrical discontinuities ...............................................................48  1) Intrinsic characteristics of weldolets and socket weld connections prone to fatigue ..............................................................................................................48  2) Inadequate piping supports inducing fatigue in the piping or tubing .......................48  4.2.5. Fatigue evaluation in SSC ..................................................................................49  4.3.  General corrosion .........................................................................................................54  4.4.  Stress corrosion cracking (SCC) ...................................................................................56  4.4.1.  Transgranular Stress Corrosion Cracking ........................................................58  4.4.2.  Primary water stress corrosion cracking (PWSCC) .........................................59  4.4.3.  Irradiation Assisted Stress Corrosion Cracking (IASCC)................................60  4.4.4. 61  4.5.  Flow assisted corrosion ................................................................................................61  4.6. THERMAL AGEING ...........................................................................................................63  5. 

STRUCTURES, SYSTEMS AND COMPONENTS ..................................................................66  5.1.  5.2. 

Degradation of steel containment .................................................................................66  5.1.1.  Ageing Mitigation Methods .............................................................................68  RPV Ageing Degradation in PWRs ..............................................................................69 

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5.2.1.  RPV Beltline region irradiation embrittlement ................................................70  5.2.2.  Under clad cracking: ........................................................................................70  5.2.3.  Primary water stress corrosion cracking (PWSCC) in RPVs ..........................71  5.2.4.  Bottom Mounted Instrumentation (BMI) penetration leaks ............................72  5.2.5.  Pressurizer Relief Line cracks .........................................................................72  5.2.5. Thermal Ageing: .................................................................................................73  5.2.6.  Fatigue .............................................................................................................73  5.2.7.  Corrosion .........................................................................................................74  5.3. PRESSURIZER IN PWRS ...................................................................................................74  5.4. Ageing monitoring and mitigation methods in PWRs ..........................................................75  5.4.1. Mitigation of Radiation Embrittlement ..............................................................75  5.4.2. Mitigation of CRDM penetrations SSC ..............................................................76  5.4.3. Vessel Head repairs and replacement: ................................................................76  5.4.4. Inspection and monitoring: .................................................................................77  5.4.5. Surveillance Programmes ...................................................................................78  5.5. RPV, Internals and Piping Ageing Degradation in BWRs....................................................78  5.5.1 Radiation Embrittlement in RPV and Overlay Clad ........................................79  5.5.2 IASCC, IGSCC, TGSCC and ODSCC in Core Shroud, Piping and Spent Fuel Pool ................................................................................................79  5.5.3 SCC in Dissimilar Metal Weld ........................................................................80  5.5.4 Environmental Fatigue in Nozzle and Piping ..................................................80  5.5.5 Flow Assisted Corrosion in Carbon Steel Piping ............................................80  5.5.6 SCC and Fatigue in Pump and Valve ..............................................................81  5.5.7 Thermal Ageing Embrittlement of Duplex Stainless Steels ............................81  5.6. Ageing degradation in CANDU type reactors ......................................................................81  5.6.1. CANDU Reactor assemblies: Calandria and End shields ..................................81  5.6.2. Pressure Tubes – Ageing deformations ..............................................................82  5.6.3. Pressure tube geometry monitoring: ...................................................................83  5.6.4. Flow-Accelerated Corrosion (FAC) in feeders ...................................................84  5.7. Steam Generators in PWRs and PHWR plants ....................................................................85  5.7.1. Steam Generator tubing materials and their degradation mechanisms ...............85  5.7.2. Degradation mechanisms in Steam Generators ..................................................86  5.8. Reactor coolant and interconnected system piping ...............................................................89  5.8.1. Thermal Stratification .........................................................................................89  5.8.2. Thermal striping .................................................................................................90  5.8.3. Thermal fatigue management .............................................................................90  5.8.4. Fouling ................................................................................................................90  5.8.5. Leak-Before-Break (LBB) ..................................................................................91  5.8.6. Water-hammer and steam-hammer.....................................................................91  5.8.7. High-energy pressure breakdown orifices ..........................................................92  5.9. Emergency feedwater system................................................................................................95  5.10. Buried Piping ......................................................................................................................95  5.11. Secondary Side piping in PWRs - Flow-Accelerated Corrosion ........................................96  5.12. Cables and I&C systems .....................................................................................................98  5.12.1. Cable testing .....................................................................................................99  5.12.2. Cable acceptance criteria ................................................................................106  5.13. Concrete and non-metallic AMP(s) ..................................................................................106  5.13.1. Containment Designs ......................................................................................106  5.13.2. Ageing management for concrete structures ..................................................106  5.13.3. Ageing mechanisms for concrete structures ...................................................107  5.13.4. Inspection and monitoring of concrete and metallic structures ......................108  5.14. Metallic Structures ............................................................................................................109  6. 

ASSESSMENT FOR AGEING MANAGEMENT ...................................................................114  6.1. 

Inspection and assessment methods............................................................................114 

6.2. 

7. 

REGULATORY FRAMEWORK..............................................................................................135  7.1.  7.2.  7.3.  7.4. 

8. 

Current Practice ..........................................................................................................143  Important Issues ..........................................................................................................145  Recommendations ......................................................................................................145 

INNOVATION TECHNIQUES AND R&D .............................................................................146  9.1.  9.2.  9.3.  9.4.  9.5.  9.6. 

10. 

Various national approaches .......................................................................................135  Time limited Ageing analysis .....................................................................................137  International common approach .................................................................................141  Codes and standards ...................................................................................................142 

ORGANIZATIONAL STRUCTURE ........................................................................................143  8.1.  8.2.  8.3. 

9. 

6.1.1.  Non-destructive methods ...............................................................................114  6.1.2.  Destructive methods ......................................................................................116  Monitoring ..................................................................................................................127  6.2.1. REACTOR PRESSURE VESSEL NDE/ISI ....................................................133  6.2.2. STEAM GENERATOR NDE/ISI ....................................................................133  6.2.3. REACTOR PRESSURE VESSEL HEAD NDE/ISI ........................................133 

R&D and mitigation technologies ..............................................................................146  Japan ...........................................................................................................................148  CHINA .......................................................................................................................149  Germany .....................................................................................................................152  International collaboration on Ageing Management ( REPEAT)...............................154  International activities on Ageing Management .........................................................155  9.6.1.  AM initiatives of IAEA’s activities ..............................................................155  9.6.2.  AM related initiatives of other international organizations ...........................157 

SUMMARY ...............................................................................................................................158 

REFERENCES .....................................................................................................................................160  ABBREVIATIONS ..............................................................................................................................169 

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1. INTRODUCTION 1.1. BACKGROUND This Ageing management handbook has been developed in compliance with the requirements of the IAEA safety standards and in the IAEA safety guide on ageing management for nuclear power plants [1] [2]. As stated in the IAEA safety standards the primary safety goal of an ageing management programme should be to ensure the availability of all required safety functions throughout the service life of a nuclear power plant (NPP), taking into account the changes that occur with time and use [2-3]. Consequently ageing Management should be planned primarily with the safety principles in mind, not only during the operating phase of a plant, but from the very first basic conceptual design as stated in requirement 31 of IAEA Safety Standard on Design and continuing into the operating period as stated in requirement 14 of the IAEA Safety Standard on commissioning and operation [3]. Existing basis of ageing management has been re-examined after Fukushima accident [ref..IAEA NISA] It has been concluded that lessons-learned from Fukushima accident showed that there is no impact on existing ageing management standards or procedures. Since political debates often deflect technical basis it becomes increasingly important to develop a basic handbook on ageing management. The IAEA produced a comprehensive series of supporting documents on ageing to capture the accumulated knowledge, including R&D findings, engineering support and operating experience on the diverse aspects of ageing management and plant life management. [4-38] This Handbook has been written with the intent of providing young engineers and graduate students with both underlying principles and practical knowledge for Ageing Management and Plant Life Management for nuclear power plants. It is also a compendium and an update of the ageing information on structure, systems and components (SSC’s) collected by the Agency over years. The information is presented as much as possible in a concise manner by relying on flowcharts with descriptive explanation on underlying principles in order that the handbook may be also used as a “what, why and how to” manual and as a reference book for classrooms. In adition, on-going research and development on innovative technology and procedures are also complied by experts of participatig IAEA member states in close collaboration with a dedicated organization in the area, known as the International Forum on Reactor Aging Management (IFRAM). The Handbook will be further edited into a web-friendly format so that IFRAM can utilize the product for its members’ benefit and cooperation. 1.2. TERMINOLOGY AND DEFINITIONS Ageing is the continuous time dependent degradation of SSC-materials during normal service, extending to standard power production and transient conditions. Postulated accident and postaccident conditions are excluded. The concept of ageing is also extended to the spheres of component obsolescence, to the state of update of the NPP documentation and of staff training. Ageing Management Programme (AMP) is a set of policies, processes, procedures, arrangements, and activities for managing the aging of structures, systems and components (SSCs) for a nuclear power plant (NPP) Beyond Design Basis Accident (BDBA) comprises accident conditions more severe than a design basis accident that may or may not involve core degradation. They are not fully considered in the design process. However “beyond design-basis" accident sequences are analysed to the extent needed to fully understand their consequences and the robustness of the design Condition assessment (CA) is defined as an ageing assessment methodology applied to systems, as well as components and structures, or groups of components with similar characteristics (commodities).

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Control Rod Control System (CRCS) is a system capable of providing dynamic control of core power by actuating rod movements in and out of the ore. The system is distributed across several control groups. Design Basis Accident (DBA) is defined as accident conditions against which a facility is designed to withstand, according to established design criteria, and for which the damage to the fuel and the release of radioactive material are kept within authorized limits. Finite Element Analysis (FEA) is a computational method used in engineering for finding an approximate solution to a problem involving a complex geometrical figure by using a simplified model solvable by a finite number of numerical operations called discretization which is in turn characterized by a finite number of parameters N, called degrees of freedom Institute of Electrical and Electronics Engineers (IEEE) is a professional association based in New York City that is dedicated to advancing technological innovation and excellence in the electrical and electronic fields Integrated Life Cycle Management (ILCM) is a project integration management methodology utilized to enhance the viability and sustainability of localized technologies, namely the localization programme of a certain level of nuclear power generation technology International Generic Ageing Lessons Learned Report (IGALL) is the international version managed by the International Atomic Energy Agency of compilations of lessons learned on ageing which builds on the GALL (Generic Aging Lessons Learned) report in the U.S. (see NUREG-1801, Vols. 1 and 2). Irradiation Assisted Stress-Corrosion Cracking (IASCC) : Certain materials such as Austenitic steels for examples are known to degrade while in service as core components in a nuclear reactor. Among the many factors characterizing susceptibility of these materials to IASCC is hardening, and at the microstructural level, the dislocation of the material microstructure and changes in the grain boundary composition. Large Early Release Frequency (LERF) is defined as the frequency of those accidents leading to significant, unmitigated releases from containment with the potential for “early” health effects. Such accidents are generally the consequence of containment failure, containment bypass events and loss of containment isolation Life assessment (LA) is defined as an ageing assessment methodology applied to critical and/or complex components and structures that involve mainly passive functions and typically are not expected to be replaced within the original design life of the plant. Long Term Operation (LTO) of a nuclear power plant may be defined as operation beyond an established time frame set forth by, for example, licence term, design, standards, licence and/or regulations, which has been justified by safety assessment, with consideration given to life limiting processes and features of systems, structures and components (SSCs). Maintenance, Surveillance, In-Service Inspection (MSI) is a programme in NPPs designed to ensure that the levels of reliability and availability of all plant structures, systems and components continue to meet the assumptions and the intent of the design through the programme duration and the plant life. Non-Destructive Examination (NDE) is generally understood to mean using non-destructive testing methods to inspect and characterize materials and structures that are not destroyed by the testing medium or the form of energy used. On-Line Monitoring (OLM) is the application of instrumentation and observation techniques in order to evaluate equipment performance by assessing its consistency with other plant or baseline indications. Industry experience at several plants and R&D have shown this overall approach to be

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very effective in identifying equipment that exhibits degrading or inconsistent performance characteristics Periodic Safety Review (PSR) is a comprehensive safety review of all important aspects of safety, carried out at regular intervals, typically every ten years. In addition, a PSR may be used in support of the decision making process for licence renewal or long term operation, or for restart of a nuclear power plant following a prolonged shutdown Plant Life Management (PLiM) is the integration of ageing and economic planning to maintain a high level of safety, optimize the operation, maintenance and service life of SSCs, maintain an acceptable level of performance, maximize return on investment over the service life of the NPP; and provide NPP utilities/owners with the optimum pre-conditions for long term operation (LTO). Probability of Detection (POD) of degraded portions of cables or of buried piping in nuclear power plants is a technique used to help optimize inspection and repair programmes. Curves of detection probability versus range of stressors and stressor fluctuations can help predict location of high degradation probability. Safe Long Term Operation (SALTO) is an IAEA peer review service of LTO feasibility studies offered to IAEA member states who request it.

1.3. OBJECTIVE The Ageing Management Programmes for structures, systems and components (SSCs) in Nuclear Power Plants are aimed at preserving the licensing basis envelope, to adequately manage safety margins, to successfully control performance degradation, to improve system availability and capacity factors. Given the large interests at stake, the nuclear industry has established owners groups to share their operating experience and resources, line up the best tools and R&D available to help them manage SSC ageing. This handbook is intended as a general reference document on NPP ageing, covering topics such as degradation mechanisms, ageing management techniques. It should be used as a basic reference book for new comer countries as they set up their ageing management programmes, as a learning tool for new staff, as a common base at an international level to facilitate information exchange and understanding of ageing management issues. The IAEA recognizes that plants in the future will be operated for longer periods of time and that advanced ageing management techniques will become even more important if operators are to improve and maintain the safety and availability of their nuclear power plants, even beyond the originally design life of their facilities. This handbook intends to also help plan for long term operation as early as possible in the life cycle a nuclear power plant. 1.4. SCOPE This handbook builds on the IAEA engineering documentation and safety guides on ageing management and incorporates the knowledge base acquired by member states on large individual components. It also shows how to integrate this knowledge with plant life management insights for a better ageing management of SSCs which are never taken in isolation but with appropriate consideration given to the associated and interfacing systems to improve their functional availability and ultimately the entire unit capacity factor. Beyond large passive components, the scope of this manual extends also to active components and sub-components as well as to classes of smaller devices and groups of similar constituents. Topics are treated from condition-based assessments and reliability-centred maintenance techniques, of online monitoring data and processing techniques to databases with statistical data tracking capability on

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ageing related issues. At the same time the information is presented in a very practical easy-to-use consultative way, typical of handbooks. End Users IAEA’s handbook on ageing management of NPPs is designed for the training and support of nuclear power plant staff, maintenance managers, vendors, research organizations, and regulators to assist them in their work related to SSC ageing in operating NPPs. 1.5. DOCUMENT STRUCTURE (TO BE UPDATED AFTER ALL SECTIONS ARE COMPLETE) The Handbook is derived from both the successful experiences and the failures in aging management practices for systems, structures and component (SSCs) in PWRs BWRs and PHWR plants with particular emphasis on the proactive aging management and plant life management principles. For the better understanding of majority of readers who are young engineers and graduate students with college education in science and engineering, this Handbook is structure that is amenable to becoming a textbook as well as a handbook . Chapter 1 is an introductory chapter. It contains sections on the background the circumstances under which the handbook is being published, one on the scope and objectives and one on the target audience. Section 1.3 contains a list of definitions used in the areas of AM, PLiM, MSI.It provides an overview of the basic concepts of ageing management, an analysis of the impact of ageing on plant safety and performance, common weaknesses of traditional ageing management programmes and operation feedback, international forums, task committees, R&D and other activities on ageing management including the IAEA action plan and the member state initiatives on exploring the implications and drawing lessons learned, configuration changes and implementation plans derived from the fallout of the Fukushima accident. Section 2 concludes with the evolution of ageing management into integrated Plant Life Management programmes. Chapter 2 provides upfront information on the proactive approach to ageing Management which includes a proactive strategy for AM and the development and optimization of AM activities for critical structures, systems and components in a nuclear power plant. Chapter 3 is intended to be a reference manual on material degradation and degradation mechanisms. It includes definitions, examples, R&D and the underlying science, the knowledge base and the state of the art discoveries particularly in the areas of monitoring and detection technologies applied to phenomena such as radiation embrittlement, fatigue, stress corrosion cracking, general corrosion, and specifically erosion-corrosion and flow accelerated corrosion. More phenomena applicable more specifically to the SSCs of BWRs, PWRs and CANDU type reactors are also discussed. In chapter 4 the ageing phenomena and the degradation mechanisms related to critical components, such as the RPV its vessel head, walls and internals in PWRs and BWRs are presented in a concise and systematic way. Beyond the phenomena and their causes ageing management guidelines are given for the implementation of ageing programmes. In parallel for PHWRs ageing phenomena typical of natural uranium fuelled reactors affecting pressure tubes, fuel channels, feeders within the reactor face region are summarized together with the best recommendations from the extensive and lengthy R&D conducted over the years in support of the reactors life cycles. Beyond the core components, ageing phenomena affecting the reactor coolant system piping are discussed, notably thermal stratification and thermal fatigue, thermal striping and thermal fatigue management programmes, erosion-corrosion, FAC, fouling, leak before break (LBB), mechanical and acoustic vibration damage, water and steam hammer, degradation of supports, locking of snubbers, pressure breakdown orifice, pump and valve cavitation, vibration, leakage. Other major components discussed in this chapter are the steam generators, power and low voltage cables, I&C components, concrete ageing and ageing programmes for non-metallic components. Chapter 5 describes degradation experiences and mitigation options made available through extensive international cooperation and conferences such as IAEA’s Plant Life Management (PLiM) Conference. Dissemination and identification of both phenomena and remedial measures allow for industrial decision-making actions on maintenance intervention and mitigation activities and their prioritization on the basis of risk and SSC criticality. Prioritization can also be established by means of single point of vulnerability (SPV) analyses. Depending on their categorization SSCs will be assigned to a particular aging assessment programme (time based assessment, condition assessment or life

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assessment). As a result of these assessments, a critical spare part list can be prepared, spare parts inventories can be defined in support of repair/replace (R/R) strategies. It should be noted that the obsolescence in either design and configuration is also dealt with in Chapter 5 as a part of ageing assessment programme for Systems, Structures and Components(SSC’s). Chapter 6 deals with the assessment of ageing and ageing management in operating plants using inspection and evaluation tools and the processes and regulations developed to assess the effectiveness of AM programmes with regards to safety, safety margins, reliability and performance management. Chapter 5 deals also with topics such as the ageing assessments done for Long Term Operation using engineering tools such as computational analysis to revalidate margins and Time Limited Ageing Analysis (TLAA), predictive tools for ageing and remaining life prognosis, condition monitoring and engineering tools to proactively develop preventive maintenance programmes aimed at mitigating ageing degradation, EQ programmes, pressure boundary programmes, the use of probabilistic techniques and cost-benefit assessments to help design and optimize maintenance, surveillance and in-service inspection. Chapter 5 also provides examples of component specific programmes for single unit, multi-unit and fleet management in terms of lessons learned and recommendations. Chapter 7 deals with the regulatory framework concerning ageing and ageing management in the various jurisdictions and the common approach and best practice by the international community and the IAEA. Legal criteria and procedures for the safety assurance of long-term operation are both extensive and complicated in real world where different countries have different approach and priorities in achieving the same safety goals. Therefore readers are recommended to understanding materials described in earlier chapters of this Handbook in order to better rationalize the vast diversities in technical criteria, as described herein. Chapter 8 treats of organizational structures necessary to manage ageing effectively, to implement guidance documents and methodologies, to establish data bases, periodic updates, reviews and reporting systems and to develop effective corrective action plans. Chapter 9 discusses innovative techniques in ageing management and includes a survey of international experience. Demography of world nuclear power plants is such that the Plant Life Management is relatively new field with rapidly intensifying demands for improvement in both technology and approach. New international cooperations are burgeoning under various programs including IAEA’s PLiM, OECD-NEA’s SCAP, International Forum on Reactor Ageing Management (IFRAM) as well as European Commission’s LONGLIFE and related programs. Chapter 9 introduces as much of these activities as possible in order to promote early dissemination of new body of information and to facilitate cooperation. Chapter 10 contains a summary of the main elements of modern NPP ageing management programmes. Conclusions and recommendations are also provided. In the Annex, country reports can be found where the various national approaches and experiences related by the participating experts and authors of this handbook are summarized for the benefit of all users.

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2. OVERVIEW OF AGEING MANAGEMENT

2.1. BASIC CONCEPTS Primary objective of an ageing management programme is to maintain the reliability and availability of required safety functions throughout the service life of the NPP, in accordance with the licensing basis and of preserving SSC performance and the plant performance targets as a whole. [1] This requires that the ageing management programme addresses:    

Physical ageing Degradation of the ssc performance characteristics, Obsolescence and spare part availability Progressive slipping out of date with current standards, regulations, technology and knowledge.

An ageing management programme during the operating phase of a nuclear power plant should require an active involvement in monitoring, inspection and surveillance activities to follow and compare SSC degradation against predictions specifically of mainly large and passive components that cannot be replaced and that, if not checked, may limit the useful service life of a plant. When ageing processes are known, they can be monitored and mitigated through an appropriate ageing management programme (AMP). New or unexpected degradations, if remained undetected, on the other hand, can lead to accidents. In terms of documentation, an aging management program (AMP) is a set of policies, procedures, instructions, activities and planning designed to successfully manage SSC aging in a NPP. Ideally ageing management should be present at each stage of the life cycle beginning with first design, and on through construction, commissioning, operation (including long term operation), design changes, refurbishments, extended shutdowns and decommissioning. Maintenance activities are usually separate from ageing control activities, although the programmes may be integrated. Maintenance is routinely applied mainly to active and replaceable components. It includes run to failure constituents and can help better manage even bulk materials and part inventories. Degradation and ageing are terms used to describe SSC deteriorations. They should not be used interchangeably. There is a distinction.  

Degradation is a gradual deterioration of one or more characteristics of a SSC that could at one point impair its ability to function within acceptance limits. Ageing is a general deterioration process in which the characteristics of an SSC gradually change with time and use.

Ageing management is a key element of the safe and reliable operation of nuclear power plants and as such it is the object of particular attention of regulators and industry auditors. The cumulative effects of ageing and obsolescence on the safety of NPPs are re-evaluated periodically by license holder and reviewed by regulators using for example the periodic safety review (PSR) or equivalent processes of safety evaluation. The PSR method is typically used in Member States with unlimited or continuing licenses. For example European countries apply the PSR process as their main safety assessment method against the effects of ageing and obsolescence. A second approach is based on a limited term license coupled with a license renewal process. The USA use the License Renewal Application (LRA) concept that implies the continuing regulation of the current licensing basis (CLB) through the promulgation of updated licensing requirements as appropriate and the mandatory implementation of regulatory action items. This process implies CLB compliance checks, a thorough review of the maintenance programmes, and requirements on the

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monitoring of safety performance parameters and requires updates to the Final Safety Analysis Report (FSAR) and supporting analysis. A third approach is a combination of the former two as happens in Korea, where PSR is used as the basic safety assessment method every 10 years, complemented with updates to safety requirements when requested on the international scene and judged appropriate. The PSR method calls for ageing management and critical component life assessments as pre-requisites to validate the operating license of nuclear power units. Finally, Plant Life Management is another term that should be clarified. PLiM is a decision maker support programme that uses systematic review and screening techniques and a set of advanced tools including integrated R&D knowledge bases, operation feedback data, elements of probabilistic, reliability and economic analysis and other engineering tools to help manage the whole plant life cycle. PLiM analysts study both passive and active structures, systems and components. PLiM can help optimize ageing management, maintenance, inspection and surveillance programmes and effectively support management decisions regarding long time operation, large component replacements, modernization and refurbishment plans. 2.2. GENERAL FEATURES OF AGEING PROGRAMMES 2.2.1.

Data collection and record keeping

Without a good data collection and a good record keeping system the management of nuclear power plant ageing can only be reactive and is eventually bound to fail. In order to prevent gaps and weaknesses in maintenance programmes downstream, the data required and if not available to be recreated, is essentially of three types: baseline data, operating history data and maintenance history data. The baseline data should include:      

Component data sheets with descriptive and functional information, its interfaces, its boundary conditions, its relationship to the system and the unit in which it operates. Operating and performance requirements including its design service conditions and any other operational limits. Initial, un-degraded material conditions such as its wall thickness profile especially in the case of critical pressure boundary components subject to wall thinning Installation records including specific manufacturing defects and exceptions, any welding specs used, any deviations, the component orientation, as-built conditions, supports, and any specific modifications from the original design due to construction deviations or field changes. Functional capabilities as tested during commissioning, Component calibration, if applicable

The operating history should include: Actual service conditions experienced including the relevant evolution of process parameters such as,     

Water chemistry, Environmental conditions inside and out, The actual number and profile of the transients endured (such as pressure-temperature transients for pressure retaining components) Data on the component availability Testing results

The maintenance history records should include: 

Condition monitoring of critical components including data collection, testing, inspection and surveillance activities aimed at understanding the degree, speed and modality of material and functional degradation

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  2.2.2.

Repair or modification history (history dockets) Overall maintenance cost history (including cost in terms of influence on critical path and system down time) Continuing equipment reliability improvement programmes

A continuing SSC reliability improvement programme is key to high plant capacity factors and long term plant performance. The main constituents of such a programme are the implementation of a continuing operation feedback assessment, a corrective action optimization process and a technology watch programme to study and implement new technologies and new diagnostic tools. Reliability improvement relies on failure prevention, which relies on the capability to:   

Measure the symptoms of incipient failures, Analyse operations feedback, failures and event at other plants identify premonitory signs

Reliability improvement is possible if we acquire the capability to map the probability of failure locations and this is possible if we know the vulnerability to degradation mechanisms of all our SSCs, if we effectively use inspection, surveillance, testing and monitoring equipment as for example:     

Installed instrumentation assigned to multiple use (e.g. Process parameter measurement and degradation monitoring) Ad-hoc instrumentation such as vibration detectors Oil sampling, Thermography, Electro-magnetic signature of motorized pumps and valves

Figure **** shows a process used to screen the equipment and apply monitoring where it can be effective in improving equipment reliability. The process begins with a review of the ageing monitoring techniques applied, its limitation and its compatibility and accuracy with the ageing observed in the field and with the ageing programme in the plant. If gaps are found, the ageing management programme needs to be complemented or further developed. Otherwise, the degradation mechanism observed simply needs to be monitored and a method to diagnose its growth needs to be implemented. Finally the ageing management programme for this specific mechanism needs to be documented and input into the design basis and into the periodic safety review documentation.

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Start Clarification of related technique standard/ limitation and evaluation of compatibility

Consider assessment result of degradation for object equipment

Analysis of aging management program applying to nuclear power plant

Is it effective aging management program?

NO

Complement and development of aging management

YES

Consider monitoring and diagnose method for aging mechanism Input the aging management DB and PSR documentation

End Figure :**** Program Monitoring and Managing the Degradation and Aging Process [ref: PSR report – Korea] 2.2.3. Relation between ageing management and maintenance programmes of safety critical components If plant life management programme is not well established systematically, ageing and maintenance activities on safety critical components typically display the following issues:  

Duplication of effort between ageing management and maintenance activities Conflicts in planning activities and re-work

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

 

 

             

Deficiencies in the capacity to identify incipient failures and age related degradation (e.g. routine walkdown and on-line monitoring strategies). This deficiency may be due to the inadequacy of the administrative processes to monitor system and component performance such as missing periodic walkdowns and lack of walkdown procedures to implement proper visual observations for the detection of non-conformances. Lack of an effective inspection programme and of inspection procedures. For example inadequate procedures at the craft level to report the “as found” and the “as-repaired” condition of the component being worked Lack of a user-friendly condition feedback system which should include a condition grading code to characterize the equipment state and a component or system condition evaluation instruction to regulate the evaluation, the selection of prevention or mitigation actions to be taken by the system engineer or equivalent functions and the approval process. Errors in condition evaluation: Was there a failure to recognize what existing barriers could have prevented the failure (procedure completeness, craft training, post-maintenance testing, tag-out restoration, troubleshooting, unavailability management, and human performance) to prevent reoccurrence? Was due consideration given to the risk/benefit of the addition or change. Were the downstream effects of the failure and its extent correctly evaluated? Was the susceptibility of other components to the same failure mechanism correctly determined? Did the feedback reach the reliability improvement process (if available)? Is the frequency of the monitoring and other control activities correct can they prevent reoccurrence? Is the scope of the preventive maintenance tasks comprehensive enough? Is the possibility of obsolescence been addressed? Was the root cause analysis, if any, correctly conducted? Is staff training and qualification adequate? Inadequate performance monitoring: For critical components, performance monitoring is usually required by the regulators. Performance trending should be conducted to ensure that corrective action is taken before the component or the system under scrutiny fails to meet its performance limit requirement. The lack of an adequate equipment history and corrective action database containing performance limits will not allow the systematic detection of failure trends of similar components across the plant and in other plants. The corrective action data base should preferably have the capability to set up specific alert values for condition-monitoring data and send the alerts to the appropriate action centres. Lack of a long-term maintenance strategy to regulate for example how to proactively maintain components for as long as it is economically feasible Lack of systematic condition assessment of safety critical components and use of diagnostic techniques to prevent failures Lack of a systematic inaccessible equipment life assessment programmes Lack of metrics on the effectiveness of ageing management programme, its activities and the overall strategy: the most efficient way to measure the effectiveness of an ageing management programme is to set up a good universal work order database system capable of providing a historic record by component and by system of all activities and activity types (preventive, testing, design change, reactive with the type of failure, etc.) of the start and completion dates, of labour and material costs. This data, particularly failure data can then be sorted by component and component groups or types, by year, by cost and failure trends can be obtained from which a reliability assessment can be made and the entire ageing management programme can be evaluated.

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2.2.4.

Ageing Management Assessment and Screening of SSCs

Whether an operator has implemented PLiM techniques and probabilistic tools or not in support of its ageing management programme, still the first step remains that of the screening the relevant SSCs to determine those that are going to be enrolled in the NPP Ageing Management programme. It is neither practicable nor necessary to evaluate and quantify the extent of ageing degradation in every individual SSC. Structures include both simple and complex structures consist of structural elements. A systematic approach should therefore be applied to focus resources on those SSCs that can have a negative impact on the safe operation of the plant and that are susceptible to ageing degradation. This should include SSCs that do not have safety functions but whose failure could prevent other SSCs from performing their intended safety functions. A safety based approach, such as the one outlined in the following, should be applied to the screening of SSCs for review of the management of ageing:

Fig. XXX Flow chart of SSC screening for a NPP Ageing Management Programme Key elements of a SSC screening exercise is that of determining whether the ageing mechanisms and their effects on the SSC safety, reliability and performance are all well-known and understood. Post-service examination and testing of structures or components (including destructive testing) may substantially improve the understanding. An assessment of the component examination and of the test results should be undertaken to establish a characterization of the environment, to obtain an estimate of the stressors on the materials. The assessment should include a description of the ageing mechanisms, the type of degradation and the mapping of the degradation sites, a theoretical or empirical model (if tests have been conducted). Results can be quantified to predict future development of the degradation. From a list of all systems and structures, those that are important to safety should be identified, on the basis of whether or not a component malfunction or failure could lead to the loss or impairment of a safety function. An outline of this screening process is illustrated in Fig. XXX. Also the structural elements and components of SSCs that are important to safety and whose failure could lead to the loss or impairment of a safety function should be identified and

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grouped by type for the sake of resource usage optimization. The term ‘structural element’ is used only for this screening process for better understanding; in subsequent paragraphs structural elements are again referred to as ‘structures’ If probabilistic tools are available, a combination of probabilistic safety analysis and deterministic approaches is recommended. This method is called a risk informed review. It allows a prioritization of the ageing management review of all SSCs on the basis of their safety significance. The method can yield a SSC grading based on the SSC impact on the core damage frequency. Attention should be paid to common cause degradation across SSCs with similar vulnerabilities when conducting the probabilistic safety analysis.

Lists of systems, Structures and Components

Safety Related? R

Regulations

Normal Maintenance Safety Support?

Identify The Structure and Components

Prioritization

Phase I Structures and Components

Safety Failure Rate Economics

Phase II Structures and Components

Lifetime Management

Fig OOO – CRI-KHNP – Structure & Component screening After the SCC screening step a design and operation data collection step follows for all the screened components. The selected components are then classified in terms of importance to safety and production, When the SSC selected is a complex component such as a steam generator, sub structures and sub-components such as for example the internals of the steam generator are examined successively and also classified. Then the ageing related degradation mechanisms are identified. If unknown, failure data are sent to R&D for interpretation and research. As this step is completed all the elements are now available to undergo a prognosis which is hence undertaken.

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It is to be noted that active less critical components (such as rotating machinery, valves, etc.) have received more attention in the past with regard to diagnostics and prognostics and are more advanced than early detection of material degradation in pressure vessels, concrete, reactor structures and piping. In particular the physics of failure for structural materials (from damage initiation to failure of the component) is still incomplete. The factors that impact the growth of a crack are reasonably well understood but the dynamics of incipient crack growth are less well known, as are also the impact of stressors, the sensitivity of the diagnostic tool to the degradation mechanism and the capability of determining the current damage level from the degradation growth rate measurements. Early detection of incipient degradation relies on the knowledge of precursors. Precursor characterization is key to a successful prognosis together with the availability of instruments sensitive to precursors, and of analysis tools capable of interpreting from the precursor states, the evolution cycle of SSC degradation, from initiation to component failure. The development of advanced measurement techniques, such as phased-array ultrasound, acoustic emission and guided waves, can be deployed today in ISI programmes, and these methods can also be used for online monitoring. Data from ISI are then imported into the AMP to obtain SSC prognostics. Accurate SSC life prognosis is never easy. It is a function of the degradation state of the SSC and of the assumed development of the stressors in time, but also of their past history, if at all known. They must contain the thresholds for early warning of incipient component degradation and be capable of reading the marks of the material type and degradation state. The optimal selection of the sensor locations and the parameters to be monitored must be decided on the basis of a riskinformed SSC analysis. The SSC ageing data such as stressors, degradation mechanisms, time in service etc. are used to calculate the likelihood of failure. The curve normally follows the well-known “bathtub curve” pattern as shown in figure XXXY. At the beginning of life, when the new SSC enters service it experiences the so-called “run-in” period, The likelihood of failure during that time is relatively high, but the failure curve decreases as the SSC enters a stabilization time with a lower and somewhat flat failure rate. If the maintenance programme is optimized the stabilization period can be The probability of failure in this flat period depends on the quality and timely implementation of preventive maintenance and the extent and quality of the condition monitoring programme. As the SSC ages, the stressors become more aggressive and the likelihood of failure is statistically bound to increase. If the maintenance programme is less than optimized, the wear out period begins earlier and its slope is steeper. In well managed maintenance programmes at nuclear power plants, the flat portion of the bathtub curve is low and may even show declining probabilities of failure. With time however it will eventually increase and enter the wear-out period. The starting point of the wear out period and its rate of increase with time should be determined to plan corrective action, allocate budget and insert it into the schedule.

2.2.5.

Components Life evaluation

A preliminary screening is done from the top down, beginning with plant-life limiting components requiring a life assessment and a customized ageing management programme, including R&D where necessary. These are usually critical, mainly passive or structural SSCs, The screening can then continue at the life system level, where active but still critical components are considered and so on until AMPs are developed for all safety-critical and economically critical SSCs. In this instance regulatory requirements regarding ageing management must be considered because of the requirements of the regulatory framework. Two phases of SSC prioritization are followed,  

One covering criticality to safety and A second phase for SSCs with a failure rate threshold linked to economic requirements.

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By screening critical systems, structures, and components, owner/operators can reduce their ageing management analysis effort to a manageable scope. Depending on the grade and criticality of the component or group of components, ageing studies can be optimized to help develop the most appropriate and cost effective programme. An example can be the three assessment methods for three criticality dependent categories of SSC: 





Life assessment (LA): This technique is typically applied to the most critical structures and components that are generally passive in nature and designed not to be replaced. Predicting the remaining life of these critical systems requires prognostics tools and methods. This begins with a rigorous assessment of all plausible ageing related degradation mechanisms. The methodology entails a detailed review of plant data in order to establish current condition and to evaluate ageing degradation at a sub-component level. The LA report provides a prognosis for attainment of design life and/or long term operation. Recommendations provide the technical basis for ageing management planning and may be used for economic planning and decisions on PSRs for long term operation or on license renewal applications enabling LTO even beyond 60 years. While life assessment activities are focused on passive systems and structures, active systems (such as pumps and valves) are, and can continue to be, well managed and routinely monitored, diagnosed, analysed and upgraded with less onerous practices. However, it must be said that passive systems and structures such as large pressure vessels and reactor structures are not easily or economically upgraded when degradation is detected in an in-service inspection (ISI), and therefore, managing passive systems and structures degradation is likely to be the key to determining the economic viability of LTO. Condition assessment (CA): Typically done for less critical systems, structures and components. The CA process assigns components to commodity groups such as instrument categories, valve types, small rotating or reciprocating machinery, etc. where they are evaluated by group. The methodology entails a general review of plant data in order to establish current condition and to evaluate ageing degradation. The CA report provides a prognosis for attainment of design life and/or long term operation with associated recommendations. Recommendations provide the technical basis for on-going ageing management of the subject structure, component or commodity and may identify a need for further assessment. Systematic assessment of maintenance (SAM): This form of assessment makes use of Failure Mode Effect Analysis (FMEA) methodologies and information from internal and external feedback and R&D findings. It utilizes streamlined Reliability Centred Maintenance techniques, as modified for nuclear plant applications. It is 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

PLiM provides the opportunity to optimally deploy online condition-based monitoring and maintenance practices including the use of remote monitoring which altogether allow a better prognosis capability and more reliable predictions of remaining useful life (RUL). This proactive approach, known as prognostics-based health management (PHM) or integrated system health management (ISHM) requires online tools to detect degradation early in the lifecycle (diagnostics), and estimate RUL of the degraded component or system (prognostics). One methodology developed by KEPRI in Korea for life time assessments is shown in Figure xx through which a remaining life prognosis is conducted in 6 steps.

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Figure xx Remaining Life Evaluation A similar process with a somewhat different process flow is followed in Korea. Figure **** shows the Structure and component screening process developed by Central Research Institute of KHNP in Daejeon Korea.

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Figure XXXY: Ageing Bathtub Curve The bathtub curve of an SSC is in reality the envelope of the probability of failure curves of each of the stressors acting independently. In order to determine the beginning of the SSC wear-out period, one method is to determine the dominant stressors and their degradation mechanisms. This is usually done by consulting PLiM data bases containing the degradation mechanisms and related R&D data. At this point the PLiM data bases will generate a credible prognosis for the SSCs, based on their bathtub curves and plan mitigating actions to extend the SSC lives. There are other methods beside the stressor analysis method. Below are listed three of the most used prognostics methods in PLiM analysis:   

Reliability data analysis uses historical time-to-failure data to model time to failure predictions. This method does not rely on actual plant operating data and may be less pertinent to the actual plant conditions. Stressor-based methods that take specific operating conditions into account. They allow trending of the stressor values and the development of a distribution: stressor value versus predicted failure times. This correlation is then used in the life prognostics. Effects-based methods compute a degradation or damage index and correlate this quantity with the probability of failure. The RUL is typically estimated based on the time for the damage index to exceed some predefined threshold

The integration of conventional ageing management programs and PLiM insights provides the capability to select the ad hoc monitoring programmes and support the design and installation of automatic on-line monitoring instrumentation and surveillance devices. These programmes are crucial to alert the crossing of thresholds and hence allow the optimization of preventive maintenance activities. With the implementation of a PLiM programme, the safety aspects of ageing management are also better served, since integration provides a better awareness of safety margins and a better management of pressure boundary integrity throughout the plant life cycle. If advanced PLiM

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techniques are not implemented, degradation in safety-critical components is usually only detected at periodic inspection time (every 10 years), and potentially only after significant degradation has already occurred. This has potential implications in terms of safety if a serious failure occurs and in terms of economics, possibly resulting in higher maintenance costs and even extending an outage when a defect is found. In addition, to physical ageing, the plant equipment qualification programmes are reviewed in terms of their effectiveness. Up to date methods and practices are proposed to preserve and upgrade, where necessary, equipment qualification. Finally recommendations are formulated including the most appropriate organizational model for implementation. Finally, it is possible to envisage a future transition from current labour intensive ISI to more online monitoring and prognostics for passive components in parallel with regulatory relief from some burdensome ISI inspections, as they become more capable, automated, digitized and better remotely managed.

2.3. IMPACT OF AGEING DEGRADATION CAUSED BY THE ACCIDENT OF FUKUSHIMA DAI-ICHI NUCLEAR POWER PLANT

1. Background of review There has been a public concern that aging of the Fukushima Dai-ichi nuclear power plant facilities might have impact on causing or expanding the accident of Fukushima Dai-ichi NPPs. It was described in “Report of Japanese Government to the IAEA Ministerial Conference on Nuclear Safety” (June, 2011) that the impact of aging would be assessed in detail, and the relationship between aging and causes of the accident would be examined. . 2. Results of Review (i)

Assessment of each aging event

Regarding low-cycle fatigue cracking and irradiation-induced stress corrosion cracking of upper grid plate which might have the aging impact of this earthquake, it was assessed in the past technological assessment that the impact of the seismic ground motion was sufficiently small. As a result of assessment on the low-cycling fatigue cracking using the seismic ground motion of this earthquake, the impact on the safety margin to the acceptable limit was small and the acceptable limit had some safety margin to the actual damage limit. Therefore, as a result of comprehensive review, it was difficult to consider that there was the impact of aging such as losing the functions by the seismic ground motion of this earthquake. In the assessment on the neutron irradiation degradation of the RPVs using the seismic ground motion of this earthquake, the impact on the safety margin to the acceptable limit was small and the acceptable limit had some safety margin to the actual damage limit. Therefore, it was difficult to consider that there was the impact of aging such as losing the functions by the seismic ground motion of this earthquake. (ii)

Seismic impact assessment on major safety-related facilities

In light of the seismic safety assessment, regarding the whole-surface corrosion of the basic volt of the pump of the reactor shut-down cooling system which might have the aging impact of this earthquake, the assessment was taken using the seismic ground motion of this earthquake, and as a result, it was assessed that the impact on the safety margin to the acceptable limit was small.

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Regarding the low-cycle fatigue cracking of the shroud support, the low-cycle fatigue cracking of main steam system piping, and the low-cycle fatigue cracking of nuclear reactor recirculation piping, the assessment was conducted using the seismic ground motion of this earthquake, and as a result, it was difficult to consider that the seismic ground motion of this earthquake caused the impact of aging such as losing the functions, because the impact on the safety margin to the acceptable limit was sufficiently small in the past technological assessment, and the acceptable limit still had some margin to the actual damage limit. 3. Summary Utilizing the method and results of the aging technological assessment conducted in the past, it was confirmed whether there was any impact of the seismic ground motion of this earthquake on the safety-related equipment in terms of conservative assessment of aging from the commissioning to the 60-year-old operation. As a result of assessment by the knowledge obtained to date, it is difficult to consider that the safetyrelated equipment lost its functions because of an impact of aging. Also, it is difficult to consider that the aging events contributed to occurrence and enlargement of Fukushima Dai-ichi NPP accident during the time period between the accident occurrence and the time when the accident developed beyond the design basis. However, at this point, it is difficult to conduct the on-site confirmation of the equipment, this report is on the theoretical assessment of the impact of aging by the analysis, etc. utilizing the aging technological assessment in the past. Therefore, if new findings are available in the future through the on-site confirmation, etc., an additional review will be needed on the impact of aging. Ref. NISA report on Feb. 16, 2012, Regarding Impact of Ageing on the Accident of TEPCO Fukushima Daiichi Nuclear Power Station.

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3. PROACTIVE AGEING MANAGEMENT IN OPERATING NPPS A proactive management strategy for ageing management in a NPP implies the adoption of two non-conventional guiding principles: [***] Recognizing common weaknesses in conventional ageing management of NPP, especially in inability to trace root causes, to develop lessons-learned and to disseminate resources of ageing of SSCs important to safety and reliability; (b) Application of a proactive and systematic ageing management process, by pursuing continuing improvement of the ageing management programme as one of the core values. It has been conventional to place resources at higher priorities on the plant performance and leaving the long-term ageing management as one of the secondary goals. In this mode of operation, plants were operated as long as technical specifications are satisfied and corrective measures are made in reactions to component failure(s). The corrective or reactive management of ageing has led to surprises in the safety consequences, as exemplified in steam generator tube rupture events of Palo Verde NPP and the CRDM penetration nozzle failure at Davis-Besse NPP Preventive maintenance method has evolved in oredr to prevent such mishaps that can be anticipated to occur by plant condition monitoring. A proactive ageing management programme is one that takes into account not just the present and anticipated events but the whole life cycle of a plant in ageing management planning and systematic reviews of world-wide experiences at each stage of the lifetime. Proactive ageing management of SSCs starts, whenever possible, at the design stage, continues through the component manufacture/fabrication phase, the plant construction, and system installation, commissioning, and it extends throughout the operation period (including long term operation and extended shutdowns) to end with and including decommissioning. Ageing considerations must extend as well to all associated external activities such as engineering, procurement, fabrication, transportation, installation etc. Proactive ageing management entails a continuous learning process and the continuing improvement of methods and processes from the experience accumulated during operation of the plant itself and from that of other nuclear power plants, in order to avoid recurrent problems and to prevent lagging behind in plant operation techniques and in maintaining the plant performance targets. This implies that the operating organization is responsible for identifying ageing issues that are generic and known internationally from operations feedback, from vendor recommendations, from manufacturer operating manuals and from the plant designers. In addition, ageing concerns specific to the plant should be addressed in the plant operating instructions and procedures. Each known ageing issue should be addressed with an adequate ageing management programme. [2] The primary aim of most ageing management programmes is to help ensure the availability of required safety functions throughout the service life of the nuclear power plant. However, it is recognized that AM is also essential to the achievement of the desired plant performance and profitability. A proactive ageing management programme will therefore extend to both the safety and economic aspects. Underpinning practice of proactive ageing management cannot be better described by the Plan-Do-Check-Act (PDCA) protocol. Planning step can be better applied as a stage as early as design and as far in the future as decommissioning, as described in this Chapter. Regulatory requirements for ageing management at the national level must be developed before an operating license can be released. It may be that the national regulatory authority adopted the ageing regulations from the regulator of the NPP technology originating country. If this were the case, the national regulator should nevertheless review and update and adapt ageing management requirements and guidance to the local conditions. The national regulator should require that all ageing issues be identified and documented in the safety analysis report and periodically updated throughout the plant lifetime. [1]

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3.1. AGEING CONSIDERATIONS IN DESIGN Beginning with the conceptual design phase of any nuclear power technology, ageing considerations of all SSCs must be taken into account in the high level design target (or objective) documents, in safety design guides and in the specific design requirement documents of the SSCs. In the design requirement documents and in the basic assumptions and inputs of safety, thermo-hydraulic and stress analysis, as well as in the SAR reference should be made to the appropriate articles on ageing in the applicable codes and standards. A proper EQ programme provides effective means for ageing management. The EQ process is illustrated in Fig. XXX. Activities at the design input stage provide important information that is needed before EQ can be established for specific plant applications. The stage during which EQ is established involves all those activities necessary to the Design inputs, to Establishing an EQ programme and to Preserving the EQ of SSC requiring it.

FIG. XXX. The equipment qualification(EQ) process; PIE: postulated initiating event. In the pre-contractual phase in the bid specification the plant owner or operator must clearly communicate ageing criteria and requirements to the plant supplier, design organizations and manufacturers in order to receive a design that complies with all criteria and requirements for reliability, performance and even long term operation (if necessary). During the bid evaluations, ageing considerations and equipment qualification (EQ) are important issues and complementary to each other in designing SSCs. A proper EQ programme provides effective means for ageing management. Ageing should be one of the items being evaluated for each of the technologies under scrutiny. During the contractual phase, following the purchase of a generic design, the detail design phase and the plant deployment are negotiated and agreed upon. It is important that all ageing considerations and assumptions made by the designer on how the plant should be operated, including operating margins are crucial and should be handed over by the technology Vendor to the operator. An example could well illustrate the importance of design assumptions. There may be parts of associated and auxiliary systems to the primary reactor coolant system that are not evaluated for thermal fatigue

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under the assumptions that certain temperature ranges will be respected during operations. If the operator is not aware of such assumptions then fatigue failures may occur. While ageing considerations are given at the stage of the conceptual design, they must not be underestimated or worse be compromised during the site specific detail design phase. Detail design is normally governed by a division of responsibility by chapters in the contract. Equally important is the ageing consideration in procurement. When the plant supplier and the plant operator select the equipment manufacturers, they should determine whether ageing considerations and EQ tests are in compliance with their own EQ programmes. In this respect, they should guard against non-conformances, interfacing discontinuities or material incompatibilities. If the plant generic design has been certified by the regulator of the technology originating country, ageing issues should have been addressed in the generic design and in the generic design documentation. The national regulator may decide to accept the certification to the extent of the generic design. However, the national regulator should still require that the plant specific detail design meet the same standards as those applied to the generic design in the technology originating country and that the interfaces and the programmes be fully compatible. Ageing management should be included in the SSC design criteria, in design guides, in design requirement and should be addressed in safety design guides and in the safety analysis report. The use of IAEA’s PLiM data bases, tools and techniques as ageing issue awareness tools and design support to improve material selection, equipment specifications, qualification requirements, maintenance and access envelopes, vibration reduction, I&C and monitoring, inspection and testing requirements etc. The appropriate design document should address ageing management and include as a minimum the following topics [1]:  A recommended strategy for ageing management and prerequisites for its implementation;  A list of all safety significant SSCs of the plant that could be affected by ageing;  Proposals for appropriate materials monitoring and sampling programmes where ageing may affect the capability of critical SSAs to perform their function throughout the lifetime of the plant;  Appropriate consideration of operating experience with respect to ageing;  Ageing management recommendations for critical SSCs (concrete structures, mechanical & electrical, I&C components, cables, etc.) and measures to monitor and mitigate their degradation;  Environmental qualification requirements of SSCs important to safety, including equipment lists, functions required for normal operation and postulated initiating events;  General principles stating how the environment of an SSC is to be maintained within specified service conditions (location of ventilation, insulation of hot SSCs, radiation shielding, damping of vibrations, submerged conditions and water chemistry, selection of cable routes, requirements for stabilized voltage centres, etc.). Design basis accidents anticipated following all postulated initiating events are also to be considered part of the operating life and hence part of the design margin calculations. Following the Fukushima accident, additional severe accident mitigating features, including portable SSCs may have been introduced in the context of increasing robustness and/or flexibility of the plant in mitigating the consequences of severe accidents. They may or may not be aligned and energized and some of these may be in storage. Stress tests are employed for this purpose. Particular attention should be paid to assure appropriate technical margins. The margins consist of design margin, operational margin and safety margin to prevent the component failure. The operator should ensure and the regulator should verify that margins adequately include the ageing effects on SSCs, particularly on those important to safety. Margins should include the operational margin considering effects of wear mechanisms and other age related degradation to ensure their safety and commercial functions are ensured for the duration of the SSC mission time. If the mission time extends to the plant service life, as is the case for passive non-replaceable components, these capabilities should also analysis the duration of the entire plant service life. This of course includes all normal and transient operating conditions, testing, maintenance, and outages. Another area that requires time and attention and that should include ageing considerations is the component environmental qualification programme. This programme ensures that the design basis,

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including transient conditions and postulated event conditions are considered in the design of environmentally qualified components. Although not directly part of the ageing management programme the environmental qualification programme should nevertheless be compatible and certainly not in contradiction with it. It should be ensured that the design basis conditions, including transient conditions and postulated initiating event conditions, are taken into account in the design and operation of the equipment qualification programmes. All potential ageing mechanisms for passive and active SSCs should be identified, evaluated and taken into account. Potential ageing mechanisms that could affect the safety functions of the SSCs during their design life include thermal and radiation embrittlement, fatigue, corrosion, environment assisted cracking, creep and wear. [1] Consideration should also be given to the following:    

Use of advanced materials with greater ageing resistance properties. Materials testing programmes to monitor ageing degradation. Inspection and/or on-line monitoring, of areas with high risk of degradation leading to failure of sscs and where the consequences of failure could be significant to safety. Plant layout and design of SSCs that facilitate inspection, maintenance and ease of access for inspection, testing, monitoring, maintenance, repair and replacement, and minimize occupational exposure during these activities.

3.2. AGEING CONSIDERATION DURING FABRICATIONS & CONSTRUCTION. During fabrication and construction installation non-conformances or quasi non-conformances may occur. Consequently, residual stresses may be induced that the design did not foresee. For example sometimes in welding during construction an installer may apply a certain amount of force to bring two parts (e.g. pipe spools to be welded) closer together. Such a use of force may induce important residual stresses in the weld, thus increasing the risk of cracks. Although force application in pipe fitting is not forbidden, it is important that the plant operator be informed of the practice so that appropriate measures may be taken to manage the component usage and ageing. Other issues during construction could be the use of inappropriate materials, loose parts in pipelines or vessels, insufficient gaps for thermal expansion in pipe supports, high temperatures in repairs of stainless steel welds or in multi pass welding, inducing unacceptable or accumulated heat input, geometrical discontinuities that are not properly taken into account or modelled in stress analysis; damage during transportation etc. [2] The operating organization or the contracting agency responsible for quality control and procurement should ensure that the suppliers adequately address factors affecting ageing management and that all suppliers provide sufficient information and data to facilitate ageing management during plant operation. [1] The owner/operator should ensure as a minimum that:    

Current knowledge on the factors affecting ageing is available to SSC manufacturers and is properly taken into account in the fabrication and construction of sscs; Current knowledge about relevant degradation mechanisms and mitigation measures are taken into account during the manufacturing and installation of sscs; The appropriate amount of reference (baseline) data are collected at the factory and documented; Surveillance specimens for specific ageing monitoring programmes are specified, made available and installed in accordance with design specifications.

3.3. AGEING CONSIDERATIONS DURING COMMISSIONING During commissioning, proactive ageing management considerations should also be applied. The functional capability of SSCs is checked and baseline data are collected to record conditions at start up for later reference in ageing management during operation. Also environmental conditions in the various buildings during the hot tests are checked for consistency with the design and qualification assumptions. Errors during commissioning may induce accelerated ageing in some systems. If the

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wrong resins or wrong chemicals are used, even though corrective action and a thorough clean up may be undertaken, residual aggressive agents may remain in the system and cause early corrosion and cracks. Abnormal temperatures should be notified during commissioning and corrective action taken to prevent accelerated ageing in adjacent SSC’s including concrete structures, cables etc. Transient cycles and extreme conditions including hydro testing and hot conditioning tests should be recorded to ensure they are not out of bounds from the assumptions of the SSC fatigue design and the records kept as baseline data to facilitate future actual usage factor assessments during operation or when investigating root causes of possible cases of premature ageing. The operating organization should establish a systematic programme for “as installed” measuring and recording baseline data relevant to ageing management for critical SSCs. This includes mapping the actual environmental conditions in each critical spot of the plant to ensure that they are in compliance with design requirements. Special attention should be paid to the identification of hot spots in terms of temperature and dose rate, and to the measurement of vibration levels. All parameters that can influence ageing degradation should be identified and controlled during commissioning, the information made available to operations in order that it could be tracked throughout the plant life. The regulatory body, as part of its review and inspection programme, should ensure that the operating organization collect required baseline data and confirm that critical service conditions (as used in equipment qualification) are in compliance with the design requirements and analysis assumptions. [1] 3.4. AGEING MANAGEMENT IN OPERATION A systematic approach to managing ageing such as PLiM techniques if applied during plant operation can help mitigate the effects of ageing. Application of PLiM since the early stages of plant operation will automatically apply a proactive approach to ageing management because PLiM is based on understanding degradation mechanisms, on prevention and mitigation, rather than supposedly minimizing expenses with a reactive approach by responding only to SSC failures, which instead will inevitably result in increased cost and performance reductions. A proactive approach to ageing management during operation will also look at continuing education in ageing management and plant life management, at increasing the awareness, familiarization, the motivation and sense of ownership of all operations, maintenance and engineering support staff. In order to ensure the correct application of proactive ageing management principles, appropriate procedures, tools, materials and qualified staff should be made available, appropriate storage of spare parts and consumables should be required to minimize degradation and control shelf life, as well as the use of multidisciplinary teams for dealing with complex ageing management issues. The use of feedback of operating experience should be required, and the use of on-line and ad hoc monitoring and of NDE should be encouraged. Figure YYYY shows the elements necessary in the development of an ageing management strategy during operation of an NPP.

28

To keep safety and reliability of nuclear power plants for long term operation In Lab

Establishment of information Basis

Database for Degradation for Materials

Systematic Ageing Management Program

3.

2.

Evaluation Technology For Degradation of Components

Database on Regulation Procedures IASCC In Other IASCC Countries

RPV Radiation Embrittlel ment

4. Codes and Standards

Technical Development

Standardization of Ageing Management Procedures

Schemes to apply new techniques

Performance Index

Systematic Maintenance

Optimization of Maintenance

Human Risk-based maintenance Resources

Fig. YYYY Strategy Map for Ageing Management [ ref : Prof. Sekimura presentation in TWG meeting in 2013] This diagram was first presented by the Atomic Energy Society of Japan in 2009 and updated yearly until 2011.It constitute a strategy map for ageing management in that it points at the four major principles that should govern the implementation of an ageing management programme in NPPs: the establishment of a material degradation database, the development of a technical understanding and an evaluation of component degradation in the plant, the application of codes and standards and the setting up of a systematic approach to maintenance which leads to maintenance optimization through HR training and selection and the use of tools such as risk-based guidance in the maintenance planning process. Maintenance during operation should also be mindful of ageing and should be conducted proactively. The best way to achieve this is to integrate maintenance programmes with ageing management activities and if the plant has a Plant Life Management set up integrate maintenance and ageing management into the PLiM programme as well. Figure XXX shows the interconnectivity of maintenance and ageing management.

29

Fig XXXX Integration of a routine maintenance programme and ageing management activities. Figure XXXX was first presented at the NEA Stress Corrosion Cracking and Cable ageing project in Tokyo, Japan May 25-26, 2010. The blue boxes indicate the interfacing activities of the large maintenance envelope and ageing management. This interfacing gives rise to additional programmes indicated by the yellow boxes which become part of the overall maintenance policies and goals and of the routine maintenance activity chart. The IAEA developed PLAN, DO, CHECK and ACT cycle to apply a systematic ageing management process. For example, Fig. VVV represents the key elements of an ageing management programme utilizing PDCA cycle for the reactor pressure vessel in a PWR. Similarly, Fig PPP represents the key elements of an ageing management programme utilizing a systematic ageing management process for the reactor pressure vessel internals in a pressurized light water cooled reactor. The peripheral boxes (PDCA cycle) describe the four steps of a systematic implementation programme: The PLAN box at the top is the first of these steps and involves the planning of activities aiming at optimizing ageing management of a SSC and so improving its effectiveness. The second peripheral box to the right is the DO box is the collection of all activities aimed at minimizing the expected degradation during the operational phase of the SSC involving the careful operation of the SSC within its technical specifications and according to the applicable procedures including the best chemistry and environmental control possible and the recording of transients and historical events. The box at the bottom is the CHECK box that requires the periodic inspection, the monitoring and the condition assessment of the SSC’s functional capabilities. The box in the middle is the first step which is the determination of the ageing mechanisms of a structure / component. If one of the mechanisms is new or not known, then R&D is required to develop an understanding. For the well-known mechanisms only recognizing them and listing them or

30

extracting them from a PLiM data base will be sufficient. Other information on the SSCs themselves such as material properties, fabrication methods, stressors, operation experience, R&D findings is also crucial to complement the understanding of the degradation mechanisms. PLAN 2. Co-ordination of RPV Ageing management Programme (AMP)

Improve AMP

Co-ordinating ageing Management activities:  Document regulatory  requirements and safety criteria   Document relevant activities   Describe co‐ordination  mechanism   Optimize understanding, periodic  self‐assessment and peer reviews

Minimize Expected degradation

DO

ACT 5. RPV Maintenance

1. Understanding RPV Ageing

3. RPV Operation/Use

Managing ageing affects:  Radiation embrittlement  ‐ Thermal annealing   SCC of Alloy 600 CRDM  penetrations  ‐ Penetration repair  ‐ RPV closure head replacement   Corrosion and pitting of flanges  ‐ Grinding repair   Wear of closure head studs and  threads  ‐ Repair by grinding and  threaded sleeves 

Key to effective ageing management:  Materials and material properties  Stressors and operating  conditions   Ageing mechanism   Degradation sites   Condition indicators   Consequences of ageing  degradation and failures under  normal operating and DBE  conditions 

Key to effective ageing management:  Operation acc. to procedures and  technical specifications  ‐ water chemistry  ‐ heatup/ cooldown(P‐T)  ‐ design transients   Mitigation of radiation  embrittlement  ‐ Fuel management (LLC)  ‐ RPV wall shielding   Mitigation of Alloy 600 SCC  ‐ Zinc addition to primary water   Operating history  ‐ Water chemistry  ‐ Inadvertent cool down and  pressurization  ‐ Inlet water temperature  ‐ Transients (inadvertent safety  water injection, etc.) 

CHECK 4. RPV Inspection, Monitoring and Assessments Mitigate degradation

Detecting & assessing ageing effects:  Surveillance specimen programme   In service inspection (NDE)   Monitoring pressure, temperature,  power distribution, water chemistry   Leakage monitoring   Assessment of:  ‐ Radiation embrittlement  ‐ Flaw assessment  ‐ SCC of Alloy 600 components  ‐ Fatigue usage  ‐ Thermal ageing 

Check for degradation

Fig. VVV – Key elements of a PWR pressure vessel ageing management programme This can be done by testing and calibrating, inspecting to detect ageing symptoms such as leaks, vibration and by following up by means of continuing record keeping. When conditions grant it a fitness for service investigation may be required. The fourth peripheral box is the ACT box that deals with mitigating degradation. This may involve preventive and corrective maintenance action, spare parts inventory management and record keeping of all maintenance history.

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Fig. PPP Systematic approach to managing ageing of a reactor pressure vessel internal structure for a PWR At the end of the cycle, the experience gained going systematically though the four peripheral boxes should be used to improve the effectiveness of the ageing management programmes. This experience should therefore close the peripheral loop by acting on the planning box at the top with a view to further improve the ageing management programme effectiveness. The operating organization should also identify and address the following potentially significant common issues of ageing management programmes:   

Premature ageing of SSCs caused by more severe service conditions than assumed by design or by errors in design, fabrication, installation, commissioning, operation and maintenance, or by lack of coordination or unforeseen ageing phenomena; by reactive ageing management Lack of a technology watch function causing lack awareness of operating experience feedback and research results; Unexpected stress loading to structures or components by external events (e.g. earthquakes).

The contribution and the relationship among all stakeholders in Ageing Management for a generic nuclear power plant in Japan is well described as an example in the diagramme of figure WWWW from the Japanese regulator.

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Nuclear Safety Regulatory Standard Committee

Nuclear and Industrial Safety Agency (NISA) Reporting

Technical Information Coordination WG for Technical Information Basis

JNES Each technical Review committee - Irradiation Embrittlement - SCC, IASCC - Flaw Inspection - Flaw Evaluation - Electrical Equipment Ageing

Regulatory Organization

WG for Safety Research

WG for international Collaboration

General Review Committee Sub-committees

- Fatigue - Irradiation Embrittlement - IASCC - Cable - Concrete - Pipe Wall Thinning - SCC - Seismic Safety - Technical Information Infrastructure - Internal Cooperation

PLM Research Promotion Conference - JANTI - CRIEPI - MHI - HITACHI GENE - TOSHIBA - JPOWER - 9 Utilities

Industries

NISA Ageing Management Project

Academia-Academic Societies Figure WWWW: Intervention and relationship among all stakeholders of a NPPA

A multi-representative Nuclear Safety Regulatory Standard Committee assumes a coordination role for all stakeholders involved in ageing management of a typical nuclear power plan. It interfaces with the regulatory organization, academia and academic societies in Japan and wit the industrial groups providing services to the pant ageing management and maintenance groups. Coordination is executed through working group operating in each of the three larger stakeholder groups. These stakeholders carry out their ageing management review benefiting from communications and action coordination among themselves and through their Working Group report to the Technical Information Coordination Committee which in turn reports to the Nuclear and Industrial Safety Agency (NISA) whose Nuclear Safety Regulatory Standard committee receives the information and disposes of it appropriately. Out of this process comes a continuous updating of the ageing management strategy. These are translated in requirements and procedural changes in the specific Nuclear Power Plant. If the changes are generic and apply to all reactors of a certain type, then a Generic Action Item to update the Ageing Management strategy map is issued to all NPPs for Implementation. 3.4.1.

Refurbishments, Modernizations and Power Uprating

In the event of reactor power uprating, refurbishments, important modifications or equipment replacements, the operating organization should identify and justify possible associated changes in process conditions (e.g. flow pattern, velocity, vibration) that could cause accelerated or premature ageing and failure of some components. Also if a components must be replaced because of obsolescence or for other reasons, with similar but not same specifications and if a thorough assessment is not made or if the operator is not aware of such changes, even though form and

33

functions remain the same, different specs may produce material compatibility and ageing issues, and there could be safety consequences. In the case of guide tube pins made of a nickel based alloy, very minor differences in the fabrication process could affect component sensitivity to stress corrosion. If a new ageing mechanism is discovered (e.g. through feedback of operating experience or research), the operating organization should prepare contingency plans or exceptional maintenance plans to deal with the condition The availability of spare or replacement parts and their shelf life, of consumables should be continually monitored and controlled and protected from ageing. [1] The rate of ageing degradation can often be reduced by optimizing operations. In the case of primary system components, proper recording of transients allows the fatigue usage factor to be improved using actual operating transients and stressors. Also optimizing the fuel pattern in the core to reduce fluence on the vessel and mitigate embrittlement is a good example of operation optimization to mitigate the effects of ageing. [2]

Ref : NES series – Power Uprate in Nuclear Power Plants: Guidelines and Experience, IAEA Nuclear Energy Series NP-T-3.9,

3.4.2.

Long Term Operation:

At the end of service life, the ageing management instructional content builds on a plan of approach developed by the Swedish nuclear safety authority during the preparation phase of the applications for long term operation beyond the plant lifetime as postulated in the original design. This methodology has proven successful in the preparation to the last periodic safety review or licensing renewal application prior to a NPP’s long term operation.

34

LTO Demonstration Ringhals 2

Feasibility (3.1)

Phase prior to LTO assessment (SR57, fig1)

IAEA Safety report Nr 57 Safe Long Term Operation of Nuclear Power Plants

IAEA guidelines

Regulatory framework

IAEA IGALL International Generic Ageing Lessons Learned (TLASSs)

Verification of preconditions (3.2) Scoping (4.1)

TLAAs(6)

Phase LTO Assessment (SR57, fig1)

Screening (4.2)

Not (explicity) In SR57

Active

RPV AMR Passive

Check of Existing Maintenance And surveillance programs

Primary System

Mechanical

Electrical And I&C

Civil / structural

Fatigue

EQ

LBB

Containment Tendons

Underclad cracks

Concrete creep

CASS Fatigue Flywheel HCF SG

RPV internals vibrations Thermal ag martensitic

Documentation of Basis for LTO

Phase LTO Approval & Implementation (SR57, fig1)

Regulatory Oversight

Implementation of plan commitments For LTO

Fig.1 Plan of approach for LTO projects (Swedish Nuclear Safety Authority)

35

PSR

PSR is the standard approach for those Member States that do not follow the License Application process. The plan of approach shown in Figure 1 illustrates a systematic phased approach: The first phase at the top is the preparatory phase prior to the application for Longer Term Operation. The second block represents the LTO assessment phase and the third block is the implementation of the actual license renewal application commitments. A good ageing management programme conducted from day one in the service life of a plant will guarantee that the plant will be in its best possible condition and will require the minimum possible changes at the time of its longer term application or its end of life safety review which is usually more demanding than regular periodic safety reviews. Beyond IAEA safety report 57 [***], the diagram shows that the country’s nuclear regulatory framework complements its content with the local regulatory framework which in terms of ageing management, relies on the periodic safety review framework. In the context of a licensing framework that requires periodic safety reviews, the last safety review cycle leading to a regulatory permit for an operating period beyond the plant life assumption as first postulated in the original design is illustrated in figure ZZZ provided by the Qinshan phase 1 nuclear power plant in Qinshan, China. Previous PSR

Over all AMP Review of existing programs and procedures

AMP organization and Responsibilities

Equipment specific AMPs

Topical AMPs

Equipment Screening

AMP for SG

- Degradation mechanisms - Ageing detection - Ageing mitigation - Ageing assessment - Amp procedures

AMP for T2

Topic for FAC - Investigation - Detection - Mitigation - Assessment - AMP procedures

AMP for Tm

- Degradation mechanisms - Ageing detection - Ageing mitigation - Ageing assessment - Amp procedures

. . .

. . .

AMP for En

Topic for FAC - Investigation - Detection - Mitigation - Assessment - AMP procedures

AMP Database

AMP for RPV

- Degradation mechanisms - Ageing detection - Ageing mitigation - Ageing assessment - Amp procedures

AMP for T1

Topic of obsolescence

New PSR

Life extension

Figure ZZZ Systematic Ageing Management Model for Qinshan-1

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Evaluation points Clustering/Screening -

LTO issues

LTO-issues to existing improvement projects non-issue

Selection

LTO issues short

-

Consolidated issues

Safety impact assessment Expert judgement Probabilistic insight

Improvement orientations Potential improvements

Agreed design upgrade

-

Feasibility Safety benefits

Integrated assessment

Improvement plan

Figure HHH: Design Upgrade Assessment Process When a design upgrade is deemed necessary to enhance safety or performance, the process to assess the project is shown in figure HHH. First a screening or clustering step is undertaken to review all LTO issues, the currently proposed improvement projects and the non-issue. A selection is undertaken based on a safety impact assessment, expert judgment supported by probabilistic insights (if available). A cost benefit analysis is conducted to orient the proposed improvements in a feasibility context as well as the view point of safety benefits. All inputs are then collected and an integrated assessment is conducted. The process leads eventually to an improvement plan which includes both the LTO drivers for the changes which are short listed and the work plan for the agreed upon design upgrades. 3.5. AGEING MANAGEMENT PRE-REQUISITES FOR DECOMMISSIONING [MORE INPUT REQUIRED] Appropriate arrangements, including ageing management, upgrades, replacements, etc. should be made to ensure that SSCs and tools (e.g. containment system, cooling equipment, lifting equipment and condition monitoring equipment) remain available and functional to facilitate decommissioning activities. [1] 3.6. EFFECTIVE AGEING MANAGEMENT PROGRAMMES So far we have been focusing on SSCs of safety significance and the screening used criteria focused on safety significance. Naturally the Owner/operator and licensee may wish to include in his AM programme also those SSCs that may not directly be of safety significance, but whose failure would have a significant financial impact, for example those leading to long down time and/or substantial repair or replacement cost penalties. The attributes of an effective ageing management programme are summarized in table 3:

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TABLE 3 EFFECTIVE AGEING MANAGEMENT PROGRAMMES 1. Scope

of the ageing management programme based on understanding ageing

 

 2. Preventive actions to minimize and control ageing degradation

 

3. Detection of ageing effects





4. Monitoring and trending of  ageing effects



Structures (including structural elements) and components subject to ageing management Understanding on ageing phenomena (significant ageing mechanisms, susceptible sites): - Structure/component materials, service conditions, stressors, degradation sites, ageing mechanisms and effects - Structure/component condition indicators and acceptance criteria Quantitative or qualitative predictive models of relevant ageing phenomena  Identification of preventive actions Identification of parameters to be monitored or inspected Service conditions (i.e. environmental conditions and operating conditions) to be maintained and operating practices aimed at slowing down potential degradation of the structure or component  Effective technology (inspection, testing and monitoring methods) for detecting ageing effects before failure of the structure or component Condition indicators and parameters monitored Data to be collected to facilitate assessment of structure or component ageing Assessment methods (including data analysis and trending) Operations, maintenance, repair and replacement actions to mitigate detected ageing effects and/or degradation of the structure or component Acceptance criteria against which the need for corrective action is evaluated  Corrective actions if a component fails to meet the acceptance criteria

5. Mitigating ageing effects

 

6. Acceptance criteria



7. Corrective actions



8. Operating experience feedback and feedback of research and development results



Mechanism that ensures timely feedback of operating experience and research and development results (if applicable), and provides objective evidence that they are taken into account in the ageing management programme

9. Quality management



Administrative controls that document the implementation of the ageing management programme and actions taken Indicators to facilitate evaluation and improvement of the ageing management programme Confirmation (verification) process for ensuring that preventive actions are adequate and appropriate and that all corrective actions have been completed and are effective Record keeping practices to be followed 

 



Administrative control of the development and execution of an ageing management programme is an important parallel activity aimed at maintaining quality control over the process. Fig JJJ shows a four step procedural control of the ageing management programme at the PAKS nuclear power plant. Control extends to the scoping, development, execution and feedback steps.

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Fig JJJ Administrative control of the AMP at the NPP in PAKS, Hungary

39

4. AGEING MECHANISMS Degradation mechanisms due to ageing are usually classified into two main categories:  

Those affecting the internal microstructure or chemical composition of the material and thereby change its intrinsic properties (thermal ageing, creep, irradiation damage, etc.). Those imposing physical damage on the component either through metal loss (corrosion, wear) or through cracking or distortion (stress-corrosion, deformation, cracking).

Corrosion

Water

Stress Corrosion Cracking Fracture Mechanics

IASCC

Stress

Radiation

Radiolysis Radiation creep Radiation corrosion

Material

Radiation induced

Figure KKK the concurrent action of all stressors present on the material.  From the point of view of the stressors Figure KKK illustrates the concurrent action of all stressors present on the material.  Under the most demanding operating condition near the core of water cooled nuclear reactors, metallic materials are subject to the concurrent influence of the cooling medium, of stress fields and of high radiation levels. Every coloured square in the diagramme represents a different combination of stressors on SSCs or parts of SSCs. The severity of the consequences depends on the specific function, the SSC is called to fulfil, on its location with respect to the source of degradation and on the mitigation measures, the shielding or other protective covers the material is endowed with. The square located at the centre of the diagramme indicates a condition known as irradiate-assisted stress corrosion cracking of austenitic stainless steels, a condition to which are typically subjected light water reactor core internals. The green square to the left represents the conditions under which light or heavy water radiolysis occurs more frequently. Exposure of the reactor cooling water to ionizing radiation induces high-energy radiolysis of H20 water molecules into H+ and OH- radicals. These radicals are themselves chemically reactive, and in turn recombine to produce a series of highly

40

reactive compounds such as superoxide (HO2) and peroxide (H202), which enhance oxidative damage to metallic components. The green square to the right represents the region where the effect of water radiolysis on corrosion of stainless steel, particularly in the presence of oxygen dependant intergranular stress corrosion cracking. The effect of radiation and water radiolysis on activity transport in nuclear power systems has been the subject of numerous investigations. Radiation-induced segregation is the consequence of irradiation at elevated temperatures where depletion of alloying elements (typically, chromium, molybdenum and iron) and enrichment of others (typically nickel, silicon) occur in regions near the component surface. In extreme cases there can be dislocations, void creation, newly formed grain boundaries or phase boundaries of segregated elements within the alloy matrix. Irradiation creep is the changes in physical properties due to radiation exposure such as the strength, ductility, elastic moduli that may lead to swelling, to fatigue and even creep-related ruptures. Swelling is exacerbated by the effects of temperature, cold work and stress levels. Stress corrosion cracking (SCC) is the presence and of cracks typically in austenitic steels and aluminium alloys in a corrosive environment such as the presence of chlorides, alkali, nitrates, ammonia. It can lead to unexpected sudden failure especially at elevated temperatures The metal surface may appear shiny and unaffected but the cross section can harbour cracks which often go undetected but progress rapidly producing devastating and unexpected failures. When cracks are detected through in-service inspection (ISI) the analytical methods called fracture mechanics can be applied to make predictions. It applies the theories of elasticity and plasticity, to the microscopic crystallographic defects found and is able to predict with acceptable accuracy mechanical failures. Degradation mechanisms include metallurgical phenomena such as irradiation embrittlement, fatigue, corrosion, interaction and a combination of such mechanisms. Table 1 illustrates some of the the interaction of such phenomena. It should be noted however that the table shows only the main degradation mechanisms for different NPP components. There are many more mechanisms, some of which are shown in more detail in table 1 TABLE 1. MAIN DEGRADATION MECHANISMS IN PWR COMPONENTS Component Reactor pressure vessel Control rod drive mechanisms Internals structures Reactor coolant pump casing Piping and safe ends Pressurizer

Irr. Emb. 



Surge and Spray lines Steam generator tubing Steam generator shell and nozzles

Fatigue 

Corrosion Fatigue

SCC  *















Thermal Ageing

**

***

 +

Wear

  

 

Corrosion

 ++

 



 







 

Note : Irr. Emb : Irradiation Embrittlement

41



 

SCC : Stress Corrosion Cracking * ** *** + ++

In closure head bolts When there is no cladding (VVER 230) or cladding defect corrosion is possible. RPV corrosion can occur on the outside, if leaking head penetrations allow borated water to contact the RPV vessel. Depending on RPV material In instrumentation nozzles and heater sheaths The combination of flow and corrosive fluid can lead to erosion-corrosion or flow-accelerated corrosion and thus reduce the wall thickness of piping.

Fig.SSS. Ageing factors, ageing mechanisms and consequences Figure SSS correlates the stressors present in the material to the setting up of ageing mechanisms which in turn determine the observable physical consequences. Irradiation assisted stress corrosion cracking (IASCC) is an ageing degradation mechanism affecting mainly the reactor internals. Field experience and experimental work have shown that several austenitic stainless steels, such as types 304, 316L, 316CW and 347, are susceptible to IASCC. However, from a practical point of view, it may be difficult to decide whether cracking in the field is caused by IASCC or by other types of IGSCC. Neutron irradiation causes atom displacements from their equilibrium crystallographic locations thereby creating atomic scale point defects, i.e. vacancies and interstitials. Neutrons generate large cascades of point defects as energy transfer to the displaced atoms is significant so that the displaced atoms in turn continue and cause even more atom displacements. Subsequent diffusion of point defects to various sinks such as grain boundaries, dislocations and surfaces, leads to significant changes in microstructure and mechanical properties in metallic materials. Neutron irradiation effects are primarily thermal. However, in the case of thick section components, significantly higher temperatures than the surrounding aqueous coolant can be generated within the material by gamma heating. Such higher temperatures can have a significant effect on the likelihood of void swelling occurring.

42

4.1. RADIATION DAMAGE Radiation‐induced microstructural changes significantly degrade material properties. Radiation damage in RPV steels is usually correlated with the neutron fluence irradiated given material. Two different threshold energies are used for characterization of irradiation conditions - neutrons with energies larger than 1 MeV for PWR and BWR RPVs of western design and with energies larger than 0.5 MeV for WWER RPVs. Unfortunately, ration between both neutron fluences is not constant for all irradiation conditions, thus any comparison of data must be carefully analyzed. Radiation damage materials for reactor internals is best correlated to displacement damage to quantitatively characterize it. A physical unit called displacements per atom or (dpa) has been used. It is a damage‐based or consequence-based radiation exposure unit and represents the number of atoms displaced from their normal lattice sites as a result of continuous subatomic particle hits. Although radiation damage cannot be fully characterized by a single parameter dpa is well suited to correlate radiation to physical property alterations Reactor pressure vessels of both PWR and BWR plants are fabricated from low-alloy steels of Mn-Ni-Cr-Mo types (western designs) or Cr-Mo-V and Ni-Cr-Mo-V types (WWER design). The most important damage mechanism is radiation embrittlement (together with radiation hardening) that practically determines the RPV lifetime. Radiation embrittlement is presented by the shift of material transition temperature from brittle-to-ductile fracture to higher temperatures and depends on material type, content of alloying (like nickel) and detrimental elements (like phosphorus and copper)and neutron fluence. Reactor vessel internals of both BWR and PWR plants are mainly fabricated from austenitic stainless steels. Unlike the fuel elements, which are removed after a few years of service, the internals are intended to remain for the full life of the plant and consequently can be exposed to very high radiation doses, typically 5–10 dpa in a BWR and up to 80 dpa in a PWR (assuming a 40-year life cycle and depending on fuel management). With such high radiation doses, the material microstructure and mechanical properties change considerably; which have a significant impact on the stress corrosion susceptibility in both reactor types.

43

44

Neutron irradiation is the cause of several degradation processes shows the major degradation processes affecting austenitic steels under neutron irradiation and their effects on the material physical and mechanical properties.

Radiation Swelling Vacancy voids)

New phases Generation: G-phase, carbides

Partial → Transformation, Resulting in Brittle-to-ductile Transition in Austenitic steel

Generation of dislocation loops

Helium and Hydrogen generation

Redistribution of alloy And impurity elements

1. Increase of yield strength Generation: G-phase, carbides

Decrease of cohesive Strength of grains boundaries

2. Decrease of strength of Interphases 3. Decrease of strain hardening

Additional internal Stresses in the Components of internals

Sharp decrease of Ultimate strength for Ductile fracture (S >15%)

Increase of voids nucleation Rate under deformation Increasing of strain localization

Decrease of plasticity and sharp Decrease of fracture toughness

Grain boundary Sliding under Radiation creep

Decrease of stress Corrosion cracking resistance

Figure EEE – Effects of neutron irradiation in austenitic steels This diagramme correlates the physical phenomena in the orange boxes at the top due to neutron irradiation to the degradation mechanisms at the microstructural level in the blue boxes, through to the changes in the material physical properties in the pink boxes. In addition, neutron capture reactions induce transmutation reactions and hence changes in chemical composition. From a materials viewpoint the following radiation induced changes should be considered in relation to IASCC: Microstructure 

High irradiation induced dislocation

45

 

loop density; Cavities (bubbles and voids).

Mechanical properties    

Increased tensile properties (e.g. yield and ultimate tensile strengths up to 1000 MPa); Decreased uniform and total elongation (e.g.