Considering the best international practices and also tendencies in the development of management of ageing in general (including obsolescence), one can conclude that the required condition and functioning of all SSCs relevant to safety should be ensured via

• analyses defining time limits of safe operation (time-limited ageing anal­yses (TLAAs)) and corresponding monitoring of time limiting assump­tions (e. g. fatigue monitoring)

• ageing management programmes (AMPs)

• environmental qualification and programmes for maintaining the qualification

• maintenance and control of effectiveness of maintenance with respect to safety criteria

• scheduled replacements and reconstructions.

This system should be comprehensive in the sense that ageing of any item in the list of safety-related SSCs, should be covered by at least one of the methods. The required safety function and performance of any selected SSC has to be ensured by one of the approaches listed above or a combination of the methods/programmes (e. g. AMP and TLAA). The safety functions are properly ensured if the non-safety classified items, which may affect the safety functions, are also covered by one of the programmes.

The operator should pay specific attention to those structures and com­ponents, the function and performance of which directly limit plant life­time. These are the non-replaceable or not-to-replace SCs to which either an effective ageing management programme should be applied, or the required functions should be demonstrated for the extended operational time by analysis (fatigue, embrittlement, etc.) or by qualification (e. g. in case of cables).

There are different approaches to how the operator defines which method/ programme or combination thereof is applicable for particular SSCs; the optimization of plant efforts may have economical aspects too. The exist­ing plant programmes might be credited as appropriate for ensuring the required plant condition in the long-term, if they are reviewed and found to be adequate. The concept outlined above is illustrated in Table 8.3.

K. L. MURTY, North Carolina State University, USA and K. RAMASWAMY, Bhabha Atomic Research Center, India

DOI: 10.1533/9780857097453.1.3

Abstract: A typical light water reactor (LWR) has components like the clad, the internals, the reactor pressure vessel (RPV), the heat exchanger tubes, etc., made from different materials. Some of these components experience pressure and temperature effects while others experience an additional contribution from high neutron flux. These components undergo degradation to various extents based on the severity of service conditions and their inherent material properties. This chapter presents an overview of the various deformation modes that materials are known to undergo under reactor operating conditions, and the known theoretical or empirical relations between the crucial material and environmental parameters are outlined. Materials degradation phenomena briefly described in the chapter include radiation damage, plastic deformation, fracture and fatigue, following which radiation effects on these phenomena, as well as corrosion are enumerated. Degradation mechanisms of concern to specific nuclear reactor structures are detailed in the last section with emphasis on fuel, cladding and internals.

Key words: nuclear reactor, damage, degradation, hydride embrittlement, life prediction, mechanical property, creep, fatigue, fracture, irradiation creep, corrosion, irradiation assisted stress corrosion.

1.1 Introduction

This chapter provides a review of materials ageing and degradation encoun­tered in light water reactors (LWRs). Ageing of any engineering structure — through exposure to pressure, temperature and environment — can manifest as changes in the material properties which may be classified into three major categories: (1) changes in dimensions or shape of the structure, (2) changes in material weight due to oxidation, corrosion and erosion and (3) changes in physical or mechanical properties without any noticeable change in dimensions. The in-service component(s) may undergo more than one of the above changes simultaneously, and when these changes affect plant safety, production efficiency or economy they are viewed as degradation. In a thermal energy based power plant (nuclear or fossil fired), various energy transfer stages with complex heavy engineering are involved before the final stage generation of electric energy is achieved. At each stage of energy trans­fer, the machinery involved undergoes ageing and the material properties undergo degradation with continuous use. The severity of degradation may vary from simple and minor to serious and complex. For the core components of a nuclear reactor in a nuclear power plant (NPP), there is an additional influence of the severe radiation environment that accelerates the ageing. The types of nuclear reactors vary in their design features according to the type of fuel and coolant used. The choice of materials for their construction differs according to the reactor design as well as to previous experience in operating nuclear reactors. The components in the reactor core must toler­ate exposure to the coolant media (high-temperature water, liquid metals, gas or liquid salts), stresses and vibrations as well as an intense field of high energy neutrons. Ageing of materials under this extreme environment can lead to reduced performance and, in the worst cases, sudden failure of the components.

A common consideration given in a power plant design at any installation (nuclear/thermal) is the safety requirement. The concern for safety increases as the material properties get degraded from their initial values with pro­longed exposure to service conditions. Thus, intermittent surveillance cam­paigns are mandatory in an operating installation for the evaluation of the health of the components — even if the initial design adhered to strict safety norms. For this, we must be able to identify the critical components that can possibly undergo ageing degradation and decide the frequency of the inspection campaigns. The outcome of such campaigns can forewarn of any impending failure and suggest replacement of components such that the designed life of the plant can be reached — and if reached, the campaign can advise if the life of the component can be extended beyond the design life. In the worst case, the campaign outcome may suggest shutdown of the plant if safe continued operation of the component cannot be ensured. The cost of such campaigns and subsequent component replacements should be recov­erable by putting the plant back into operation. The following statement with regard to nuclear installations is pertinent:1

With the present 60-year licenses beginning to expire between the years of 2029 and 2039 for the first group of NPP that came online between 1969 and 1979 utilities are likely to initiate planning of base-load replacement power by 2014 or earlier. If the option to extend current plant lifetimes is not available, strategic planning and investment required to maintain the current LWR fleet may not happen in a sustainable manner. The research window for support­ing the utility’s decisions to invest in lifetime extension and to support NRC decisions to extend the license must start now and is likely to extend through the following 20-year period (i. e. 2010 to 2029), with higher intensity for the first 10 years. The LWR’s R&D Program represents the beginning of timely

collaborative research needed to retain the existing nuclear power infrastruc­ture of the United States.

Our understanding of the behaviour of the service material in that envi­ronment is based on years of operating experience of a reactor. Sustained research and development is required to develop newer materials. Further, it is from the examination of the ageing/aged materials we learn the role of new environmental parameters that were unthought-of during the design stage, and allow us to modify our safety codes in future designs. Specific ageing and degradation mechanism depends on the component in question and the various conditions such as temperature, load and environment to which the materials are exposed. A typical NPP can be considered to con­sist of seven different components: (i) fuel, (ii) structural components, (iii) moderator/reflector, (iv) control, (v) coolant, (vi) shields and (vii) safety systems. Each of these components has specific requirements and selection criteria based on which suitable and economic materials are chosen.2

Fission based nuclear reactors can be classified as thermal and fast, based on the energy of the neutrons and the thermal reactors can further be cat­egorized as boiling water reactor (BWR) and pressurized water reactor (PWR). The latter type can further be classified as light water cooled and heavy water cooled. We will confine ourselves here to the LWRs that use the steam-cycle conversion system wherein the steam produced by nuclear fis­sion drives a conventional turbine generator to produce electricity. A steam generator is used in PWRs to produce steam while the direct cycle BWRs generate steam in the reactor core thereby not requiring a separate steam generator; Fig. 1.1a and 1.1b are schematics of PWR and BWR, respectively, with important structural components indicated.3

In a typical PWR which uses ceramic fuel, the fuel is separated from the coolant by a physical barrier that prevents their direct contact. The barrier, called the clad, has adequate thermal conductivity to transfer the fission — heat to the coolant and has a low thermal neutron absorption characteris­tic to allow fission neutrons to sustain a chain reaction. The cladded fuel is immersed in a pressurized pool of coolant flowing at an average temper­ature of ~300°C under a pressure of around 16 MPa thus preventing the water from boiling. Two separate water systems, the primary and secondary, are contained in the steam generator which is a heat exchanger consist­ing of a large number (~3000) of nickel-based super alloy tubes in a large steel shell. Depending on the vendor, PWRs may have two, three or four loops with respective coolant circuits, each with its own steam generator. In BWRs, on the other hand, water is circulated through the reactor core producing saturated steam that runs the turbine generator. Nuclear reaction in BWRs is controlled using steel-clad boron carbide control rods that are inserted from the bottom of the core while control rod cluster assemblies

containing either B4C or AgInCd are inserted from the top in PWRs. While the PWRs contain about the same weight of fuel and cladding as BWRs, the number of assemblies is one third of that in the BWRs since the number of rods per assembly is greater in the former (17 x 17 in PWRs vs 8 x 8 in BWRs — these numbers vary slightly depending on the vendor and the gen­eration type). Table 1.1 summarizes the important characteristics of these reactors that include Russian VVER and RBMK.

As outlined earlier and presented in Fig. 1.1a and 1.1b, different materi­als are selected for the manufacture of various components and the selection

criteria are based on physical, mechanical, thermal and nuclear characteris­tics including the chemical and nuclear stability as well as the resistance to radiation damage and induced radioactivity. Table 1.2 summarizes the various components and major requirements along with possible materials. Based on

these various possibilities the reactor vendors select different materials for the construction of the structures in LWRs and it is very interesting to follow the history of the development of LWR fuel cladding that resulted in the adoption of Zr-based alloys such as Zircaloy-2 for BWRs and Zircaloy-4 for PWRs.4

Table 1.3 provides a summary of the reactor components, the key materi­als, the major materials-related problems and the primary causes of failure.3 These materials face different environmental parameters whose severity varies with location. This leads to differences in their ageing mechanisms and the intensity of their degradation. It is possible that one mechanism can be a precursor to another, leading to unexpected early failure of the material. The common ageing-related degradation can be classified as due to radiation embrittlement, loss of toughness, time dependent deforma­tion (creep), fatigue, radiation growth, thermal ageing, corrosion, oxida­tion, stress corrosion cracking (SCC), intergranular and irradiation assisted stress corrosion cracking (IGSCC/IASCC). The dominant mechanism for the different structures will vary with environments. These phenomena are described in the sections which follow.

and application

In this section, we discuss in further detail management techniques for both reactor vessels and internals as well as steam generator tubes, pressurizer nozzles and the CDRM and finally, look at applied management practice around the world.

7.1.1 Management techniques for a reactor vessel

In a pressure vessel and its internals, degradation areas are welds of beltline regions, inlet-outlet nozzles, CRDM, instrumentation nozzles and flange closure studs. The degradation mechanisms are largely radiation embrittle­ment, fatigue, IGSCC and boric acid corrosion.

Embrittlement of pressure vessels is a more significant problem in PWR than in boiling water reactors (BWRs). This is, because in a PWR the layer of coolant around the core is thinner, so the PWR core generates a 20-100 times greater neutron fluence. The current design of RPVs does not feature welded joints in the beltline region, as this is the most radiation-embrittled zone, but in older vessels there are both circumferential and axial welds in this area since vessels were manufactured from plates. Current materials regulations describe the application of low copper materials and low-alloy steel of SA533B-1 for the fabrication of pressure vessels, so that the parent metal in the beltline part of the shell is damage resistant. In older type ves­sels, the most important issue is radiation embrittlement around the weld zone of the beltline area. The weld zone can easily become more embrittled than the parent metal not only because copper, nickel and phosphorous impurities are present, but also because it is the point of connection of var­ious metals and the heat-affected zone (HAZ). When materials have been embrittled, the nil ductility transition temperature or the ductile-brittle transition temperature increases, and the upper shelf energy (USE) value from the Charpy impact test decreases. As a result, the permissible pres­sure temperature (PT) of power plants is limited. Damage by fatigue crack­ing occurs in the beltline weld zone (under normal operating pressure/ heat cycle and abnormal events), closure head studs (during loading cycles in normal operation and repair), primary coolant entrance and exit noz­zle (under heat cycling) and penetration and CRD housing (under heat cycling). Heat cycling can occur in normal operation during the heat-up or cool-down phases associated with servicing, or may be unexpected (Morgan and Livingston, 1995).

Degradation management strategies can be categorized as: mitigation, inspection and surveillance, or repair, as in Table 7.2.

RPV components	Ageing mechanisms	Management categories	Management techniques Description of techniques

For thermal stress reduction, a device that inspects the state should be installed or risk abnormal events. To reduce damage from neutron radia­tion, the low leakage fuel loading technique can be applied which minimizes the influence of the neutrons on the materials of the pressure vessel through appropriate arrangement of burned and fresh nuclear fuel. The thermal annealing technique which returns the hardened pressure vessel materials to the nature of their raw materials can be carried out near 343°C (650°F) in water or 430°C in air.

There are two kinds of in-service inspections for vessels: ultrasonic testing and acoustic emission testing. The ultrasonic testing is described in ASME Section XI, and it is used to characterize cracks of the HAZ and weld zone. The uncertainties in this method, especially when it is used on cracks under cladding, have resulted in conservative regulatory requirements for use of these flaw estimates to set the permissible PT limits and evaluate pressurized thermal shock (PTS) events. ASME Section XI requires four inspections every ten years, and during this period, it recommends 100% volumetric inspection on repair welds on all shells, heads and flanges in the shell, nozzles in the vessel and beltline parts (Morgan and Livingston, 1995). This enables closer monitoring at the beginning and growth of potential fatigue cracks. The sharp cracks found on the surface of the vessel or in the embrittled beltline are most important to PTS but it is difficult to detect or inspect these cracks. Some studies have developed advanced ultrasonic techniques for this purpose (Shah and MacDonald, 1993). Acoustic emis­sion monitoring can be used in online monitoring the growth of cracks if the surface of a vessel is accessible (Morgan and Livingston, 1995).

A system for the verification of the effectiveness of AMPs and feedback of experience has to be in place at plants. In the case of any damage dis­covered, the degradation mechanism should be identified followed by an evaluation of whether the given degradation mechanism is appropriately managed by the AMP(s).

Following the introduction to various degradation phenomena, fundamen­tals of radiation damage and radiation effects, we now turn our attention to

specific structures and materials that experience degradation during reactor operation.

The V-179, V-230 and V-213 types of VVER plants are equipped with a six-loop VVER-440 reactor. In each loop, there are main isolating valves (MIV) on the cold and hot legs, one main circulation pump (MCP) per loop and horizontal steam generators (SG). The pressurizer, with safety valves, is connected to the primary loop. The two generations of the VVER-440 type of reactors have very similar layouts in their primary systems. Typical oper­ating parameters are Thot=297°C, Tcold=266°C, p=12.3 MPa. However, the design bases of the VVER-440/230 and the VVER-440/213 are essentially different, which manifests in the design of safety systems and confinement (IAEA, 1992; 1996a).

There are 16 nuclear power plant units of type VVER-440/213, namely, four in Hungary, four in the Czech Republic, four in Slovakia, two in Russia and two in Ukraine. The owners of these plants are preparing for the LTO of these units.

The design bases for the VVER-440/213 safety systems are similar to those used in Western PWRs, including the postulating of a double-end guillotine break of the main circulation line in the reactor coolant sys­tem. The safety systems exhibit triple redundancy and the reactors have bubbler condenser-type, pressure suppression containments capable of withstanding the imposed loads and maintaining containment func­tionality, even following large break LOCA events. The design of the VVER-440/213 plants considered internal and external hazards to some extent. Protection against single failures in the auxiliary and safety sys­tems has generally been provided in the design. The safety concerns with VVER-440/213 plants are discussed in the IAEA report (1996a). The VVER-440/213 has essentially inherent safety characteristics, for example robustness of the design, low heat flux in the core, large water inventory in the primary system and a large containment volume, which compen­sates to a large extent for other deficiencies in the containment concept. At all of the plants, most of the safety deficiencies have been addressed by retro-fitting and plant modifications. Due to the robustness of the design, it was feasible to upgrade the safety of the original VVER-440/213 design to a level comparable with the PWR plants of the same age. The latest constructed units of VVER-440/213, such as Mochovce NPP Units 1 and 2, had several improvements and modifications made during the design and construction phase.

There are specific modifications of the VVER-440 design: the Finnish nuclear power plant at Loviisa, represents a combination of the VVER-440/230 basic design and nuclear island equipment with a Westinghouse-type, reduced pressure, ice-condenser containment and sev­eral other western-designed and manufactured systems, like the complete

instrumentation and control (I&C) systems. These units have a very suc­cessful operational history and excellent safety features. A comprehensive lifetime management programme was launched in the very early stages of operation, and has allowed LTO of the Loviisa units. The Armenian reac­tor also represents a modification of VVER-440 with an enhanced seismic capacity. The shut down Units 3 and 4 at Kozloduy NPP, Bulgaria represent an intermediate type between 230 and 213 series.

I t should be noted that, the VVER-440s have certain inherent safety characteristics that are superior to most modern PWR plants, for example robust design, large water inventory in the primary system relative to the reactor power and large volume of the confinement.

There is a synergy between the possibility of LTO and different plant actions and measures implemented for safety upgrading, power up-rate, improving reliability and plant programmes. Implementation of the safety upgrading programme for ensuring the compliance with national and international requirements is a precondition for LTO. At the same time, safety is the most important aspect of public acceptance. The operator commitment in relation to safety is and will be the decisive point of judgement by the public.

Most of the safety upgrading measures result in positive technical effects too. Due to these modifications, the safety systems or essential parts thereof had been practically renewed or reconstructed. Consequently, a large part of safety systems is un-aged. In some cases, the safety upgrading measures have a direct influence on the lifetime limiting processes. For example, the new relief valves installed on the pressurizer for the cold over-pressurization protection eliminate the danger of brittle fracture of the reactor vessel. Some of the VVER plants implemented an extensive seismic upgrading programme involving the addition of a large number of new seismic fixes and other strengthening measures (see papers in IAEA, 1993). Fixing the building structures, the anchorage equipment, cabinets, racks and also the structural support of cable trays can be considered as reconstruction of these SCs.

The most important economic condition for LTO is preserving the present cost advantage of nuclear electricity generation within the mar­ket conditions. By exploiting reserves and advantageous features of the VVER-440/213 reactors, the electrical output of the plants can be safely increased up to approximately 500 MWe by improving the efficiency of the secondary circuit/turbine and increasing reactor thermal power via imple­mentation of modernized fuel assemblies. Obviously the power up-rate should not result in a decrease of the plant safety level and should not cause stressors of ageing which affect the lifetime extension perspectives and the plant availability.

The VVER plants replaced the frequently criticized, obsolete I&C sys­tems. The new I&C systems have proper environmental qualification. Aside from their obsolescence, the lack of environmental qualification was the basic issue in the case of the old systems at practically all plants.

One of the major causes of corrosion in the steam generator heat-exchange tubes local is the high concentration level of corrosion activators (chloride ions, sulphates, copper oxides, etc.) in the secondary circuit and partially in the hidden surfaces of the SG secondary side locations. This can be critical in the case of VVER-440 plants where the steam generators are not prac­tically replaceable. To limit local corrosion, the high levels of deposition on tube surfaces should be eliminated to reduce the concentration of the cor­rosion activators. The most important measure implemented was to replace the main turbine condenser, for example at Paks NPP (Katona et al, 2003). Unlike the old condensers with a copper alloy tube bundle, the new con­densers with stainless steel tubing are leak tight. They in turn allowed the introduction of the high pH water regime in the secondary circuit providing better operational conditions for components of the feed-water system and for the steam generators as well.

Some AMPs are based around addressing a particular degradation mecha­nism, examples of which are shown in Table 8.6.

Structure — or component-oriented AMP

Applying the graded approach, the SCs can be separated into two categories:

1 Highly important from a safety point of view, items with complex fea­tures and ageing mechanisms.

2 Items which have the same type, safety class, identical design features, materials, operating circumstances and dominating ageing mechanism could be grouped into commodity groups and for each commodity group a designated AMP can be implemented, for example pipelines, pipe ele­ments, valves, heat exchangers, etc.

The highly important SCs like the reactor pressure vessel together with internals or components of main circulating loop (SCs of Safety Class 1 and some SCs of Class 2) can have dedicated, individual AMPs, for example:

[17] Reactor pressure vessels

• Steam generators

• Reactor pressure vessel internals

• Pressurizers

• Main circulation pipeline

• Main coolant pumps

• Main gate valves.

[18] the analysis remains valid for the period of LTO;

• the analysis has been projected to the end of the period of LTO via removing the conservatism used in the TLAA analysis by less conserva­tive assumptions and methods for analysis; or

• the effects of ageing on the intended function(s) will be adequately man­aged for the period of LTO.

Scope of ageing management

The scope of the review covers the following SSCs:

• SCs relevant for safety — Classes 1, 2 and 3

• those non-safety SCs which can jeopardize the safety functions.

The non-safety-related SSCs which can jeopardize the environment (oil pipelines and tanks, containers for storing different chemical substances)

Table 8.3 Concept for ensuring long-term operation

Ensuring

Safety functions/ Production/ Functioning of

performance economy the operating

organization

Review, assessment and amendment of the plant programmes Reconstitution of the TLAAs

— All systems, structures and components (SSCs) have to

be covered by certain plant programme(s), for example preventive/corrective maintenance, ageing management, scheduled replacement

— All ageing mechanisms have to be considered

— All plant activities have to be considered, that is the

routine activities should be integrated with those specific to LTO

— Synergies have to be utilized PLiM Programme for LTO [14] [15]

• Temperature: in the case of organic materials, commonly used as insu­lation and/or sealing parts of components, high temperature is the main factor of ageing

• Radiation: inside the containment, y-rays are mainly taken into account. The most sensitive material is PVC and the least sensitive is XLPE. Therefore, PVC insulated cables are not used for safety-related func­tions inside the containment. Neutron radiation is be considered only for copper parts located next to the reactor, where these parts may be activated

• Pressure changes: extreme pressure changes may occur in loss-of-cool — ant conditions and may endanger the proper operation of systems and components by affecting the sealing materials of some equipment

• Humidity: humidity in the containment may change for several reasons, for example leakage or pipe breakage, unintended operation of fire extinguishing appliances. Penetrating humidity may result in malfunc­tion of electrical and I&C equipment.

• Steam: under LOCA conditions, steam may condense on the surface of equipment causing rapid temperature rise and it may also penetrate into the equipment.

• Chemicals: the applied chemicals (boric acid, hydrazine, etc.) may pene­trate into seals of electrical equipment, reducing dielectric strength, and causing corrosion.

• Seismic events: seismic effects and vibration may degrade the func­tionality of certain electrical and I&C equipment (relays, transmitters, motors, etc.)

Identification of the ageing mechanisms for civil structures and structural

components is discussed by Katona et al. (2009a). Examples are given in

Table 8.5 on the basis of Hungarian regulatory guide No. 1.26.

As shown in Tables 1.2 and 1.3, we note that a large number of materials are used for various components in nuclear power systems. It is important to state here that relatively large structures can only be fabricated using welded joints and the designers need to account for the varied properties of the dif­ferent materials and their welds; often welds are known to be more sensi­tive to radiation and corrosive environments. As pointed out by Roberts,3 in many cases, nuclear grade materials are fabricated to more stringent speci­fications than those for other technologies and are subjected to inspection and surveillance following in-reactor exposures. Stress states experienced by different components vary depending on locations, for example biax­ial stresses in thin-walled cladding tubes and more complex ones in pipes, elbows and their welds; a time dependent constant load leads to creep failure in out-of-core structural components while in-core materials experience irra­diation enhanced damage thereby further shortening their life. All structures, especially massive ones such as reactor pressure vessels (RPVs), invariably contain flaws and cracks that need to be taken into consideration through fracture mechanics and structural integrity analyses. It is therefore necessary to develop appropriate constitutive laws and models taking into account the individual or combined effect of: (i) instantaneous elastic and plastic defor­mation, (ii) time dependent recoverable deformation (anelastic strain), (iii) time dependent plastic deformation (creep), (iv) strain accumulation due to cyclic loading (fatigue), (v) corrosive environment effects, (vi) compositional effects such as dynamic strain ageing leading to premature failures and finally (vi) radiation damage and effects. The common ageing-related degradation mechanisms are described in the subsections which follow while more details are given in various chapters of the book — Part I covers major phenomena and Part II pertains to specific structural components and varied NSSSs.