Category Archives: Materials’ ageing and degradation in. light water reactors

Creep of RPV and internals

For RPV steels which undergo a damage of about 0.1 dpa, deterioration due to irradiation creep is much less in comparison to toughness loss. But creep crack growth studies indicate that the HAZ, with a different micro­structure and coarser grain size than the base metal, can lead to lower life after prolonged neutron exposure in the temperature range 320-420°C.101 Many of the components of PWR internals (screws, core barrel and baf­fle assembly) are made of austenitic stainless steels and undergo an aver­age damage rate of about 1 dpa/year (=5 x 1013 n/cm2s) at a temperature which may reach a maximum of 400°C due to gamma heating. They undergo irradiation-induced creep and stress relaxation.102

In some PWRs the core baffle consists of sheets and formers. The sheets are separated by small gaps (0.2-0.4 mm). The connection between the core baffle sheets and the formers, and between the formers and core bar­rel is completed by a large number of bolts (about 900). During the core baffle manufacturing process the bolts are tightened with well defined pre-stress to guarantee the geometrical and mechanical stability of the structure. During operation the pre-stress of the bolts becomes reduced as a consequence of thermal and mechanical loads aided by neutron irra­diation which can possibly affect the flow induced vibrations of fuel rods in the outer fuel elements.1 03 Biaxial creep rates measured in solution annealed (SA) 304L (used as baffle plates) and cold-worked 316 (used as bolts) in the temperature range 280°C to ~380°C and irradiated to a dose level of 120 dpa indicate that SA 304L creeps faster than CW 316.104 These results suggest that the correct grade of steel and optimum metal­lurgical conditions can reduce the creep rate and prolong the life of these components.

1.4 Conclusions

Structural components in NPPs undergo ageing with continuous opera­tion and eventually reach the end of life. The rate of degradation depends on their inherent ability to withstand the stress, temperature and service environment. To get the best potential from a material the acumen of the designer, the alacrity of the operator and dexterity of the surveillance per­sonnel should play a non-compromising role. The engineering structures in a NPP can be broadly classified into two categories: (i) the components of steam generators, turbines, etc., which experience thermal and mechanical environment and (ii) in-pile components such as fuel clad, reactor pressure vessel, etc., which are subjected to an added condition of intense neutron irradiation. Materials in both categories also have to face high-temperature flowing water, the energy transfer medium, which corrodes/erodes the mate­rial. The feedback data on the performance of materials in these environ­ments help material scientists to modify the materials and to manoeuvre their properties to perform better. This closed cycle needs to be kept active to meet the required technological advancements.

The properties of materials used in LWR power plants are evaluated for the service they have to render: a fluctuating load requires material with good fatigue strength, constant pressure at elevated temperature demands good creep strength and stress relaxation, good toughness is needed even after prolonged neutron irradiation, low tendency to absorb hydrogen so as to minimize hydrogen related problems, etc. It is difficult to have one mate­rial endowed with all these properties and hence more than a dozen materi­als are used inside a reactor — these need to be joined in some way and this adds to corrosion-related problems.

The elastic and plastic deformations of a material, whilst obeying a generic relationship, will show a marginal difference in their properties because of its metallurgical condition. The constants used in these rela­tionships are material — or microstructure-specific. Despite such variability it is still possible to isolate a material with the required microstructure to serve under specified environmental conditions, and above all, for a known life expectancy. An indication of the crack length in a material helps to keep a check on its degradation if its fracture toughness property is known. Charpy impact tests provide an easier alternative to LEFM tests and are used to grade the deterioration of the material. In situations where the initial toughness of a material is unknown for comparison, the master curve technique is convenient to evaluate the irradiation embrittlement of steels. The growth rate of a crack can be estimated from the known rela­tion between the crack length and applied stress. As irradiation is known to benefit HCF and, as the material behaviour under HCF is well under­stood, a prudent design for longer life becomes possible. Knowledge on the creep rate of a material alerts for corrective measures as the dimensional changes are predictable. The activation energy for creep indicates which metallurgical parameter is crucial in limiting the life. Resolving the yield stress into a source hardening and frictional terms helps understanding of the flow response of the material to nuclear irradiation. It is now known that synergistic effects of neutron irradiation and DSA could lead to bene­ficial effects on strength and ductility in certain temperature and strain-rate regimes. By making a judicial choice of the temperature and fluence, a steel can be safely used in the blue brittleness range. Understanding the metal­lurgical treatment and the material response has helped in choosing the right material such as SA 304L instead of CW 316 for better creep resis­tance for baffle plates. In Zr-2.5%Nb alloy, the stable в phase (80%Nb) is seen to be less creep resistant than the в phase (35% Nb) and the pressure tubes (in Pressurized Heavy Water Reactors (PHWRs)) can have a longer life with this modification.

Corrosion is another major problem in nuclear reactors. Uniform, nod­ular and shadow corrosion that affects the reactor components, and which are not influenced by any external stress, are controlled by modifying alloy and water chemistries. Routine surveillance test programmes enable better understanding of material behaviour. This has helped to substitute some of the components which suffer from SCC with those having better resistance (e. g. Alloy 690, 52,152). IASCC is known to occur in almost all materials and in components at low stress levels and this phenomenon is yet to be under­stood well to come out with effective solution.

This chapter serves as an introduction to the various materials degrada­tion phenomena as summarized above while the subsequent chapters dwell on various details with Part I on various fundamental phenomena, Part II on specific and varied components of LWRs while Part III covers manage­ment strategies adopted by various nuclear utilities/vendors.

[1] Carbon steels are cheap iron-base metals with less than 1% of alloying element present. These materials exhibit a poor resistance to corrosion but their forming, machining and welding are superior.

• Low-alloy steels are iron-base metals containing a few percent of, for example, nickel, chromium, molybdenum, vanadium, which are usually

[2] solubility of H in the specific alloy at its specific burnup that will deter­mine the amount of H in solution at the maximum temperature and the amount of circumferential hydrides;

• microstructure features such as grain size, amount of cold work and dis­location structures;

[3] Stress free axial elongation due to irradiation growth.

• Anisotropic creep (before pellet/cladding contact) due to external reactor system pressure. Because of the tubing texture, axial elongation generally results from creep down of the cladding diameter; however for

[4] Long storage time — 40 years or more.

• Inert gas, helium (He) storage atmosphere instead of pressurized or boiling water (decreased heat transfer, but no corrosion).

• Decay heat that can raise the cladding temperature to 400°C or higher, then decreases over time.

• Atmospheric storage pressure that, combined with high fuel rod temper­ature and internal gas pressure, results in a high clad AP and clad stresses that decrease with time as well.

• No external radiation (no additional radiation damage).

• Dry cask storage containers dissipate the fuel decay heat by natural con­vection of the cask He atmosphere and conduction through the cask container walls; there are no moving parts or forced cooling in this sys­tem. As a result, the cladding can reach temperatures of several hundred degrees Celsius. The pressure differential across the cladding can be sig­nificant since the fuel rod internal gas pressure is made up of (Adamson et al.,2010):

[5] Hoop stress.

• Maximum temperature.

• Cool-down rate and final temperature.

• Solubility of H in the specific alloy at its specific burnup that will deter­mine the amount of H in solution at the maximum temperature and the amount of circumferential hydrides.

[6] In PWRs it is found that Zircaloy-4 no longer meets corrosion and hydriding needs therefore virtually all current PWR cladding uses a zir­conium alloy containing Nb.

[7]ln-situ:Test that can be performed without disconnecting the end device while the cable is installed in its normal environment.

[8] Copyright information: Please note that some material in the following sections has been published previously in the cited articles:

Section 7.2 (IAEA 2011. Chapter 3) Reprinted with permission from the International Atomic Energy Agency.

Sections 7.2.1 and 7.3.3 (Shah and MacDonald 1993. Chapter 3.6 and 3.7) Reprinted with per­mission from © Elsevier 1993.

Sections 7.2.1, 7.2.2 and 7.2.3 (Morgan and Livingston 1995. Chapter 2.1.2, 3.1, 4.2 and 4.5) Courtesy Pacific Northwest National Laboratory, operated by Battelle Memorial Institute for the U. S. Department of Energy.


[9] Copyright information: Please note that some material in the following sections has been published previously in the cited articles:

Section 7.2 (IAEA 2011. Chapter 3) Reprinted with permission from the International Atomic Energy Agency.

Sections 7.2.1 and 7.3.3 (Shah and MacDonald 1993. Chapter 3.6 and 3.7) Reprinted with per­mission from © Elsevier 1993.

Sections 7.2.1, 7.2.2 and 7.2.3 (Morgan and Livingston 1995. Chapter 2.1.2, 3.1, 4.2 and 4.5) Courtesy Pacific Northwest National Laboratory, operated by Battelle Memorial Institute for the U. S. Department of Energy.


[10] symptoms of leakages

• condition of the insulation

• paint condition

• condition of unpainted surfaces

• condition of welding

• condition of components at junction points of different materials

• condition of bolted j oints.

[11] Reducing neutron flux on the RPV, low leakage core design, dummy shielding assemblies.

• Annealing, that is effecting a change of material properties.

[12] Replacement of the condensers: the new condensers have austenitic stainless steel tubes.

• Removal of copper and copper-bearing alloys from the secondary circuit.

• Replacement of the feed-water distributor (the old one was manufac­tured from carbon steel).

• Cleaning the heat exchanging surface of the SGs.

[13] cracking at headers of the cold collectors of the heat-exchange tubes

• degradation of the welded zone at hot collector headers

should also be considered when identifying the scope of AMPs. In some VVER operating countries, the SSCs which are important for production are also within the scope of AMP (e. g. turbine, cooling water system).

Materials management strategies for VVER reactors

T. J. KATONA, MVM Paks Nuclear Power Plant Ltd, Hungary

DOI: 10.1533/9780857097453.3.335

Abstract: The strategic goal of the VVER operator is to extend its operational lifetime beyond the design life. Here, technical and regulatory conditions and methods for ensuring long-term operation of the VVER plant are presented plus an overview of the basic technical design features of VVER relevant to long-term operation. Degradation mechanisms of structures and components which limit the operational lifetime of the plants are identified. The method for evaluating ageing of the plant, a review of existing plant activities for ensuring the required performance of safety-related systems, development of ageing management programmes and other related plant programmes are described. The integration of plant programmes into a system that ensures safe long-term operation is shown through examples. Trends and need for future research are presented.

Key words: VVER, ageing mechanism, ageing management, long-term operation, in-service inspection, maintenance, environmental qualification, time-limited ageing analyses.

7.2 Introduction

The VVER reactors (Vodo-Vodyanoi Energetichesky Reaktor, which trans­lates as Water moderated Water Cooled Energetic Reactor or WWER) are light water moderated and cooled, that is, pressurized water reactors (PWRs). A summary of basic features of VVER reactors is given by Katona (2010, 2011). VVERs were developed in the 1960s. The first three were built in Russia and Eastern Germany in the period 1964-1970, and operated up to 1990. There are 52 Russian-designed, VVER-type, pressurized water nuclear power plants operating in the world today, out of a global total of 443 nuclear power plants (for the latest operational statistics on VVER plants, see IAEA PRIS database) (IAEA PRIS, 2011). The cumulative time of safe operation of VVER reactors currently exceeds 1200 reactor-years.

The first standard series of VVERs had a nominal electrical capacity of 440 MW (and are therefore referred to as 440 units, 440 reactors, 440 designs, etc.) and reactors in the second standard series have a capacity of 1000

MW (and are thus referred to as 1000 units, etc.). There are two basic types of VVER-440 reactors, which are based on different safety philosophies. The VVER-440/230 design comprise the Generation I reactors, while the VVER-440/213 represents the Generation II reactor design with reduced pressure containment. There are two specific VVER-440 designs currently in operation: the Finnish Loviisa NPP with reduced pressure western-type containment and the Armenian Medzamor NPP. In the VVER 1000 MW series, there was a gradual design development through the five oldest plants (small series), while the rest of the operating plants represent the standard­ized VVER-1000/320 model. More VVER-1000 units were commissioned recently and those currently under construction are improved versions of the VVER-1000/320. For example, the Tianwan (China) plant with AES-91 type units and the Kudankulam (India) plant with AES-92 type units. New VVER models, such as the AES-2006 design, are being considered for future bids; these new evolutionary models of large VVERs already exhibit Generation III features.

The design operational lifetime of the VVER plants is generally 30 years, with the exception of the new VVER-1000 type units which have 50 or 60 years of designed operational lifetime. A great majority of VVER plants are quite old, nearing the end of their design lifetime, except for some in Russia. The VVER operating countries are dependent on nuclear power production, for example the Paks Nuclear Power Plant in Hungary pro­vided 40% of domestic production in 2010. The nuclear power capacities in these countries ensure the necessary diversity of power generation and contribute to the security of supply. Therefore, the VVER owners in Central and Eastern Europe are keeping their plants in operation via implement­ing plant lifetime management (PLiM) programmes, with the intention of ensuring a safe and financially viable operation in the long term. The PLiM practice of VVER plants is presented by Katona (2010) and Katona and Ratkai (2008,2010).

The possibility of extending the operational lifetime of VVER-440/213 plants was recognized in 1992. It was based on an assessment of the robust­ness of the design, good technical condition of the plants and synergy between safety upgrading measures and overall condition of the plants (Katona and Bajsz, 1992). In all VVER operating countries, lifetime man­agement had the explicit goal of ensuring the extension of operational life­time (Rosenergoatom, 2003).

The operational licence of the four VVER-440/213 units at Paks NPP in Hungary, is nominally limited to the design lifetime of 30 years. Extension of the lifetime of this particular plant by an additional 20 years is feasible. The first formal step of licence renewal of the Paks NPP was made in 2008 and the relicensing process is still ongoing. In Ukraine, the nuclear share of domestic production of electricity is approximately 48%, while this nuclear power plant comprises 26.6% of total installed capacity. There is a keen interest in extending the operational lifetime of all Ukrainian NPPs. The operational licence of the VVER-440/213 type Units 1 and 2 at Rivne NPP in Ukraine has been renewed by an additional 20 years with the condition of performing a safety assessment after ten years of prolonged operation. The extension of operational lifetime is a generic strategy of operators of VVER-440/213 plants in the Czech Republic and Slovakia. The Loviisa NPP in Finland (a non-standard VVER-440 design) has been allowed to prolong operation up to its next Periodic Safety Review (10 years).

The operational lifetime of the VVER plants in Russia will be extended by 15-25 years. The four oldest VVER-440/230 units, Novovoronezh NPP Units 3 and 4 and Kola NPP Units 1 and 2, have already received a 15 year licence for extended operation. The VVER-440/213 type units (Kola NPP Units 3 and 4) are also prepared for 15 years extension to the operational licence. Among VVER-1000 plants, Novovoronesh Unit 5 is prepared for a 25-year extension of operation, after an extensive safety upgrading and modernization programme.

The VVER operators performed a comprehensive assessment of plant condition and safety, while making their decisions about the extension of operational lifetime. A decision on the preparation of feasibility studies for long-term operation (LTO), was based on the recognition of the following VVER features and experiences:

• robust design of VVER plants

• good plant condition due to well-developed maintenance, in-ser­vice inspections, careful operation and extensive modernization and reconstruction

• implementation of safety upgrading measures, resulting in an acceptable level of safety.

Safety of the plants and compliance with international standards has been considered as the decisive precondition for LTO. The comprehensive modernization and safety upgrading programmes (Vamos, 1999) imple­mented by the VVER operators during the last two decades, resulted in gradual decreases in the CDF of these plants. The level 1 probabi­listic safety analysis (PSA) study establishes the resulting CDF for all VVER-440/213 units at Dukovany NPP of 1.47-1.67 x 10-5/a, as stated in national reports compiled under the Safety Convention (Czech National Report, 2010). The same achievements are published for other VVER plants. Extensive modernization and safety upgrading programmes have been implemented in Ukraine (2011), Russia (Rosenergoatom, 2003) and Bulgaria (Popov, 2007). The safety deficiencies do not inhibit the LTO of the VVER plants; the VVER operators have a strong commitment to continuous improvement of safety and are ready to meet the future chal­lenges in this respect.

One of the issues related to the current licensing basis at VVER plants out­side of Russia was the inadequate knowledge of the design basis. The design of VVER-440/213 and the older VVER-1000 plants was generally based on the former USSR regulations of the early 1970s, the General Requirements on Safety of NPP Design, Construction and Operation (OPB-73) and the General Safety Rules for Atomic Power Plants (PBYa-74). OPB-73 marked the beginning of a transition to the generally accepted international practice in nuclear safety (e. g. defence in depth, single failure criterion). Knowledge of the design base is absolutely critical for the preparation of LTO and licence renewal, especially for the review of time-limited ageing analyses. Operators of VVER-440/213 units have to perform a specific project for design base reconstitution. In many countries, the design base has to be entirely recreated, taking into account all essential changes in the licensing requirements. For example, in the case of the Paks NPP, seismic loads had not been considered in its design. The current design/licensing base includes safe shutdown during an earthquake with 0.25 g horizontal acceleration. Availability of a state-of-the-art Final Safety Analysis Report (FSAR), and regular updating thereof is required for the control of compliance with the current licensing basis and configuration management.

The condition of the plant and appropriate plant programmes are also preconditions for LTO, especially surveillance of reactor pressure vessel (RPV) embrittlement and monitoring the condition of long-lived pas­sive structures and components. The most important ageing management (AM) activities are performed at the VVER plants from the very begin­ning of their operation. The early AM activity was focused on known degradation of the main systems, structures and components (SSCs), like the RPV embrittlement, or on the early recognized issues, for example leaking of the confinement due to the liner degradation, outer surface corrosion of the steam generator heat-exchange tubes. Most of the early AM programmes were state-of-the-art, for example the RPV surveillance programme. In the course of the first periodic safety reviews, the defi­nition of the most critical SSCs for operational lifetime and the domi­nating ageing mechanisms were explained. Adequate assessment of the aged condition and forecast of safe lifetime of structures and components (SCs) can only be performed if the ageing process is monitored properly from the very beginning of the operation. The operational history of SCs has to be documented in sufficient detail for the trends in ageing to be discovered.

There are several non-technical conditions which affected the strategy of VVER operators and can be considered as motivation for the decisions on LTO. The positive international tendencies, with regard to LTO of existing nuclear power generation capacities, stimulated the LTO of VVERs too. (This tendency might be changed by the Fukushima nuclear accident fol­lowing the Great Tohuku earthquake in Japan March 2011.) Accumulation of the experiences and scientific evidence for justification for longer than designed operation of NPPs, provides a good basis for LTO of the VVER. Good market positions of NPPs overall in the VVER operating countries, with high levels of public acceptance and positive public attitudes, help in supporting the operation of NPPs in these countries.

Considerable progress has been achieved at VVER plants with respect to the improvement of the performance and plant reliability. The load factor of the majority of VVER plants is over 80%; in some places for example at Paks and at Dukovany NPP it is around 90%.

The national regulation for allowing the approval of an extension beyond designed operational lifetime is also a condition of the LTO. According to Svab (2007) and IAEA (2006, 2007a), there are two prin­cipal regulatory approaches to LTO, depending on the legislation for the operational licence. The operational licence in VVER operating countries may be either limited or unlimited in time. In countries where the oper­ational licence is not time limited, the basis of regulatory approval is the periodic safety review (PSR). In those countries where the operational licence has a limited validity in time, a formal renewal of the operational licence is needed.

The internationally accepted rules and requirements regarding PSR are documented in the IAEA Safety Guide NS-G-2.10 (IAEA, 2003). One of the objectives of the PSR is to review the condition of the SSCs, and whether it is adequate to meet their intended safety functions. This includes knowledge of any existing or anticipated ageing and obsolescence of plant systems and equipment. In particular, the objective of the review of PSR Safety Factor 4: ‘Ageing,’ is to determine whether the ageing of SSCs is being effectively managed. This means whether or not the required safety functions are main­tained, and whether an effective ageing management programme is in place for future plant operation (NS-G-2.10 para 4.21 of IAEA, 2003). The design lifetime is a technical limit for the operation, which is based on assumptions by the designer regarding time limit of performance and functionality of systems, structures and components due to ageing. The PSR used for jus­tification of extension of operational lifetime beyond the design lifetime has to demonstrate that the prolonged operation is safe, despite expiration of the design lifetime. It means the PSR has to review all the time limiting analyses made by the designer. When reviewing the ageing of the plant, both programmatic aspects and technical aspects of ageing management should be evaluated. Rules for developing and establishing and attributes for ade­quacy of ageing management programmes are given in the IAEA Safety Guide NS-G-2.12 (IAEA, 2009).

Examples of the licence renewal approach are the Russian and Hungarian cases. For licence renewal, the regulations require the perfor­mance of integrated plant assessment, focusing on the review of plant con­dition, effectiveness of ageing management programmes and validation of time-limited ageing analyses for the extended period of operation. In Hungary, the national rules for licence renewal have been developed on the basis of 10CFR54, the licence renewal rule of the U. S. Nuclear Regulatory Commission. In Russia, the rules are defined within the context of national regulation.

In this chapter — after an overview of the basic technical features of VVER plants — the basic issues and methods for ensuring LTO of VVER plants will be presented. The dominating degradation mechanisms of structures and components limiting the operational lifetime of the plants will be identified, on the basis of operational experience and research results. The method for evaluating the condition of the plant; review of existing plant activities for ensuring the required performance and functionality of safety-related sys­tems, structures and components; and development of ageing management programmes and other related plant programmes are described. Integration of particular plant programmes into a system that ensures safe LTO is shown on the basis of particular examples. Trends and needs for future research are also presented.

The presentation of the ageing issues will focus on the older VVER — 440/213 and VVER-1000 plants. The VVER-440/230 plants (Kozloduy NPP, Bulgaria and Bochunice V1 NPP, Slovakia) are already on permanent shutdown. In contrast to this, the Kola 1 and 2 and Novovoronesh 3 and 4 units in Russia have already received licences to operate for a further 15 years. This was after implementation of modernization and safety enhance­ment programmes (Rosenergoatom, 2003) to cope with the safety issues relevant to this design (IAEA, 1992). The LTO and plant lifetime manage­ment of VVER-440/230 is not a generic practice and will be discussed below, although only to a limited extent. The LTO of the VVER-440/213 plants requires specific engineering effort and will be discussed in detail. From the point of view of LTO, the newly designed and constructed VVER plants are also of less interest. Obviously, they have been designed and manufactured taking into account the ageing lessons learned from operational experience. The question about the need and possibility of longer than designed opera­tion of these plants is not on the agenda today.

Maintenance programmes

Maintenance is the subject of Maintenance Effectiveness Monitoring (MEM), the purpose of which is to control the effectiveness of maintenance on SSCs ensuring they are capable of performing their intended functions. This means ensuring that safety-related SSCs are capable of performing their intended functions; that failures of certain non-safety-related SSCs that could affect safety-related functions will not occur; and failures that could result in scrams or unnecessary actuations of safety-related systems are minimized. The systems within the scope of MEM might be divided into high and low risk-significant categories. The risk significance has been defined quantitatively by PSA or qualitatively by expert judgement. The MEM is an adaptation of 10CFR50.65 for the VVER-440/213 design fea­tures, Hungarian regulatory environment and plant practice.

There are two basic methods applied in the MEM: deterministic method, that is control of maintenance via testing/measuring performance parame­ters of components, and probabilistic method, that is assessing the effective­ness of maintenance via comparison of reliability/availability parameters at the level of component/system or plant. Performance parameters are defined in accordance with safety class and risk significance. The determin­istic method is based on ASME OM Code. For example in case of pumps the performance criteria to be checked are the head, flow-rate and vibration level. Plant level deterministic performance parameters include the capac­ity factor, thermal efficiency of the unit and leakage of the containment (%/day). Risk significance and the probabilistic performance criteria are set on the basis of PSA. High risk significant SSCs are those which are in 90% cut-set, have a high contribution to CDF or high Fussell-Vessely rank. Performance criteria for MEM are based on the reliability or unavailability data of performing safety function. System level performance parameters are, for example, failure rates per demand (failure/start) or run failure rate (failure/time) during operation. Plant level performance parameters are the CDF or some selected contributors to the CDF and other safety fac­tors (unplanned reactor scrams or safety system actuations per year). The MEM is being implemented at Paks; for the implementation of ASME OM Code, the existing in-service and post-maintenance testing programmes of the Paks NPP have to be modified and amended. Probabilistic performance criteria are under development at present. It is expected that the MEM will improve the safety factors and capacity factors for the plant while the main­tenance effort will be optimal. MEM is a prerequisite for license renewal in Hungary, since it provides assurance for the correct functioning of active components.

Ageing of electrical systems and I&C

Electrical components and I&C are replaceable and the required perfor­mance of these commodities can be ensured via maintenance and scheduled replacement. The qualified condition of the electrical and I&C equipment has to be ensured.

Full scope ageing studies had been prepared for the Paks NPP for the fol­lowing electrical and I&C items:

1 Equipment of electric power and transmission systems:

• Bus cabinets

• Overhead-line towers, medium- and high voltage insulators

• LV and HV cables of power supply systems

• Cables for containment electrical penetration

• Cable joints and assemblies

• Enclosed electrical equipment

• Battery packs.

2 Equipment of the technological systems:

• Fixtures for transmitters

• Impulse pipes and assemblies

• Operation monitors

• Relay boards

• Cables for E, I&C equipment

• Cables of containment electrical penetration for E, I&C

• Cable joints and assemblies

• Terminal boxes.

The basic issue at all VVER plants regarding electrical and I&C equip­ment is the lack of or insufficient environmental qualification. Lack of initial qualification of the VVER equipment was recognized in the 1980s at all VVER-440/213 plants as well as at VVER-1000 “small series” and VVER-1000/320 models.

Establishing the initial qualification is understood as a current licensing basis requirement at all VVER plants. This consists of the following steps:

• Definition of environmental parameters characteristic of the installation site.

• In the case of safety equipment, definition of environmental parameters characteristic to the installation site under accidental (loss-of-coolant) conditions.

• Definition of accelerated thermal and radiation ageing test parameters.

• Performing laboratory tests with the above parameters (accelerated thermal and radiation ageing, radiation exposure with accident condi­tion and simulation of loss-of-coolant conditions).

• Performance checks on tested samples to verify conformity with accep­tance criteria.

The maintenance of qualified condition of the cables for harsh environmen­tal conditions is a critical issue at VVER plants.

In regard to the cables, the technical task of qualification is rather dif­ficult. For example at the Paks NPP there are 130 000 cables and among them, several hundred types. The first necessary measure related to the cables was to develop a comprehensive database, instead of having the cable sheets on paper. The database identifies for each cable the safety clas­ses, types and routes. The environmental conditions to which the particular cables are exposed are identified in the database. It also shows whether these safety-related cables are affected by the harsher conditions after accidents. Examples of ageing mechanisms of important cables are shown in Table 8.2.

Table 8.2 Examples of ageing mechanisms of NPP cables


Site of

Mechanism of





XLPE I&C cables

Cover and core

Thermal ageing;

Crack/loss of

in harsh


change of

function under


the material


properties due to heat or irradiation.


6 kV PVC power

Metal structure



cables in

of cables




corrosion of



metal structure

of function/

Cable connection

Corrosion of metal


Increased transit

in harsh


corrosion of joints

resistance of



In its current state, the database at the Paks NPP covers approximately 19 000 safety-related cables. The database allows the formation of commod­ity groups for cables. Currently there are 45 commodity groups related to safety cables. A similar approach is implemented as an example at VVER plants in Ukraine. For each group of cables, a particular sample cable is identified which is under worst-case condition. The condition of the sample cable is monitored during the operation.

The VVER plants replaced the frequently criticized, obsolete I&C sys­tems. At the Paks NPP, nearly all safety-related I&C systems have been replaced: the reactor protection system, Engineered Safety Features Actuation Systems (ESFAS) protection system and load sequencer pro­gramme of diesel generators. The new system is a digital one (Siemens TELEPERM XS) with multiple redundancy and diverse software features, and physical separation of hardware of different trains. The reactor protec­tion logic was also reviewed and modified to assure diverse physical signals for detecting each postulated initiating event and to eliminate unnecessary input and output signals. Similar reconstruction programmes have been implemented in Slovakia and at Russian plants entering into an extended period of operation.

Other nuclear reactor systems

The remainder of the nuclear plant is similar to any other power plant whether the steam is generated using coal, oil, natural gas or nuclear fuel. There will be issues with cooling tower decay, turbine corrosion, generator moisture, condenser corrosion, buried piping, etc. There are some issues that will be unique to nuclear power generation, however. For instance, decay of cable insulation in cable trays due to irradiation and high temperature is an important issue. Replacement of cabling in a nuclear plant is very expensive and takes a very long time and therefore should be avoided if at all feasible. The effects of soil and groundwater on buried cabling and the miles of bur­ied piping on any nuclear plant site are also at issue (INL, 2009).

Another issue that arises comes from the use of secondary sources of water. For instance, so-called gray water (sewage that has been through a waste treatment plant) has been used and is being looked at as cooling water to reduce overall water use by large power plants. This water has the poten­tial for higher and different salt concentrations than drinking water sources and can therefore cause unexpected corrosion or SCC issues (EPRI, 2007). Corrosion and decay of materials in the spent fuel pools, such as the boron containing structures that allow tighter packing of the spent fuel but which suffered unexpected degradation, also needs to be considered.

Outside of the nuclear plant itself but within the nuclear cycle, materials issues arise in such varied areas as zirconium metal production (the graphite receptors, ceramic gas injection nozzles, and ceramic linings in the chlorina­tion of either ZrO2 or zircon sand); the manufacture of nuclear fuel due to the common handling of mixtures of HF and nitric acid for UO2 dissolution; solvent extraction systems; and incineration of radioactive waste materials. For the most part, material issues have been solved in other portions of the cycle, but the ones listed above continue to be very resistant to reasonable materials solutions.

In summary, the issues outside of the nuclear island that require research and development are:

• The effects on cable insulation decay of temperature and radiation inside the containment and due to groundwater or soil for buried cable outside the containment.

• Methods to monitor and repair buried piping and cabling.

• Understanding of the interaction between older materials used in the initial construction of the plant and newer materials that may be used to repair or upgrade current plant systems including spent fuel pools.

• Components in the zirconium conversion from oxide or silicates to chlorides.

• Systems that handle mixed HF and HNO3 in the UO2 fuel manufacture area.

• Development of phenomenological understanding of the behavior of these materials in their respective environments that can be used to pre­dict their behavior over time.

Corrosion-related problems

Corrosion is a major concern for reactor structures because in their con­struction many different materials are used which corrode at different rates by electrochemical effect and the corrosion (pitting, cracking, etc.) is accel­erated by neutron radiation. More importantly, the corrosion products from steam generators, piping and other components are transported through the core and deposit on the fuel rods leading to formation of crud, in turn lead­ing to increased fuel temperature and fuel failure. Corrosion can also lead to deposition of radioactive corrosion products on out-of-pile surfaces of the primary loop (e. g. heat exchanger) which becomes a safety concern for maintenance personnel. Further, the flow of the medium replenishes the concentration and pH at the corroding site which aggravates the corrosion (flow-assisted corrosion).

An important aspect in the degradation of the Zircaloy clads, the sole bar­rier between the hot fuel and coolant, is the oxidation and corrosion problem. The various oxidation processes in Zircaloys have led to major degradation phenomena that are described in detail in subsequent chapters. Stable, adher­ent oxide films form which act as a protective coating and offer resistance to environmental cracking as in the case of stainless steels. However, in the case of Zircaloys long exposures lead to the film flaking — that results in wall thinning — which in some cases may result in through-wall failures. More and specific details can be found in the chapters on Zr-alloys in Part II.

The addition of transition metals (Fe, Cr, Ni) to zirconium was aimed at reducing the severe oxide growth stresses and localized cracking of the grain boundaries that exposes more grains to the corrosive environment. Formation of second phase precipitates (SPPs) containing the transition elements help by aiding uniform oxidation of the grains and preventing localized cracking and spalling as the oxide grows.3739 It has been noted in Zircaloy-4 that the fraction of SPPs decrease and correspondingly the oxide thickness increases with fluence (Fig. 1.27).40


1.27 Dissolution of SPPs with fluence and the increase in oxide layer thickness under BWR condition in Zircaloy-4 at 290°C.41

A few factors under irradiation conditions are recognized as affecting the corrosion of clad material: (1) The increase in thickness of the oxide film (water side) decreases the thermal conductivity of the film and increases the temperature of the zirconium matrix which results in a higher corrosion rate;41 (2) the changes that occur in the microchemistry of the Zircaloy-4 matrix due to irradiation are seen to accelerate weight loss (when com­pared with unirradiated material);3 942 (3) accumulation of lithium in the oxide layer reduces the protective nature of the oxide at the metal oxide interface and enhances the corrosion rate;43 and (4) high concentrations of zirconium-hydride are found in locations where the oxidation of Zircaloy is high.3144 The vast research work conducted over the years has led to some understanding of these problems: contrary to the expectation that irradiation-induced defects can cause breakdown of the protective film on Zircaloy clads,45 the examinations of irradiated clads did not show any evi­dence of oxide damage;4647 though the exact mechanism of lithium induced corrosion of zirconium is not clear, the general understanding of the lithium effect is that Li gets incorporated in solid solution in the ZrO2 and alters the vacancy concentration and distribution in the oxide film; since zirconium alloy corrosion proceeds by oxygen diffusion through the film, the increased number of vacancies should increase the diffusion and, hence, the corrosion rate; further, it has always been noted in irradiated Zircaloys that hydride density is high in locations where the oxide thickness is high; this clearly indi­cates that the cathodic hydriding and anodically favoured oxidation occur


1.28 Temperature profile along the length of a PWR fuel clad results in increased oxide layer thickness.48

independent of each other. Change in fabrication route, which can result in the second phase with a different composition, can reduce the susceptibility to nodular corrosion but can lead to increased uniform corrosion.

The most commonly encountered corrosion types in the nuclear reactor are uniform corrosion, nodular corrosion and shadow corrosion. Uniform corrosion, as the name suggests, is associated with uniform oxide thicken­ing and is commonly seen in PWRs and BWRs. Unlike the PWR environ­ment where dissolved hydrogen is present in the coolant and the oxide films remained uniform over a very large thickness, the intermetallics present in Zircaloy, under the BWR environment, promoted nodular corrosion. The mechanism of oxide layer growth on Zircaloy under an irradiation envi­ronment is complex. Uniform corrosion starts with low burnup and the thickness of the grey oxide layer increases with burnup and operating tem­perature. Unlike in BWRs where the outlet temperature and pressure are limited, PWRs can operate at higher outlet temperatures but with the risk of increased corrosion, and this effect is explicitly seen from the increased oxide thickness with the elevation of the fuel rod. Figure 1.2848 depicts the profile of the oxide thickness layer with the elevation in a typical PWR fuel rod, indicating the increased oxide layer thickness with increase in tempera­ture. The increased turbulence in the coolant close to the spacer grids (which increases the cooling efficiency) and their parasitic absorption of neutrons result in the suppression of clad oxidation at these locations whereas the fuel rod temperature along the length, and hence the oxide thickness, is fairly uniform under a BWR environment.49

A more crucial parameter than metal coolant temperature is the metal oxide interface temperature which is difficult to measure but may be calcu­lated with large uncertainty. A further complexity arises as the thermal con­ductivity reduces with burnup (due to penetration by the coolant into the porosity, cracking and spalling of the oxide films and crud deposition). The measurement of the thermal conductivity of the loose or non-adherent crud layers, which modifies the metal oxide interface temperature, is extremely difficult as the properties of subsequent layers deposited may not be the same. It is known that the thermal conductivity of the crud is higher than the zirconia layer or water or steam50 which in a way increases the heat transfer characteristics.

Uniform corrosion occurs in both PWRs and BWRs. The oxide that forms is uniform in thickness, consists of several different layers and depends on many factors such as initial SPP size, extent of cold work and irradiation, alloy and water chemistries, temperature, local thermohydraulics, etc. The microstructure of the Zircaloys used in BWRs is continuously evolving, leading to dissolution of the SPP and formation of small and thin patches of white oxide on the otherwise black uniform oxide layer, which thicken at an accelerated rate. The sensitivity of nodular corrosion can be related to the second phase particles present in the alloy, though the number of nodules may not bear a one-to-one relation with the number of particles. Nodular corrosion is encountered in BWRs and starts appearing after a few to 100 days from the start of operation and usually saturates at higher expo­sure times. Nodules, in general, do not form in Zircaloys with small SPP sizes (<0.1 pm) but initiate early in materials with large SPPs and grow at a decreasing rate with fluence. Figure 1.2951 shows the appearance of nodular corrosion on the fuel clad of a BWR fuel pin. The shape of nodular corro­sion can be lenticular or spherical and growth in Zircaloy-2 decreases at high burnups.52 The nodular corrosion problem can be eliminated (or delayed) by judiciously controlling the second phase particle sizes through appro­priate в quench treatment although this may enhance uniform corrosion.53 Nodules, whose thickness greatly exceeds the uniformly growing film, are prone to spalling and promote hydrogen pick-up. They can also be a cause of introducing zirconia particles to the coolant. Though PWR and WWER structures are not prone to the nodular corrosion attack, nodular corrosion can be a problem if steam forms at the oxide-coolant interface.54


1.29 Nodular corrosion on the fuel clad of a BWR fuel pin.51

There is another type of localized corrosion in Zircaloys: the enhanced in-reactor corrosion when Zircaloy is placed close to a noble metal (under BWR conditions it is stainless steel or a nickel alloy), and where Zircaloy ‘mimics’ the noble metal corrosion. This is termed as ‘shadow corrosion’.55 The oxide thickness is unusually large and often appears to be particularly dense and uncracked. This localized corrosion is a special case of crevice corrosion and is predominantly seen in BWR components, although there is no direct electrical contact between Zircaloy and the material producing the shadow effect. The oxidation of H2O2 to HO2+H+ at the noble metal surface is balanced by the regeneration of H2O2 on the ZrO2 surface and the coupling between the two metals (Zircaloy and nickel) is maintained by the ionic transport under a concentration gradient. The driving force is the potential difference between the two metals and radiolysis of water is required to sustain this reaction.55 Shadow corrosion is invariably noted in BWRs and not in PWRs where the coolant is high in hydrogen concentra­tion, which in turn reduces or eliminates galvanic potentials between dis­similar alloy components.

Environmentally assisted cracking is another manifestation of corrosion — related problems and is very often encountered in the steam and feed water piping as well as in condensate systems, RPV feed water nozzles and the secondary circuit of LWRs. This process is accelerated by stress (i. e. SCC) and neutron flux (i. e. IASCC). A typical fracture surface of IASCC is shown in Fig. 1.30.56 Attempts are being made to reduce IASCC. Figure 1.31 shows the effect of hydrogen injection into the BWR environment on IASCC of 304 SS. The mechanism of crack growth mitigation by hydrogen injec­tion could be explained by analyzing the corrosion potential of the system. The presence of molecules like H2 O2 and O2 increases the free corrosion potential which falls into the cracking range and hence the crack velocity is enhanced following the slip dissolution model and Faraday’s law. Whereas, when hydrogen is introduced into the environment it helps the recombi­nation of species and thus reduces the corrosion potential far below the cracking range.57

Austenitic stainless steels (e. g. blade sheathing in BWR) at high tem­perature and in a neutron-rich environment (>0.7 dpa), further influenced by higher oxygen levels in the water (BWR environment), exhibit IASCC. Other steels and nickel-base alloys also undergo IASCC at lower stress levels. Another aspect is that IASCC occurs in almost all materials and is known to occur in components at low stress levels. It is an expensive process to detect and repair the affected component. SCC is a major issue of PWR components like steam generator tubes, RPV penetrations, pressurizer noz­zles, etc. While SCC can be controlled by modifying the water chemistry and the composition of the alloy, that is by replacing components with those resistant to SCC (e. g. Alloy 690, 52, 152), IASCC is more complex. Though


1.30 I ntergranular fracture surface morphology of IASCC (304 SS, 3 dpa). Corrosion debris and cracks along grain boundaries can be seen.56


Time, days

1.31 Typical reduction in crack growth rate by the addition of hydrogen in annealed 304 SS.57

both IASCC and IGSCC require external stress, temperature and dissolved oxygen in the water environment, the former is accelerated by neutron radiation. It is recognized that the influence of water chemistry becomes weaker and disappears at high doses (>50 dpa) suggesting that mechanical processes dominate chemical processes in IASCC.58 The major distinctions between IASCC and other environmental cracking phenomena (e. g. SCC) are that in the former (1) the microstructure is modified by fast neutrons with time and (2) the chemistry of the environment is altered by the ionizing radiation. However, the overall stability of water increases with increasing temperature and the yields of molecular decomposition products (H2 , O2 and H2O2) correspondingly reduced.59 Austenitic stainless steel is the major material that has been the subject of IASCC investigation as compared to other grades. In the case of in-core structures, radiolysis increases the elec­trochemical potential in that region where the SCC susceptibility is high. Among the various radiolytic products, H2O2 is the most concentrated spe­cies present in irradiated, aerated water which gives rise to high corrosion potential for stainless steels. However, the critical potential to mitigate SCC of irradiated materials has not yet been established.6061

Slow strain-rate tests have been carried out on type 304 stainless steel with prior thermal sensitization of the grain boundaries (to produce grain boundaries with chromium depletion) that show that the electrochemical potential of stainless steel increased significantly on irradiation in oxygen­ated water but decreased slightly in the hydrogen treated water. Though the mechanism is not fully understood, it is now realized that neither Cr depletion near grain boundaries22 nor RIS (of S or P) at the boundar­ies63 alone plays the detrimental role in IASCC. Further, the low stacking fault energy (SFE) of the matrix leads to localized deformation through dislocation channelling and irradiation has been found to accelerate the IASCC process.64 Studies done under simulated BWR environments to examine the susceptibility of four steels with varying SFE under irradia­tion showed that the one with the highest SFE exhibited good resistance to cracking whilst that with the lowest SFE was seen to be susceptible to cracking at all of the doses studied25 (Fig. 1.32). In 316 SS, the initiation and propagation of IASCC in a water environment depends on the dis­solved hydrogen and the stress required decreases with (a) increase in dissolved hydrogen and (b) decrease in the rate of straining.66 Corrosion problems are equally important in storage and disposal of nuclear wastes where long term safety and reversibility act as guidelines in designing the basic layout of a geological repository. Unlike conventional engineering structures, the ageing and degrading clad tubes should not only serve trou­ble free all through their service life but also maintain their integrity in repository conditions.67

Corrective actions

Any damage not in compliance with the acceptance criterion should be repaired if possible. In the case of fatigue cumulative usage factor (CUF) > 1.0, appropriate fatigue monitoring and a focused in-service inspection programme can be implemented.

Administrative control

The administrative and organization arrangements have to be defined for the performance of ageing management programmes. Appropriate plant procedures have to ensure the planning, staffing, performing, documenting and management control of the AMPs. Proper systems for documentation and reporting have to be established. A proper quality assurance plan also has to be developed for AMPs.

Identification of ageing mechanisms

The identification of ageing mechanisms and their effect on the safety is based on the following information:

• Analysis of the operational experience

◦ Experience at the individual plant: events-related ageing, for example

load cycles.

◦ Experience at plants of the same design.

• Generic industrial experience.

• Research results.

• Analysis of results of destructive and non-destructive tests.

[15] Review of the design assumptions regarding ageing.

After analysing these sources of information, the dominating ageing mecha­nisms, critical locations and measures for ensuring the required status of SCs can be identified. A list of important mechanical systems and compo­nents and the relevant ageing mechanisms are given in Table 8.4.

Examples for the identification of the ageing mechanisms of cables are given in Table 8.2 and more fully listed below in the table.

given structure or component and plant lifetime limiting character of the given ageing mechanisms.

Description of operating VVER reactors

In the sections below, the basic design characteristics of VVER plants are presented. The design and manufacturing features which are relevant from the point of view of LTO are discussed.

Materials management strategies for VVER reactors 341

Maintenance of environmental qualification

Performance and functioning of active systems can be tested during oper­ation and can be ensured via maintenance under maintenance rule (MR), that is evaluation and assessment of the effectiveness of the maintenance along safety criteria, and/or via implementation of the programme for main­taining the environmental qualification (EQ).

For I&C components operating under harsh conditions environmen­tal qualification should be implemented. When the older VVER-440 and

VVER-1000 NPPs were built, a large part of the originally installed elec­trical and I&C equipment did not have initial qualification or the qualifica­tion was not certified properly. The issue was recognized in the first safety reviews (see IAEA, 1992; 1996a; 1996b; 2000). The issue can be resolved in two steps:

• Restoring the initial qualification.

• Maintaining the qualified state of equipment.

The maintenance of the qualification means:

• Control of the capability of equipment to fulfil its safety function through:

◦ periodic testing of systems and components

◦ testing of the equipment following maintenance

◦ results of service routes by maintenance personnel

◦ diagnostics measurements.

• Development and implementation of a scheduled replacement pro­gramme taking into account the requirements for environmental quali­fication in purchasing the new equipment.

• Preventive maintenance of the equipment.

The environmental qualification should be reviewed and validated for the extended operational lifetime. There are different possible outcomes of the review:

• The qualification remains valid for the period of LTO.

• The qualification has been projected to the end of the period of long-term operation.

• The effects of ageing on the intended function(s) have to be adequately managed for the period of LTO via introducing new ageing management programmes.

• Replacement of the equipment.