Materials used for PWR internals include ferritic steel, wrought austenitic stainless steel, cast stainless steel (CASS) and Ni alloys. The internals main­tain the soundness of the geometrical core. The core consists of the upper core structure, core baffle/former/barrel, thermal shield and lower core sup­port structure. The factors which influence degradation of these parts are: thermal plant transient, flow-induced vibration, radiation, high temperature, mechanical and thermal stress, and corrosive coolant. The main degradation mechanisms are: fatigue; radiation and thermal embrittlement; void swell­ing; and irradiation assisted stress corrosion cracking (IASCC). IASCC is a type of SCC indicated by a large quantity of neutrons in a material. The main objective of degradation management in the case of reactor internals is to ascertain if the internals support the core and can protect the CRDM (Morgan and Livingston, 1995).

In-service inspection and surveillance and changing of the materials are some of the measures used to manage degradation of the internals.

It is difficult to inspect the inside of a nuclear reactor, but it is possible to obtain information on physical damage, leakage and mechanical and structural states through visual inspection of the accessible zone. When an in-service inspection is being conducted, all flange closure studbolts and heads are removed. At this time, damaged equipment can also be removed. Equipment moved to the pool or which remains in the pressure vessel can be inspected using a remote control camera. It is difficult to conduct ultra­sonic testing on this equipment or to interpret the results, but eddy-current testing is effective in measuring reduced thickness of pipes. Inspecting the inaccessible zone using the monitoring systems is complicated. Therefore, more effective remote control inspection equipment is needed (Morgan and Livingston, 1995).

The general regulation of in-service inspection of reactor vessel internals in ASME Section XI requires a visual inspection every ten years. Recently the requirement has become a visual inspection (VT-1, VT-3) supplemented with ultrasonic inspection of the baffle former bolts. The baffle former bolts comprise the weakest part of the internals. Supplementary ultrasonic exam­ination is carried out in accordance with ASME Section XI subsection IWB, examination category B-N-3 in the United States and some other countries. Development of the ultrasonic examination equipment used for inspecting these bolts should take into consideration the existence of locking bar style bolts and the accessibility problems.

Attributes of adequate ageing management programmes are defined by the regulation; see for example the NUREG-1801 (US NRC, 2010) adapted by several VVER operating countries. An adequate AMP has to have the fol­lowing elements:

1 Definition of SSCs that are subject to ageing management

2 Actions to prevent or mitigate specific ageing processes

3 Surveillance, monitoring and testing of all parameters related to the deg­radation of the function or serviceability of the SSCs

4 Investigation of ageing factors that may cause degradation or loss of function of SSCs

Building or part of the building Structure Aim of the monitoring Measurements Evaluation and criteria Main building complex: Reference points all buildings, including reactor and auxiliary buildings, stacks, diesel-building and other structures Building movements; settlement; contro of stability of cracks caused by movements Fixed geodetical measuring points; I 3D evaluation of building movements; correlation with groundwater table; allowable declination of vertical axis of reactor pressure vessel defined by functioning of CRDM Reactor building: floor Heavy reinforced Interaction with Investigation of Control of mechanical and chemical slabs and walls concrete boric acid media samples; inspection of check-holes properties and comparison with reference values Reactor building, Reinforced Control of possible Investigation of Control of mechanical and chemical turbine building, intermediate building and galleries: floor slabs and walls concrete leakages and consequent leaching samples; inspection of check-holes properties and comparison with reference values Reactor building: floor slabs and walls Carbon steel liner Control of corrosion rate, identification of possible leakages Ultrasonic control of liner wall thickness at the identified places Control of corrosion rate and thickness; focused investigation if the overall leak tightness is less than the reference value for the given unit (comparison with allowable leak rate) Reactor building and auxiliary building: floor slabs and walls Decontaminable coating and painting Control of condition of coating and painting Walk-down and visual control according to checklist Expert judgment Main building complex: all building parts Hatches, gates, penetrations, fire protection doors; Control of condition of doors Walk-down and visual control according to checklist, fluorescent test Expert judgment

5 Trend analysis to predict degradation processes and to perform correc­tions in time

6 Acceptance criteria to assure that the functions of the SSCs are maintained

7 Correction measures to prevent or solve problems

8 Feedback process to ensure that preventive actions are effective and appropriate

9 Administrative control of the processes

10 Information retrieval from operational practice to ensure that ageing management is properly carried out.

The same attributes can be applied while reviewing the adequacy of existing plant programmes.

The fuel U-235, used in PWRs in its oxide form UO2, is enriched up to 5% (maximum) and is used as dried pellets of about 10 mm in diameter and 10 mm in height with theoretical density (TD) ~95%. The stoichiometry of UO2 is kept such that the ratio of O:U is never allowed to go beyond 2:1. The surface temperature of the fuel can reach ~1400°C with the centre temperature still higher. The oxide pellets are enclosed in a Zircaloy clad tube and the tube is capped on both sides to make a fuel rod. The fuel-clad gap is filled with high thermal conductivity helium gas and the conductiv­ity degrades slowly with the fission gases diluting helium. Many such fuel rods (17 x 17) are bundled to form a fuel assembly. Many such fuel assem­blies (~200) are immersed in a pool of light water, flowing at a pressure of ~16 MPa, which is the heat transfer fluid in the primary loop of a PWR. During start up, the pellet-clad gap gets reduced due to thermal expansion of the fuel but soon increases as the fuel densifies under irradiation. At high burnups, the gap slowly reduces and eventually an intimate contact is estab­lished with the fuel swelling and the clad collapsing under creep due to the coolant pressure (clad creep-down). All cladding tubes have some ovality which increases with creep-down and the direct impact of creep-down is the increased gap (between rods) and increased water volume. This increased water volume increases the moderation effect and gives rise to power peaks in the neighbouring fuel pellets.68 The swelling of the fuel applies a severe hoop stress on the clad which slowly gets embrittled by irradiation and by absorption of hydrogen. The thin oxide layer on the clad breaks and bonding between fuel and clad becomes established, leading to a condition called pellet clad interaction (PCI). In extreme cases, these factors lead to cracking of the clad and to release of the fission gases into the coolant. Such situa­tions may demand reduction of reactor power or the shutdown of the reac­tor if safe discharge limits are crossed. The strain concentration produced in the cladding by the sharp corners of radially cracked pellet during a power increase is increased by (i) the increasing coolant pressure, (ii) the pellet/ cladding friction coefficient, (iii) the pellet radius and (iv) the circumferen­tial temperature gradient. This severity can be marginally reduced by using a stronger clad material and by reducing the number of radial pellet cracks. For burnups greater than 70 MWd/kgU, a high burnup structure (HBS) with bubble size much larger than the grain size forms and traps the fission gases. A rapid rise in temperature can lead to shattering of the HBS and can put severe stress on the cladding.69

The VVER-1000 model exists in several versions. The ‘small series’ plants could be considered as pioneers of this model. The VVER-1000/320 is the large series version of the design. Developed after 1975, VVER-1000/320 type plants are operated in Bulgaria, the Czech Republic, Russia, Ukraine and China. Modernized versions of VVER-1000 plants are under construc­tion in five countries (Bulgaria, China, India, Iran and Russia).

In regard to lifetime management, the VVER-1000/320 plants have great­est practical importance. The ‘small series’ plants show some specific design features, but the lifetime management practice of these plants does not dif­fer essentially from the VVER-1000/320 version.

The VVER-1000 is a four loop PWR with horizontal steam generators. Each loop consists of a hot leg, a horizontal steam generator, a main circu­lating pump and a cold leg. Main isolating valves on the hot and cold legs of each loop equip the non-standard VVER-1000 primary loops. The standard V-320 design and the new clones of the VVER-1000 do not have isolating valves on the primary loop. A pressurizer is connected to the hot leg of one of the loops and the spray line to the cold leg. Operating conditions are Thot=322°C, Tcold=290°C, p=15.7 MPa. The reactor, the primary and the safety systems are all placed within a full pressure, dry, pre-stressed concrete containment.

The design bases, and also the technical solutions applied, are very similar to the PWRs operated in Western countries. The safety concerns about the VVER-1000 plants are discussed in detail in IAEA reports (1996b; 2000). The main safety concern regarding the VVER-1000 plants lies in the qual­ity and reliability of the individual equipment, especially the I&C equip­ment. The plant layout has weaknesses that make the redundant system parts vulnerable to hazardous systems interactions and common cause fail­ures by fires, internal floods or external hazards. At all plants, many of these deficiencies have been addressed by plant modifications and an acceptable safety level has thus been achieved.

There are several advanced VVER-1000 plants presently under con­struction, more than 20 new projects of advanced VVER design are under preparation or consideration and several are in the bidding phase. The most advanced versions of VVER design, showing features of Generation III reactors, are being considered for future bids for large generating capacity reactors.

A complex picture of ensuring and justification of LTO of VVER plants has been given in this chapter. The VVER-440/213 model especially is discussed in detail.

In VVER operating countries, proper regulatory frameworks and compre­hensive plant lifetime management systems have been developed to ensure the safety of LTO of VVER-440 and VVER-1000 type plants. Detailed stud­ies and already approved cases of extension to plant operational lifetime demonstrate the feasibility of LTO of VVER plants.

Generally accepted principles for safety have been followed while devel­oping plant systems, which ensure that any SSCs will be covered by some of the plant programmes, and within the framework of the LTO programme, all conditions of safe operation will be ensured. Structures, systems and compo­nents are identified which are significant for safe LTO. An appropriate level of understanding of the ageing phenomena has been reached and adequate age­ing management programmes developed for ensuring the required status and intended function for the long term. Revalidation of time-limited ageing anal­yses also justify the safety of LTO, which is completed by the monitoring of maintenance on performance criteria combined with the maintenance of envi­ronmental qualification and replacement and reconstruction programmes.

Best international practice and state-of-the-art methodologies have been applied while performing the particular tasks for preparation and justification of LTO and licence renewal. However, as demonstrated, any good examples and experiences should be adapted in creative ways, taking into account the design features, national regulations and existing plant practice. In this way the strategy of VVER operators to operate safely for as long as possible with economic advantage and at higher power levels will be ensured.

The identified TLAAs cover the usual areas: fatigue calculations, assessment of embrittlement, changes of material properties, etc. However, the scope of TLAAs for some VVERs differs from the usual one either because of the peculiarities of the design or because of national regulation. For example, in

[19] Analysis of buildings classified into safety category for the verification of the design.

• Fatigue analysis for the containment penetrations.

• Fatigue analysis for the hermetic liner of the containment (welding, tran­sition welding, area of anchors).

• Fatigue analysis for the liner of the spent fuel pool (welding, transition welding, area of anchors).

• Stress and fatigue analysis for the safety-classified crane in the reactor hall with capacity of 250/32/2 tons.

[20] Structural changes in the UO2 due to collection of solid fission products and the effect these structural changes have on the mechanical stability of the fuel during normal and abnormal operation.

[21] How fission product gases are held up within the pores of the pellet and how they are released during upset events.

• Effect of coolant on fuel structure and stability under operating condi­tions when fuel cladding failure occurs.

[22] The oxidation behavior of zirconium alloys at all temperature condi­tions (including high temperature accidents).

• The hydriding of zirconium alloys at all temperature conditions.

[23] Resistance to accidents and departure from nucleate boiling (DNB) or dry-out incidents because of higher operating temperature (~2000°C) capability.

• Minimal hydrogen production due to a much lower rate of reaction with water.

• Ability to operate in much longer cycles due to the very low corrosion rate of SiC in water.

• Enrichment savings due to ~75% lower thermal neutron absorption.

• Uprate capability of ~30% due to the ability to operate at DNB or dry-out conditions.

• Immunity to debris or fretting failures.

[24] Property standards for SiC/SiC-composite matrix ceramic (CMC) mate­rials as applied to LWR’s.

• Mechanical properties as a function of time, temperature and irradiation and use.

• Corrosion properties at high temperatures in oxidizing (steam/air) atmospheres.

• Thermohydraulic response under design basis (LOCA, RIA) and severe accident scenarios.

• Core melt progression and relocation during beyond-design basis accidents.

The VVER plants developed different types and systems of ageing manage­ment programmes:

• Overall plant AMP

• AMPs addressing a degradation mechanism

• Structure — or component-oriented AMP.

Overall plant AMP

An AMP for an overall plant can be developed and implemented for: defi­nition of goals of the operating company, distribution of responsibilities in

the organization and policy level activities, and definition of the programme system structure for ensuring the required plant condition, that is the imple­mentation of the concept described in the introduction of section 8.4.

Several operating VVERs have utility — or even industry-level or umbrella type ageing management programmes. For example, in Ukraine the plant level programme has to be deduced from the overall one and the unit level programme from the plant level one. The overall plant AMP also includes the categorization of the SCs in accordance with safety relevance, importance and complexity. In considering the structuring and organization of AMPs, a graded approach should be applied according to the safety relevance of the

Table 8.6 Degradation mechanisms which an AMP may address

Low-cycle fatigue Irradiation damage Boric acid corrosion Local corrosion

Irradiation-assisted stress corrosion Swelling

Thermal stratification fatigue Erosion-corrosion Water hammer Deposition [16] [17]

The pipelines, pipe elements (elbows, T-pieces), valves and heat exchang­ers can be grouped into commodity groups according to type, material and working environment. The SCs within a group have the same degradation mechanism and approximately the same operational and maintenance history. It is very reasonable to develop specific ageing management pro­grammes addressing the ageing of commodity groups. The definition of the commodity groups is decided by applying the attributes given in Table 8.7 in all reasonable combinations.

High energy neutron exposure results in accumulation of many defects like point defects (vacancies and interstitials) and dislocations, and causes redistribution in the chemistry (phase change or radiation-induced segre­gation (RIS)) in the materials. These modifications lead to deterioration in mechanical and corrosion properties of the exposed materials. The micro­scopic defects produced in materials due to irradiation are referred to as radiation damage. The crystal defects thus produced modify the macro­scopic properties (physical, thermal and mechanical) of materials which are referred to as radiation effects.

The high energy neutron knocks out a stationary atom from its equilib­rium position and transfers some kinetic energy (KE) to it which in turn displaces more atoms to cause a cascade effect, resulting in a number of interstitials leading to Frenkel defects.5 This process continues until the energy of all primary and secondary knock-on atoms is insufficient (<25 eV) to displace those from the lattice sites. The atomic displacements per atom (dpa), defined as the number of times each atom is displaced from the lat­tice site, is estimated using various models, among which the Kinchin-Pease model is often used:

ЛЕ 4 A

dpa = o-el n Ф, with Л = , [1.1]

el 4Ed (1 + A)2’ L J

where Ф = 0t is the fluence, ф is neutron flux (/m2s), t is irradiation time, En is the mean neutron energy, Ed is the atomic displacement energy, A is atomic mass (in amu) and oe is the elastic neutron scattering cross section. This dpa is a better damage evaluation unit than the commonly used fluence (Ф) since it takes into account the spectral distribution of neutron energy in a given reactor.5 This process continues until the KE falls below that needed to cause further displacements. The interstitials thus formed segregate as small disc shaped clusters.6 They can either dissolve by vacancy emission or coalesce to large nano-voids. Clearly, the creation of excess point defects not only changes the physical dimensions due to a reduction in density but also enhances the diffusion kinetics and the phenomena controlled by atomic diffusion such as creep, while the production of line, areal and volumetric defects result in radi­ation hardening and embrittlement. In RPV steels, the excess vacancies pro­duced by irradiation favour long range diffusion of copper atoms to form a copper-rich-phase, which has a BCC structure and is coherent with the steel matrix. This phase gets enriched in iron, nickel and manganese, and increases in volume.7 These phenomena pertaining to RPV steels are further discussed in detail in a later chapter of the book. Before discussing the radiation effects a brief description of various materials properties and phenomena is presented. It should be noted that materials issues specific to various reactor compo­nents as well as detailed descriptions on various phenomena are included in specific chapters later in the book: Part I covers various fundamental materi­als phenomena and Part II covers reactors and components.

Degradation management in steam generators is possible with the help of research and development or a technical support programme. Strategies can be established by supplementing inspection and repair programmes based on operating experience in power plants. Management of ageing can be divided into the understanding, prevention, detection, monitoring and mitigation of ageing. A measure to systematically combine the management strategies is needed for steam generators which are widely used globally. An effective strategy could also be established and efforts to reduce duplication made through the cooperation of the equipment vendors and energy utility companies as shown in Fig. 7.1.

With the techniques developed to date, such as shot peening, rotopeen — ing and heating and temperature reduction of the hot leg side, it is possible to reduce the tensile stress inside steam generator tubes. These measures markedly postpone PWSCC initiation. Plugging, sleeving or changing the affected pipes is effective in terms of repair. Secondary water chemis­try control is the best defence against ageing damage on steam gener­ator tubes. Measures to expand the life of the steam generator include

PLAN

7.1 General structure of a steam generator ageing management strategy. (Reproduced with permission from the Electric Power Research Institute © 2008).

controlling impurities (chloride, iron and copper ions in the primary side) and oxygen (in the secondary side) and to prevent the accumulation of sludge. Ingress of chlorine-containing inorganics through condenser leak­age, resin releases from condensate polisher and make-up water, are fac­tors that compromise water chemistry control. It is proven that certain chemical additives (e. g. boric acid or morphine) decrease intergranular attack (IGA), SCC and denting of the tubes. However, it is not known whether such additives influence the equipment of other power plants (Morgan and Livingston, 1995). A continuous monitoring and control programme should also be followed to reduce impurities in the secondary water (Wood, 1990).

Denting of the tubes, fretting wear and erosion-corrosion can be detected through a normal in-service inspection before leakage occurs, whereas it is difficult to detect regional pitting corrosion and cracking (fretting-fatigue, stress corrosion and formation of intergranular) before leakage occurs. The fundamental cause of fretting is related to the design of the stream gen­erator. As a result, the most effective management option is dependent upon the design. In most cases plugging of the affected tubes is an effec­tive solution when damage is found in a particular part of a certain design. The occurrence of erosion-corrosion and corrosion fatigue is limited to once through steam generators (OTSGs), and management options vary depending on the characteristics of particular power plants (Morgan and Livingston, 1995).

To avoid maintenance cost increases, suspension of operation or reduc­tion of output, it has become possible to replace the existing steam gener­ator with one using corrosion resistant alloys (Alloy 690). As of 2011, over 100 steam generators have been replaced around the world. For most of the replaced stream generators, thermally treated Alloy 690TT has been used. The power plant Cook-2 used this alloy for the first time in 1989. With advanced methods and greater experience, it no longer takes much time to replace a steam generator. Developments in design and material allows newer steam generators to have a long service life. Crevices can be removed, allowing a steam generator to have low residual stress. New generator designs also have improved accessibility for secondary lancing and chem­ical cleaning (Morgan and Livingston, 1995). Improved corrosion-resistant materials for SG tubes include high temperature mill annealed Alloy 600 (Alloy 600 HTMA), mill annealed Alloy 690 (Alloy 690 MA) and Alloy 690TT. Alloy 690TT has only recently been used in new steam generators. Ferritic stainless steel is used for tube support structure.

TLAAs and their role in the justification of LTO

Although the wording is sometimes different, the term ‘time-limited ageing analyses’ is understood by the VVER operators in a very similar way to its definition in 10CFR54.3 (Requirements for Renewal of Operating Licenses for Nuclear Power Plants). The role of the review and revalidation of the TLAAs in the justification of LTO is also the same as international practice.

Existing TLAAs should be reviewed and revalidated with an assumed extended time of plant operation. The evaluation of each identified TLAA should justify that the safety function of the SC will remain within design safety margins during the extended period of operation. The plants have to demonstrate either in the context of the PSR or in the licence renewal application that: [18] case of Paks NPP, the scope of fatigue calculations is extended to the Safety Class 1 and 2 piping and components and includes analysis of thermal strat­ification, too. In regard to its RPV, besides of PTS analysis, the limits and conditions of safe operation, that is the p-T curve has to be re-analysed in the frame of revalidation of TLAAs.