The standard cladding material in a LWR is a dilute zirconium-base alloy containing some other elements such as tin, niobium, iron, nickel, chro­mium and oxygen (Zircaloy-2 for BWR and Zircaloy-4 for PWR). Being a hexagonally close packed (HCP) crystal structure and hence inherently anisotropic, zirconium acquires further anisotropic properties after fabri­cation due to induced texture (Fig. 1.25b) with the <c> axis of the HCP crystal oriented at ~30° from the radial direction of the tube. Minor modi­fications in the chemistry of the alloys are made to reduce the water side corrosion in the clad tubes. Formations of inter-metallic precipitates (which increase the corrosion rate) are avoided by giving the clad material a beta quench (fast cooling from the beta phase). The fuel and the pressure bound­ary (clad tube) experience time-related ageing and degradation; the former may affect linear power rating while the latter can lead to catastrophic clad failure. The burning of the fuel leads to release of fission gases and to fuel swelling. The fuel makes contact with clad which has picked up hydrogen from the coolant and leads to degradation of the clad tube.7 0 In order to improve the structural rigidity, spacer grids made of Inconel/Zircaloy are placed at definite intervals along the length of the assembly which holds the fuel elements in the assembly with spring forces. When these forces relax (due to creep) a gap is created between the grid and rod, and the rod can vibrate. This fretting may result in a breach in the clad integrity.71

Operational experience provides the basis for preparing the strategy of age­ing management. The experiences of the plants regarding degradation of the lifetime-limiting structures and components have primary importance. These are the non-replaceable long-lived structures and components. In the VVER-440 plant design, lifetime-limiting structures and components are the containment building, reactor pressure vessel and the steam gen­erator (Katona et al. , 2005, 2009b; Katona and Ratkai, 2008). Unlike the VVER-1000 and PWRs, the steam generators are practically irreplaceable in the VVER-440/213 design. In the case of the VVER-1000, the most impor­tant lifetime-limiting structures and components are the containment and the reactor pressure vessel. The proven design solutions of the VVER-440 were incorporate in the VVER-1000 design: the horizontal steam generator and also materials selection.

Alongside this, ageing the mechanical commodities, structures other than containment and electrical equipment, are also important for the develop­ment of an ageing management strategy. With this in mind, the operational experience of the plants varies because of the design variation of these com­ponents and structures at different plants.

S. RAY and E. LAHODA, Westinghouse Electric

Company LLC, USA

DOI: 10.1533/9780857097453.3.385

Abstract: This chapter provides background on current materials-related problems faced by the nuclear industry. These issues have become more important as the current fleet of nuclear plants ages and as life extensions of 20 years each are added onto the current 40 year life. Materials issues requiring research and development are presented in terms of the fuel, the primary boundary, the containment and other general issues.

Key words: nuclear, power plant, industry, materials, issues, fuel, rods, cladding, containment, primary, secondary, corrosion, cracking, buried, pipe, wiring, concrete, steel.

9.1 Introduction

As the light water reactor (LWR) nuclear fleet reaches and surpasses the original 40 year lifespan that it was licensed for and embarks on its next 20 years with visions of yet another 20 years beyond that, the need increases for a scientific underpinning of the understanding of the degradation of materials in a nuclear environment. The need to generate this scientific underpinning becomes more compelling when one considers that the orig­inal 40 year life had no scientific basis (INL, 2009) and that the materials designs were not based on irradiated materials in real life chemistry con­ditions (Majumdar, 2011). There are indeed many existing models and correlations for determining what may happen as materials in a nuclear environment age, but most are based purely on empirical data. The ability to extrapolate these models is under question by the industry and the NRC, which will not grant licenses based on extrapolated models. Far too often, researchers have discovered unexpected effects, both good and bad, which should not have occurred based on extrapolation of models.

This chapter focuses on the materials in a nuclear system from the inside out of a nuclear plant. First the initial fission boundary of the fuel (Fig. 9.1),

Zircaloy Zr based alloy

BWR Zry-2 (Zr-1.5Sn-0.12Fe-0.05Ni-0.1Cr) PWR Zry-4 (Zr-1.5Sn-0.15Fe-0.00Ni-0.1Cr)

9.1Nuclear fuel pellets and rods (Kazuya Idemitsu, Genshiryoku Zumenn Syuu, JAERO, p. 4, used by permission of the author).

both the fuel itself which contains most of the fission products, and then the fuel cladding which normally contains the rest of the fission products. Occasionally, the cladding leaks and the second boundary comes into play — the primary system of the pressurized water reactor (PWR) (Fig. 9.2) and the steam system of the boiling water reactor (BWR) (Fig. 9.3). After that, the boundary is the secondary system and the containment of the PWR or BWR. During postulated accident conditions, the innermost system (fuel and cladding) fails and then all that remains to avoid exposure to the public is the concrete and steel containment structure. Thus, a good understanding of all the containment systems is also needed.

A peculiarity of the VVER-440/213 design is the extremely large number (over one hundred thousand) of safety-classified SSCs because of the design features and methodology of safety classification.

After screening out the active and short-lived systems from the total safety-classified SSCs, approximately 38 000 mechanical, 6500 electrical and 2000 structural SCs have been identified to be in scope at the plant in Paks, Hungary.

Ageing management of mechanical commodities might be ensured approximately by nine vessel specific, nine pump-specific, 14 valve-specific, 22 heat-exchanger-specific, 15 piping specific, nine filter-specific programmes. There are also 15 special components requiring individual AMP. The num­ber of structural commodities exceeds 25. The AMPs and their hierarchical structure is plant specific, demonstrating that Paks NNP practise an adapta­tion of best international practice to VVER-440/213 instead of a copy-paste approach. At the same time, the Paks NPP is utilizing the ageing experi­ence of other plants and elements of an adequate ageing management pro­gramme are in line with international practice.

The specific approach practicable in the case of the VVER-440/213 plants can be shown in the example of ageing management of civil structures. The VVER-440/213 design differs very much from the usual architecture of PWRs. In the example of the Paks NPP, practically all buildings, earth structures, etc., at the plant are within the scope. Most of these building structures are complex, and heterogeneous from the point of view of struc­tural design, layout, manufacturing and construction of members, material composition and contact with environment (Katona et al, 2009a).

In the case of the Paks NPP, it would be difficult to adopt the AMPs described in the GALL Report (US NRC, 2010), where nine groups of building structures and seven groups of structural components are defined, and ten ageing management programmes cover the whole scope. At the Paks NPP the large number and variety of building structures and struc­tural components requires establishment of a hierarchical structure of age­ing management programmes.

Type A programmes have been developed for foundations, reactor sup­port structures, building movement, reinforced concrete structural members, high temperature concrete, equipment foundations, steel and reinforced con­crete water structures, liners (Carbon-steel), prefabricated panels, masonry walls, earth structures, doors and hatches, steel-structures, cable and pipe supports, paintings and coatings, SS-liners, cable and pipe penetrations, fire protection structures, main building settlement, support structures of cabi­nets, seals and isolation and corrosion in a boric acid environment. These programmes are related to specific structures, that is structural commodities or specific ageing mechanisms (e. g. building settlement due to soft soil con­ditions). An exceptional A-type programme is the control of leak tightness of the containment, which is related to the containment only.

The buildings having identified safety functions are composed from struc­tural commodities. Using these type A programmes for specific structures (commodities), 30 type B programmes have been developed which cover all plant building structures. These AMPs contain the identification of ageing effects and mechanisms to be managed, the lists and details of the proper application of type A AMPs to be applied, while managing the ageing of the given building. The type B AMP also contains logistical type information since the accessibility of certain buildings is limited.

Structural materials experiencing complex stresses due to varied external forces may suffer elastic, anelastic and plastic deformations. Elastic strain is an instantaneous and completely recoverable deformation, the extent of which depends on the elastic modulus of the material and, in a simple uni­axial loading case,

Ze = |, [1.2]

E

where a is the stress (load per unit area), E the modulus of elasticity (also known as Young’s modulus) and eE is the instantaneous elastic strain (change in length per unit length). Anelastic strain is time dependent, completely reversible and generally small in magnitude — albeit non-negligible in some cases — as will be discussed in detail in Chapter 3. On the contrary, plas­tic strain is permanent and remains even after removal of the stresses; it is generally time — and rate-dependent. A typical stress vs strain curve under uniaxial loading is shown in Fig. 1.2a8 and the important design parameters are the yield strength, tensile strength, uniform elongation and ductility or total elongation to fracture. The deformation beyond the elastic limit obeys a power relation between the true stress (a) and the true plastic strain (ep):

a = K(£p)n, [1.3]

where K is the strength coefficient and n is the strain-hardening exponent. The area under the stress-strain curve represents the energy to deformation and fracture (referred to as resilience and toughness in the elastic and plastic regime, respectively), and this grades a material as brittle or ductile (Fig. 1.2b). The various mechanical properties of a material are also rate dependent and the flow stress is often characterized by the strain-rate sensitivity (m):

Ir, e = A m. [1.4]

The higher the n value, the higher is the uniform elongation, while a higher m value means a higher total elongation to fracture. The maximum possible value for m is unity which corresponds to viscous flow as seen in fluids, and this is generally noted in metals and ceramics at relatively high tempera­tures and at low strain-rates (or stresses).

Time dependent plastic deformation that occurs under constant load or stress (creep) becomes important above homologous temperature (T/TM > 0.4, where TM is the melting point in absolute temperature). The reader is referred to Chapter 3 for more detail on the underlying creep mechanisms and phenomenological descriptions of the creep rupture life. A typical creep curve is illustrated in Fig. 1.3 and design allowances are limited to the total strain accumulation in the primary and secondary regimes. Thus the strain at any instant of time is given by the sum of instantaneous recoverable elastic

strain, instantaneous plastic strain and time-dependent strain component from primary and secondary creep regimes:

£ = £0 + £ (l-) + £st, [1.5]

where £0 is the instantaneous strain (the majority from elastic deforma­tion), £t is the extent of primary creep strain, r is the rate at which strain decreases with time during primary creep regime and subscript ‘s’ stands for steady-state creep rate. The steady-state creep rate is a unique function of the applied stress and temperature for a given material

£s = Aone-Qc/RT, [1.6]

where Qc is the activation energy for creep, n is the stress exponent, R is the gas constant and T is absolute temperature. The activation energy for creep can generally be matched with that for self diffusion and the above relation­ship can be rewritten as

£s = A’Don, [1.7]

where D stands for appropriate diffusion coefficient and A’ could be grain size dependent (see Chapter 3 for details). In general, lattice diffusion is temperature dependent:

D = ea2vDCVe ~Qm/RT, [1.8a]

6

and

CV=e ~Qv /RT, [1.8b]

where в is the coordination number, a is the atomic jump distance, vD is Debye frequency, CV is vacancy concentration and Qm and QV are the acti­vation energies for migration and formation of a vacancy, respectively. It should be noted that higher stress increases the diffusion and leads to higher creep rate with reduced rupture time.

To minimize occurrence of damage to pressurizer nozzles and the CRDM, basic management strategies consist of: operation within operating guide­lines; inspection and monitoring; assessment of any degradation that is detected; and maintenance. The main degradation mechanisms which can occur in pressurizer nozzles and the CRDM are thermal fatigue, vibratory fatigue, SCC and boric acid corrosion. Because coolant leakage through the heater sheath, instrument penetrations or manway cover gasket can cause corrosion and SCC of other equipment of the pressurizer system, this must be controlled. Molybdenum di-sulphide lubricant should also not be used in a steam exposed environment, because experience suggests that MoS2 has a pronounced tendency to decompose in the presence of high temperature and moisture conditions releasing sulphide which is a known promoter of SCC. In the United States, to manage ageing of vessel head penetrations and nozzles, the utilities are forced to conduct a regular inspection under ASME Code Case 694. Some plants also conduct supplementary inspections on the PWSCC sensitive zone. Many power plants conduct supplemented inspec­tion and replace their RPV heads with new ones (IAEA, 2007).

Review and validation of TLAAs is a rather complex task for the majority of VVER plants. The issue is related to the availability of design base infor­mation and incompleteness of the delivered design documentation. Often only the final results of the analyses are known; in some cases, the analyses are presumably obsolete. For the majority of VVER plants outside Russia the TLAAs have to be performed anew using state-of-the-art methods in accordance with the recent requirements. In comparing the practice of dif­ferent VVER operating countries, the most complex cases are probably the Eastern-European VVER-440/213 plants since these plants have to over­come this issue. For instance, in the case of the Rivne NPP in Ukraine, full scope stress calculation and fatigue analysis had to be performed for the VVER-440/213 type units.

The case of Paks NPP Hungary will be discussed below on the basis of Katona, Ratkai and Pammer (2007), Katona et al. (2010) and Katona, Ratkai and Pammer (2011). TLAAs have to be reviewed and verified for the most important structures and components (SCs). Developing a methodology for TLAA reconstitution and defining the method of adaptation of ASME BPVC for a Soviet designed plant has been reported by Katona, Ratkai and Pammer (2007) and Katona, Ratkai and Pammer (2011). Hungarian regu­lations require application of state-of-the-art methods and standards in the time-limiting ageing analyses. ASME Boiler & Pressure Vessel Code, Section III, edition 2001 (ASME BPVC) had been selected for the reconstitution of TLAAs and associated strength verification. The code selection requires understanding of both the Russian (Soviet) design standards and the ASME BPVC code. Different studies were performed for ensuring the adequacy of ASME BPVC implementation for VVER-440/213. Calculations were per­formed for a 50 year extended operational lifetime with an additional margin of 10 years. The use of ASME BPVC is not a generic approach used by VVER operators. In some VVER operating countries the conservative PNAE G-7- 002-86 standard is used by the operators and accepted by the regulators.

Hydride related problems are viewed as an issue to be considered for clad failure. The main sources of hydrogen for the clad are: the corrosion reaction of metal with water, hydrogen released by radiolysis of water and hydrogen gas that is added in the coolant to keep the oxygen potential low.72 The defects present in the clad (such as manufacturing defect, PCI crack, debris fretting, etc.) can aid the pick-up of hydrogen and the coolant could surge in through these defects when they grow through-thickness and form steam. The steam reacts with the fuel and hydrogen is released. When the hydrogen-to-steam ratio crosses a critical value (steam starva­tion), the growing oxide layer on the ID of the tube finally breaks down and the hydrogen diffuses into the matrix of the tube. The hydrogen thus picked up can reduce the toughness of the zirconium matrix in three ways: (i) hydride reorientation, (ii) delayed hydrogen cracking and (iii) forma­tion of a hydride blister.

The solubility limit of hydrogen in zirconium at the reactor operat­ing temperature is about 100 ppm. When the temperature is reduced (for instance, during reactor shutdown), the excess hydrogen precipitates in the form of hydride. The hydride precipitates along the radial direction of the tube owing to the texture and the hoop stress in the tube. The hoop stress required for reorientation (in an unirradiated and recrystallized Zircaloy-2) is about 80 MPa.73 The differential temperature between ID and OD of the clad wall (either during service or at wet repository) drives the hydrogen to the OD side which is at a lower temperature. The concentration of hydrides found in the tube after irradiation is higher near the water side than at the fuel side which is attributed to the corrosion reaction between the clad OD and the coolant.74 The threshold stress for failure of irradiated and hydride-reoriented spent fuel cladding is significantly higher than the stress due to the internal pressure of the fuel rod. The degree of oxidation and hydriding in the more advanced fuel claddings commonly used these days in LWRs, such as low-Sn Zircaloy-4 (Sn content around 1.3 wt.%), optimized Zircaloy-4, Zirlo (a Zr-1Sn-1Nb-0.1Fe alloy), M5 (a Zr-1Nb alloy) and opti­mized Zircaloy-2, is relatively low even at high burnup.

Frequently, a hydride blister is produced when a fuel rod that contains spalled oxide is operated continuously to high burnup. During steam star­vation, hydrogen ingress is faster than its diffusion into the tube matrix. This leads to excess amounts of hydrogen getting localized at the inner wall of the clad tube forming a large hydride called a blister. The hydrogen atoms, diffusing down the temperature gradient, form radial hydrides in a sun-burst pattern.

Delayed hydrogen cracking (DHC) is important for spent fuel in either wet or dry repository and is a two-step process. Hydrogen migrates up the stress gradient towards a stressed crack-tip and precipitates as hydride that cracks and extends further. There is an incubation time for the hydrogen to arrive and the concentration to build to the required level, so that the solubility limit is exceeded for the hydrides to form and grow, before the crack extends further. The crack-tip can undergo a corrosion reaction and the hydrogen released can either be absorbed by the matrix close to the crack-tip or the hydrogen can diffuse through the matrix to the crack-tip. The former is called corrosion hydrogen cracking and the latter is known as DHC. Knowing the crack velocity enables prediction of the failure time of the tube. Since there is no way to measure the crack velocity in the reactor, it is assumed that the crack starts at the centre of the tube and proceeds in both directions with a velocity in the range of 2.5 x 10-7-6.6 x 10-7 m/s which is determined from out-of-pile unirradiated laboratory specimens.75 It is possible that the velocity may be much higher in reactor as the stress state is more severe. The stress arises partly from the increased pellet volume because of increased temperature (due to reduced thermal conductivity as the pellet cracks up with burnup) and partly from the increased volume of the oxide layer of the Zr-liner on the interior of the tube. The increase in stress due to this volume expansion is faster than the creep relaxation by the clad.

Evaluation of actual/aged condition in safety-critical SCs is the basic method for identifying the ageing mechanisms and their effects on the intended functions. Plant condition has to be reviewed for the feasibility study of LTO. Review and evaluation of plant condition is an obligatory part of both the periodic safety review (Safety Factor 2 in the PSR, see IAEA, 2003) and the justification of safe operation in the licence renewal process.

The scope of review and evaluation of actual plant condition covers the safety — and seismic-classified SSCs and non-safety SSCs, failure of which may in turn jeopardize the safety functions. The review of plant condition is based on the information related to the health of components from the following sources:

• Results of operational information, records of the operational events.

• Failure data, root-cause analysis, failure statistics.

• Outage and maintenance records.

The inspection programme for safety Class 1 SCs is the most rigorous. It includes the following:

• Data of the non-destructive testing of the SCs.

• Evaluation of the results of the in-service inspections.

• Evaluation of the results/findings of the maintenances.

• Evaluation of the results of the ageing management programmes.

• Evaluation of failure data and other lifetime information.

• Evaluation of operational information.

Non-destructive testing is a regular activity at nuclear power plants. However, in the context of the plant review for the justification of LTO, some additional tests might be necessary. Individual programmes can be useful and developed for the Class 1 SCs, the reactor, main isolation valves (if such exist), main pipelines of the primary loops, steam generators and pressurizer.

In the case of SSCs in safety Classes 2 and 3, the most practical review method is visual on-site inspection. Application of the graded approach is useful, so that, in the case of higher importance or safety relevance, the inspection has to be performed for each particular item, whereas the review can be limited to the inspection of a representative sample of the commod­ity. The selection of the representative sample has to be made taking into account its type, the material, dominating degradation mechanism, environ­mental stressors, etc.

There are minor aspects to be checked during the inspections, for example: [10]

After performing all of the on-site inspections, the findings have to be eval­uated and any corrective measures identified. The final result of their eval­uation can result in:

• modification of the maintenance procedures

• modification of the periods of maintenance

• introducing new diagnostic/monitoring measures in order to determine the necessary additional actions

• performing additional evaluation of the situation

• modifications such as implementing new sealing

• replacement of the component for a different type.

The information obtained has to be taken into account while reviewing and developing the ageing management programmes. Review and revali­dation of the time-limited ageing analyses can also be considered as part of the evaluation of the part conditions. Feedback from experience of other VVER plants and the research results provide some guidance and back­ground information for the review and evaluation of plant condition.

The first boundary for fission products is the fuel and the cladding (Fig. 9.1). The fuel normally holds up about 90% of the fission products during the normal operating cycle of an LWR. The main exceptions are elements that are gaseous at normal fuel operating conditions (~400°C to ~1200°C) such as I2 , Kr, and Xe which are held up by the cladding. Even under normal

I

Demineralizer

9.2

Graphic of the typical pressurized water reactor. (Published in NUREG -1350, Volume 23, August 2011.)

operation the cladding and other fuel assembly components operate under very extreme conditions. These components are constantly bombarded by neutrons (~1012 neutrons/cm2/s), at moderate temperatures (~250-340°C), pressures (~15.5 MPa), mechanical conditions (boiling surfaces, high veloc­ity two-phase flow with large amounts of vibration), and chemical condi­tions (up to ~2000 ppm of boron and up to 10 ppm Li for PWRs).

9.2.1 Understanding the fuel

A stable UO2 fuel structure holds up most of the fission products and even a fair amount of the fission gases (up to ~90%) in the pores of the UO2 pellet. Even though UO2 pellets undergo a significant amount of cracking due to

Emergency water supply systems

9.3 Graphic of the typical boiling water reactor. (Published in NUREG -1350, Volume 23, August 2011.)

the low thermal conductivity and the resulting stresses from the very steep temperature gradients experienced under normal operating conditions, the resulting pieces are relatively large and most of the pores of the UO2 fuel act as reservoirs for fission gas products and the fission products themselves. Under circumstances where the cladding leaks and coolant enters the rod, the UO2 fuel can further disintegrate releasing much of the soluble fission product (primarily Cs and Sr) and most of the gaseous fission products. This effect becomes more pronounced as the burnup of the fuel increases sig­nificantly above the current levels of about 50 MWtd/kgU. Therefore basic research work on fuel that is exposed to burnups >~60 MWtd/kgU is needed and includes: [20]

Steam dryer and shroud head alignment and guide rods

Steam separator and standpipe assembly

Feedwater inlet Feedwater sparger

Core spray sparger Fuel assembly Control rod Fuel support

Core shroud Core plate

9.4 BWR/3 or BWR/4 reactor vessel (G. E. Technology Advanced Manual Differences/Introduction, USNRC Technical Training Center Rev 1195). [21]

• Extension of the understanding of these effects to new fuel pellet mate­rials (e. g. uranium nitride (UN)).

• Effects of long-term wet and dry storage, as well as environmental con­ditions in any potential disposal site, on the integrity of the fuel in terms of its holdup of long-lived radioactive components (mainly U, Pu, Am, Cm, and Np).

• Interactive effects of non-homogeneous portions of the fuel (such as the rim after high burnups) on the performance of the fuel and its interac­tion with the cladding during upset events such as reactivity insertion accidents (RIAs).

Understanding of these effects in a mechanistic way is at the frontier of nuclear fuel research. Since the current level of burnup between 50 and 60 MWd/kgU is just about the practical and economic upper boundary of 5% enriched U-235 fuel today, this is the limit of our empirical knowledge. There is a need to extend this boundary if enrichments above 5% become accepted or if higher density fuels (such as UN) come into use. Phenomenological models, not correlation of empirical data, will be needed to allow predic­tions to be made without the huge cost associated with totally empirical approaches.