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.

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[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.

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[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).