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

DOI: 10.1533/9780857097453.1.3

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

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

1.1 Introduction

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

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

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

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

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

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

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

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

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

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

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

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