We proceed here with descriptions of corrosion phenomena, which often limit the lifetime of core components.

4.5.1 Types of corrosion and comparison between PWRs and BWRs

Corrosion of zirconium alloys used in the core of nuclear power plants (and the accompanying absorption of hydrogen in the zirconium metal matrix)

4.40 Comparison of hydride precipitation between CWSRA Zircaloy-4 and RXA Zircaloy-2 cladding (Ito et al., 2004). Copyright 2004 by the American Nuclear Society, La Grange Park, Illinois.

is of prime interest when considering performance of the core components and therefore the performance of the entire reactor. For instance, for PWRs a practical corrosion limit exists (about 100 pm oxide thickness which is associated with a critical amount of hydrogen absorption) that has driven a material change away from Zircaloy-4.

The technical literature and many conferences are full of papers deal­ing with corrosion issues. Of particular interest is the series ‘Zirconium in the Nuclear Industry: International Symposium’, ASTM STP, American Society of Testing and Materials, which provides most relevant details. Also reviews of the topic are given in ZIRAT reports: Adamson et al. (2002), Cox et al. (2004), Adamson etal. (2007); and ‘Waterside corrosion of zirco­nium alloys in nuclear power plants’, International Atomic Energy Agency (IAEA)-TECDOC-996, January 1998. The most recent open literature review of mechanisms is by Cox (2005).

Corrosion of zirconium alloys is an electrochemically-driven process affected by the microstructure and microchemistry of the alloy surface, the nature of the oxide layer that forms, the temperature at the metal/oxide interface, the chemistry and thermohydraulics of the corrodent water, the effects of irradiation and the effects of time. Table 4.7 gives information on the various types of commercial power reactor systems currently being used throughout the world. In comparing BWRs with PWRs, with corrosion mechanisms in mind, the main features are:

• BWR coolant boils; PWR coolant does not. This has an important effect at the oxide/water interface.

• PWR coolant contains a high concentration of hydrogen; BWR coolant does not. Complementarily, BWR coolant contains a high concentration of oxygen, PWR coolant does not. This has an important effect on cor­rosion processes.

• PWR components generally operate at higher temperatures than BWR components. Corrosion processes are temperature dependent.

• Both reactor types employ chemical additions to the coolant which may affect corrosion and buildup of deposits on fuel rods.

It also should be noted that BWR zirconium alloys continue to be primarily Zircaloy-2 or slight variants of Zircaloy-2. PWR zirconium alloys no longer tend to be Zircaloy-4, for reasons of insufficient corrosion resistance (and hydriding resistance) at high burnup, but have moved toward zirconium alloys with Nb additions.

The type of oxides which form during corrosion in reactor water can be classified into several categories. The two most basic are uniform and nodu­lar corrosion. The ‘uniform’ category has an extension — ‘patch’ or acceler­ated uniform. The fourth category is ‘shadow corrosion’, which can look like thick uniform corrosion but has some characteristics of nodular corrosion. The fifth category is crud-related corrosion, which is a temperature driven process induced by poor heat transfer in crud-impregnated corrosion layers. These categories will be discussed later, but are introduced here. Table 4.8 (Garzarolli in Adamson et al, 2002) gives a useful summary of characteris­tics of various corrosion types.

Uniform corrosion occurs in both PWRs and BWRs. The oxide itself is uniform in thickness and consists of several different layers. For either in or out of reactor, the initial shape of the corrosion-versus-time curve is as shown in Fig. 4.41 in the pre-transition region. The first transition point occurs at around 2 pm oxide thickness in PWRs. The shape of the post-tran­sition curve in PWRs depends on several variables: initial SPP size, irradia­tion, amount of cold work, specific alloy, water chemistry, temperature, local thermohydraulics and hydride concentration. For Zircaloy-4 the corrosion

Parameter Western type PWR VVER (440/1000) MW CANDUb BWR RBMK0 1. Coolant Pressurized H20 Pressurized H20 Pressurized D20 Boiling H20 Boiling H20 2. Fuel materials (Pressure Zry-4, ZIRLOe, DUPLEX, Zr-alloy E110 Zry-4 (Zr-2.5Nb) Zr-2, Zry-4, Zr-alloy E110, tube materials) M5, Inconel, SSe Inconel, SS (Zr-2.5Nb) 3. Average power rating, kW/l 80-125 83-108 9-19 40-57 5 4. Fast neutron flux, Average, n/cm2.s 6-9 x1013 5-7 x1013 1.5-2 x1012 4-7 x1013 1-2×1013 5. Temperatures, °С Average coolant inlet 279-294 267-290 249-257 272-278 270 Average coolant outlet 313-329 298-320 293-305 280-300 284 Max cladding OD 320-350 335-352 310 285-305 290 Steam mass content, % 7-14 14 6. System pressure, bar 155-158 125-165 96 70 67 7. Coolant flow, m/s 3-6f 3.5-6 3-5 2-5f 3.7 8. Coolant chemistry9 Oxygen, ppb <0.05 <0.1 200-400 <20 Hydrogen (D2), 2-4 (3-10) 0.05-0.30 — PPm 25-50 30-60 cc/kg Boron (as boric acid), ppm 0-2200 0-1400 Li (as LiOH), ppm 0.5-3.5 0.05-0.6 1 — — К (as KOH), ppm — 5-20 — — NH3, ppm 6-30 NaOH, ppm 0.03-0.35

aVodaVoda Energo Reactor (VVER). bCanadian Deuterium Uranium (CANDU).

°Reaktor Bolshoi Mozhnosti Kanalov (RBMK). dZirconium Low Oxidation (ZIRLO). eStainless Steel.

‘Variation from lower to upper part of the core and from plant to plant.

9Zn in ppb quantities may be added for BWRs and PWRs; Pt and Rh in ppb quantities may be added for BWRs. Source: A. N.T. International (2011).

4.41 Schematic of PWR corrosion kinetics (Adamson et al., 2006/2007). (Source: Reprinted, with permission, from Garzarolli et al. (1996), copyright ASTM International, 100 Barr Harbor Drive, West Conshohocken, PA 19428.)

thickness at high burnup may reach or exceed 100 pm, while some of the newer alloys have less than half of that. As noted in Table 4.8, only uniform corrosion occurs in PWRs under non-boiling, hydrogenated conditions.

The study of uniform corrosion in the laboratory at temperatures rele­vant to reactor operation suffers from three problems: (1) irradiation effects are absent, (2) corrosion rates are very low at 280-360°C (553-633K) and (3) very long times (hundreds of days) are required to differentiate between material variables. The most widely used test appears to be 360°C (633K) pressurized water for very long time periods. An example is given in Fig. 4.42 where a series of Zircaloy-type alloys with different Sn contents are compared. Garde et al. (1994), show that in-reactor results after high burnup give the same ranking as the autoclave laboratory tests. Also, adding Li to the water is thought to improve comparisons for PWRs (Sabol et al., 1994 ).

Both Zircaloy-2 and -4 form a protective uniform oxide in typical BWRs and PWRs. In BWRs the various types of oxidation kinetic are shown in Fig. 4.43.

Uniform corrosion continues at a very low rate out to high burnups. During reactor exposure the microstructure of the Zircaloys used in BWRs is continually evolving due to irradiation damage and SPP dissolution. In the range 30-50 MWd/kgU (6-10 x 1021 n/cm2, E > 1 MeV), changes in the microstructure induce an acceleration of uniform corrosion, as described earlier. First, patches of white oxide appear in the otherwise black or grey uniform background, as illustrated in Fig. 4.44. These patches remain very

Type of corrosion Comment Observed Irradiation effect in oxygenated coolant (BWR) Irradiation effect in hydrogenated coolant PWR Uniform Normal mode Out-of-pile, in-BWR and in-PWR 5-10x increased from beginning, low flux dependency 2-4x increased after 1st corrosion rate transition, low flux dependency Nodular Local breakdown of oxide protectiveness In-BWR and out-of-pile >500°C, Zry with large SPP Increases almost linearly with fast flux, little temp, dependency Not observed Shadow corrosion Probably driven by potential differences Only under irradiation and oxidative coolant conditions (BWR) Increases probably almost linearly with fast flux Not observed Crevice corrosion Change of environment in small gaps Out-of-pile, in-BWR and in-PWR Observed Observed Increased corrosion with fine SPP Reduction of protectivity Out-of-pile, in-BWR and in-PWR Increases almost linear with fast flux Increases almost linearly with fast flux, little temp, dependency Increased corrosion at high fluences Reduction of protectivity In-PWR ? Increases with increasing fluence above a critical threshold Increased corrosion at high hydride concentrations Lower corrosion resistance of hydride Out-of-pile, in-BWR and in-PWR Observed Flux dependent but less temperature dependent Source: A. N.T. International (2011) and Garzarolli in ZIRAT7 STR ‘Corrosion in Zirconium Alloys’, Adamson eta!. (2002).

4.42
Corrosion weight gain as a function of laboratory autoclave test exposure time for several Zircaloy-4 tube variants. (Source: Reprinted, with permission, from Garde et al. (1994), copyright ASTM International, 100 Barr Harbor Drive, West Conshohocken, PA 19428.)

4.43
Characteristics of different types of corrosion observed in BWRs. Oxide layer thickness in arbitrary units versus fuel assembly burnup.

5 4

3

4.44 Patch oxide formation on Zircaloy-2 after an exposure of 8.5 x 1021 n/cm2, E > 1 MeV. The oxide is thinnest at uniform black oxide (5) and patch oxide (4) and is thickest at coalesced patches (3). (Source: Reprinted, with permission, from Huang et al. (1996), copyright ASTM International, 100 Barr Harbor Drive, West Conshohocken, PA 19428.)

thin, about the same as the black uniform corrosion film. At some point the patches cover 100% of the surface and oxide thickening occurs at an accel­erated rate, labelled ‘late increased corrosion’ in Fig. 4.43.

Importantly, at the same or somewhat higher fluences, a marked increase in hydrogen pickup fraction occurs. This hydriding is potentially a more seri­ous issue than the corrosion increase and is discussed later.

Figure 4.45 gives optical micrographs and schematics of various types of corrosion. Schematics A and C show normal and increased uniform oxide, and the others illustrate nodular corrosion. Zr-Nb alloys and small-SPP Zircaloys (average SPP size less than about 0.1 pm) in general do not form nodules. In large-SPP Zircaloys, nodules initiate early in life and grow at a decreasing rate with fluence. In some cases, for aggressive water chemistries and susceptible material, nodules can coalesce to cover the entire surface. Nodular oxide thickness does not generally cause performance problems; however in severe cases spalled oxide can be a source of ‘grit’ in control drive mechanisms. Serious fuel failure problems can be induced by a combi­nation of heavy nodular corrosion and copper-zinc-laden crud, resulting in the crud-induced localized corrosion (CILC) phenomenon discussed later.

Susceptibility of Zircaloys to in-reactor nodular corrosion can be identi­fied by laboratory high temperature steam tests. The most effective testing procedures are variations of the ‘two-step test’ described by Cheng et al. (1987) where Zircaloys are exposed to steam at 410°C/1500 psi (683K/102 bars) for 8 h followed by 51071500 psi (783K/102 bars) for 16 h. In such tests, Zr-Nb alloys do not exhibit nodular corrosion.

The fifth type of corrosion, indicated in Fig. 4.43, is so-called shadow cor­rosion, described in more detail later. Shadow corrosion is induced on all zirconium alloys when they are in close proximity to many non-zirconium

Zircaloy

ZrO.

Normal

uniform

Nodular

Increased

uniform

4.45 Corrosion morphology for Zircaloy in BWRs (Adamson et al., 2007).

alloys such as stainless steel or Inconel. The oxide thickness is unusually large and often appears to be particularly dense and uncracked. For exam­ple shadow corrosion oxide induced by a stainless steel control blade bun­dle is shown in Fig. 4.46 . Shadow corrosion has ‘always’ been present in BWRs, but not in PWRs primarily related to the high PWR hydrogen con­centration which reduces or eliminates galvanic potentials between dissim­ilar alloy components. In BWRs shadow corrosion caused no performance issues until recently when at one reactor fuel failures were induced by unusually severe ‘enhanced spacer shadow corrosion’ (Zwicky et al., 2000). More recently, shadow corrosion has been alleged to be involved in BWR channel bow problems (Mahmood et al. , 2010). Both issues are addressed later in this chapter.