2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

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1 ZIRCALOY-2 CORROSION AND HYDROGEN PICKUP NEAR BWR CORE INLET David Schrire 1, Erik Mader 2, Aylin Kucuk 3, Ron Adamson 4 1 Vattenfall Nuclear Fuel, SE Stockholm, Sweden, david.schrire@vattenfall.com 2 Electric Power Research Institute, 342 Hillview Avenue, Palo Alto, CA 9434 USA, emader@epri.com 3 Electric Power Research Institute, 342 Hillview Avenue, Palo Alto, CA 9434 USA, akucuk@epri.com 4 Zircology Plus, Montecito Dr, Fremont, CA 94536, USA, rbadamson@gmail.com ABSTRACT: Both corrosion behaviour and hydrogen uptake of Zircaloy components in BWR environments are complex phenomena that show large variability. Various types of corrosion have been seen under BWR conditions, including nodular corrosion, shadow corrosion, and several forms of uniform corrosion. Although mechanisms of these various types of corrosion phenomena are different, uniform corrosion phenomena are similar to the corrosion of Zr-based alloys in PWR environments. The corrosion environment of BWRs is a relatively oxidising water chemistry as well as having bulk boiling and a high void fraction in the middle to upper part of the core compared to that of PWRs. Additions of hydrogen, noble metals and zinc in some plants, as well as plant-to-plant differences in levels of crud-forming impurities in the coolant, add to the complexity. In this paper, some specific experimental observations relating to corrosion and hydrogen pickup in Zircaloy-2 fuel cladding and non-heat transfer components near the bottom (inlet) of the core are discussed. In the lower part of the core, there is no bulk boiling and thus no stable steam phase or void near the fuel rod surface to facilitate preferential stripping of volatile species from the liquid phase. On fuel rods with a lower blanket region of natural uranium, a sharp increase in the cladding outer surface uniform corrosion oxide thickness that coincides with the interface from the blanket region to the higher enriched region of the fuel is commonly observed. Intuitively, it might be assumed that this is an effect of the increased surface heat flux in the higher enriched zone. However, exactly the same type of sharp uniform corrosion increase has also been seen on adjacent water rods. This observation precludes the surface heat flux as an explanation. Shadow corrosion occurs on Zr-based alloys in contact with, or in close proximity to, dissimilar metallic alloys in the reactor core radiation field in BWRs or other reactors with similarly oxidising water conditions. The fuel rod shadow corrosion oxide thickness varies roughly inversely with the distance between the metal surfaces, with the peak thickness occurring at contact positions. Typically, the peak oxide thickness at the spacer contact positions is approximately the same over most of the fuel rod length. This is unlike the shadow oxide thickness at positions where there is a small gap between the metal surfaces, that tends to decrease from the lower part of the fuel rod towards the top. The decreasing shadow oxide thickness in these non-contacting areas appears to follow the variation in the bulk void fraction in the coolant. However, in the case of spacers right at the lower end of the fuel rod where there is a steep axial gradient in burnup but no bulk boiling/void, a clear axial gradient is also observed in the shadow oxide thickness. Additionally, it has been observed that the corrosion and the hydrogen uptake exhibit different behaviour with respect to the change in enrichment at the transition to the blanket zone in the lower end compared to the upper end of the fuel rods. These observations from several fuel rod hot cell examinations are discussed and compared with the local variations of some parameters (such as neutron and gamma fluxes, heat flux, boiling, radiolysis) that characterize the in-core environment, in order to identify possible correlations with any such parameters, and thereby to try to elucidate possible mechanistic effects on the uniform corrosion, shadow corrosion and hydrogen pickup phenomena. The results and discussion include comparisons with earlier published work. KEYWORDS: Zircaloy, Corrosion, Hydrogen, BWR, Core Inlet. 1

2 I. INTRODUCTION I.A. Background - Corrosion and hydrogen pickup phenomena in BWR conditions Both the corrosion behaviour and the hydrogen uptake of Zircaloy components in BWR environments are complex phenomena that show large variability. Various types of corrosion have been seen under BWR conditions, including nodular corrosion, shadow corrosion, and several forms of uniform corrosion. 1 Nodular corrosion is a form of localised accelerated corrosion that has mainly been observed in BWRs, and has led to fuel failures in the past. Nodular corrosion has been found to be sensitive to cladding heat treatment (microstructure) and plant water chemistry. Shadow corrosion occurs on Zr-based alloys in contact with, or in close proximity to, dissimilar metallic alloys in the reactor core radiation field in BWRs or other reactors with similarly oxidising water conditions. Although the mechanism of shadow corrosion is not understood in detail, it has been proposed that it is a form of irradiation induced galvanic corrosion 1. The hydrogen pickup fraction (HPUF) of Zircaloy-2 in BWR environments is very variable, depending on the material, form of corrosion and level of irradiation. This behaviour is observed both on cladding and non-heat transfer components such as water rods, spacers and channels. Although mechanisms of these various types of corrosion phenomena are different, some aspects of the uniform corrosion phenomena appear to be similar to the corrosion of Zr-based alloys in PWR environments. Unlike in a PWR the coolant bulk temperature in a BWR is constant (about 286 C) from the onset of bulk boiling which typically starts roughly half a meter above the bottom of the fuel rods. Above this level, there is bulk boiling with a void fraction (vapour phase) that increases with elevation in the core, although the surfaces of the fuel rods and other components are always covered with a liquid film of water. As a result of the efficient heat transfer due to the boiling, the cladding (oxide) surface temperature in the two-phase region is only slightly above the bulk boiling temperature irrespective of the local power in the fuel rods. The bulk temperature is so low throughout the core (roughly 275 C to 286 C) that purely thermally activated corrosion processes are relatively slow compared to in a PWR. The corrosion environment of BWRs is a relatively oxidising water chemistry compared to that of PWRs. Radiation in the core generates oxidizing radiolysis species in the coolant. Even though hydrogen is injected in some BWR plants, since dissolved hydrogen would leave the liquid phase and follow the vapour phase to the steam plant (turbine system) and result in high dose rates due to 16 N generation, it is not possible to add large quantities of hydrogen as is done in the PWR primary system. Once two-phase flow occurs any remaining H 2 in the coolant is stripped into the vapour phase while several oxidizing radiolysis products stay in the liquid water phase. The oxidising water chemistry is thought to play an important role in the shadow corrosion phenomenon. 2-7 Nodular corrosion, when it occurs, is also believed to be promoted by oxidising conditions 1, although it also possible to provoke nodular corrosion (in sensitive material) out-of-pile in high temperature steam under less oxidising conditions 8. Within the core of a BWR the liquid phase of the coolant is estimated to contain on the order of several hundreds of ppb of O 2 as well as significant amounts (tens of ppb) of H 2 O 2. 9 However, there are substantial uncertainties in the calculation of the in-core concentrations of radiolysis products, 15 and there are also significant differences between plants depending on their water chemistry regimes. Additions of noble metals and zinc in some plants, as well as plant-to-plant differences in levels of crud-forming impurities in the coolant, add to the complexity of the corrosion and hydrogen pickup phenomena. In particular, nodular corrosion has been found to be strongly dependent on certain impurities in the water I.B. Specific characteristics of BWR core near-inlet conditions Radiolysis of water tends to generate a surplus of oxidising species in the liquid phase and any injected hydrogen is essentially already consumed by the time the coolant enters the core 9. So radiolysis products can form in the water near the inlet of the core without being suppressed by hydrogen, but (unlike higher up in the core) without a stable vapour phase or void to allow the volatile species to be partitioned from the liquid phase. The highest concentration of some oxidising species like O 2 is therefore predicted to occur just before the onset of bulk boiling 9 The inlet section of the core is also characterised by a steeply increasing radiation field from below the core up to the region where the neutron flux starts to flatten out, usually a little above the onset of bulk boiling. The axial profile of the radiation field varies depending on the type and energy of the ionising radiation for instance the fast neutrons have a larger range than the thermal neutrons and the fast neutron flux gradient thus varies more gently than the thermal neutron fluence. Similarly the gamma radiation intensity, especially for higher energy gamma, has a longer range than beta radiation. So the axial flux profiles of the different types of radiation differ in the bottom of the core. Another aspect of the inlet section, 2

3 compared to higher up in the core, is the fact that the entire flow cross-section is occupied by the liquid phase. This has an important impact on shadow corrosion. The void (vapour) phase has a negligible electrical conductivity compared to the liquid phase. Good electrical conduction through a continuous conductive liquid-phase path may be the reason why shadow corrosion in the lower part of the core (and in non-boiling bypass regions like the outside of the channels) occurs over a longer range (distance between the dissimilar metals) compared to the upper part of the core where the shadow corrosion is more localised to the vicinity of positions with contact between the dissimilar metals. The temperature-dependent mechanisms of the corrosion process are usually assumed to be controlled by the temperature at the metal-oxide interface which is essentially the same as the surface temperature early in life when the oxide layer thickness is negligible. As the oxide layer builds up, its thermal resistance increases. The temperature increase across the oxide layer is proportional to the surface heat flux (SHF) or the linear heat rating (LHR), and the oxide thickness. In the single-phase region, the temperature increase from the bulk coolant temperature to the surface is also proportional to the SHF. In this paper, near inlet conditions are taken to mean the lower part of the core where for most or all of the lifetime there is no stable bulk void. II. OBSERVATIONS IN POST-IRRADIATION EXAMINATIONS II.A. Corrosion and hydrogen pickup at enrichment step On fuel rods with a lower blanket region of lower enrichment or natural uranium, a sharp increase in the cladding outer surface uniform corrosion oxide thickness that coincides with the interface from the blanket region to the higher enriched region of the fuel is commonly observed 14. In Figure 1 below, the axial profile of the oxide thickness in the lower part of two fuel rods symmetrically positioned in the same fuel bundle is shown together with the measured 137 Cs intensity of one of the rods (the rods have identical burnup profiles). Both rods had experienced essentially identical conditions (power history, void distribution, neutron fluence) during their operation. However, they had cladding with different heat treatments (but the same chemical composition) that exhibited different corrosion behaviour. The clad wall thickness was about,66 mm. Both fuel rods had a natural uranium lower blanket zone of just over 15 mm in length, with a higher enriched zone (2.4%) above that. Rod A3 had an oxide thickness profile (blue curve) that closely matched the 137 Cs (burnup) profile in the enriched region. Rod C1 had an almost identical oxide thickness profile to rod A3 in the blanket region but a very different profile in the enriched region. For both rods the corrosion profile had a flatter axial slope than the burnup profile in the blanket region. Both rods also showed a marked increase in the oxide thickness in the region of the interface from the blanket to the enriched pellets Rod C1 circumf av oxide Rod A3 circumf av oxide Spacer 1 Cs-137 (burnup profile) 4E+9 3E+9 Oxide thickness [µm] E+9 1E+9 Cs-137 activity 5 Blanket zone (nat U) Higher enrichment zone E Distance from rod bottom end [mm] Fig. 1. Oxide thickness and 137 Cs: lower part of two fuel rods (identical duty but different cladding material) 3

4 The burnup is equivalent to the power integrated over the entire lifetime of the fuel rod, so the 137 Cs axial profile reflects the lifetime-averaged power profile in the fuel rods. The regular small dips in the 137 Cs reflect the reduced amount of fissile material at the pellet interfaces (due to chamfers and dishing at the pellet ends), and are qualitatively the same as the variation in the local SHF at pellet ends. The very sharp step increase in the burnup at the interface from the blanket region occurs at the interface between the uppermost blanket pellet and the lowermost enriched pellet. The heat transfer from the pellets is almost perfectly radial, so the change in the SHF at the outer surface of the cladding also occurs sharply (since small chamfers at the pellet ends effectively prevent axial heat flow between adjacent pellets). However, as seen in Fig. 2a, the increase in the oxide thickness above the enrichment interface appears to lag the jump in burnup and SHF by some tens of millimetres axially. The hydrogen concentration in the cladding does not increase quite as much as the oxide thickness does from the blanket to the enriched zone; from the 4 th pellet below the transition to the 4 th pellet above the transition the hydrogen concentration in the cladding of rod A3 increased by about 5% whereas the oxide thickness increased by about 7% (Fig. 2b), corresponding to 35 % and 32 % HPUF (assuming 1 % dense oxide) Rod C1 circumf av oxide Rod A3 circumf av oxide Rod A3 Hydrogen 4 The oxide thickness appears to lag the burnup jump at the beginning of the higher enrichment zone Oxide thickness [µm] Wall-average hydrogen [ppm] 5 Blanket zone (nat Higher enrichment Distance from rod bottom end [mm] Fig. 2a. Detail from Figure 1 at the enrichment zone interface Fig. 2b. Hydrogen in Rod A3 either side of transition. Another example of a clear increase in the oxide thickness on the outer surface of a Zircaloy tube at the axial elevation of the transition from the natural uranium blanket to the enriched fuel zone is shown in Fig. 3. However, in this case the tubing is an unfuelled (hollow) water rod the change of enrichment from the lower blanket to the enriched zone occurs in the fuel rods surrounding the water rod. The axial oxide thickness profile in the water rod has several qualitative similarities to the profiles of the fuel rods shown in Fig. 1. In the blanket region the oxide thickness increases steadily from the bottom end of the tube until the transition from the blanket to the enriched fuel. The factor 2 to 3 increase in the oxide thickness on the water rod at the enrichment transition is even larger than for the two fuel rods shown in Fig. 1, possibly due to the higher enrichment step in the fuel rods neighbouring the water rod (from.7 % to 3.2 % 235 U). As was seen in the fuel rods the slope of the increase in the oxide thickness is gentler than the sharp transition between the fuel enrichments in the neighbouring fuel rods, although in this case the different fuel rods in the immediate vicinity could have had very slight variations in the exact location of the transition. There is negligible heat production in the water rod so there is no positive surface heat flux on the outside of the tube. The inside of the water rod is filled with liquid phase water. Although the change in hydrogen concentration in the water rod from the lower end to a position well above the enrichment transition (i.e. from ~2 ppm to ~18 ppm) is even more dramatic than the change in oxide thickness, it should be remembered that an unknown fraction of the hydrogen has been picked up from the inner surface of the water rod in addition to the uptake from the outer surface, which makes direct comparison with the fuel rods difficult. The wall thickness of the water rod was,8 mm. 4

5 Fig. 3. Oxide thickness profile in lower part of a BWR Zircaloy-2 water rod operated for 8 annual cycles. (Note that the interior oxide thickness profile was not measured, but hydrogen pickup occurs from both sides). The fuel bundles from which the above examples are taken also had a natural uranium blanket at the top of the fuel rods, with a step change in enrichment at the transition from the enriched zone similar to that at the bottom. However, the corrosion behaviour vs. the change in enrichment at the transition to the upper blanket zone is very different to the lower end of the fuel rods, with no significant drop in corrosion coinciding with the decreased enrichment and burnup, see Fig Rod C1 circumf av oxide Rod A3 circumf av oxide Cs-137 (burnup profile) 4E+9 3E+9 Oxide thickness [µm] E+9 1E+9 Cs-137 activity 5 Spacer Higher enrichment zone Blanket zone (nat E Distance from rod bottom end [mm] Fig. 4. Oxide thickness and 137 Cs profiles in upper part of same two fuel rods as in Figure 1 5

6 II.B. Shadow corrosion in radiation gradient Shadow corrosion occurs on Zr-based alloys in contact with, or in close proximity to, dissimilar metallic alloys (including other Zr-based alloys) in the reactor core radiation field in BWRs or other reactors with similarly oxidising water conditions 2. The shadow oxide thickness varies roughly inversely with the distance between the metal surfaces, with the peak thickness occurring at contact positions. Significant shadow oxide usually occurs on BWR fuel rod cladding in proximity to Ni-alloy spacers or spacer springs. Typically, the peak oxide thickness at the spacer contact positions is approximately the same over most of the fuel rod length. This is unlike the behaviour at non-contact positions where there is a small gap between the metal surfaces, where the shadow oxide thickness tends to decrease from the lower part of the fuel rod towards the top. The decreasing shadow oxide thickness in these non-contacting areas appears to follow the variation in the bulk void fraction in the coolant higher up in the core there is a higher void fraction. However, in the case of spacers right at the lower end of the fuel rod where there is a steep axial gradient in burnup but no bulk boiling/void, a clear axial gradient is observed in the shadow oxide thickness even where there is a small gap between the spacer and the cladding. This suggests that one or more of the key environmental factors that cause shadow corrosion (e.g. radiation fields, radiolysis, etc.) increases relatively steeply over this fairly short distance at the lowest spacer position. In Figure 5 the cladding surface oxide thickness axial traces at 8 orientations around the circumference are plotted together with the relative 137 Cs activity (which is proportional to the local burnup) and the ratio of 134 Cs / 137 Cs activity (which is proportional to the thermal neutron fluence in the fuel rod). The fuel had operated in an assembly with Ni-base alloy spacers, resulting in shadow corrosion at the locations that had been within the spacers. The gap between the spacer and the cladding varies between the 8 different orientations such that the traces with the thickest shadow oxide are believed to be those with the smallest gaps (and at some points actual contact) between the spacer material and the cladding surface. The shadow oxide thickness on all the traces increases significantly over the height of the first spacer (roughly 25 mm axial length), with the largest relative increase at the angles with the lowest absolute thickness. In the location of the first spacer the shadow oxide was intact at the time of the measurements, while some of the shadow had flaked off the cladding at the spacer 2 position. In the figure, the 134 Cs / 137 Cs ratio has been scaled to match the peak (intact) shadow oxide thickness at spacer 2. The 134 Cs / 137 Cs ratio (slope and absolute level) then appears to also match fairly well with the measured shadow oxide at spacer 1, for the angles that had the thickest shadow oxide. By comparison, the burnup profile (corresponding to the 137 Cs profile) clearly does not match the slope of the shadow corrosion at spacer 1 when scaled to match the spacer 2 peak shadow oxide thickness. Hence, the neutron flux may be more important than burnup for shadow corrosion kinetics. Oxide thickness [µm] Spacer 1 Shadow Oxide EC Oxide Trace Angle Oxide thickness [µm] Oxide vs burnup and thermal neutron fluence profiles EC Oxide angle Shadow oxide from spacer #1 Shadow oxide from spacer #2 Relative thermal neutron fluence Relative burnup profile Cs-137 (Burnup) Av oxide thickness Cs-134/Cs-137 (therm. neutron fluence) 1,4 1,2 1,,8,6,4,2 Cs-134/Cs-137 (rel. thermal neutron fluence) Distance from rod bottom end [mm] Spacer 2, Distance from rod bottom end [mm] Fig. 5. Oxide thickness vs. 137 Cs (burnup) and 134 Cs / 137 Cs (thermal neutron fluence) profiles, detail at spacer #1 on left. 6

7 III. DISCUSSION OF OBSERVATIONS AND LIKELY CONTROLLING FACTORS The two specific examples of Zircaloy-2 corrosion behaviour observed close to the inlet of the BWR core described above (the marked increase in the uniform oxide at the elevation corresponding to a sharp increase in the fuel enrichment, and the axial profile in the shadow oxide thickness very low in the core) will now be discussed in terms of likely environmental factors controlling or affecting these specific phenomena. By its nature, this section is to some extent speculative, and it should also be borne in mind that in general it is possible to disprove but not to confirm a hypothesis by experimental observation. III.A. Role of surface heat flux and temperature on the uniform corrosion The corrosion rates of many zirconium alloys, including the Zircaloys, exhibit a strong temperature dependence in PWRs and in out-of-pile autoclaves 1. Intuitively, it might thus be assumed that the marked increase in the uniform oxide on BWR fuel rods corresponding to a step increase in the fuel enrichment, as illustrated in Fig. 1, is an effect of the increased surface heat flux (SHF) in the higher enriched zone. However, closer inspection of the oxide thickness profiles in the immediate vicinity of the enrichment step shows that the oxide thickness doesn t increase as sharply as the local burnup (proportional to the local lifetime-averaged power, and thus SHF and cladding surface temperature). Instead, the increase in the oxide thickness above the enrichment interface appears to lag the jump in burnup and SHF by some tens of mm, see Fig. 2a. This behaviour therefore does not support the increase in SHF being the primary cause of the increase in the oxidation rate. More importantly, exactly the same type of sharp uniform corrosion increase has also been seen on water rods operated adjacent to fuelled rods with a step change in enrichment. Fig. 3 shows a water rod with a very marked increase in oxide thickness at the elevation of the enrichment transition in neighbouring fuel rods. There is no net surface heat flux from the water rods, so there is no step change in either the SHF or the temperature rise from the bulk coolant to the outer surface of the water rod in connection with the change in enrichment in the neighbouring fuel rods. This observation clearly precludes the SHF as an explanation for the increase in oxide thickness observed at the same elevation as the change in enrichment in the surrounding fuel rods. Finally, could the change in the oxide thickness be a result of a change in the coolant water bulk temperature? Again, such an effect is not supported by the flat oxide profile above the transition in fuel rod C1 (Fig. 1) and the water rod (Fig. 3), while there is actually an accelerating rate of increase in the bulk water temperature in the single-phase region (up to roughly 5 mm elevation) that is proportional to the increasing average power of the fuel rods. These observations therefore clearly preclude both the SHF and the clad wall surface temperature as an explanation for the increase in oxide thickness observed at the same elevation as the change in enrichment. III.B. Role of irradiation damage effects in cladding metal on the uniform corrosion Irradiation damage is known to affect the microstructure of the zirconium alloys during in-pile operation 28. For some alloys (like Zircaloy-2) the irradiation has been found to result in an increase in the corrosion rate 1, 12. The increase in corrosion rate and hydrogen pickup appears to coincide with the dissolution of intermetallic particles This type of irradiation damage, involving atomic displacements, depends far more on the fast neutron fluence than on lower energy neutrons and other types of radiation (beta and gamma) dose to the fuel cladding and water rod tubing 1. However, the fast neutron fluence axial profile/gradient is far smoother than the transition observed in the uniform oxide thickness. There is essentially no change in the fast neutron fluence corresponding to the enrichment transition in the fuel pellet stack. In fact, even the thermal neutron fluence, which reacts more than the fast neutron fluence to local changes in the material environment like the absorption, scattering and fission cross-sections, behaves relatively smoothly as shown by the 134 Cs / 137 Cs ratio in Fig. 5. It thus seems that irradiation damage effects in the underlying zirconium alloy metal due to neutron fluence does not play a significant role in the marked increase in the uniform oxide thickness at the transition in enrichment seen in both fuel rods (Fig. 1) and water rods (Fig. 2). And since other forms of ionizing radiation cause less displacements of atoms in the metal 1, it can be concluded that irradiation-induced damage in the metal is not the explanation for the change in the uniform oxide thickness at the enrichment transition. 7

8 III.C. What irradiation effects have a profile compatible with the observations? It has been shown above that neutron fluence profiles/gradients (both for fast and thermal neutrons) are far smoother than the change in uniform oxide observed at the enrichment transition. Since the thermal neutron fluence has a relatively smooth axial profile it follows that beta and gamma radiation resulting from neutron activation of material outside of the fuel rods should also have a similar, smooth axial profile. So, the question is then: which (if any) type and source of radiation exhibits an axial gradient at the enrichment transition similar to that observed in the uniform oxide on both fuel rod cladding and neighbouring water rods? The change in enrichment corresponds also to a step change in the fission density within the pellets, as seen from the burnup profile derived from the 137 Cs intensity. So, a candidate source of relevant radiation could be particles/radiation created within the fuel pellets proportional to the fission rate, with sufficient range to escape the fuel rod but at the same time only a relatively limited range of some cm in water. Examples of forms of radiation that might have roughly the appropriate range are beta with an energy above about 1 MeV, which should be able to penetrate the cladding but not travel too far in the water. Similarly, characteristic k-α X-rays for uranium (with an energy of about 95 and 98 kev) might have a suitable range with an attenuation µ in UO 2, Zr and water of about 2, 7 and.15 cm -1 respectively 16. Characteristic X-rays are generated by many types of radiation in the pellet, and uranium k-α X-rays emitted close to the surface of the pellets can easily escape. Another important factor is that the pellet stack itself has such a high density and attenuation, that the beta, gamma and X-rays escaping from the fuel rods are predominantly emitted in the radial direction, which further reduces their effective range in the axial direction 17. The intensities, ranges and directions of different forms of radiation has not been studied in this work, but it is proposed that such a study could be performed in order to identify particles or radiation with a range or intensity profile that matches the observed variation in the uniform oxide at the enrichment transition. Such studies have been performed for calculating Cherenkov light generation from irradiated fuel rods, which also involves the reaction of ionizing radiation emitted from fuel rods with water 17. This could help in identifying the mechanisms and factors responsible for the in-pile uniform corrosion of Zr-alloys in a BWR environment. Although both the fuel rods shown in Fig. 1 and the water rod shown in Fig. 3 exhibited an increase in the uniform corrosion at an elevation corresponding to the enrichment step, a detailed examination also showed that the oxide thickness profile lagged the burnup profile at and immediately above the enrichment transition (Fig. 2a). This observation qualitatively supports the idea that radiolysis of the water plays a key role in the observed uniform corrosion behaviour, since the upward flow of the coolant (with a lifetime averaged velocity on the order of 1-2 m/s) would tend to entail a small but finite lag in the effects of the radiolysis products on the cladding surface corrosion relative to the axial position at which the ionizing radiation causing the radiolysis is emitted from the fuel. Such a lag would likely be due more to the transport processes for the active radiolysis species than to the reactions involved in producing these species, which occur on time-scales of 1-15 to 1-6 seconds 12. III.D. Apparent saturation behaviour of environmental factor Another aspect of the marked increase in the uniform corrosion seen in fuel rod C1 in Fig. 1, and in the water rod in Fig. 3, is the essentially flat oxide thickness profile above the transition zone despite a large further increase in the overall fluence and burnup above the transition. So maybe the effect of the environmental factor that strongly affects or controls the corrosion in these cases tends to saturate at a level that is higher than the level in the blanket region, but lower than in the enriched zone. One possibility is that the factor or mechanism that determines the increase at the enrichment transition no longer operates higher up; however, in that case it would be a remarkable coincidence that the effect stops directly above the enrichment transition. A more likely type of factor in that case would be some aspect of the water chemistry, such as an equilibrium concentration of certain active species that are produced (e.g., by radiolysis) and at the same time consumed or eliminated in competing processes; this topic is discussed further in section III.E below. Interestingly, the saturation effect in the oxide thickness above the enrichment transition is not seen in fuel rod A3 in Fig. 1, where the oxide thickness profile above the transition instead closely follows the burnup profile. Yet rod A3 also showed a marked oxide thickness increase at the enrichment transition, and the oxide thickness in the natural uranium blanket zone below the transition does not match the burnup profile but is very similar to the oxide profile in rod C1. This suggests that the uniform corrosion, at least for rod A3, may have involved more than one rate-determining mechanism (relating to the material behaviour) over its lifetime. 8

9 III.E. Discussion on elusive environmental factor controlling the uniform corrosion in the core inlet region In section III.A the surface heat flux and metal/oxide interface temperature were discussed and precluded as the explanation for the marked increase in the uniform oxide thickness at the level of the enrichment step. This conclusion contrasts with the conclusion of Franklin and Li regarding nodular corrosion in BWRs, where they discussed observations of both increases in the nodular corrosion at enrichment steps from the blanket zone for low-powered rods as well as drops in the nodular corrosion at enrichment steps from the blanket zone for high-powered rods 14. Since the water rod shown in Fig. 3 in this paper is a fully unheated surface, yet shows a clear increase in the uniform corrosion corresponding to the enrichment step, a conceivable explanation for the difference in observed behaviour compared to the earlier reported work is that the uniform corrosion mechanism is controlled by different factors than the nodular corrosion described in that earlier study 14. In all the cases reported in this paper there were no clear indications of nodular corrosion. In section III.B it was concluded that damage to the underlying metal caused by neutron flux is not consistent with the observed oxide thickness increase at the level of the enrichment step. In section III.C it is argued that beta, gamma and X-ray radiation from the pellets might be consistent with the observed correlation between the oxide and the burnup profiles close to the enrichment step. Rishel and Kammenzind have compared Zircaloy-4 in-pile corrosion with different gamma to neutron flux ratios and concluded that gamma contributes to enhancing post-transition corrosion rate 19. The authors proposed that this was due to the impact of the high-energy photon (gamma and X-ray) radiation on the properties of the oxide layer. Similar mechanisms have been proposed earlier even for lower energy photon radiation, e.g. by Cox and Fidleris 2. Despite consensus with the earlier proposals 21, 22 that beta, gamma and/or X-ray radiation could correlate with the observations in section III.C above, it is also concluded that the slight lag in the corrosion enhancement compared to the sharp step increase in burnup at the enrichment transition better supports a mechanism involving radiolysis effects in the flowing water than a direct impact on the cladding material or oxide properties. Franklin 21, based on a re-interpretation of inpile electropotential measurements 22 has argued that radiolysis of oxygenated single-phase water results in a substantial increase in the electrochemical potential (ECP) between the surface of the oxide and the coolant. He also proposed that changes in the potential due to different radiolysis product concentrations could explain the difference in nodular corrosion seen on either side of the enrichment transition described before 14, invoking a critical range of potential to account for both the increase and decrease in nodular corrosion at the transition to higher enrichment. III.F. Possible environmental factors controlling the shadow corrosion in the core inlet region Unlike the case for the sharp increase in the uniform corrosion at the elevation of the enrichment transition, the shadow corrosion profile in the single-phase region illustrated by the shadow corrosion at the first spacer in Fig. 5 appears to be much more similar to the expected irradiation flux profiles in the core. It seems that the (thermal) neutron fluence profile derived from the 134 Cs / 137 Cs ratio, when scaled to the peak remaining (i.e. unflaked) shadow oxide at the second spacer also matches reasonably well with the absolute level and profile of the shadow corrosion at spacer 1. There is no shadow corrosion in the enrichment transition zone so there are no experimental data on how the shadow oxide would behave at the jump in the enrichment and burnup. However, the slope (gradient) of the shadow oxide thickness in spacer 1 does not match with the burnup gradient which is flatter in the blanket zone. It has been proposed that shadow corrosion is a form of radiation-enhanced galvanic corrosion, where the in-bwr conductivity of both the zirconium oxide film and the water gap are critical factors 23. Other studies have supported this hypothesis 24. While the water conductivity becomes relatively more important where there is a larger liquid-filled water gap, the conductivity of the zirconium oxide is believed to be the determining factor where there is direct contact or a very thin water gap. The electrical conductivity of both the zirconium oxide and the water during operation are, in turn, determined by radiation effects; however, their dependence on different types of radiation and energies may differ. When considering the shadow oxide profile in spacer 1 as well as its relationship to the peak oxide at spacer 2, it should also be remembered that the shadow corrosion rate expected from the galvanic corrosion hypothesis not only depends on the conductivity of the oxide layer, but also (inversely) on its thickness. In other words, all else being equal, the thicker the shadow oxide layer is the slower it is expected to grow. So, it would be helpful to use a model of the shadow corrosion mechanism to perform a more realistic analysis of the relationship between the shadow oxide and radiation fields, which was outside of the scope of this work. 9

10 V. CONCLUSIONS This paper describes an attempt to elucidate possible mechanistic effects on the uniform corrosion, shadow corrosion and hydrogen pickup phenomena in Zircaloy-2 in BWR conditions. Four specific observations from several fuel rod hot cell examinations have been discussed and compared with the local variations of some parameters (such as neutron and gamma fluxes, heat flux, boiling, and radiolysis) that characterize the in-core environment. Some limited (but important) clad hydrogen pickup data were presented but not discussed in this paper in terms of parametric sensitivities of the hydrogen pickup and migration mechanisms. A marked increase in the uniform corrosion was observed at the elevation corresponding to a step increase in fuel pellet enrichment in the lower part of the fuel assembly. It has been shown that this phenomenon is not related to a change in heat flux or neutron flux or the onset of boiling. Instead, it is suggested that the effects of radiolysis in the single-phase coolant, due primarily to beta, gamma and/or X-ray emissions from the fuel, could potentially be consistent with the observations. The observed axial variation in oxide thickness due to shadow corrosion near the core inlet, however, appeared to have a different behaviour to the uniform oxide. The slope of the shadow oxide thickness profiles for the angles without contact with the spacer (i.e. with a small open water gap) in the non-boiling zone appeared instead to be consistent with the neutron flux profile. This implies that the rate-controlling kinetic processes are different for uniform and shadow corrosion films, and likely changes the hydrogen pickup process too. These observations, in addition to others, offer an excellent opportunity to better understand corrosion and hydriding mechanisms and should be included in future industry-sponsored R&D. ACKNOWLEDGMENTS The authors would like to acknowledge Studsvik for performing the hot cell PIE described in the paper. Vattenfall financed the PIE on the fuel rods, and Vattenfall, OKG and WSE jointly financed the PIE on the water rod. REFERENCES 1. "Waterside corrosion of zirconium alloys in nuclear power plants", International Atomic Energy Agency, Vienna, IAEA-TECDOC-996, (Jan. 1998). 2. Chen, J.-S. F. and Adamson, R. B., "Observations of Shadow Phenomena on Zirconium Alloys," ANS Int. Topical Mtg on Light Water Reactor Fuel Performance, West Palm Beach, FA, USA, April 17-21, Lefebvre F. and Lemaignan C., Irradiation Effects on Corrosion of Zirconium Alloy Cladding, J. Nucl. Mat. 248, , Adamson R. B., Lutz D. R., Davies J. H., Hot Cell Observations of Shadow Corrosion Phenomena, Proceedings Fachtagung der KTG-Fachgruppe, Brennelemente und Kernbautelle, 29 Februar/1 Marz 2, Forschungszentrum Karlsruhe, 2 5. Y.-J. Kim, R. Rebak, Y.-P. Lin, D. Lutz, D. Crawford, A. Kucuk, B. Cheng, Photoelectrochemical Investigation of Radiation Enhanced Shadow Corrosion Phenomenon, Zirconium in the Nuclear Industry 16th International Symposium, ASTM STP 1529, pp BWR Zircaloy Shadow Corrosion and Hydriding Meeting July Freeport, Maine: Impact, Mechanism, Testing and Mitigation. EPRI, Palo Alto, CA: B. Andersson, M. Limbäck, G. Wikmark, E. Hauso, T. Johnsen, R. G. Ballinger and A.-C. Nystrand, Test Reactor Studies of the Shadow Corrosion Phenomenon, Zirconium in the Nuclear Industry: 13th International Symposium, ASTM STP 1423, pp , Cheng, B.et al., "Development of a Sensitive and Reproducible Steam Test for Zircaloy Nodular Corrosion", Zirconium in the Nuclear Industry: 7th International Symposium, ASTM STP 939 (1987). 9. E. Ibe et al., Radiolytic Environments in Boiling Water Reactor Cores, Journal of Nuclear Science and Technology, 24:3, (1987) 1. Marlowe M. O., Armijo J. S., Cheng B. and Adamson R. B., Nuclear Fuel Cladding Localised Corrosion, Proc. ANS Int l Topical Meeting LWR Fuel Perform., Orlando, FL, USA, April 21-24, 1985, , April 21-24, Shimada S., Cheng B., Lutz D., Kubota O., Ichika N. and Ibe H., In-Core Tests of Effects of BWR Water Chemistry Impurities on Zircaloy Corrosion, Journal of ASTM International, Vol. 2, No. 5, Paper ID JAI12373, May 25 1

11 12. Garzarolli F., Schumann R. and Steinberg E., Corrosion Optimized Zircaloy for Boiling Water Reactor (BWR) Fuel Elements, Zirconium in the Nuclear Industry, Tenth Int l Symposium, ASTM STP 1245, A. M. Garde and E. R. Bradley, Eds., American Society for Testing and Materials, Philadelphia, , Ito K., Shimada S., Levin H., Adamson R. B., Chen J. S. F., Oguma M., Cheng B., Ikeda T., Takei K. and Ishii Y., Effects of Water Chemistry Impurities on Corrosion of Zr-Alloys under BWR Condition, Proc. ANS Int l Topical Meeting LWR Fuel Perform.; West Palm Beach, FL, USA, April 17-21, 1994, , Franklin, D., & Li, C.Y. (1987), Effects of heat flux and irradiation-induced changes in water chemistry on Zircaloy nodular oxidation, Zirconium in the Nuclear Industry: 7 th International Symposium, ASTM STP 939 (1987). 15. H. Christensen, Fundamental Aspects of Water Coolant Radiolysis, SKI Report 26:16, April 26. (ISSN ). 16. National Institute of Standards and Technology (NIST), E. Branger,1 S. Grape, S. Jacobsson Svärd, P. Jansson and E. Andersson Sundén, On Cherenkov light production by irradiated nuclear fuel rods, J. Instrumentation, Vol. 12, June S. Le Caër. Water Radiolysis: Influence of Oxide Surfaces on H2 Production under Ionizing Radiation, Water, 211, 3, pp D. M. Rishel and B.F. Kammenzind, The Role of Gamma Radiation on Zircaloy-4 Corrosion, Zirconium in the Nuclear Industry: 18 th International Symposium, Charlotte, USA (216). 2. Cox, B., and V. Fidleris, Enhanced Low-Temperature Oxidation of Zirconium Alloys Under Irradiation, Zirconium in the Nuclear Industry: 8 th International Symposium, ASTM STP 123, (1989), pp D. G. Franklin, Performance of Zirconium Alloys in Light Water Reactors, Kroll Award Paper 22. Rishel, D. M., et al., In Situ EIS Measurements of Irradiated Zircaloy-4 Post- Transition Corrosion Kinetic Behavior, Zirconium in the Nuclear Industry: 15 th International Symposium, ASTM STP 155, 29, pp G. Lysell et al., On the shadow corrosion mechanism for Zirconium alloys. Proc. TopFuel N. Treeman, Electrochemical study of corrosion phenomena in zirconium alloys, M. Sc. Thesis, MIT, June T. Miyashita et al., Corrosion and hydrogen pickup behaviors of cladding and structural components in BWR high burnup 9x9 lead use assemblies, ANS LWR Fuel Performance Conference, San Francisco, 27, paper Etoh, Y. et al., The Effect of Microstructure on the Corrosion Behavior of Zircaloy-2 in BWRs, Zirconium in the Nuclear Industry: 12 th International Symposium, ASTM STP 1354, 2, pp S. Valizadeh et al., Effects of Secondary Phase Particle Dissolution on the In-Reactor Performance of BWR Cladding, J. ASTM International, Vol. 8, No. 2, Paper ID JAI R. M. Kruger and R. B. Adamson, Precipitate behavior in zirconium-based alloys in BWRs, J. Nucl. Mater., Vol. 25, 1993, pp