contribution of research to the understanding and...
TRANSCRIPT
1
Henri Coriou Award Keynote Lecture
Contribution of Research to the Understanding and Mitigation of
Environmentally-Assisted Cracking in Structural Components in Light
Water Reactors
David TICE
Amec Foster Wheeler
Walton House, Birchwood Park, Warrington WA3 6GA, United Kingdom
Environmentally-assisted cracking (EAC) is a potential threat to the safety and integrity of water-wetted components in operating water-cooled nuclear power plant. Two forms of EAC are commonly distinguished, depending on the form of loading contributing to damage: stress corrosion cracking and corrosion fatigue. A number of instances of in-service degradation due to EAC have occurred in operating plants worldwide, often leading to unplanned plant outages. Understanding the causes of EAC is essential to minimise the loss of plant availability due its occurrence and to avoid the possibility of catastrophic failure, for example, if a crack grew to a critical size in a major pressure boundary component. This paper will describe some examples of these phenomena in the main materials of construction of pressure boundary and other critical components in pressurised and boiling water reactors. Over the last several decades, substantial research programmes have been carried out in a number of laboratories worldwide, aimed at furthering understanding of the processes leading to EAC in order to manage occurrences in plant and minimise future failures. Selected areas of research on EAC in light water reactor environments are discussed. Corrosion fatigue in low alloy pressure vessel steels was the subject of considerable attention in the 1980s and early 1990s because of its potential threat to pressure vessel integrity and the publication of data suggesting that a major influence of environment on fatigue crack growth in some laboratory tests. The author’s research provided insight into the conditions under which the major environmental effects occur and contributed to the development of an AMSE Code Case for PWR conditions which provided a means of screening based on steel sulfur content and loading conditions. More recently the research focus in this area has moved to austenitic stainless steels, again providing support to Code Case development and furthering mechanistic understanding. A recent review of knowledge gaps for EPRI provides a basis for future research on environmentally-assisted fatigue and will inform the development of new assessment methodologies. A key area of current study concerns differences in loading conditions between specimens in laboratory tests and plant components subject to transient loading. In the case of stress corrosion cracking (SCC), stainless steels have shown the greatest propensity to cracking in BWRs, whilst Alloy 600 has been a major cause of in-service failures in PWRs, both on the primary side, as recognised by Coriou in the early 1960s, and in secondary environments where a number of different corrosion-related failure processes have been identified. High strength alloys such as Alloy X-750 used for fastener applications have also caused failure in both reactor types. For austenitic materials, SCC susceptibility is enhanced by irradiation, resulting in failures in core internals components. Ferritic steels also undergo SCC under some specific circumstances but are generally more resistant than the lower chromium austenitic materials.
Keywords: Environmentally-assisted cracking, EAC, stress corrosion, corrosion fatigue, PWR, BWR
2
INTRODUCTION
A range of materials are used for the construction of nuclear power generation plants and, in most cases, adequately fulfil their design intent. Nevertheless, some failures have occurred, either due to fatigue (including thermal and corrosion fatigue) or stress corrosion cracking (SCC), both of which are forms of environmentally-assisted cracking (EAC). A number of different reactor types are operational worldwide but the majority of commercial power generation plants are water-cooled, including pressurised water reactors (PWR), boiling water reactors (BWR) and pressurised heavy water reactors (PHWR). The focus in the current paper is on PWRs and BWRs, collectively referred to as light water reactors (LWRs).
Figure 1: The three prerequisites for EAC
As shown in Figure 1, there are three pre-requisites for environmentally-assisted cracking, a susceptible material, an appropriate environment and a tensile stress. EAC encompasses both stress corrosion cracking (SCC), which occurs under steady or slowly-changing load, and corrosion fatigue which occurs under cyclic loading. It is often convenient to distinguish two stages of EAC: crack initiation and sub-critical crack growth, Figure 2 [1]. When EAC occurs, crack initiation often occupies a substantial portion of the total time to failure. As shown in Figure 2, EAC initiation encompasses several distinct processes: the development of material damage (precursor formation), nucleation of small defects, and the growth and coalescence of small cracks. The subsequent phase of damage is sub-critical crack growth which describes the growth of a dominant crack under fracture mechanics control up to a critical defect size, beyond which structural failure may occur by fast fracture. Figure 3 illustrates crack growth behaviour for crrosion fatigue and stress corrosion cracking, respectively.
Figure 2: Stages of EAC. Adapted from reference 1
3
a) b)
Figure 3: Sub-critical crack growth for (a) corrosion fatigue and (b) stress corrosion cracking
EAC IN LIGHT WATER REACTOR PLANT
EAC is a problem that has affected both BWRs and PWRs. A number of plant failures due to stress corrosion cracking or corrosion fatigue have resulted in coolant leakage rather than catastrophic failure and many other instances have been detected by in-service inspection. Many of the reactor circuit structural materials have been affected by EAC to some extent, although there are significant differences between PWRs and BWRs in terms of the materials most affected and factors influencing EAC susceptibility. The main classes of materials of relevance are:
• Ferritic steels. These include low alloy steels such as A533B and A508-III used for main pressure vessels, carbon steels used for PWR steam generator tubesheets, as well as for piping in some BWRs and various other piping applications, and high strength quench and tempered steels used for bolting applications.
• Stainless steels. Type 300 series austenitic stainless steels are widely used in PWRs and many BWRs for piping, and for pump and valve bodies and a variety of other applications. Martensitic stainless steels are employed where higher strength is required such as in valve stems and some fastener applications. Precipitation-hardened alloys such as A-286 and 17-4PH are also used for high strength applications. Alloy 800, with a much higher nickel content of 30-35%, is used for steam generator tubing in many PWRs of German manufacture and for some steam generator replacements.
• Nickel-based alloys. Alloy 600 has been extensively used in PWRs, for steam generator tubing and for a variety of other applications such as control rod drive penetrations, bottom head penetrations and steam generator divider plates. Applications in BWRs include the core support structure. The weld metal alloys 182, 82 (and, in Japan, 132) are also widely employed. Alloy 690 (and weld metal alloys 52, 152 and variants) are being widely used for replacement and new build due to relatively poor service experience of Alloy 600. Precipitation-hardened alloys X-750 and 718 are employed for higher strength requirements such as internal bolting, fasteners and springs.
The main focus in the current paper is on carbon and low alloy ferritic steels and austenitic stainless steels. The main factors influencing in-service degradation by SCC or corrosion fatigue will be discussed and some examples of plant failures will be described.
4
STRESS CORROSION CRACKING
Austenitic stainless steels Type 300 stainless steels have suffered more extensively from SCC in BWRs than in PWRs. This is because of the inherent differences in water chemistry between these two reactor types. Due to the high radiation field, oxidizing species, including oxygen and hydrogen peroxide, are produced from the aqueous coolant by radiolysis; this increases the electrochemical potential (ECP) within the reactor circuit. In PWRs, hydrogen is added to the coolant to remove these radiolytically generated species so the ECP is normally reducing (around -700mV SHE). Initially, BWRs operated without oxygen addition (so-called “normal water chemistry”, NWC), which resulted in a much higher potential, in excess of +100mV SHE. More recently, many BWRs have operated with “hydrogen water chemistry” (HWC), resulting in potentials intermediate between PWRs and BWRs which operate NWC. It is not possible to add sufficient hydrogen to fully mitigate radiolysis in some regions of the BWR circuit due its partitioning to the steam phase in the boiling coolant. Many BWRs operating HWC also use noble metal addition which reduces ECP further by catalyzing hydrogen-oxygen recombination [2]. A substantial number of instances of cracking of Type 304 stainless steels have occurred in BWR plants and a number have resulted in coolant leakage, Most of the early failures, in the 1970s and 1980s, were associated with thermal sensitization of the steel [3]; chromium carbide precipitation occurs at grain boundaries within a critical temperature range which results in chromium depletion at grain boundaries and increased susceptibility to localized corrosion in the presence of the oxidizing conditions present in BWR normal water chemistry. The presence of sufficient tensile stress under these conditions can produce SCC. In the heat-affected zone (HAZ) of welds, both sensitization and high residual stress can be present. In the early years of BWR operation, the importance of close control of contaminant species such as sulfate and chloride which further enhance SCC was not adequately appreciated. The subsequent operation of many BWR plants with hydrogenated water chemistry and improved water chemistry control substantially decreased the occurrence of SCC in sensitized stainless steel components. Moreover, following early failures, many plants replaced high carbon Type 304 stainless steel with lower carbon grades such as 304L, 316L, 316L(N) or 316NG which are much less susceptible to sensitization. Stabilized grades such as Type 321 and 347 which are widely used in German designed BWRs, are also less prone to sensitization and have shown much reduced susceptibility to SCC; nevertheless, some failures have occurred, attributed mainly to inadequate levels of the stabilizing elements, titanium or niobium [4]. More recently, a number of SCC occurrences have been observed in non-sensitized, low carbon stainless steel components in BWRs, most extensively, but not exclusively, in plants not operating hydrogen water chemistry [5]. Most of these more recent events were not caused by sensitization but by local cold work. Several sources of cold-work are significant. In many cases cracks appear to initiate from a cold-worked surface layer produced by machining or grinding; an example is shown in Figure 4 [6]. Another source of cold work is welding-induced shrinkage which can induce plastic strain levels in excess of 20% under some circumstances [7]. Figure 5 shows a microhardness scan within the weld heat-affected zone (HAZ) of a BWR piping weld showing elevated hardness levels due to plastic strain which extend several millimetres from the weld fusion line into the parent stainless steel [8]. A further source of cold work is cold-forming, sometimes encountered in pipe bends as a result of inadequate heat treatment, or inadvertent deformation produced by fit-up operations, for example.
5
Figure 4: Crack in a BWR core-shroud initiating
in a cold-worked surface layer [6]
Figure 5: Microhardness variations
in HAZ of BWR PLR piping [8]
In contrast to BWRs, only a relatively small number of stress corrosion failures have occurred in austenitic stainless steels in PWRs. The majority of these have occurred in occluded locations where oxygen may be trapped following plant fill and anionic impurities such as sulfate and chloride may accumulate due to inadequate refreshment from the main primary circuit. Most instances of cracking in occluded environments have occurred in components that operate below the operating temperature of the main primary circuit, usually below 200°C. Examples include control rod drive mechanism (CRDM) canopy seals that ensure the leak tightness of threaded joints, dead-leg or low-flow pipework connected to the primary circuit such as sample and drain lines, components of the residual heat removal (RHR) system and chemical and volume control system (CVCS) [9 ,10]. Excessive cold work due to material deformation during manufacturing or post-manufacturing is a contributing factor in many cases [10], but is not essential in the presence of oxygen together with anionic contaminants. Sensitization has been implicated in a only a few of instances of cracking in occluded locations in PWRs. A much smaller number of SCC occurrences in stainless steel components in PWRs have been observed under nominally free-flowing conditions, all of which appear to be associated with high levels of local cold work, Figure 6 [10]. Surface cold-work appears be present in many cases which suggests that it may be a pre-requisite for crack initiation. A single incidence of cracking has been reported in a stainless steel safe-end in a PWR primary circuit, but cracking was confined to the cold-worked surface layer [11]. However through-wall cracking has been reported in non-isolable stainless steel piping elbows in the residual heat removal system of a PWR in China, in the vicinity of a weld. The material appeared to be of unusually high hardness suggesting a degree of cold work and evidence of surface grinding may have contributed to initiation after 17 years plant operation [15].
Figure 6: Influence of cold work on cracking events observed in PWR plant [10]. Note that all
occurrences under free-flowing conditions are associated with substantially elevated hardness levels
6
A significant proportion of the PWR failures under nominally free-flowing conditions have occurred in pressurizer heater sheaths [12, 13]. Several factors may make these components more susceptible to IGSCC in non-contaminated primary coolant compared to most other stainless steel components in the primary circuit. These factors include the higher operating temperature compared to the rest of the plant, the use of cold swaging which introduces cold work and tensile residual stress, thermal cycling introducing dynamic strain and possible fatigue damage which may enhance initiation, heat flux which may cause pH elevation in creviced locations such as at heater support plates or under surface laps, and the presence of surface cold work. Experimental studies have shown that all the above factors can increase susceptibility of austenitic stainless steels to IGSCC [12]. Two other failures affecting pressurizer heaters occurred close to the weld connecting the heater to the pressurizer penetration, one at Braidwood Unit 1 in the US [14] and one in Sizewell B in the UK [16]. In the former case, SCC occurred in a creviced location rather than free-flowing coolant and the presence of local oxidising conditions was suspected. In the case of Sizewell B, leakage at the heater well insert location appears to have been a consequence of prior heater sleeve failure due to SCC, thereby allowing water ingress to the magnesium oxide insulation around the heater which caused swelling of the insulation and deformation of the surrounding stainless steel. Despite the small number of reported occurrences of SCC in stainless steel components exposed to flowing primary coolant, a number of studies have shown that propagation of a pre-existing crack can occur in the presence of levels of cold work of 15-20%. In an early study, Shoji et al. [17] examined the effects of levels of deformation imparted by either cold or warm rolling Types 304L and 316L stainless steel. Specimens tested were deformed to nominal yield strengths of 450, 750 or 1,000MPa. Figure 7 shows the effect of yield stress and material type on crack propagation rates of austenitic stainless steels. No SCC was found below a yield stress of 450MPa, but increasing the deformation level to 750MPa produced SCC with growth rates around 2x10-10 m/s in PWR conditions and higher in the more oxidising BWR environment. Growth rates increased further for a yield stress of 1,000MPa.
There was no significant effect of temperature between 290 and 340°C.
Figure 7: Effects of yield strength on SCC propagation susceptibility of Types 304L and 316L austenitic
stainless steels [17].
A number of subsequent studies have found that specimen orientation plays a significant role in the SCC susceptibility of cold worked stainless steel in PWR environments. Figure 8 shows data reported by Nouraei et al. which show the crack growth rates of specimens tested parallel (S-L), transverse (T-L) and orthogonal (L-S) to the original cold work direction. In this study, the S-L orientation, in which crack growth is parallel to the cold working direction produces the greatest susceptibility to SCC and the L-S orientation (crack propagation orthogonal to cold work) the least [18]. Crack propagation in the S-L direction is relatively uniform whereas, for the T-L orientation, crack propagation frequently appears as fingers of intergranular cracking ahead of the main crack front, with secondary intergranular cracking perpendicular to the main fracture surface, Figure 9 [18, 19].
A further major factor influencing SCC propagation in cold worked stainless steel is the material composition, especially the sulfur content. A number of tests have clearly shown that the presence of
7
sulfur near the upper specification limit for Type 304 or 304L stainless steel severely retards crack growth as shown in Figure 10 [18]. More recent tests have shown that even sulfur contents in the range 0.004 to 0.006% have an appreciable effect on crack growth rate, Figure 11 [19].
Figure 8: Crack Growth Rate of 20% cold rolled stainless steel, in different orientations, as shown on
right. (UDR= unidirectional rolling, BDR = bidirectional rolling.)
All specimens 0.001% sulfur except R581 and R583 (0.004%) [18]
(a)
(b)
(c)
(d) Figure 9: Optical images of fracture surfaces of S-L specimen (a) and T-L specimen (b). Corresponding
SEM images show intergranular morphology for S-L (c) and mainly transgranular but with secondary
intergranular cracking for T-L (d) [18, 19]
The reasons for the effect of sulfur have been the subject of considerable discussion. There is now clear evidence that enhanced corrosion occurs as a result of dissolution of manganese sulfide inclusions from the steel. Injection of 10ppm sulfide (as the sodium salt) close to the tip of the crack was found to arrest a growing crack in a very low (0.001%) sulfur steel [21]. The fracture surfaces of cracks in high sulfur steels were found to be covered with a much thicker layer of oxide (magnetite/spinel) crystallites than was observed for low sulfur steel, although the inner oxide layer appeared similar. A similar thick oxide layer was present for the specimen from the sulfide injection experiment. Cracks in both the high sulfur steel and in the sulfide injection specimen were
8
considerably more blunt than for low sulfur steels. There is still a continuing debate, however, as to whether the retardation observed is due to a reduction in crack driving force due to corrosion blunting, or whether other processes may play a role, such as enhancement of creep due to corrosion-induced vacancy injection or reduced passivity due to an effect of sulfur on oxide adhesion.
Figure 10: Crack growth rates for steels with different sulfur content under constant load. No
macroscopic crack extension occurred for the material with the highest sulfur content [18]
Figure 11: Results of SCC growth rate tests conducted in 338°C DW (a) with an initial K of 38
under constant load, showing reductions in the SCC growth rate with increasing sulfur content
[20].
11.2
11.4
11.6
11.8
12.0
12.2
12.4
12.6
500 1,000 1,500 2,000 2,500 3,000 3,500 4,000 4,500 5,000 5,500
Cra
ck
Le
ng
th (
mm
)
Time (hours)
1,000s
9,000s
hold
Constant load Constant load
Dosing with
sulfide
Stop dosing
0.001Hz
cycling
9,000s
hold
Constant
load
Dosing with
LiOH only
Constant
load
Dosing
with
LiOH
+ sulfide
Constant
load
Dosing
with
LiOH
only
Figure 12: Retardation of crack growth observed in a low sulfur steel by injection of sulfide close to the
crack tip [21]
9
The observation of relatively high rates of crack propagation in cold worked stainless steel in good quality PWR primary water appears at first sight to be at variance with the good performance of stainless steel components in PWR service. However, other laboratory studies indicate that, even for relatively high levels of cold work, SCC initiation is very difficult except in the presence of higher levels of surface cold work combined with tensile residual stress, or under dynamic straining. Increased susceptibility of austenitic stainless steels to SCC is observed in both BWRs and PWRs in the presence of irradiation; this phenomenon is termed irradiation-assisted stress corrosion cracking (IASCC) [22]. In BWRs, IASCC has been observed in a number of components, whereas in PWRs it is restricted primarily to baffle former bolts. Radiation has an effect on the environment through radiolysis which increases the electrochemical potential and hence SCC susceptibility of stainless steel. The presence of dissolved hydrogen in PWRs suppresses this effect almost entirely. Hydrogen water chemistry in BWRs substantially reduces ECPs compared to normal water chemistry, especially if combined with noble metal additions but not sufficiently to completely prevent SCC in some parts of the circuit. Radiation has a number of effects on materials including radiation-induced segregation (enhancement or depletion) of species at grain boundaries (RIS), and microstructural changes such as the formation of interstitial and vacancy loops, resulting in hardening and non-uniform deformation behaviour (channelling), and, at higher doses, the formation of voids and precipitates. The influence of these individual changes on SCC is difficult to investigate because the changes occur concurrently with irradiation. Irradiation-induced depletion of chromium effectively results in sensitization, even in low carbon steels, which would be expected to promote SCC under oxidizing (BWR NWC) conditions. However, correlations of IASCC susceptibility in BWR environments with measured chromium depletion levels are not good [24] and this mechanism does not explain cracking under PWR conditions. Irradiation hardening would be expected to increase SCC susceptibility in both BWRs and PWRs by analogy with the effects of other hardening processes such as cold work and precipitation hardening. High levels of bulk silicon are known to enhance SCC of stainless steels under both PWR and BWR conditions and it is possible that silicon enrichment at grain boundaries due to irradiation may have a similar effect. The interactions of these numerous effects and their results on IASCC are not yet fully understood. Recent work suggests that the localisation of deformation due to irradiation may play a significant role in IASCC due to discontinuous slip where a dislocation channel is arrested at a grain boundary [24]. Nickel-based alloys In PWRs, the most extensive in-service failures due to stress corrosion cracking have occurred in the nickel-based material, Alloy 600, both on the primary side, in steam generator tubing, reactor head penetrations and other components, and on the secondary-side [25, 26]. Primary water cracking (PWSCC) occurs in a narrow potential range which is controlled by the dissolved hydrogen content in PWR primary coolant, with maximum IGSCC susceptibility close to the Ni/NiO equilibrium. Cracking susceptibility is also very sensitive to the microstructure, composition and strength level of material, with the presence of intergranular carbides being beneficial. SCC is also highly sensitive to the level of applied or residual stress and the temperature of the reactor coolant. Lower temperature components are thus less affected, as are those made from material given a thermal-treatment at around 700°C which improves the microstructure by precipitating carbides on grain boundaries. Nickel-based weld metals such as Alloys 182 and 82 are also susceptible to PWSCC [27], as are higher strength precipitation-hardened alloys such as Alloy-X-750 [28]. A variety of mechanisms can cause secondary side cracking including caustic environments or acidic conditions, both of which may occur in crevice locations under certain conditions [29]. The higher chromium content Alloy 690 is much more resistant to PWSCC and most of the secondary-side cracking mechanisms. Although laboratory tests have shown that it is possible to propagate an SCC crack in Alloy 690 in primary coolant in the presence of moderate (10-20%) levels of cold work, the alloy possesses a high resistance to PWSCC initiation and no in-service failures have been reported over 20 years of plant operation. Some SCC failures of Alloy 600 have also occurred in BWRs which are usually attributed to thermal sensitization. Further discussion of SCC of nickel-based alloys is outside the scope of the current paper; a useful overview is provided by Staehle [29].
10
Carbon and low alloy steels
Compared to austenitic stainless steels in BWRs and Alloy 600 in PWRs, carbon and low alloy steels have generally performed well in both types of LWR plant. An exception is a number of instances of cracking which occurred in ferritic steel components in German designed BWRs. Many of these occurrences have been attributed to the relatively high strength of the steel employed which was selected to reduce section thickness, together with the presence of active plastic straining during transients such as start-up and shutdown and the oxidizing nature of the BWR coolant. As a result of the requirement for active plastic straining, the cracking process is sometimes referred to as strain-induced corrosion cracking (SICC) [30]. Cracking of girth welds in steam generator shells has affected a small number of US PWR plants [31]. This was considered to be due to a combination of factors including inadequate post-weld heat treatment (resulting in over-hard heat affected zones), intermittent cyclic loading which introduces a corrosion fatigue component to cracking, inadequate feedwater deoxygenation and ingress of oxidizing copper species. Mechanistically, these failures appear similar to SICC in German BWRs. A number of experimental studies have addressed SCC in ferritic steels, prompted by work carried out by Speidel and Magdowski in the 1980s which claimed to show that BWR pressure vessels were at high risk of failure by SCC [32]. A large number of tests were carried out on pre-cracked, bolt loaded specimens of A533B or A508 pressure vessel steel in a static autoclave system with a gas overpressure to control the dissolved oxygen concentration. It was concluded that measured values of KISCC of
~20MPa√m and crack growth rates up to 10-7m/s at 288°C did not vary significantly with dissolved oxygen concentration, water chemistry or steel sulfur content within the ranges tested, Figure 13. These observations were reviewed by Scott and Tice [33] who concluded that the observations were inconsistent with plant experience and that other factors such as high levels of dynamic straining were necessary for cracking to occur, as pointed out by Hickling and Blind [30]. Subsequent laboratory studies have confirmed that low to medium strength carbon and low alloy steels are resistant to SCC propagation in the absence of dynamic straining in good quality BWR coolant chemistry [34], Figure 14a. SCC can however occur under constant load in BWR chemistry in the presence of relatively low levels of chloride contamination, or in pure water with ripple loading or with periodic partial unloading, Figure 14b [35]. Higher strength ferritic steels are significantly more susceptible to cracking under conditions which promote hydrogen embrittlement.
Figure 13: Effect of stress intensity factor on SCC growth rate in pressure vessel steel exposed to
oxygenated water at 288°C [11]
11
a) Very low crack growth rates observed under constant load in absence of anionic contaminants [34]
b) Increased crack growth rates due to chloride contamination, cyclic loading or periodic partial unloading (PPU) [35]
Figure 14: Crack growth of low alloy steels in simulated BWR normal water chemistry conditions
CORROSION FATIGUE
A number of plant failures in BWRs and PWRs have been attributed to fatigue or corrosion fatigue. Cyclic loading can occur from a variety of causes including pressure cycling during start-up and shutdown, flow-induced vibration and a variety of thermal transients. Because plant is designed against fatigue, failures are often due to unanticipated transients which may arise, for example, due to flow stratification at feedwater nozzles [36] and at locations where fluids at different temperatures meet such as mixing tees [37]. Because of the potential consequences of failures of large pressure vessels and pipework systems in nuclear plant, major research studies have been performed to understand the factors influencing environmental enhancement of fatigue, both in low alloy, pressure vessel steels and in austenitic stainless steels. The corrosion fatigue crack growth behaviour of low alloy steels was a major topic for research in the 1980s and 1990s and is now relatively well understood. There are, however, some outstanding concerns relating to stainless steels for which plant experience appears better than might be expected from predictions based on laboratory data and this is still an area for active study. A review for EPRI in 2011 identified a number of knowledge gaps regarding corrosion fatigue, many of which are the subject of ongoing research [38]. The effect of environment on fatigue crack growth for nickel based alloys is significantly lower than for stainless steels and SCC is therefore the dominant failure mechanism for the lower chromium alloys (600, 182 and 82), especially at higher temperatures. Two distinct approaches to evaluating fatigue (and corrosion fatigue) exist. Fatigue endurance testing is carried out to produce plots of numbers of cycles to failure versus applied stress range or strain range. These so-called S-N plots are used in plant design and are incorporated in Section III of the ASME Boiler and Pressure Vessel Code. Fatigue crack growth testing using pre-cracked specimens
produces relationships between applied stress intensity factor range (∆K) and cyclic crack growth rate (da/dN), e.g. Figure 3a. These form the basis of reference crack growth curves in ASME Section XI. The current paper will focus on sub-critical crack growth rather than fatigue endurance. Carbon and low alloy steels
Early studies into the influence of high temperature LWR environments on fatigue crack growth for ferritic steels revealed that a substantial enhancement of crack growth could occur under some circumstances for both low alloy pressure vessel steels and carbon steels used for piping in some BWRs [39]. Reference curves for sub-critical crack growth in carbon and low alloy steels in high temperature LWR environments were accordingly introduced into a revised Section XI of the ASME Code (Appendix A, Article A-4300) in 1980. Figure 15 shows the reference curves together with
12
some of the data (in PWR environments) on which they are based [40]. These reference curves were intended to apply to both PWR and BWR environments.
Figure 15: ASME Section XI Reference Curves for sub-critical fatigue crack growth in high
temperature reactor water environments and some of the data in PWR environments on which they are
based (adapted from reference 40)
Despite the introduction of these reference curves for water-wetted ferritic steel components, a major research effort continued to further understanding of the reasons for the degree of variability observed in laboratory tests examining the extent of environmental enhancement of fatigue crack growth. It was quickly established that the loading frequency substantially influenced the crack growth rate in LWR
environments, with plateau shaped curves being frequently observed on a da/dN – ∆K plot, Figure 16a [41]. Decreasing the loading frequency increased the plateau crack growth rate but also increased the
threshold ∆K for environmental enhancement. Significant enhancement of fatigue crack growth occurred only under specific material and environmental conditions: in particular, high steel sulfur content, high dissolved oxygen content and low water flow rates were detrimental, as shown in Figure 17 [42].
a) Cyclic crack growth rate as a
function of ∆K
b) Time-based growth rate as a function of rate in air
Figure 16: Corrosion fatigue crack growth of low alloy steels in PWR environment [41]
13
Based on statistical analysis of a substantial body of data available by the mid 1990s, Eason and co-workers prepared a proposal for a new set of reference curves which recognized the significant influence of loading frequency (or transient rise time, tR) on the degree of environmental enhancement of crack growth rate by plotting the rate of environmental crack growth with respect to time against the ASME air rate under the same loading conditions [43]. As shown in Figure 16b, this representation rationalizes the dataset showing a large effect of loading frequency. Additionally, a separate study by James [44] indicated that enhanced growth rates could only be sustained in PWR environments if the crack growth rate exceeded a critical value for a minimum crack growth interval; this was rationalized in terms of allowing sufficient access of water to outcropping sulfide inclusions in the steel to maintain a local crack tip environment [45]. Eason and Heys [46] drew on these various studies to develop an alternative assessment approach, applicable to PWR environments only, which formed the basis of ASME code case N-643 [47]. Numerous studies had shown that steel sulfur content was an important factor influencing environmental enhancement of fatigue crack growth in PWR environments, e.g. Figure 18 [48]. Different crack growth curves are therefore provided in the Code Case depending on the sulfur content of the steel: no enhancement for <0.004%S, intermediate enhancement for 0.004%≤S≤0.013%, and the largest degree of enhancement for S>0.013%. For medium and high sulfur content, the crack growth curves are a function of the rise time of the loading transient. The enhanced crack growth curves are only required to be used if both the incremental time
based crack growth rate (da/dt) and crack growth increment (∆a) exceed critical values.
Figure 17: Influence of controlling material and environmental variables on corrosion fatigue crack
growth of low alloy steels [42]
Figure 18: Summary of corrosion fatigue crack growth data for low alloy steels in PWR and BWR
environments showing that steel sulfur content has a much larger effect under PWR conditions [48]
14
No alternatives to the 1980 ASME Section XI reference crack growth curves for ferritic steels are available for BWR environments for which increased electrochemical potentials relative to PWR environments (especially in normal water chemistry) enhance the ability to maintain a sulfur-rich crack enclave environment. As a result, the benefit of a low steel sulfur content is substantially reduced as is evident from Figure 18; it would therefore not be expected that the James criterion for critical values of da/dt and crack increment based on measurements under PWR conditions would be applicable to the oxidizing conditions in BWR normal water chemistry. Data reported by Seifert and Ritter indicate that the 1980 ASME XI fatigue crack growth evaluation curves are non-conservative for certain critical combinations of material, environmental and loading conditions relative to BWR NWC but excessively conservative under some other conditions, Figure 19 [49]. They suggest that there is no fundamental reason why a similar procedure to ASME Code Case N-643 should not be applied to BWR HWC conditions, provided that appropriate criteria are experimentally developed and verified for the BWR ECP range.
Figure 19: Fatigue crack growth rates of low alloy steels in BWR type environment at different oxygen
concentrations [49]
Austenitic stainless steels In contrast to ferritic steels, Section XI of the ASME Code does not include fatigue crack growth curves for wetted flaws for austenitic steels. Shack and Kassner reviewed the data available up to 1994 and developed crack growth reference curves for oxygenated environments considered to be relevant to BWRs; curves for 200ppb and 8ppm dissolved oxygen were developed [50]. The data analysis was based on the time domain approach discussed above, with the time dependent crack rate, åe, being given by a superposition model:
SCCfatiguecorraire aaaa &&&& ++= .
For PWR conditions, most of the data available covered only relatively high frequency loading conditions, and consequently the observed effect of the environment appeared small; the use of the relationship for 200ppb oxygen BWR conditions was recommended as an interim measure. More recent studies have addressed lower loading frequencies and there is now a significant body of evidence that fatigue crack growth rates in stainless steels can be enhanced appreciably in PWR coolants. Figure 20a illustrates data for Type 304L stainless steel in a PWR environment which shows a significant effect of the rise time of the loading waveform [51]. As the rise time of the loading waveform was increased, the cyclic crack growth rate also increased. The maximum enhancement of crack growth rate in PWR water at 250ºC occurred at very long rise times (~500 minutes) and, at low
∆K; it is approximately 80 times the ASME Section XI reference line for air data. Similar observations
15
have been reported by other researchers [52, 53]. Figure 17b shows that the time domain approach rationalizes the time dependent nature of corrosion fatigue crack growth for this system.
a) Cyclic crack growth rate vs. ∆K b) Time domain plot
Figure 20: Crack growth data for Type 304L stainless steel in PWR environment at 250°C [51]
Mills has recently performed an analysis of fatigue crack growth data of type 304 and 316 stainless steel in high temperature deaerated water relevant to PWR coolant environments [54]. The majority of
the data included were from the UK, USA and Japan and covered a range of ∆K values, stress ratios (R>1), rise times and temperatures. The cyclic crack growth rate is described by the following equation:
[ ] 3.2
)(
'/50003.0
' KFeFtFCdN
daNAT
TR
RRLMAT ∆= −
where C is a constant, FMAT’L is a material dependent constant, FR is a function of R ratio and FT is an Arrhenius temperature term. Figure 21 compares the results of the model prediction with the UK , US and Japanese data available at the time. ASME Code Case N-809 has recently been developed based on the above model []. A number of experiments have recently been conducted to investigate the accuracy of Code Case N-809 predictions of environmentally assisted fatigue crack growth for plant-informed loading waveforms, e.g. Figure 22. The data obtained indicate that, for some waveforms, the simplistic application of Code Case N-809 based on total rise time can lead to significant over-estimation of the environmental fatigue crack growth rate. The contribution to the effective rise time of a particular section is dependent on its position within a waveform, with the loading ramp rate close to the maximum load of the waveform contributing more heavily to the effective rise time. It was argued that the damage accumulated during different portions of the waveform (and hence their contribution to the effective rise time) is in proportion to the localized plastic strain at the crack tip.
16
Figure 21: Comparison of Mills model predictions with crack growth data for austenitic stainless
steels in PWR environments [54]
Figure 22:Typical plant-informed waveforms used to assess relevance of ASME Code Case N-809 for
corrosion fatigue crack growth in austenitic stainless steel
More recent data in BWR NWC environments are generally consistent with those reviewed by Shack and Kassner but extend to longer rise times [57]. Data in BWR HWC conditions are up to a factor of 5 lower than for BWR NWC, Figure 23a, but are similar to PWR data, , Figure 23b.
17
a) Comparison of NWC and HWC b) Comparison of BWR HWC and PWR, including effects of impurities
Figure 23: Crack growth rates for austenitic stainless steel in BWR environments [57]
Although enhanced crack growth rates are generally similar for different heats of Types 304 and 316 stainless steel, under some conditions, crack growth rates in a PWR environment sometimes decrease towards the air rates as the loading frequency is reduced. Such ‘retarded’ growth data are ignored in Mills’ analysis and ASME Code Case N-809. Retardation is observed to be promoted by higher temperatures, high water flow rate, Figure 24a [51] and high sulfur content in the steel, Figure 24 b [58]. Studies continue to understand the mechanisms influencing enhancement and retardation for the stainless steel/high temperature water system.
a) Effect of water flow rate b) Effect of steel sulfur content
Figure 24: Retardation observed due to (a) high flow rate and (b) high steel sulfur content [51, 58]
SUMMARY
Either sensitization or cold work may result in sensitivity to stress corrosion cracking of austenitic stainless steels under BWR normal water chemistry conditions. Sensitization can be fairly easily prevented by the use of low carbon or stabilized materials whereas cold work can be introduced by a variety of fabrication processes such as surface finishing and welding-induced shrinkage. Various surface treatment processes can be used to reduce surface cold work and residual stresses. Modification of the environment by use of hydrogen water chemistry substantially reduces the risk of SCC and has been widely adopted; many plants also use noble metal additions to further reduce electrochemical potentials. Improved control of anionic contaminants in recent years has also been effective. Stainless steels in PWRs are less susceptible to SCC than under BWR NWC conditions and most observed failures have been in occluded or deadleg locations where full flow chemistry control is not always maintained. The presence of oxygen is commonly implicated in failures in these locations, frequently combined with chloride and/or sulfate contamination. If sufficient levels of impurities are
18
present, cracking is possible in annealed stainless steel, but many of the failures appear to be associated with cold work and a smaller number with sensitization. There have been a much smaller number of failures reported under nominally free-flowing conditions. Many of these have been in pressurizer heaters where higher temperatures, heat flux or cyclic loading may have an influence. Most, if not all, instances of SCC under free-flowing conditions are associated with high levels of surface cold work. Despite the good plant experience, laboratory tests indicate that crack propagation can occur in cold worked stainless steel in good quality PWR primary coolant but initiation is difficult and appears to require elevated levels of surface cold work and/or dynamic straining. In the presence of irradiation, IASCC has affected both BWRs and PWRs, although more components have been affected in the former. The critical fluence for the onset of cracking in BWR normal water chemistry is lower than in PWRs or BWRs operating HWC which may indicate that more than one cracking mechanism exists. Low strength ferritic steels are generally much less susceptible to SCC than austenitic materials. Cracking appears to require oxidizing conditions and high levels of dynamic or plastic strain and most occurrences have therefore been in BWRs. Elevated hardness following inadequate heat treatment enhances susceptibility and was implicated in cracking in the girth welds of a smaller number of PWR steam generator shells. Laboratory data reveal substantial influences of LWR environments on corrosion fatigue crack growth for both ferritic and austenitic steels. In both cases, decreasing the applied loading frequency initially enhances the effect, although retardation back to the air rate can occur under some circumstances for very low frequency loading. A high steel sulfur content is a major enhancing factor for ferritic pressure vessel and piping steels under PWR (and BWR hydrogen water chemistry) conditions which is attributed to creation of a local aggressive environment in the crack enclave. This effect is smaller in BWR normal water chemistry because the elevated external potential serves to maintain the local high sulfur environment. Austenitic stainless steels behave differently, with retarded growth rates in austenitic steels occurring more readily in steels with high sulfur contents under PWR conditions. In contrast, sulfur in stainless steels appears to be detrimental under oxidising conditions although studies in BWR normal water chemistry are limited. Application of laboratory data obtained using triangular or sawtooth loading waveforms to plant which is subject to more complex loading transients is not straightforward. Recent data show that the contribution to the effective rise time of a particular section of a transient is dependent on its position within a waveform, with the loading ramp rate close to the maximum load of the waveform contributing more heavily to the effective rise time.
REFERENCES
1. R W Staehle, Workshop on Detection, Avoidance, Mechanisms, Modeling and Prediction of Stress Corrosion Cracking Initiation, Beaune, France, 2008 (pub. EPRI 1018908, 2009)
2. Hettiarachchi, S., Cowan, R.L., Law, R.J., Miller, W.D. and Diaz, T.P. “Full cycle performance of a NobleChem treated BWR, Proc. 9th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 1999, Newport Beach, USA
3. Jones, R.L., Gordon, G.M. and Neils, G.H. “Environmental degradation of materials in boiling water reactors”, Proc. 4th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 1989, Jekyll Island, USA
4. Kilian, R. et al, “German experience with intergranular cracking in austenitic piping in BWRs”, Proc. 6th International Symposium. on Contribution of Materials Investigations to Improve the Safety and Performance of LWRs (Fontevraud 6), 2006, Fontevraud, France
5. Kumagai, K., Suzuki, S. and Asano, K. “Detailed examinations of PLR weld joints for the evaluation of reasonable SCC growth rate”, Proc. 5th International Symposium. on Contribution of Materials Investigation to the Resolution of Problems Encountered in Pressurized Water Reactors (Fontevraud 5), 2002, Fontevraud, France
19
6. Suzuki, S., Takamori, K., Kumagai, K., Sakashita, A., Yamashita, N.,Shitara, C., and Okamura;Y., “Stress Corrosion Cracking in Low Carbon Stainless Steels in BWRs”, E-Journal of Advanced Maintenance, Vol 1 (2009) 1 – 29.
7. Angeliu, T.M., Andresen, P.L., Hall, E., Sutliff, J.A. and Sitzman, S, “Strain and microstructure characterization of austenitic stainless steel weld HAZs, NACE Corrosion 2000, Paper 00186
8. Ooki S. et al. “Study on SCC growth behaviour of BWR core shroud”, Proc. 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2005, Snowbird, USA
9. Boursier, J.M., Gallet, S., Rouillon, Y. and Bordes, P. “Stress corrosion cracking of austenitic stainless steels in PWR primary water: an update of metallurgical investigations performed on French withdrawn components”, Proc. 5th International Symposium on Contribution of Materials Investigation to the Resolution of Problems Encountered in Pressurized Water Reactors (Fontevraud 5), 2002, Fontevraud, France
10. Ilevbare, G., Cattant, F. and Peat, N. “Assessment of cold work effects in stress corrosion cracking of stainless steels under PWR service conditions”, EPRI International Pressurized Water Reactor Materials Reliability Conference and Exhibition, June 2010, Colorado Springs, USA.
11. “Report of Kansai Electric Power Company on the cause of, and measures taken in response to, damage to a weld on the primary coolant inlet nozzle stub in the steam generator A of Unit 2 of Mihama Power Station, and NISA’s review results”. 4 Feb 2008, Ministry of Economy, Trade and Industry, Nuclear and Industrial Safety Agency, Japan
12. Thebault, Y. et al, “PWSCC of Thermocoax pressurizer heaters in austenitic stainless steel and remedial actions to preventing SCC, Proc. 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2011, Colorado Springs, USA
13. Hall, J. F., Falvo ,G. E. & Weiss, S. “Analysis and Mitigation of Stress Corrosion Cracking of High-Watt-Density Pressurizer Heater Sheaths”, EPRI International Pressurized Water Reactor Materials Reliability Conference and Exhibition, 2010, Colorado Springs, USA
14. Chynoweth, J., “Metallurgical Examination of a Leaking Pressurizer Heater Tube from Braidwood Unit 1”. Proc. 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2007, Whistler, Canada
15. “Recent field experience with elbows at Daya Bay Nuclear Plants”, http://wenku.baidu.com/view/5e649987e53a580216fcfe85.html
16. Health and Safety Executive, HM Nuclear Installations Inspectorate: ND Division 1 Project Assessment Report 61-2010: Forced outage (FO43) – Leakage of primary coolant from a Pressurizer Heater Well Insert
17. Shoji, T., Li, G., Kwon, J. Masushima, S. And Lu, Z., “Quantification of Yield Strength Effects on IGSCC of Austenitic Stainless Steels in High Temperature Water” Proc. 11th International Conference on Environmental Degradation of Materials in Nuclear Power Systems, 2003, Stevenson, WA, USA
18. Nouraei, S., Tice, D. R., Mottershead, K. J. and Wright, D. M., “Effects of Thermo-mechanical Treatments on Deformation Behavior and IGSCC Susceptibility of Stainless Steels in PWR Primary Water Chemistry”. Proc. 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2011, Colorado Springs, USA
19. Tice, D R, Stairmand, J W, Fairbrother, H J and Stock, A, “Crack Growth Testing of Cold Worked Stainless Steel in a Simulated PWR Primary Water Environment to Assess Susceptibility to Stress Corrosion Cracking”, Proc. 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems. 2007. Whistler, Canada
20. West, E. A., Tackes, C., Newsome, G. and Lewis, N., “Influence of sulfur and ferrite on SCC and corrosion fatigue behaviour of model heats of stainless steel”, Proc. 17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems. 2013. Ontario, Canada
21. Panteli, A., Platts, N. and Tice, D.R., Amec Fostter Wheeler unpublished work
20
22. Bruemmer, S.M., “New issues concerning radiation-induced material changes and irradiation-assisted stress corrosion cracking”, Proc. 10th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2001, Lake Tahoe, USA
23. Was, G.S., “Recent developments in understanding irradiation assisted stress corrosion cracking”, Proc. 11th International Conference on Environmental Degradation of Materials in Nuclear Power Systems, 2003, Stevenson, WA, USA
24. McMurtrey, G. and Was, G.S., “Relation between irradiation assisted stress corrosion cracking and discontinuos slip at grain boundaries”, Proc. 17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2015, Ottawa, Canada
25. Berge, P. “Can PWR U-tube steam generators work safely?”, Proc. 3rd International Symposium on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, 1987, Traverse City, USA
26. Bamford, W. and Hall, J, “A review of Alloy 600 cracking in operating nuclear plants including Alloy 82 and 182 weld behaviour”, Proc. International Conference on Nuclear Engineering, 25-29 April 2004, Arlington, USA, Vol. 1
27. Pathania, R.S., McIlree, A.R. and Hickling, J. “Overview of primary water cracking of alloys 182/82 in PWRs”, Proc. 5th International Symposium on Contribution of Materials Investigation to the Resolution of Problems Encountered in Pressurized Water Reactors (Fontevraud 5), 2002, Fontevraud, France
28. Hänninen, H. and Aho-Mantila, I., “Environment sensitive fracture of reactor internals”, Proc. 3rd International Symposium on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, 1987, Traverse City, USA
29. Staehle, R W and Gordon, G M, “Progress in understanding and mitigating corrosion on the secondary side in PWR stream generators”, Special Paper in Proc. 10th International Conference on Environmental Degradation of Materials in Nuclear Power systems – Water Reactors, 2001, Lake Tahoe, USA
30. Hickling, J. and Blind, D. Nuclear Engineering and Design 91, 1986, pp.305-350 31. Czajkowski, C.J., Int. J Press. Vessels & Piping 26, 1986, pp.97-110 32. Speidel, M.O. and Magdowski, R.M. “Stress corrosion cracking of nuclear reactor pressure
vessel and piping steels”, Int. J. Press. Vessels & Piping, 34, 1988, pp.119-142 33. Scott, P.M. and Tice, D.R. “Stress corrosion in low alloy steels”, Nuclear Engineering and
Design, 119, 1990, 399-413 34. Seifert, H-P and Ritter, S, New observations about the SCC crack growth behaviour of low-
alloy RPV steels under BWR/NWC conditions, Proc. 11th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2003, Stevenson, USA
35. Seifert, H-P and Ritter, S, Mitigation effect of hydrogen water chemistry on stress corrosion and low-frequency corrosion fatigue crack growth in low-alloy steels, Proc. 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2005, Snowbird, USA.
36. Enrietto, J.F., Bamford, W.H. and White, D.F., “Preliminary investigation of PWR feedwater line cracking”, Int. J. Pressure Vessels and Piping 9, 1981, pp. 421-443
37. Faidy, F. and Le Duff, J.A., “High cycle thermal fatigue in PWRs: Lessons learned from field experience and consequences”, Proc. 6th International Symposium on Contribution of Materials Investigations to Improve the Safety and Performance of LWRs (Fontevraud 6), 2006, Fontevraud, France
38. Tice, D.R. Green, D. and Toft, A., “Environmentally Assisted Fatigue Gap Analysis: A Roadmap for Future Research Activities”, EPRI, Palo Alto, CA: 2011. Product ID 071128
39. “BWR Environmental Cracking Margins for Carbon Steel Piping.” EPRI, Palo Alto, CA and General Electric Company, San Jose, California: May 1982. NP-2406.
40. Bamford, W.H. "Technical Basis for Revised Reference Crack Growth Rate Curves for Pressure Boundary Steels in LWR Environments", ASME J. of Pres. Vessel Tech. 102, 1980, pp.433-442
21
41. Data cited by Tice, D.R. Corrosion Science 25, 1985,pp.705-743 42. Tice, D.R., Atkinson, J.D. and Scott, P.M. Proc. IAEA Specialists’ Meeting on Sub-critical
Crack Growth, Sendai, Japan, May 1985. NUREG/CP-0067 Vol. 1, 1986, p251 43. Eason, E.D. Nelson, E.E. and Gilman, J.D. “Modeling of Fatigue Crack Growth Rate for
Ferritic Steels in Light Water Reactor Environments”, Nuclear Engineering and Design 184 (1998) pp.89 – 111
44. James, L.A. “Environmentally-assisted cracking of ferritic steels in aqueous environments: an interpretive review”, Weld. Res. Counc. Bull. 404, , 1995,pp1–20
45. Wire, G.L. “Cessation of environmentally-assisted cracking in a low-alloy steel: theoretical analysis”, Nuclear Engineering and Design 197, 2000, pp25–44
46. Eason, E.D. and Heys, G.B. “Fatigue crack growth rates of medium and low sulfur ferritic steels in PWR primary water environments”, ASME PVP Vol. 453, Pressure Vessel and Piping Codes and Standards -2003, PVP2003-1776
47. Cases of ASME Boiler & Pressure Vessel Code, Case N 643-2; “Fatigue Crack Growth Rate Curves for Ferritic Steels in PWR Water Environment”, Section XI, Div. 1, May 2004
48. Van der Sluys, W.A. and Emanuelson, R.H. “Enhancement of fatigue crack growth rates in pressure boundary materials sue to light water reactor environments”, Trans. 9th International Conference on Structural Mechanics in Reactor Technology, Lausanne, August 1987, Vol. F, pp.239-244 (Balkema, Rotterdam)
49. Seifert, H.-P and Ritter, S. “Corrosion fatigue crack growth behaviour of low-alloy reactor pressure vessel steels under boiling water reactor conditions”, Corrosion Science, 50, 2008, pp.1884-1899
50. Shack, W.J. and Kassner, T.F. "Review of Environmental Effects on Fatigue Crack Growth of Austenitic Stainless Steels", NUREG/CR-6176, ANL-94/1, May 1994
51. Tice, D., Platts, N., Rigby, K. and Stairmand, J. “Influence of temperature and flow rate on the corrosion fatigue crack growth of austenitic stainless steel in a PWR primary coolant environment”, Proc. 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2007, Whistler, Canada.
52. Wire, G.L., Evans, W.M. and Mills, W.J. “Fatigue Crack Propagation Tests on 304 Stainless Steel in High Temperature Water – Accelerated Cracking Rates and Transition to Lower Rates”, Pressure Vessel & Piping Codes and Standards ASME PVP 2004, PVP Vol. 480, p71
53. Nomura, Y., Tsutsumi, K. Kanasaki, H., Chigusa, N., Jokati, K., Shimizu, H., Hirose, T. and Ohata, H. “Fatigue Crack Growth Curve for Austenitic Stainless Steels in PWR Environment”, Pressure Vessel & Piping Codes and Standards, ASME PVP 2004, PVP Vol 480, p63
54. Mills, W.J. “Critical review of fatigue crack growth rates for stainless steel in deaerated water – Parts 1 and 2, EPRI International Pressurized Water Reactor Materials Reliability Conference and Exhibition, June 2010, Colorado Springs, USA
55. ASME Code Case N-809, “Reference Fatigue Crack Growth Rate Curves for Austenitic Stainless Steels in Pressurized Water Reactor Environments”, The American Society of Mechanical Engineers, New York, 23 June 2015
56. Platts, N., Tice, D.R. and Swan, D.I, “Effect of loading waveform and spectrum loading on the fatigue crack growth rate in simulated light water reactor environments” Proc. ASME 2016 Pressure Vessels and Piping Conference, PVP2016-63148, July 2016, Vancouver. Canada
57. Seifert, H.P., Ritter, S and Leber, H. “Corrosion fatigue crack growth behaviour of austenitic stainless steels under light water reactor conditions”, Corrosion Science 55, 2012, pp61–75
58. Platts, N., Tice, D.R,, Mottershead, K., McIntyre, L., and Scenini, F. “Effects of material composition on corrosion fatigue crack growth in high temperature water”, Proc. 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2011, Colorado Springs, USA