Comprehensive nuclear materials 5 05 corrosion and stress corrosion cracking of austenitic stainless steels

Comprehensive nuclear materials 5 05 corrosion and stress corrosion cracking of austenitic stainless steels Comprehensive nuclear materials 5 05 corrosion and stress corrosion cracking of austenitic stainless steels Comprehensive nuclear materials 5 05 corrosion and stress corrosion cracking of austenitic stainless steels Comprehensive nuclear materials 5 05 corrosion and stress corrosion cracking of austenitic stainless steels Comprehensive nuclear materials 5 05 corrosion and stress corrosion cracking of austenitic stainless steels

5.05 Corrosion and Stress Corrosion Cracking of Austenitic Stainless Steels

U Ehrnste´n

VTT Technical Research Centre of Finland, Espoo, Finland

ß 2012 Elsevier Ltd All rights reserved.

Abbreviations

BWR Boiling water reactor

CGR Crack growth rate

ECP Electrochemical corrosion potential

EPR Electrochemical (potentiokinetic)

reactivation

HAZ Heat-affected zone

HWC Hydrogen water chemistry

IGSCC Intergranular stress corrosion cracking

K Stress intensity factor

K ISCC Threshold stress intensity for SCC

LWR Light water reactor

MIC Microbiologically influenced corrosion

NG Nuclear grade

NMC Nobel metal chemistry

NMCA Noble metals chemistry addition

NWC Normal water chemistry

PLEDGE Plant life extension and diagnosis by

GE (General Electric)

PWR Pressurized water reactor

RBMK Channel type graphite moderated

reactor

SCC Stress corrosion cracking

TGSCC Transgranular stress corrosion cracking

VVER Water-water energetic reactor

5.05.1 Introduction to Austenitic Stainless Steels

Austenitic stainless steels have rendered their vast use because of their good performance in corrosive environments, in addition to their excellent ductility, formability, toughness, and weldability The good corrosion resistance of austenitic stainless steels is mainly due to chromium alloying, resulting in a pro-tective, chromium-rich passive film on the material

in many environments Molybdenum, used as an alloying element in Type 316 stainless steels, further increases the corrosion resistance Chromium and molybdenum are, however, both ferrite-forming ele-ments, and to maintain a fully austenitic structure,

a balance between austenite-stabilizing elements (C, N, Ni, Mn, and Co) and ferrite-stabilizing ele-ments (Cr, Mo, Si, Ti, Nb, Al, V, and W) in solution must be established.1To compensate for the molyb-denum addition in Type 316 stainless steels, the amount of nickel is increased In materials for nuclear environments, the cobalt content is kept as low as possible (<0.02%), because of its strong influence

on radioactivity buildup

Stress corrosion cracking (SCC) is taken into account

in the design codes for light water reactors (LWRs,

93

i.e., boiling and pressurized water reactors (BWRs

and PWRs)) through a statement that SCC should not

occur.2Intergranular stress corrosion cracking (IGSCC)

is still by far the largest damage mechanism for

austen-itic stainless steels in oxidizing BWR conditions, and

work on avoiding IGSCC is still going on.3,4 Several

factors affect IGSCC, one of them being sensitization

Sensitization is a result of nucleation and growth of

chromium-rich carbides on grain boundaries, causing

grain boundary chromium depletion

Chromium-depleted grain boundaries are prone to corrosion

and, in combination with a large enough stress and

suitable environment, to IGSCC All measures to

prevent sensitization are therefore taken in all steps

of component manufacturing and plant operation

Concerning the chemical composition, this can be

done by reducing the carbon content to levels below

0.03%, as is done in Types 316L, 316LN, and 316NG

stainless steels, where the carbon content is

typi-cally in the order of 0.02% Since carbon is a

strengthening element, nitrogen is added to these

steels to still achieve good mechanical properties

Nitrogen also reduces chromium-rich carbide

forma-tion, a concept that is utilized in the French RCCM

norms, which allow a carbon content of 0.035% in

their nitrogen-strengthened Type 316 stainless steel

with
0.08% N The other approach to avoiding

sen-sitization is to tie up the carbon into precipitates

This is utilized in stabilized stainless steels, which

are of two main categories, titanium- and

niobium-stabilized stainless steels, Type 321 and 347,

respec-tively To ensure that the carbon is tied into Ti(C,N)

or Nb(C,N) precipitates, a high enough stabilization

ratio, that is, Ti/C or Nb/C above 5 or 10,

respec-tively, is specified

Good corrosion resistance is ensured by

restrict-ing the amount of harmful elements, especially sulfur

and phosphorus, which may cause intergranular

corrosion when segregated to grain boundaries

Sev-eral standardized grain boundary corrosion tests,

such as the Strauss test5 and the electrochemical

potential reactivation (EPR) test,6are employed

rou-tinely as part of acceptance tests for materials and

components

5.05.1.1 Types, Mechanical Properties, and

Microstructures

The chemical compositions and mechanical

proper-ties of the most common stainless steels are presented

summarized inTable 2

In addition to compositional, mechanical, and cor-rosion resistance requirements, several other require-ments are put on austenitic stainless steel materials for nuclear components These include, for example, requirements on grain size A small enough grain size is needed to enable reliable nondestructive inspec-tion requirements, using ultrasonic techniques A com-mon requirement is that the grain size must not exceed ASTM number 4.0, which corresponds to an average grain size of 90mm The grain size of stabilized stainless steel components is typically smaller than this

Stainless steels are generally welded with a slightly over-alloyed filler metal to ensure good corrosion resistance of the final joint The weld shall contain

a small amount (>3% but <10%) of d-ferrite to avoid solidification and liquation cracking.7Welding induces residual stresses, which together with the operational stresses enhance crack initiation and growth The aim is, naturally, always to minimize the residual stresses by a proper choice of welding parameters, by securing a good fit between the parts

to be welded, etc The use of a narrow-gap welding technique has increased remarkably during the last few decades The narrow-gap welding method has many advances as it results, for example, in a lower level of residual stresses, a reduced weld volume, a narrower heat-affected zone (HAZ) with lower risk of sensitization, and less grain growth.8

Steels in BWRs and PWRs Austenitic stainless steel is the main construction material in nuclear power plants (NPPs) owing to its good corrosion resistance, ease to manufacture dif-ferent shapes, and good weldability In BWRs, stainless steels are used for piping and reactor pressure vessel cladding and structures inside the pressure vessel, including the core shroud (which separates the pri-mary water upward flow through the core from the downward flow in the annulus), the core plate (which supports the bottom of the fuel), the top guide (which aligns the top of the fuel bundles), the shroud dome, the steam separators, etc Austenitic stainless steel is also largely used for other components such

as pumps, valves, shafts, sleeves, and in auxiliary sys-tems such as water tanks

The material choices for PWRs are essentially the same The steam generator (SG) tubes are made of

Table 1 Chemical composition of common stainless steel alloys

alloys a

DIN 1.4301

1.4306

1.4311

DIN 1.4436/DIN 1.4401

DIN 1.4404/DIN 1.4435

1.4429

min Ti

SS 2337 DIN 1.4541

min Nb

SS 2338 DIN 1.4550 Source: Wegst, C E Stahlschluessel Key to Steel; Stahlschluessel Wegst GmbH: Marbach, 1995.

a The corresponding alloys according to the Swedish SS and the German DIN standards are also given.

Ti-stabilized stainless steel in the Russian designed

Water–water energetic reactors (VVERs), which are

PWRs with slightly different water chemistry and

horizontal instead of vertical SGs, as in Western

PWRs The SG tubes in western PWRs are made

of nickel-based materials (Alloy 600, 690, or 800)

Further, both the SG vessel and the pressurizer are

clad with austenitic stainless steel

5.05.2 Stress Corrosion Cracking

SCC is a failure mode caused by a combination of a

susceptible material, stresses, and an aggressive

envi-ronment (Figure 1) There are two modes of SCC in

austenitic stainless steel, namely intergranular and

transgranular stress corrosion cracking (IGSCC and

TGSCC) IGSCC in austenitic stainless steel is the

major failure mode in BWRs, while it has not been

considered as a plausible failure mode in PWR

pri-mary water under normal operation conditions

However, the number of IGSCC cases in PWRs has

increased by time, showing that PWRs are not totally

immune to IGSCC

Most TGSCC cases are due to chloride-induced

SCC TGSCC is rare in the primary system, and

the failure cases are typically observed in auxiliary

systems

In this section, the main factors affecting IGSCC in

BWR environment are reviewed, that is, degree of

sensitization, deformation, electrochemical corrosion

potential (ECP), water purity, and stress Also several

other parameters affect IGSCC susceptibility, but a comprehensive description of these is out of the scope

of this chapter Among these parameters are tempera-ture, hydrogen, mechanisms related to localization

of deformation and to corrosion deformation interac-tions, such as effect of strain rate, subtle differences in chemical composition, dynamic strain aging, dynamic recovery, vacancy injection, selective dissolution, grain boundary segregation, and relaxation.9–21

5.05.2.1.1 Degree of sensitization Sensitization is the result of nucleation and growth of chromium-rich carbides M23C6at the grain bound-aries, which results in a depletion of chromium at the grain boundaries because of faster diffusion rate

susceptibility

Stress

Region of potential stress corrosion cracking

Figure 1 The classic presentation of stress corrosion cracking includes the three circles: material, environment, and stress Reproduced from General Electric Company Alternative Alloys for BWR Pipe Application; NP-2671-LD, Final Report; San Jose, CA, 1982, with permission from BWR Owners Group.

Table 2 Minimum room temperature mechanical properties of stainless steels (hot finished and/or annealed forging)

(min MPa)

0.2% yield strength (min MPa)

Elongation (min.%)

Reduction

in area (min.%)

Source: Davis, J ASM Specialty Handbook; ASM International: Materials Park, OH, 1994; ASM Standard A 240.

along the grain boundaries compared to that within

the grain interior Chromium-rich carbides form

within the temperature range of 500–750C, but

continue to grow down to much lower temperatures

Sensitization can therefore occur as thermal

sensiti-zation during heat treatment and welding or as

low-temperature sensitization during long-time exposure

to LWR temperatures, below the chromium-rich

car-bide precipitation temperature.22 In the latter case,

the nucleation of carbides must have occurred

previ-ously, and the nucleated carbides grow during the

long-time exposure and deplete the grain boundaries

of chromium The degree of sensitization is typically

measured using the EPR test, which is sensitive to the

area where the grain boundary chromium content is

below 15%, and is, thus, not a true measure of the grain boundary chromium content

The time for carbide precipitation increases as the carbon level decreases as seen inFigure 2 Nitrogen alloying delays carbide precipitation (Figure 3), while deformation accelerates diffusion and precipi-tation The obvious remedy to avoid sensitization

is, thus, to decrease the amount of free carbon, as explained earlier, by reducing the carbon content, or

by tying carbon to Ti- or Nb-carbides and by restricting the degree of deformation

IGSCC in sensitized stainless steels occurs typi-cally in the weld HAZ at a distance of 4–8 mm from the fusion line, at the location where a high degree of sensitization combined with high residual stresses

Time (min)

304 (0.053% C)

1000

900

800

700

600

500

1800

1600

1400

1200

1000

Figure 2 Time–temperature–precipitation diagram for stainless steels with different carbon contents Reproduced from Shah, V N.; MacDonald, P E Aging and Life Extension of Major Light Water Reactor Components; Elsevier:

Amsterdam, 1993.

Time (h)

1000 900 800 700 600 500

1832 1652 1472 1292 1112 932

0.039% N

0.069% N 0.145% N 0.247% N

Figure 3 Effect of nitrogen on precipitation of M 23 C 6 in a 0.05C–17Cr–13Ni–5Mo stainless steel Reproduced from Peckner, D.; Bernstein, I M Handbook of Stainless Steels; McGraw-Hill: New York, 1977; pp 4-35–4-53, pp 751–757.

Corrosion and Stress Corrosion Cracking of Austenitic Stainless Steels 97

results in most severe conditions for IGSCC The

typical location of IGSCC is different in

nonsensi-tized stainless steels, where IGSCC occurs very close

to the fusion line, within the first few grains

5.05.2.1.2 Deformation

Deformation increases the susceptibility of stainless

steels to IGSCC, in sensitized as well as in

nonsensi-tized stainless steels, where the role of strain is

deci-sive Deformation occurs as bulk cold work from

rolling, bending, grinding, etc., as surface cold work

from machining, grinding, etc and as weld shrinkage

from welding Weld shrinkage can lead to up to 25%

equivalent room temperature strain in the weld HAZ

very close to the fusion boundary, and this is also the

location of observed IGSCC cracking in

nonsensi-tized stainless steels pipes.23–27 The importance of

cold deformation in IGSCC is shown in Figure 5

Deformation is estimated to be the main affecting

parameter in 50% of all IGSCC cases covered in

the survey (including sensitized and nonsensitized

stainless steels) The effect of deformation has been

studied using bulk-deformed materials,28–35and the

results show a correlation between IGSCC crack

growth rate (CGR) and yield strength (Figure 6)

Much effort is nowadays put on deformation in terms of restrictions on bulk and surface deformation and on the development of sophisticated surface treat-ment procedures to remove surface cold work at criti-cal locations.36,37Application of narrow-gap welding results both in a decrease in the degree of deformation

in the HAZ and in lower residual stresses It should be pointed out that some components, such as bolts, can

be made of intentionally cold-worked stainless steel to increase the material strength

5.05.2.1.3 Environment

As mentioned earlier, one of the main reasons for the good behavior of austenitic stainless steels in LWR conditions is the formation of a protective passive film in high-temperature water (around 300C) The oxide film formed in high-temperature water has a double-layered structure The inner layer grown on the metal surface consists of a chromium spinel or magnetite and is covered by an outer layer

of magnetite or Fe–Ni spinel precipitated from the aqueous phase.38,39The double-layered oxide struc-ture forms so that faster diffusing elements pass through the inner layer to the outer layer while the slower diffusing elements, such as chromium, remain

in the inner layer and therefore the outer layer con-tains mainly of iron and the inner layer is enriched with chromium Although consensus is not yet reached

on the mechanistic details for corrosion and SCC in LWR environments, breakage of the passive film is generally considered to be of major importance because of the fact that if the oxide film breaks, the corrosion rate is high until passivation occurs.28,40,41 The CGR of IGSCC is highly dependent on the oxidizing power of the environment, that is, the ECP

oxygen content increases in the high-temperature

0

5

10

15

20

Distance from weld fusion line (mm)

Various BWR stainless steel

weld HAZs

Figure 4 Deformation versus distance from the weld

fusion line in various stainless steel weld HAZs Deformation

is expressed in terms of equivalent tensile strain at room

temperature, and results from weld shrinkage strains

during welding Reproduced from Andresen, P L.; et al.

In Corrosion 2000, NACE 55th Annual Conference,

Orlando, FL, Mar 26–31, 2000; p 12, Paper No 00203,

with permission from BWR Owners Group.

Cold work Chemistry Material Nuclear grade Ni-base alloy Weld repair Residual stresses Sensitization

Figure 5 Cause of IGSCC in Swedish nuclear power plants Cold work is the biggest singular parameter affecting IGSCC Gott, K Personal communication, April 2010.

water, but it is not a linear relationship, and small

changes in oxygen concentration can result in large

changes of ECP and CGR Important to notice is

that the correlation between CGR and ECP is

differ-ent for sensitized stainless steels and deformed,

non-sensitized stainless steels, which show much higher

CGRs compared to sensitized materials at low

poten-tials (although lower than at high ECP)

Two solutions to lower the ECP in BWRs have

been developed, that is, hydrogen water chemistry

(HWC) developed by General Electric and Asea

Brown Boveri, and noble metal chemistry addition

(NMCA, NobleChem™), developed by General

Elec-tric The ECP in a BWR recirculation circuit during

normal operation and using normal water chemistry

(NWC), that is,200 ppb oxygen, is above 100 mVSHE

The ECP is still higher in the core because of

radio-lytic decomposition of water forming hydrogen

per-oxide, H2O2 The ECP is remarkably lower in plants

using either HWC, where 40–250 ppb hydrogen is

added to the feed water, or noble metal chemistry

(NMC™), where a small amount of platinum is

added to the reactor water either at about 130C

during startup or during full power operation

(OnLine NobleChem™), creating an electrocatalytic

surface layer.42,43 The ECP of the buffered PWR

environment is in the lower range of the ECP curve,

that is, about600 mVSHE The trend, especially for

the US BWR fleet is toward HWC and NMCA None

of the US BWR plants operates on NWC, and 75% apply NMCA.44The majority of the European BWRs operate on NWC

Water purity has a profound effect on both crack initiation and CGR in oxidizing environments The main concerns are chlorides and sulfates for SCC and additionally copper for pitting corrosion (although Cu also has a synergistic effect on SCC) Sulfate and/or chloride levels already in the ppb-range increase the IGSCC susceptibility Power plants monitor online the conductivity, which is a mirror for water purity, and analyze the amounts of impurities on regular basis from grab samples The conductivity of BWR primary water of today has been reduced from a typical range

of about 0.4mS cm1in the 1970s to 0.1–0.2mS cm1 (the conductivity of theoretically pure water is 0.056mS cm1)

Dissolved oxygen is consumed inside cracks and crevices, and the local ECP is reduced to low levels, creating a potential gradient between the outer sur-face and the crack tip This results in migration of anions into the lower potential area, which results in very high anion levels in the crack despite low levels

in the surrounding environment.45Further, the local environment in a crack can remain aggressive for a long time after, for example, short periods of higher impurity levels in the bulk environment

It is not only the environment during steady-state operation that needs attention, but also the

1E – 08 1E – 07 1E – 06

Yield strength (MPa)

Two sensitized points for comparison

Annealed cold worked

Unsensitized 304, 304L and 316L SS and A600

288 ⬚C high purity water, 2000 ppb O 2

CT tests at 27.5–30 MPa m½ Circles = High martensite SS Triangles = Alloy 600

Very high martensite

Very low or no martensite

Predicted response

Figure 6 Effect of yield strength (and martensite content) on stress corrosion crack growth rate of unsensitized stainless steels in oxygenated, high purity water The predicted response is based on the PLEDGE model (plant life extension and diagnosis by GE) Reproduced from Andresen, P L.; et al In Proceedings of 11th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, Stevenson, WA, Aug 11–14, 2003; American Nuclear Society: La Grange Park, IL, 2003; p 435, with permission from BWR Owners Group.

Corrosion and Stress Corrosion Cracking of Austenitic Stainless Steels 99

environment during shutdown, downtime periods,

and during startup The possible role of these will

increase with plant age and amount of shutdowns and

startups

5.05.2.1.4 Stress

The stresses causing IGSCC are a combination of

residual and operational stresses, although the first is

considered more decisive in IGSCC failures This is

because operational stresses are kept low by design

and components are usually designed to operate

below 80% of their yield strength The CGR of

intergranular stress corrosion cracks increases with

increasing stress intensity factor (K ) (Figure 8) The

effect of stress intensity on CGR varies depending on

the material and environment Knowledge of the

dependency betweenK and CGR is very important

for structural integrity calculations, which are made

to show that flaws, either postulated or detected using

nondestructive inspections, are tolerable and do not

pose a safety risk Huge efforts have been put on the

production of high-quality laboratory CGR data and

efforts are still going on Approved relationships (i.e., agreement reached between national safety authority and plant operators) are called disposition lines, and examples of published lines are shown inFigure 8 Several methods to mitigate IGSCC have been applied over the years, such as last pass heat sink welding, mechanical stress improvement, and weld overlay cladding.46All these aim at producing a com-pressive stress state in the HAZ However, these methods are usually applied as temporary remedies Measurement of residual stresses is an area of increased focus nowadays, and lack of knowledge can result in excessive under- or over-conservatism in design and in structural integrity calculations Also other stress-related factors affect IGSCC, such as vibratory loading, thermal loads from, for example, stratification, as well as load cycles during shutdowns and startups Much effort was earlier put on defining theKISCC; that is, the stress intensity, below which SCC would not occur With improved laboratory testing techniques, lower and lower KISCC values have been measured and a true threshold value may not exist

1E -09

1E-08

1E-07

1E-06

1E-05

–0.6 –0.5 –0.4 –0.3 –0.2 –0.1

Corrosion potential (VSHE)

–1 )

0.0 0.1 0.2 0.3 0.4

Sensitized 304 stainless steel

30 MPa m ½ , 288 ⬚C water

0.06–0.4 μS cm –1 , 0–25 ppb SO4

SKI Round Robin Data

Filled triangle = Constant load

Open squares = Gentle cyclic

42.5 28.3 14.2 μmh –1

GE PLEDGE

predictions

30 MPa m½

sens SS

0.5

0.1 0.25

0.06 μS cm –1

0.1μS cm –1

Means from analysis of

120 L sens SS data 0.06 μS cm –1

2000 ppb O2

200 ppb O2 Ann 304SS

316L (A14128, square)

304L (Grand Gulf, circle)

non-sensitized SS

50%RA 140C (black)

10%RA 140C (grey)

20% CW

A600

20% CW A600

4 dpa 304SS

GE PLEDGE predictions for unsens SS (upper curve for 20% CW)

(a)

10−1 (b)

10−6

10−7

10−8

10−9

Solution conductivity (μS cm –1 )

Predicted curves from PLEDGE code for typical range in ECP

316L stainless steel

25 mm CT specimen constant load

288 ⬚C water Test conditions:

0 ⬚C cm-2 EPR

ª27.5 MPa m½

200 ppb O2

Figure 7 Summary of crack growth rates of sensitized stainless steels versus corrosion potential, ECP (a) (reproduced from Andresen, P L.; Morra, M M J Nucl Mater 2008, 383(1–2), 97–111) and for nonsensitized stainless steels versus solution conductivity (b) (reproduced from Andresen, P L Corrosion 1988, 44(7), 450) The prediction curves for different water conductivity levels are according to the PLEDGE model, with permission from BWR Owners Group RA ¼ reduction

in area; CW ¼ cold work.

5.05.2.1.5 Components at risk

The earliest incidents of SCC in BWRs occurred in

stainless steel fuel cladding, before zirconium alloys

were used.47IGSCC plagued the BWRs in the 1970s

and caused a clear reduction in capacity factors

Crack-ing was first observed in the recirculation and water

cleanup systems in pipes with small diameter and later

also in larger diameter pipes The material was mainly

Type 304 with a high carbon content of0.6% Owing

to large efforts to solve the problem, including the

development of Type 316NG, narrow-gap welding

technique, as well as low-potential water chemistries,

the number of IGSCC incidents has remarkably

reduced.3In the late 1980s, cracking in Ti-stabilized

stainless steel piping was detected.48–50 Robust

miti-gation measures, including adoption of narrow-gap

welding, change of material to Nb-stabilized stainless

steel with higher stabilization ratio requirements, and

reduction of the amount of welds, were applied in

Germany to solve the problem Also from Russian

channel type graphite moderated reactors (RBMKs),

which operate under BWR-like conditions, numerous

IGSCC cases have been reported.51,52 In the 1990s,

the first cases with IGSCC in nonsensitized stainless

steels were reported in BWRs,23,25 first in pipings,

and later numerously in core shrouds.27Deformation

(weld shrinkage in piping and surface grinding in

the core shrouds) is considered to be of major impor-tance in these cases

PWRs operate at low corrosion potentials and very low oxygen levels, <30 ppb The risk of IGSCC

in austenitic stainless steels in nonoxidizing envi-ronment is, thus, much lower than in NWC BWR environment.31,53Incidences with IGSCC under nom-inal PWR conditions have not been reported Oxygen can, however, be enclosed in certain situations, such as startups, and can lead to IGSCC in austenitic stainless steels Although the number of IGSCC cases in PWRs

is still very low, the number seems to be increasing.54,55 IGSCC has been observed in pressurizer heater sleeves, canopy seals in the control rod drives, SG safe-ends, etc Laboratory tests on cold-worked stain-less steels show that IGSCC is possible also in normal PWR environment, indicating (although not gener-ally accepted) that more IGSCC failures may occur in the operating PWR plants with time

Environments Austenitic stainless steels are prone to TGSCC when exposed to aggressive oxidizing water, for example,

1E – 11 1E – 10 1E – 09 1E – 08 1E – 07 1E – 06 1E – 05 1E – 04

Stress intensity (MPa m ½ )

–1 )

NRC 0313, BWR NWC MD-01, BWR NWC JSME, BWR NWC JSME, BWR HWC

Figure 8 Crack growth rate versus stress intensity according to dispositions lines for sensitized stainless steels in normal water chemistry boiling water reactors environment Compiled by author from NRC Generic Letter GL880011 NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping; January 25, 1988; http://www.nrc.org; Jansson, C.; Morin, U.

In Proceedings of 8th International Symposium on Environmental Degradation of Materials in Nuclear power Systems – Water Reactors, Amelia Island, FL, Aug 10–14; American Nuclear Society: La Grange Park, IL, 1997;

pp 667–674; Kobayashi, H.; Kashima, K Int J Press Vess Pip 2000, 77, 937–944 JSME is the Japan Society of Mechanical Engineers; NRC is the Nuclear Regulatory Commission.

Corrosion and Stress Corrosion Cracking of Austenitic Stainless Steels 101

containing chlorides, under sufficiently high stresses.

As chloride levels in BWR and PWR primary systems

are kept low, TGSCC is rare under normal operating

conditions The exception is TGSCC from the

sec-ondary side in VVER SG tubes, manufactured of

Ti-stabilized stainless steels Condenser leakages,

fre-quently reported in VVERs, can result in chloride

contamination of the secondary water and eventually

in TGSCC in the SG tubes Copper release from brass

condensers and crevices formed by iron deposits

fur-ther enhance both TGSCC and pitting corrosion on

the secondary side of the SG tubes

A risk of TGSCC exists at locations where (a slow)

buildup of aggressive conditions can occur The risk

of TGSCC increases with plant age, as the buildup of

aggressive conditions can be very slow and can occur

at unknown (uninspected) locations Known chloride

sources are old insulation and sealing materials (e.g.,

asbestos), leakage from cables, polymers, paints,

con-crete, etc Wet insulation is the worst of these, as

the insulation provides crevice conditions in addition

to a chloride source Strict regulations for expendables

(grease, cleaning agents, sealing materials, etc.) allowed

in NPPs are applied to reduce the risk of buildup of

aggressive conditions All bare outer surfaces of

aus-tenitic stainless steel components, where humidity

may exist, can be at risk for TGSCC

TGSCC has been reported in valves,56for

exam-ple, where the source for the chlorides is assumed to

be asbestos sealing used early in time and in water

tanks, where chlorides probably stem from humidity

and concentration buildup at the waterline Stainless

steel bellows in the BWR reactor containment are, in

principle, at risk because of the high degree of cold

work in the bellows However, no SCC has been

reported in these New components can also be at

risk for TGSCC, if proper measures are not taken to

avoid contamination of components during

transpor-tation, storage, and installation

TGSCC can occur in oxidizing concentrated boric

acid solutions although laboratory results are not fully

conclusive whether chloride is also needed or not.56–58

5.05.3 Pitting Corrosion

Pitting corrosion occurrence has several similarities to

TGSCC, that is, it requires oxidizing conditions and

presence of water with harmful ions, such as chlorides,

fluorides, sulfates, and/or copper, but no stress is needed

The Type 304 stainless steel is more prone to pitting

corrosion than Type 316 stainless steel Pitting corrosion

is very often observed at same locations as TGSCC, but pitting corrosion can also occur without SCC and vice versa The risk of pitting corrosion under normal BWR conditions is extremely low However, pitting corrosion can occur in pressure boundary systems at locations where (slow) buildup of aggressive local conditions can occur Such locations are, for example, areas with low water flow, dead ends, and valves with sealing

As pitting occurs only in oxidizing conditions, it is not a plausible degradation mechanism in PWR pri-mary water under nominal environmental conditions However, the environment may be oxidizing both locally and/or temporarily because of startups, for example Different systems during shutdown may be filled with air, and this may cause air pockets during startup The oxygen from air will then dissolve into the primary water and local oxidizing conditions temporarily emerge until the oxygen is consumed by the oxidation of metal surfaces The risk of pitting corrosion (and TGSCC) is, however, highest in auxil-iary systems, for example, at outer surfaces, where the temperature is low enough for condensation to occur Thus, pitting corrosion can occur at nominally dry locations Accumulation of aggressive local conditions

is enhanced by crevices

The sources of chlorides were listed earlier Sulfate sources have been introduced earlier, for example, in molybdenum disulfide greases, but since the harmful influence of this material was iden-tified, it is not an allowed expendable material Again, copper can enter the system from copper-containing structural components

Pitting corrosion is seldom considered to pose a safety problem, as the wall thicknesses of pressure boundary components are usually large enough to sustain pitting corrosion for long times without leak-age However, pitting corrosion is always an indica-tion of a harmful environment existing at the locaindica-tion and is often associated with the risk of TGSCC, which can cause wall cracking in short time periods Pitting corrosion enhances the risk of SCC as the pits increase the local stress concentration and thus act as crack initiators Observation of pitting corrosion shall therefore not be omitted as insignificant

5.05.4 Microbiologically Induced Corrosion

A rather rare corrosion mode is microbiologically induced corrosion, or nowadays, microbiologi-cally influenced corrosion (MIC) MIC is normal