Comprehensive nuclear materials 5 04 corrosion and stress corrosion cracking of ni base alloys

5.04 Corrosion and Stress Corrosion Cracking
of Ni-Base Alloys
S. Fyfitch
AREVA NP Inc., Lynchburg, VA, USA

ß 2012 Elsevier Ltd. All rights reserved.

5.04.1

Introduction

70

5.04.2
5.04.2.1
5.04.2.2
5.04.2.3
5.04.3
5.04.3.1
5.04.3.2
5.04.3.3
5.04.3.4
5.04.4
5.04.4.1
5.04.4.1.1
5.04.4.1.2
5.04.4.1.3
5.04.4.1.4
5.04.4.2
5.04.4.3
5.04.4.4

5.04.4.4.1
5.04.4.4.2
5.04.4.4.3
5.04.4.4.4
5.04.4.4.5
5.04.4.5
5.04.4.5.1
5.04.4.5.2
5.04.4.5.3
5.04.4.5.4
5.04.4.5.5
5.04.4.6
5.04.4.7
5.04.5
References

Ni-Base Alloy Use in PWRS/BWRS
Wrought Ni–Cr–Fe Alloys
Age-Hardenable Ni-Base Alloys
Ni-Base Welding Alloys
General Corrosion
Water Chemistry
Flow Rates
Crevices
Mitigation
Stress Corrosion Cracking
Environmental Conditions
Temperature
Water chemistry
Sulfur intrusions

Electrochemical potential
Flow Rates
Crevices
Material Susceptibility Factors
Heat treatment
Microstructure
Grain size
Chemical composition
Product form
Stress
Operating stress
Residual stress
Surface effects
Weld geometry
Stress relief annealing
Irradiation
Mitigation
Outlook

70
70
72
73
73
75
75
75
75
75
77

77
77
80
81
81
82
82
82
83
84
84
85
85
86
86
87
87
87
88
90
90
90

Abbreviations
ASME
ASTM
B&PV
BWR
CEDM
CMTRs

American Society of Mechanical Engineers
American Society for Testing and Materials
Boiler and Pressure Vessel
Boiling water reactors
Control element drive mechanism
Certified material test reports

CRDM
ECP
EPRI
GMAW
GTAW
HAZ
IASCC

Control rod drive mechanism
Electrochemical potential
Electric Power Research Institute
Gas-metal-arc welding
Gas-tungsten-arc welding
Heat-affected zone
Irradiation-assisted stress corrosion
cracking

69


70

Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

IGSCC
INCO
LM
LWR
MSE
NRC
PWR
PWSCC
RCS
RUBs
SAW
SCC
SEM
SMAW
TT

Intergranular stress corrosion cracking
International Nickel Company
Light microscopy
Light water reactor
Mechanical surface enhancement
Nuclear Regulatory Commission
Pressurized water reactors
Primary water stress corrosion cracking
Reactor coolant system
Reverse U-bends
Submerged-arc welding
Stress corrosion cracking

Scanning electron microscope
Shielded-metal-arc welding
Thermal treatment

5.04.1 Introduction
Nickel–chromium–iron alloys (i.e., nickel-base alloys)
are widely used in the power industry in both fossil
(e.g., coal and gas) power stations and light water
reactor (LWR) nuclear power stations (i.e., pressurized
and boiling water reactors (PWRs and BWRs)). As a
result, the service behavior of these alloys has been
extensively studied,1 especially their susceptibility to
corrosion and stress-induced corrosion phenomena.
The power industry is concerned with the occurrence
of such failure phenomena because of their effect on
the safety and availability of equipment.
Corrosion and, in particular, stress corrosion failures
are not new. The power industry is well acquainted
with stress corrosion cracking (SCC) of stainless steel
in BWR piping and nickel-base alloys in PWR steam

Table 1

generators and its effect on equipment availability.2
SCC of these austenitic alloys has been known for
more than 50 years.

5.04.2 Ni-Base Alloy Use in PWRS/
BWRS
5.04.2.1

Wrought Ni–Cr–Fe Alloys

The wrought nickel-base alloys that are typically used
for nuclear applications are Alloy 600 and, more
recently, Alloy 690, which contain approximately
twice the chromium content. These materials are
used primarily for their inherent resistance to general
corrosion (i.e., oxidation resistance), strength at elevated temperatures, and a coefficient of thermal expansion very close to carbon and low-alloy steels. The
typical chemical composition and mechanical properties of these alloys are summarized in Tables 1 and 2,
respectively.
Both Alloy 600 and Alloy 690 are non-agehardenable, austenitic solid-solution strengthened
materials. No precipitation reaction is possible with
either alloy to increase strength; however, strength
can be increased by cold-working the material. They
are normally used in the annealed condition; however, a low-temperature heat treatment, or ‘thermal
treatment’ (TT), is also generally used with these
alloys, which tends to improve the resistance to
SCC in primary water chemistry conditions, which
is typically known as primary water SCC (PWSCC)
(see later sections of this chapter). This improvement
is clearly shown to be more pronounced, at least for
Alloy 600 material, in crack initiation testing.3

Chemical composition of wrought nickel-base alloys used in nuclear applications

Alloying element

Alloy 690

Alloy 600

Alloy X-750

Alloy 718

Alloy 800

Ni þ Co
C
Mn
Fe
S
Si
Mo
Cu
Cr
Ti
Al
P
Nb þ Ta
Others

58.0 min.
0.04 max.
0.5 max.
7.0–11.0
0.015 max.
0.50 max.
–

0.50 max.
28.0–31.0
–
–
–
–

72.0 min.
0.15 max.
1.00 max.
6.00–10.00
0.015 max.
0.50 max.
–
0.50 max.
14.0–17.0
–
–
–
–

70.0 min.
0.08 max.
1.00 max.
5.0–9.0
0.01 max.
0.50 max.
–
0.50 max.
14.0–17.0

2.25 – 2.75
0.40–1.0
–
0.70–1.20

50.0–55.0
0.08 max.
0.36 max.
Bal.
0.015 max.
0.35 max.
2.8–3.3
0.30 max.
17.0–21.0
0.65–1.15
0.20–0.80
0.015 max.
4.75–5.50
B 0.006 max.

32–35
0.03 max.
0.4–1.0
Bal.
–
0.30–0.70
–
<0.75
20.0–23.0
<0.60

0.15–0.45
–
–


Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

These alloys are widely used in LWRs. In BWRs,
applications include such locations as reactor vessel
nozzle safe ends, core support structures, and shroud
bolts. The PWR applications are typically within the
reactor coolant system (RCS) such as steam generator
tubing, penetrations and nozzles, control rod drive
mechanism (CRDM) and control element drive mechanism (CEDM) nozzles in reactor vessel heads, and
instrument nozzles in pressurizers and RCS piping, but
may also be found in selected non-Class 1 components
such as the Core Flood Tanks. In addition, Alloy 600
has also been used in a number of fastener applications.

Table 2

71

Figures 1–4 show typical applications of Alloy 600
material as used in the reactor coolant systems of the
four major BWR and PWR vendor designs.
Alloy 600 and 690 materials are available as plate,
barstock, tube/pipe, or forged material. The majority
of these materials were procured for the American
Society for Testing and Materials (ASTM)4 or American Society of Mechanical Engineers (ASME) Boiler

and Pressure Vessel (B&PV) Code5 specifications
(e.g., ASTM B 166 and B 167 or ASME SB-166 and
SB-167). Table 3 lists the various industry specifications that are used to procure these materials.

Typical room temperature mechanical properties of wrought nickel-base alloys used in nuclear applications

Mechanical property

Alloy 600

Alloy 690

Alloy X-750a

Alloy 718b

Alloy 800c

Yield strength, min. MPa (ksi)
Ultimate tensile strength, min. MPa (ksi)
Elongation, min. (%)

242 (35)
552 (80)
30

242 (35)
586 (85)
30

655 (95)
1103 (160)
20

1241 (180)
1034 (150)
10

334 (48)
572 (83)
30

a

Alloy X-750 HTH: Solution annealing at 1093  C (2000  F) and age-hardening at 704–718  C (1300–1324  F).
Alloy 718: Solution annealing at 1100–1400  C (1832–2000  F) and age-hardening at 720 and 620  C (1328 and 1148  F).
c
Alloy 800: Solution annealing at 1038–1066  C (1900–1950  F) and age-hardening at 760  C (1400  F) for 10 h, furnace cool to 649  C
(1200  F), hold for 20 h.
b

Feedwater recirculation
inlet/outlet welds
RPV attachments/
brackets

Shroud support
structure

Courtesy GE nuclear

CRD in-core housing
instrumentation penetrations

Figure 1 Typical applications of Alloy 600 materials in the reactor coolant systems of a General Electric Design
boiling water reactor. RPV, Reactor Pressure Vessel.


72

Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

Instrument and
vent penetrations
(both hot legs)

Decay
heat
line
weld
CRDM nozzles
Pressure
relief nozzle
safe ends/
welds (both
tanks)

PZR vent, spray,
and relief line
welds

PZR steam and
water instrument
penetrations

CRDM
motor
housings
Leak-off
monitor
lines
PZR heater
sleeves
and
diaphragm
plates

Core flood
tank
instrument
penetration
(both tanks)

PZR surge
nozzle weld

HPI/MU
nozzle
welds
(all cold legs)

Hot legsurge
nozzle
weld

Core flood
line welds
(both lines)
Core
guide
lugs
(ID)

Piping-RC pump
suction and
discharge welds
(all pumps)

RV bottom
head
instrument
penetrations

Primary drain
nozzles (both
SGs)

Instrument nozzles and
drain penetrations (all
cold legs)

SG nozzle dam
rings (both SGs)

Figure 2 Typical applications of Alloy 600 materials in the reactor coolant systems of a Babcock & Wilcox Design
pressurized water reactor. RV, Reactor Vessel; RC, Reactor Coolant; SG, Steam Generator; PZR, Pressurizer.

5.04.2.2

Age-Hardenable Ni-Base Alloys

Alloy X-750, a high-strength precipitation-hardening
alloy originally developed for gas turbines and the
aerospace industry, is widely used in internal applications for both BWR and PWR designs, such as fuel
assembly hold-down springs, control rod guide tube
support pins, jet pump beams, and reactor internals
structural bolting. This alloy is very similar in composition to Alloy 600, but contains additions of titanium and aluminum, which combine with nickel to
form the g0 precipitates, Ni3Al and Ni3 (Al, Ti), for
strengthening.6
Alloy 718 is another age-hardenable austenitic
nickel-base alloy, originally developed for the aerospace industry, that has seen much use in the nuclear
industry as a structural material due to its high strength
and corrosion resistance.7 A significant increase in
strength can be achieved by two precipitation reactions from solid solution involving g0 and g00 (Ni3Nb)

secondary phases within the austenitic matrix.8 The
addition of niobium sets this alloy apart from other
high-strength nickel-base alloys (e.g., Alloy X-750)
that are strengthened by g0 alone. Both the g0 and g00
precipitates are quite small and can only be resolved
with an SEM (scanning electron microscope) unless

gross over-aging has taken place. The microstructures
of the solution-annealed and age-hardened conditions
are indistinguishable with light microscopy (LM).
Alloy 718 has also been used extensively in PWR
primary coolant systems, predominantly for fuel assembly hardware.9 Alloy 718 is utilized for fuel assembly
hold-down springs, bolts, and spacer grids. It has been
shown to possess superior SCC initiation resistance
compared to Alloy X-750. Although Alloy 718 has
experienced some isolated failures in PWRs due to
fatigue/fretting cracking, it is considered highly resistant to intergranular SCC (IGSCC) initiation and
other forms of corrosion.


Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

73

PZR
instrument
nozzles
PZR and RC pipe-surge
line connections
CEDM motor
housing

Spray nozzlepipe weld
Safety and
relief valve
nozzle-pipe
welds and/or

flanges

CEDM/ICI nozzles to
RPV head welds
RPV top head
vent nozzle

PZR heater
sleeves

RPV head
leak monitor
tubes (2)
Surge
nozzle-pipe
welds
Charging inlet
nozzles (2 cold
legs)

Instrument nozzles
(all hot and cold legs)
Safety injection and
SDC inlet nozzle
(all hot and cold legs)
Let-down and drain
nozzles (all hot and
cold legs)

Shutdown

cooling
inlet
nozzle
(all cold
legs)

Spray
nozzles
(2 cold legs)

Primary nozzle closure
rings and welds
(both SGs)
Bottom channel head
drain tube and welds
(both SGs)

Guide lugs
flow skirt
ICI nozzles-ICI guide
tubes (system 80 plants)

RCP suction
and discharge
(all cold legs)

Shutdown cooling outlet
nozzle (1 hot leg)

Figure 3 Typical applications of Alloy 600 materials in the reactor coolant systems of a Combustion Engineering

Design pressurized water reactor. RPV, Reactor Pressure Vessel; RC, Reactor Coolant; RCP, Reactor Coolant Pump;
SDC, Shutdown Cooling; PZR, Pressurizer.

In addition, a modified Alloy 800 material, with
carbide-forming elements added to limit the solid
solution carbon content, has been successfully used for
many years in Germany for steam generator tubing.
The typical chemical compositions, mechanical
properties, and industry procurement specifications
are provided in Tables 1–3.
5.04.2.3

Ni-Base Welding Alloys

Welding of nickel–chromium–iron alloys is typically
performed using arc-welding processes such as gastungsten-arc welding (GTAW), shielded-metal-arc
welding (SMAW), and gas-metal-arc welding
(GMAW).7 Submerged-arc welding (SAW) may also
be used provided the welding flux is carefully selected.
Alloy 82, 182, and 132 are typical filler metals used to
join Alloy 600 components to carbon or low-alloy steel
vessels and other component items. These weld alloys
are also used as cladding in selected components
within the reactor coolant system. In addition, Alloy
52 and 152 (and newly developed variants such as
Alloy 52M and 152M) are filler metals that have

recently become the preferred materials used to join
Alloy 690 component items to carbon or low-alloy
steel vessels in the reactor coolant system.

Occasionally, there is a need for welded Alloy
X-750 items and the filler metals Alloy 82 or 69
(ERNiCrFe-8) were used during original fabrication.
The Alloy 69 material is no longer produced and
filler metal Alloy 718 is the currently recommended
material for welding.
The typical chemical compositions and industry
procurement specifications of these alloys are summarized in Tables 3 and 4.

5.04.3 General Corrosion
One of the main reasons that nickel-base alloys were
chosen for LWR applications is that they have the
ability to withstand a wide variety of severe operating
conditions involving corrosive environments, high
temperatures, high stresses, and combinations of these
factors. General corrosion can be defined as uniform
deterioration of a metal surface by chemical or


74

Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

Spray nozzlepipe weld

Safety and relief
nozzle-pipe
welds

Head

vent pipe
CRDM
motor
housings

CRDM
nozzles

RPV
head
leak
monitor
tube

Surge nozzlepipe welds

Thermowells (all
hot and cold legs)

RV nozzle
pipe weld
(all hot and
cold legs)

SG nozzlepipe weld
(all hot and
cold legs)

Core support
block

Bottom-mounted
instrument nozzles

Bottom channel
head drain tube
and welds (all SGs)

Primary nozzle
closure rings and
welds (all SGs)

Figure 4 Typical applications of Alloy 600 materials in the reactor coolant systems of a Westinghouse Design pressurized
water reactor. RPV, Reactor Pressure Vessel; RV, Reactor Vessel; SG, Steam Generator.

Table 3

Typical nickel-base alloy specifications used in nuclear applications

ASME B&PV Code

ASTM Standard

Material

Product form

SB-163
SB-166
SB-167
SB-168

SB-637
SB-670
SFA 5.11
SFA 5.14

B 163
B 166
B 167
B 168
B 637
B 670
–
–

Alloys 600/690/800
Alloys 600/690
Alloys 600/690
Alloys 600/690
Alloys 718/X-750
Alloy 718
Alloys 182/152
Alloys 82/52

Seamless tubing
Rod and bar
Seamless pipe and tube
Plate, sheet, and strip
Rod, bar, and forgings
Plate, sheet, and strip
Covered welding electrodes

Bare welding rods and electrodes

electrochemical reaction with the environment. Nickel
has good resistance to corrosion in the normal atmosphere, in freshwaters, and in deaerated nonoxidizing
acids, and it has excellent resistance to corrosion by
caustic alkalies. The high nickel content of these alloys
gives them resistance to corrosion by many organic and
inorganic compounds and also makes them virtually

immune to chloride-ion SCC. Chromium additions
provide resistance to sulfur compounds and also provide resistance to oxidizing conditions at high temperatures or in corrosive solutions. Details of the
corrosion resistance in these types of environments
can be found elsewhere (i.e., see also Chapter 2.08,
Nickel Alloys: Properties and Characteristics).2,10


Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

Table 4

75

Chemical composition of nickel-base welding alloys used in nuclear applications

Alloying
element

Alloy 52a
filler metal

Alloy 69
filler metal

Alloy 72
filler metal

Alloy 82
filler metal

Alloy 132
electrode

Alloy 152a
electrode

Alloy 182
electrode

Ni þ Co
C
Mn
Fe
S
Si
Mo
Cu
Cr
Ti
Al
P

Nb þ Ta
Al þ Ti
Zr
B
Others

Bal.
0.04 max.
1.0 max.
7.0–11.0
0.015 max.
0.50 max.
0.50 max.
0.30 max.
28.0–31.5
1.0 max.
1.10 max.
0.030 max.
0.10 max.
1.5 max.
–
–
0.50 max.

70.0 min.
0.08 max.
1.0 max.
5.0–9.0
0.015 max.
0.50 max.

–
0.50 max.
14.0–17.0
2.00–2.75
0.40–1.00
0.030 max.
0.70–1.20
–
–
–
–

55.0 max.
0.05
0.1 max.
0.2 max.
0.008 max.
0.1 max.
–
0.20 max.
44.0 max.
0.6 max.
–
–
–
–
–
–
–

67.0 min.
0.10 max.
2.5–3.5
3.0 max.
0.015 max.
0.50 max.
–
0.50 max.
18.0–22.0
0.75 max.
–
0.030 max.
2.0–3.0
–
–
–
0.50 max.

62.0 min
0.08 max
3.5 max
11.0 max
0.02 max
0.75 max

Bal.
0.05 max.
5.0 max.
7.0–12.0
0.015 max.

0.75 max.
0.50 max.
0.50 max.
28.0–31.5
0.50 max.
0.50 max.
0.030 max.
1.0–2.5
–
–
–
0.50 max.

59.0 min.
0.10 max.
5.0–9.5
10.0 max.
0.015 max.
1.0 max.
–
0.50 max.
13.0–17.0
1.0 max.
–
0.030 max.
1.0–2.5
–
–
–
0.50 max.

0.50 max
13.0–17.0
–
–
0.030 max.
1.5–4.0
–
–
–
0.50 max

a

Alloys 52M and 152M have controlled additions of boron and zirconium.

5.04.3.1

Water Chemistry

The water chemistry of LWRs is discussed in detail
in Chapter 5.02, Water Chemistry Control in LWRs.
Nickel-base alloys are essentially immune to general
corrosion in LWR environments due to the formation
of an adherent Cr-rich oxide on the surface.
5.04.3.2

Flow Rates

The inherent passivity of nickel-base alloys provides

them with excellent resistance to flow-assisted corrosion. They are able to withstand very high flow rates,
on the order of 18.3 m sÀ1 (60 ft sÀ1), without concern.
Corrosion rates in such flowing conditions for nickelbase materials are expected to be <2.5 mm yearÀ1
(<0.1 mil yearÀ1).11
5.04.3.3

Crevices

General corrosion of nickel-base alloys in crevices
is not anticipated to be of great concern in LWRs
because of the passive nature of these materials. Pitting
may occur occasionally in the presence of impurities,
which could lead to SCC (see Section 5.04.4.3),
particularly in the more oxidizing conditions of BWRs.
No failures in PWRs have been directly attributed to
creviced locations.
5.04.3.4

Mitigation

As general corrosion is minimal in LWR environments, mitigation is not really necessary. As noted

in this section on LWR Structural Materials, PWR
environments have reducing conditions and general
corrosion is not of concern. However, mitigation of
corrosion concerns in the more oxidizing environment of BWRs has been through the use of hydrogen
water chemistry (i.e., to make the environment more
reducing, similar to a PWR) and noble metal chemical
additions.12,13

5.04.4 Stress Corrosion Cracking
SCC of nickel-base alloys is an important age-related
phenomenon affecting LWRs. This type of failure
mechanism for nickel-base alloys typically occurs
intergranularly and is generally termed intergranular
stress corrosion cracking (IGSCC). In PWRs, IGSCC
is typically termed primary water stress corrosion
cracking (PWSCC). The occurrence of SCC of nickelbase alloys has been extensively studied since the first
reported observation of cracking in laboratory tests
using Alloy 600 in high-purity water by Coriou et al.14
in 1959. Over the last three decades, IGSCC has
been observed numerous times in LWRs and it has
affected both the safe and economic operation of
the reactors. In BWRs, cracking of nickel-base
components such as safe ends, shroud bolts, and access
hole covers has occurred; however, the predominant
failures have been identified in Alloy 182 welds. In
PWRs, PWSCC of Alloy 600 component items has
been observed in steam generators, pressurizers, and


76

Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

CRDM nozzles, and most recently in Alloy 182 and
Alloy 82 welds. The mechanism of this cracking
phenomenon is not completely understood, and prediction of crack initiation time has proven to be
extremely difficult, if not impossible, due to the uncertainty of numerous variables (e.g., heat treatment and
residual stress). In this section, emphasis will be given to

the SCC of Alloy 600 materials in PWRs, given the fact
that it has been the most prevalent; however, as noted
above, BWR conditions are not immune to IGSCC of
nickel-base alloys.
It is known, however, that SCC of nickel-base materials occurs as a result of the following three factors:
 susceptibility of the material
 a tensile stress (including both operating and residual stress)
 a corrosive environment
The synergistic effect of these three factors is
typically shown on a Venn-type of diagram (Figure 5).
As an example, the susceptibility of Alloy 600
material to PWSCC depends on several factors,
including the chemical composition, heat treatment
during manufacture of the material, heat treatment during fabrication of the component, and
operating parameters of the component. Chemical
composition and heat treatment are interrelated in
several ways. For example, one reason for annealing
Alloy 600 is to solutionize the carbon in the alloy.
As the material cools, chromium carbides precipitate from the solution at both intragranular and
intergranular locations. If the cooldown from the

Mechanical
Operational tensile stresses
residual tensile stress

SCC
Corrosive
Susceptible material
Chemical composition
microstructure

Electrochemical
corrosion potential
temperature
pH-value

Figure 5 Synergistic factors affecting stress corrosion
cracking of nickel-base materials.

anneal is sufficiently slow, a greater number of carbides
will precipitate at the grain boundaries (i.e., intergranularly) and the resistance to PWSCC will be
improved. Well-decorated grain boundaries are an
indication that an Alloy 600 material has received
proper heat treatment and that sufficient carbon
was available in the solution to combine with chromium. If adequate amounts of carbon and chromium
exist, but the anneal is not at a high enough temperature or sufficient time is not allowed to solutionize the
carbon, an adequate amount of carbon will not be
available to precipitate intergranularly as chromium
carbides, leading to minimal grain boundary decoration. Most precipitation occurs during cooldown
following annealing; however, stress relief treatments
can lead to additional precipitation. The primary goal
of stress relief, however, is to allow a local realignment
of highly strained regions to reduce internal stresses.
Carbon and chromium concentration gradients are
also reduced given the extended time at the temperature. Thus, if the anneal has not adequately solutionized carbon for chromium carbide precipitation at
the grain boundaries, stress relief treatment will not
reduce susceptibility to PWSCC.
Tensile stresses, resulting from both residual and
operating stresses, can be significant for some Alloy
600 component items. Operating stresses are produced from mechanical and thermal loading, while

residual stresses are generated as a result of fabrication, installation, and welding processes. Residual
stresses are more difficult to quantify than operating
stresses and, in many instances, are of a higher magnitude than operating stresses.
PWSCC is a thermally activated degradation
mechanism, that is, as the temperature increases, the
rate of PWSCC increases exponentially. Thus, the
hot leg temperature of the RCS creates a more
aggressive environment in which the Alloy 600 components must operate.
The cracking observed in PWRs to date is typically axially oriented (although circumferentially oriented cracks have been observed) and occurs in
an area, such as a weld heat-affected zone (HAZ),
that has high residual tensile stresses. In a cylindrically
shaped component (e.g., piping, vessel, and nozzles),
the circumferential stresses are inherently higher than
axial stresses. Thus, in a homogeneous material with no
initial flaws, cracking would be expected to occur
axially because of the higher circumferential stresses.
PWSCC has been the subject of much research
and analyses in recent years as a result of the many
failures that have been attributed to it. However,


Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

(50–70  F) temperature differences between hot and
cold legs are enough to significantly influence the
time to initiation and subsequent crack growth rate.
Temperature is generally believed to affect the
rate of SCC attack in accordance with an activation
model for thermally controlled processes (Arrhenius
equation), exp(ÀQ/RT), where Q is the activation

energy, R is the ideal gas constant, and T is the
absolute temperature. The current consensus is that
the activation energy for crack initiation falls in the
range of 188–230 kJ molÀ1 (45–55 kcal molÀ1) and
many predictions are based on 210 kJ molÀ1 (50 kcal
molÀ1). There is also evidence that the activation
energy varies with material carbon content.15

a reliable crack initiation model has yet to be developed. PWSCC of Alloy 600 components in the RCS
can lead to through-wall cracking and thus leakage of
primary water. (Catastrophic failure is not expected as
circumferentially-oriented cracks do not occur
unless very high axial stresses are generated in the
component, e.g., from roll expansion methods.)
5.04.4.1

Environmental Conditions

The major environmental conditions affecting SCC
of nickel-base material in LWR environments appear
to be temperature, water chemistry (oxygen, hydrogen, lithium, boron, and sulfur content), and electrochemical potential (ECP). Each of these factors is
evaluated as follows.

5.04.4.1.2 Water chemistry

The water chemistry of LWRs can generally be
described as essentially pure water. PWRs primarily
include hydrogen, boron, and lithium to produce reducing conditions. BWRs primarily operate with low levels
of oxygen, but in recent times, hydrogen additions have
been introduced to limit the oxidizing potential of

the environment. Additional details are included in
Chapter 5.02, Water Chemistry Control in LWRs.

5.04.4.1.1 Temperature

By far, temperature is the single most significant
environmental factor influencing the initiation of
SCC in LWR environments. This is evidenced by
the fact that the vast majority of SCC of PWR steam
generator roll expansion transitions have occurred
on the hot leg side of the tube sheet. The 28–39  C

MA + drawn
35 % area reduction

4000

SCC initiation time (h)

77

3000

2000

1000

–
–
–

–
–
Hydrogen

200 ppm B
0.7 ppm Li
hydrogen

500 ppm B
1.0 ppm Li
hydrogen

1100 ppm B
2.0 ppm Li
hydrogen

Water chemistry
Figure 6 Stress corrosion cracking initiation times for Alloy 600 steam generator tubing at 360  C (680  F). Stress
corrosion cracking initiation time as a function of primary water chemistry for as-drawn (35% area reduction) mill-annealed
tubing. Reproduced from Airey, G. P. The stress corrosion cracking performance of Inconel Alloy 600 in pure and primary
water environments. In Proceedings: 1983 EPRI Workshop on Primary-Side Stress Corrosion Cracking of PWR Steam
Generator Tubing; EPRI NP-5498, Project S303–5, with permission from Electric Power Research Institute.


78

Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

5.04.4.1.2.1 Hydrogen

The effect of dissolved hydrogen on SCC susceptibility of nickel-base alloys (e.g., Alloys 600 and
X-750) has been evaluated by numerous researchers.
Pathania and McIlree16 reviewed the influence of
hydrogen on PWSCC in 1987. At that time, the
emphasis of the work was on initiation at temperatures of 360  C (680  F) and above. The authors concluded that the susceptibility of Alloy 600 increased
when the amount of dissolved hydrogen increased.
Airey17 has shown that dissolved hydrogen increases
the rate of PWSCC of steam generator tubing in
autoclave tests at 360  C (680  F). An example of his

Temperature (ЊC)
10-5

290

325

345

365

Crack growth rate (mm s-1)

4 pts

10-6
4 pts
Ave. of
4 pts

data, shown in Figure 6, shows that the SCC initiation time for pure water is decreased dramatically
when hydrogen is added. However, in the primary
water of PWRs, the effect of hydrogen appears to be a
function of the boron and lithium content. Bandy and
Van Rooyen18 have shown a similar effect with boron
alone versus pure water and primary water (Figure 7).
They showed that 83% of the specimens cracked in
pure water with hydrogen versus only 2% in pure
water without hydrogen.
More recently, Norring has reported on tests to
determine the effect of hydrogen overpressure in
330  C (626  F) water.19 Results of these tests, shown
in Figure 8(a), suggest that the rate of PWSCC
increases with increasing hydrogen overpressure.
The most recent update on the influence of
hydrogen on PWSCC was prepared by Cassagne
et al.20 in 1997. All the data seem to indicate that the
susceptibility of Alloy 600 decreases drastically for
low hydrogen values (<10 kPa (1.45 psi)) regardless of
temperature. For hydrogen partial pressure above
100 kPa (14.5 psi), a more progressive decrease in
susceptibility seems to occur between 360 and
400  C (680 and 752  F). Data are not available in
this range for lower temperatures. Between 10 and
100 kPa (1.45 and 14.5 psi), it appears that PWSCC
initiation is not greatly affected for all temperatures
between 400 and 310  C (752 and 590  F). At 290  C
(554  F), the influence of hydrogen cannot be
assessed because of a lack of data.

5.04.4.1.2.2

10-7
Flattened specs, I.E.
Cold worked, pure H2O
As received, pure H2O + H2
As received, pure H2O + H3BO3
As received, pure H2O
As received, (0.03 % C),
pure H2O
Primary H2O
10-8

1.75

1.70

[

1.65
1
T(K)

1.60

1.55

] ´ 1000

Figure 7 Effect of hydrogen on Alloy 600 cracking in pure

and primary water environments. Reproduced from Bandy,
R.; Van Rooyan, D. Quantitative examination of stress
corrosion cracking of Alloy 600 in high temperature
water – Work in 1983. In Proceedings: 1983 EPRI Workshop
on Primary-Side Stress Corrosion Cracking of PWR Steam
Generator Tubing; EPRI NP-5498, Project S303–5, with
permission from Electric Power Research Institute.

Boron and lithium

Evaluations of boron and lithium on PWSCC of Alloy
600 have been performed by numerous investigators.
Norring et al.19 concluded that increasing the lithium
content from 2.4 to 3.5 ppm significantly decreased
the time to crack initiation (see Figure 8(b)).
The most complete evaluation was performed by
Ogawa et al.21. In these tests, hydrogen overpressure was
kept at a constant level of 30 cm3 kgÀ1 H2O. The results
of this work are shown in Figures 9 and 10. It appears
that maintaining a constant pH of 7.1–7.3 (at 285  C
(545  F)) will produce a range of crack initiation times
(Figure 9) and that increasing the boron content
greater than 1200 ppm (at any lithium level) decreases
the crack initiation times (Figure 10). Therefore,
the beginning of cycle boron concentrations appears
to be the worst condition for PWSCC initiation. Maintaining a pH level of 7.3 (at 285  C (545  F)) also
appears to be better than a pH level of 7.1.
Follow-up tests were performed at high boron
concentrations with varying lithium concentrations

Temperature...........

329.9 ЊC

329.6 ЊC

Boron......................

1471 ppm

1441 ppm

Lithium....................

2.42 ppm

2.38 ppm

Hydrogen content...

13.1 ml kg-1

25.2 ml kg-1

7.0 kPa

13.6 kPa

7.354

7.351

Activity...

99
90

50
25
10
5
1
100 200

10
5

(a)

99
90

B&W 1700 ЊF 7/8Љ (6A)

25
10
5

500 1000 2000 5000 10 000

Exposure time (h)

The influence of hydrogen on the tendency to PWSCC

Temperature...........

328.9 ЊC

329.6 ЊC

Boron......................

1241 ppm

1441 ppm

Lithium....................

3.54 ppm

2.38 ppm

Hydrogen content...

24.7 ml kg-1

25.2 ml kg-1

13.6 kPa

13.6 kPa

7.575

7.351

Activity...
pH...........................

B&W 1700 ЊF 7/8Љ (6)

90
50
25
10
5
1
100 200

500 1000 2000 5000 10 000
Exposure time (h)

B&W 1700 ЊF 3/4Љ

90
50
25
10
5
1

100 200

99
Cracked specimens (%)

Cracked specimens (%)

99

Cracked specimens (%)

99

(b)

500 1000 2000 5000 10 000
Exposure time (h)

50

1
100 200

500 1000 2000 5000 10 000
Exposure time (h)

B&W 1700 ЊF 3/4Љ

25

B&W 1700 ЊF 7/8Љ (6)

99
90

79

50

1
100 200

Cracked specimens (%)

Cracked specimens (%)

pH...........................

Cracked specimens (%)

Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

500 1000 2000 5000 10 000
Exposure time (h)
B&W 1700 ЊF 7/8Љ (6A)

90
50
25
10

5
1
100 200

500 1000 2000 5000 10 000
Exposure time (h)

The influence of lithium on the tendency to PWSCC

Figure 8 Effect of hydrogen and lithium on time to stress corrosion cracking of Alloy 600 steam generator tubing.
Reproduced from Norring, K.; Rosborg, B., Engstrom, J., Svenson, J. Influence of LiOH and H2 on primary side IGSCC of
Alloy 600 steam generator tubes, Colloque International Fontevraud II, Sept 10–14, 1990; Socie´te´ Franc¸aise d’Energie
Nucle´aire, Paris, France, 1990; pp 243–249, with permission from Socie´te´ Franc¸aise d’Energie Nucle´aire.


80

Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

5

Li (ppm)

4
1.7 1.6
1.5

3

1.4

1.3

1.2
1.1

2

1.0

1

0
6.3

6.4

6.5

6.6

6.7

6.8

6.9 7.0
pH 285 ЊC

7.1

7.2

7.3

7.4

7.5

7.6

Figure 9 Isosusceptibility diagram for primary water stress corrosion cracking of Alloy 600 as a function of lithium and pH at
285  C (545  F). The numbers (1.0, 1.1, 1.2, . . .) represent the ratio of the percent intergranular fracture referenced to the
response at B:280/Li:2.0 ppm. Reproduced from Ogawa, N.; et al. Nucl. Eng. Des. 1996, 165, 171–180.

5

Li (ppm)

4

3

1.7

1.6

1.5

1.4

1.3

2

1.2

1.1

1.0

1

0
2000

1500

1000
B (ppm)

500

0

Figure 10 Isosusceptibility diagram for primary water stress corrosion cracking of Alloy 600 as a function of lithium and
boron. The numbers (1.0, 1.1, 1.2, . . .) represent the ratio of the percent intergranular fracture referenced to the response at
B:280/Li:2.0 ppm. Reproduced from Ogawa, N.; et al. Nucl. Eng. Des. 1996, 165, 171–180.

to simulate beginning of fuel cycle conditions. Ogawa
et al.22 report that there is little effect of lithium content from 2 to 10 ppm at boron concentrations greater

than 1200 ppm (see Figure 11) and PWSCC susceptibility at 1600 ppm boron (2–10 ppm lithium) was
higher than that at 500 or 280 ppm boron concentrations (with 2 ppm lithium).
5.04.4.1.3 Sulfur intrusions

Sulfur intrusions by themselves will not produce
SCC in nickel-base material; however, sulfate will
promote intergranular attack and intergranular SCC.
A sensitized material microstructure is much more

susceptible (in terms of initiation time) to this type of
attack, although all nickel-base materials will be
attacked by sulfate.
Andresen23 has shown that the time to failure
decreased by 2–3 orders of magnitude in constant
load tests between pure water and sulfate impurities
(conductivities ranging from 5 to 55 mS cmÀ1 due to
sulfuric acid additions).
Bandy et al.24 have shown that sulfates are very
potent cracking agents for Alloy 600 materials. Temperature significantly accelerates the cracking and
most likely decreases the threshold stress for cracking
to occur.


Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

81

11
(1.53)
10

1.5

9

8

(1.46)

Li (ppm)

7

6

5
1.4
4
1.3
1.2
1.1

3

1.0
1.1

2

1.2

1

0
2000

1500

1000
B (ppm)

500

0

Figure 11 Revised isosusceptibility diagram for primary water stress corrosion cracking of Alloy 600 as a function of lithium
and boron. The numbers (1.0, 1.1, 1.2, . . .) represent the ratio of the % intergranular fracture referenced to the response at
B:280/Li:2.0 ppm. Values in parentheses are extrapolated or interpolated from test result values. Reproduced from
Ogawa, N.; Nakashiba, T.; Yamada, M..; Umehara, R.; Okamoto, S.; Tsuruta, T. In Proceedings of the Eighth International
Symposium on Environmental Degradation of Materials in Nuclear Power Systems; American Nuclear Society: La Grange
Park, IL, 1997; pp 395–401. Copyright 1997 by the American Nuclear Society, La Grange, IL, USA.

5.04.4.1.4 Electrochemical potential

100
PH2 = 0.005 MPa

IGSCC (%)

80

PH2 = 0.1 MPa

60
40
20

Alloy 600
350 ЊC
0.01 m H3BO3 + 0.001 m LiOH soln

0
-1000

-300
0
(E )
Potential (mV) Corr

300

Figure 12 Effect of electrochemical potential on rate
of primary water stress corrosion cracking in
Alloy 600 material. Reproduced from Smialowka,
S. Hydrogen induced IGSCC of Alloy 600 in high
temperature aqueous environments. In Proceedings:
1987 EPRI Workshop on Mechanisms of Primary Water
Intergranular Stress Corrosion Cracking; EPRI
NP-5987SP, with permission from Electric Power
Research Institute.

SCC is also significantly affected by the ECP.
As shown in Figure 12 by the work of Smialowska,25
small changes in the ECP can have a large effect on SCC.
Test results have demonstrated that a crack growth
rate maximum, with respect to coolant hydrogen
variation, is observed in proximity to a key phase transition, the nickel (Ni) to nickel oxide (NiO) phase transition, as shown bya Pourbaixdiagram (see Figures 13 and
14).26,27 The SCC hydrogen dependency is fundamentally described by the extent that the alloy’s corrosion
potential deviates from the potential of the Ni/NiO
phase transition. This potential difference represents
the relative stability of the SCC controlling oxide films
(e.g., crack tip oxides are often of a NiO structure).
5.04.4.2

Flow Rates

Flow rates in LWRs do not appear to have any effect
on SCC susceptibility and no testing data are known
to be available.


82

Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

Potentials typical of
deaerated environments

Oxide film and IGSCC
arrows indicate an increasing
sensitivity to cracking

RHE potential

Potential

Ni/Nio equilibrium

Potentials typical of environments
with dissolved hydrogen (10/30 bars)

No oxide film
no IGSCC

‘Neutral’ environments

‘Caustic’ environments

Stable Cr oxides

Cr very soluble
» 9/10
pH at 300/320 ЊC

Figure 13 Effect of pH and potential on the surface films and sensitivity to stress corrosion cracking of Alloy 600 material.
Reproduced from Scott, P. M.; Le Calvar, M. In Proceedings of the Sixth International Symposium on Environmental
Degradation of Materials in Nuclear Power Systems; The Minerals, Metals, and Materials Society: Warrendale, PA, 1993;
pp 657–667, with permission from The Minerals, Metals, and Materials Society.

5.04.4.3

Crevices

Nickel-base materials have been found to be very
susceptible to IGSCC in crevice areas of BWR applications, especially at welds where both weld-induced
sensitization and high residual stresses are present.
Stress corrosion crack initiation and growth of nickelbase materials in BWRs have been attributed to the
development of an acidic environment within crevices
because of the oxygen in normal BWR water chemistry, combined with high residual and applied stresses
resulting from the geometry and nearby welds. The
first widely reported occurrence of SCC of Alloy 600
material in BWRs was at the Duane Arnold unit in
1978.28 This occurred in the crevice area of a recirculation inlet nozzle safe end and was initiated in the
creviced area between the thermal sleeve and the safe
end. However, no known failures of nickel-base materials have been directly attributed to environmental
conditions within crevices in PWRs.

5.04.4.4

Material Susceptibility Factors

The PWSCC susceptibility of a particular heat of
Alloy 600 material is dependent upon a variety of
factors. The most important factors, based on results
from a number of investigators, appear to be material
microstructural features, chemical composition, and
manufacturing process (or product form). Each of
these has been evaluated below.
5.04.4.4.1 Heat treatment

Considerable research efforts have been made to

identify the mechanism responsible for PWSCC of
Alloy 600. These investigations, reviewed by Was,29
have shown that PWSCC is primarily dependent
on the heat treatment received by the material. For
instance, a heat treatment of mill-annealed (MA)
Alloy 600 in the temperature range 650–750  C
(1200–1380  F) has been found to cause a drastic
improvement in the resistance to PWSCC. This has


Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

83

1.50
O2

1.00

H2O

H

+

0.50

NiO2
1 ppb

V (SHE)

0.00

Acid
SO4
cracking

H2
1 atm
8.2 ppm

Ni3O4

Ni++
Ni
-0.50

NiO

Fe++
Fe

PWR
Secondary side
Primary side

Fe3O4
Fe

Caustic
cracking

PWSCC

-1.00

Ni(OH)3
Caustic
IGA

-1.50

-2.00

0

2

4

6

8

10

12

14

pH

Figure 14 Main domains of intergranular attack and stress corrosion cracking of Alloy 600 in aqueous solutions at
$300 C (572 F) relative to the Pourbaix diagram for nickel. (Note the stability boundary for iron relative to its lowest
oxidation states is also shown to indicate that iron can be oxidized under all likely PWR primary and secondary conditions
whereas elemental nickel can be stable under certain PWR primary conditions depending on the corrosion potential
fixed by the hydrogen partial pressure.) Reproduced from Scott, P. M.; Combrade, P. In Proceedings of the Eighth
International Symposium on Environmental Degradation of Materials in Nuclear Power Systems; American Nuclear Society:
La Grange Park, IL,1997; pp 65–73. Copyright 1997 by the American Nuclear Society, La Grange Park, IL, USA.

been termed ‘thermal treatment’ (TT). Although the
exact mechanism of enhanced resistance to PWSCC
is unclear and still under debate, there is general
agreement that the chemical composition and structure of grain boundaries are of crucial importance.
In this connection, chromium depletion, segregation
of impurities to grain boundaries, intergranular carbides and their mechanical effect on stress concentrations, and grain boundary misorientation appear to
be essential to the corrosion behavior of the material.
Annealing temperature and time are critical to the
precipitation reactions that occur. The penultimate
and final anneals appear to be the most critical
for PWSCC resistance. Stiller et al.,30 Briant et al.,31
Hall and Briant,32 EPRI NP-507233 all agree that a

high-temperature final anneal, sufficient to solutionize all carbon precipitates, followed by a slow
enough cooling rate to precipitate copious carbides
on the grain boundaries (intergranular carbides) will
produce a reasonably PWSCC-resistant material.
Norring34 tested Alloy 600 reverse U-bends (RUBs)
in high-purity water containing hydrogen. The time

to crack initiation at 320  C (608  F) increased by a
factor of 8 when the annealing temperature was
increased from 925  C (1697  F) to 1025  C (1877  F).
5.04.4.4.2 Microstructure

The precipitation obtained in the final material
microstructure is dependent upon heat treatment
and chemical composition (mainly carbon, chromium,


84

Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

titanium, and nitrogen content). During the cooling
that occurs after annealing, carbon combines with
chromium to form chromium carbides. In addition,
titanium carbides and titanium nitrides are formed.
These precipitate within the grains (intragranularly)
or at grain boundaries (intergranularly) depending
largely upon the temperature reached during annealing
and thermal treatments, the presence of prior precipitates (e.g., undissolved carbides), the time at temperature, the carbon content, and the cool down rate.
A study was conducted by the Electric Power
Research Institute (EPRI) in 1981 to evaluate carbide
dissolution and precipitation kinetics of Alloy 600.35
This study showed that final annealing temperatures
between 982 and 1010  C (1800 and 1850  F) (for
carbon contents between 0.02% and 0.048%) provide
a consistent thermal treatment response and an adequate precipitation of chromium carbides.
The most widely accepted hypothesis for the beneficial effect of intergranular carbides on PWSCC

resistance has been proposed by Bruemmer,36 who
suggests that grain boundary carbides promote crack
blunting due to their effectiveness as dislocation
sources. Another possible explanation proposed by
Smialowska25 is that Alloy 600 material passivates more
readily in the presence of grain boundary carbides.
Most recently, Fish et al.37 have proposed that intergranular carbides may improve passivity at the crack
tip by reducing the amount of carbon segregated along
the grain boundaries.
Whatever the mechanism, or mechanisms, involved,
there is general agreement that a microstructure with
copious intergranular carbides and few intragranular
carbides correlates with good resistance to PWSCC.
A range of intergranular and intragranular carbide
precipitation has been used by many investigators to
quantify PWSCC susceptibility.15 Scott et al.38 have
developed a range of material susceptibility indices,
based on minimum times to failure, for Alloy 600
material used in French steam generators. These factors
have also been used for CRDM nozzles.39,40
5.04.4.4.3 Grain size

Grain size can be related to yield strength and tensile
strength of Alloy 600 material by the general type of
Hall–Petch relationship.41 That is, a lower yield
strength material will typically have a larger grain
size. Data for PWSCC42 show that very small grain
sizes (typically ASTM grain size numbers >9) are
more prone to PWSCC than larger grain sizes (typically ASTM 4–8). Other data43 show that steam
generator tubing with a grain size number larger

than ASTM 6 is less susceptible to PWSCC. However, the effect of grain size is most likely a secondorder effect, while carbide precipitation tends to be
more important.
5.04.4.4.4 Chemical composition

The chemical composition of Alloy 600 has mainly
been correlated to PWSCC susceptibility in terms of
the carbon content. Many investigators have shown
that carbon contents near the high end of the typical
mill range (0.02–0.06 wt%) result in increased
PWSCC susceptibility. Electricite de France (EdF)
conducted a series of RUB specimen tests on steam
generator tubing, using a low mill anneal temperature
with carbon content ranging from 0.01 to 0.07 wt%.44
The tests were conducted in elevated temperature
pure water and primary water with hydrogen overpressure. These tests show that low carbon (<0.018%)
had the lowest PWSCC susceptibility. Pichon and
others45 have shown that carbon contents that are
relatively high (>0.063%) or very low (<0.012%)
are more susceptible to PWSCC.
The carbon content of Alloy 600 weld materials
(Alloys 182 and 82) has also been investigated by
several researchers.46,47 It appears that low carbon
contents (<0.02%) are most susceptible, medium
carbon contents (0.02–0.04%) are intermediate, and
high carbon contents (>0.04%) have the most resistance to PWSCC.
However, correlating the actual carbon content of
the material may not be the best approach. This is
also shown in the test data reported by Norring et al.34
A better approach may be to determine the ‘available

carbon’ content. This would entail knowledge of the
titanium and nitrogen content of the heat of material.
If it is assumed that all the nitrogen is precipitated as
titanium nitrides (TiN), then the remaining titanium
can be assumed to react with the carbon in the
material. With these assumptions (which are thermodynamically reasonable), the remaining carbon would
be the ‘available carbon’ amount to precipitate as
chromium carbides (M7C3 or M23C6 precipitates).
Unfortunately, titanium and nitrogen are not normally
reported in the certified material test reports (CMTRs);
thus, this approach cannot be used without archive
material or sampling of the heats in service.
Buisine et al.48 evaluated the PWSCC resistance of
nickel-based weld metals with various chromium
contents. The tests clearly demonstrated that weld
metals with !30% chromium were resistant to
PWSCC. The threshold for PWSCC resistance
appears to be between 22% and 30% chromium.


Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

There have been discussions in the literature of
the role of chromium depletion along the grain boundaries (‘sensitization’), as well as boron and phosphorous segregation along the grain boundaries.49 The
degree of sensitization is tied to the heat treatments
and there have been conflicting reports on the benefits
and disadvantages of sensitization. It has been shown
that a sensitized material is more susceptible to caustic
attack. In addition, boron and phosphorous may be
beneficial for retarding PWSCC, but the experimental

evidence is weak at present.
5.04.4.4.5 Product form

Alloy 600 component items in the RCS are fabricated
by press forging, hammer forging, hot rolling, and
forming and machining from bar stock in accordance
with ASME SB-166 or cold drawn and hot finished
tubing or piping in accordance with ASME SB-167.
In addition, it has been suggested that some ASME
SB-167 material was actually produced from SB-166
bar stock and recertified as SB-167 material. The
fabrication processes affect the material microstructure, yield strength, and hardness, which in turn
affect the PWSCC resistance of the material.
The PWSCC susceptibility of various product
forms has been evaluated to some degree. The most
susceptible product form appears to be cold worked
and low temperature MA steam generator tubing, as
evidenced by the vast majority of data available in the
literature. However, the other product forms (i.e.,
forgings, piping, and bar stock) have not been assessed
as extensively. One study, performed by Webb50
found differences in susceptibilities of cold-worked
and hot-worked materials with similar microstructures. The cold worked and annealed tubing was
more susceptible than hot worked and annealed
forging material. The different behavior remains
even when yield strength differences between the
materials are taken into account.
Another study on the relative susceptibility of
high and low yield strength bar and tubing, performed for EPRI,51 concluded that PWSCC susceptibility ranked in decreasing order as (1) high-yield
strength bar, (2) high-yield strength tubing, and

(3) low-yield strength bar. However, no low-yield
strength tubing was tested in this study.
Microstructural evaluation in the study described
above indicated that grain boundary carbide decoration is generally poor in bar products and better in
tubing products. However, the susceptibility of Alloy
600 material also depends on surface cold work due
to machining, grinding, and reaming. A material with

85

highly susceptible microstructure when subjected
to a large amount of cold work on the surface (i.e.,
reaming or grinding) becomes very susceptible to
PWSCC. Therefore, component items that were
machined from bar stock and with weld root grinding
should be considered highly susceptible to PWSCC.
A machined surface without reaming or grinding is
considered to have undergone moderate cold work
and somewhat less susceptible to PWSCC. The least
susceptible material is for cold drawn tubing with a
high mill-anneal temperature. This type of approach
has been adopted by Consumers Power Company
at the Palisades Nuclear Plant.52
The most recent work was performed by Briceno
et al.,53 who tested a variety of product forms.
Two groups of materials appear to exist, as shown in
Figure 15. The first group contains cold-worked
material, thick wall tube, and steam generator tubing;
whereas, the second group contains hot-worked thick
wall tubes. The forged bars showed higher crack initiation times than the tubes tested. Specimens made of

forged bars and hot-worked tube with similar grain
boundary carbide distributions ($60%) and the same
grain size (ASTM 4.0–4.5) showed a significant difference in initiation times. The same tendency is seen
for a low density of intergranular carbides when coldworked tube and forged bar with the same grain size
were compared.
5.04.4.5

Stress

Laboratory tests and operating experience in PWRs
suggest that significant PWSCC should not occur
in Alloy 600 component items at stresses less than
about 242 MPa (35 ksi) for temperatures up to around
324  C (615  F). As operating design stresses permitted by the ASME B&PV Code are much less than
242 MPa (35 ksi), PWSCC failure would not be
expected to occur due to applied pressure and thermal loadings. This is supported by the fact that
essentially all PWSCC failures have occurred at
locations where (1) high residual stresses are produced during fabrication (e.g., pressurizer nozzle
J-groove welds), (2) high stresses are produced as
a result of strains induced during operation (e.g.,
steam generator-dented tube support plate intersections), or (3) high stresses are produced by geometric abnormalities (e.g., excessive steam generator
tube ovality).
The following paragraphs address the operating
and residual stresses that may act in RCS pressure
boundary component items.


86

Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

Specimens cracked/specimens tested

1.0

0.8

0.6
FC no cracks
FB
FA
FX
V1V2
W1
W2 no cracks
GW
GV

0.4

0.2

0.0
0

Bar

Thick
wall
tube

S.G.
tube

1000

2000

3000
4000
Time (h)

5000

6000

7000

Heat

Working
method

Final thermal
treatment

YS (MPa)
after
forging

YS (MPa)
after
thermal
treatment

%C

FC

Forged bar

745 ЊC/2 h (a.c)

428

413

0.021

FB

Forged bar

800 ЊC/2 h (a.c)

491

489

0.021

FA

Forged bar

800 ЊC/2 h (a.c)

550

412

0.024

FX

Hot worked

(*)

–

291

0.03

V1

Cold rolled 1000 ЊC/3 min (w.c)

–

301.5

0.051

V2

Cold rolled 1000 ЊC/3 min (w.c)

–

274

0.052

W1

Hot worked

980 ЊC

–

320.6

0.081

W2

Hot worked

980 ЊC

–

244.7

0.067

GVW

Cold worked 927 ЊC/3–5 min (a.c)

–

393

0.04

GV

Cold worked 927 ЊC/3–5 min (a.c)

–

389

0.038

( * ) no heat treatment after extrusion of the tube.

% C = Percentage carbon content
Figure 15 Stress corrosion cracking initiation time at 330  C (626  F) for up to 10 500 h. Reproduced from Briceno, D.;
Blazquez, F; Hernandez, F. In Proceedings of the Eighth International Symposium on Environmental Degradation of Materials
in Nuclear Power Systems; American Nuclear Society: La Grange Park, IL,1997; pp 249–256. Copyright 1997 by the American
Nuclear Society, La Grange Park, IL, USA.

5.04.4.5.1 Operating stress

5.04.4.5.2 Residual stress

Operating stresses, on the order of 69–138 MPa
(10–20 ksi) in the hoop direction of J-groove, partial
penetration welds, as well as full penetration welds,
are anticipated in LWR component items. Stress
levels on this order are not high enough to cause
concern with SCC. Therefore, residual stresses must
be taken into account.

The magnitude and sign of residual stresses in most
typical RCS component items are considered to be a
function of (1) surface layer hardness produced by cold
working (e.g., rolling, machining, bending, or reaming)
and (2) residual stresses produced by welding.
Test data by Bandy and Van Rooyen (Figure 16)
indicate that the initiation time for PWSCC in


Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

87

Fraction of yield strength (RT)

10

AR

CW

345 ЊC (AR)

1.0

0.1
1

100

10

1000

Failure (days)
Figure 16 Correlation between stress and time to stress corrosion cracking for steam generator tubing tested in pure water
at 300  C (tfailure ¼ K Â Stressb, b ¼ À4). CW: cold work; AR: as-received. Reproduced from Bandy, R.; Van Rooyan D.
Quantitative examination of stress corrosion cracking of Alloy 600 in high temperature water – Work in 1983. In Proceedings:
1983 EPRI Workshop on Primary-Side Stress Corrosion Cracking of PWR Steam Generator Tubing; EPRI NP-5498, Project
S303–5, with permission from Electric Power Research Institute.

elevated temperature water varies inversely with the

fourth power of the applied stress.18 Test data by
Yonezawa et al.54 (Figure 17(a) and 17(b)) suggest
that the initiation time varies inversely with the sixth
or seventh power of the applied stress. In addition,
Figure 18 shows data from Yonezawa that suggest
that the absolute value of the stress raised to a
power is not the only significant factor. Other significant factors include the amount of cold work and the
ratio of the applied stress to the material yield
strength. In other words, the PWSCC initiation
time at a given applied stress level should increase
as the material strength is increased. Conversely, for
strain-controlled situations, such as at a J-groove
weld, where the residual stress is a fixed percentage
of the material yield strength, increasing the yield
strength can actually decrease the initiation time.
5.04.4.5.3 Surface effects

The residual stress in a component item is a function of the applied mechanical forces and the heat
input. Mechanical forces applied during cold work
tend to produce compressive residual stresses in
the surface layer. Conversely, heat input has the
potential to produce tensile residual stresses. Specifically, heat input causes expansion of the surface
layer yielding, if the thermal stresses exceed material yield strength and residual tensile stresses in
the surface layer upon cooling. This suggests that
high heat input operations such as drilling, turning,

or grinding have the potential to develop higher
tensile residual stresses than low heat input operations such as reaming. Note that these high tensile
residual stresses produced during machining are
limited to a few mils depth.

Berge et al.55 have shown, in crack initiation testing
performed in France, that the time to crack initiation with a base material tensile stress of 500 MPa
(72.5 ksi) is four times that of a material with a
cold-worked surface stress of 1000 MPa (145 ksi).
5.04.4.5.4 Weld geometry

The welding of small penetration nozzles into larger
vessel component items, by either partial penetration
or full penetration welds, can produce tensile stresses
in the nozzle. The magnitude of these stresses is
generally on the order of the yield strength of the
material. The cooling and subsequent shrinking of
the weld material pull the nozzle wall toward the
much stiffer component item to which it is welded.
This shrinkage results in residual tensile hoop stresses in the nozzle near the weld. This weld shrinkage
also results in axial bending of the nozzle. The magnitude of the bending stress, however, is typically
much lower than that of the hoop stress.
5.04.4.5.5 Stress relief annealing

The primary goal of stress relief heat treatment is to
allow a local realignment of highly strained regions
to reduce internal stresses. As mentioned above,


Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

Stress measured by X-ray diffraction analysis on the
top of U-bend specimens (kgf mm-2)

88

100

90
s-7

80

70

60

Crack

50
0.2% Offset proof stress at room temperature: 3.3 kgf mm-2
40

102

103

Applied stress (kgf mm-2)

(a)

104

Stress corrosion cracking testing time (h)

Crack
No crack
s-6
70
60
50
40
30
20

0.2% offset proof stress at 360 ЊC

10
0
(b)

102

103

104

Stress corrosion cracking testing time (h)

Figure 17 Correlation between stress and time to stress corrosion cracking for Alloy 600 steam generator tubing.
(a) Effect of stress on the stress corrosion cracking resistance of mill-annealed Alloy 600 at 360  C in high temperature
water. (b) Effect of stress on the stress corrosion cracking resistance of mill-annealed Alloy 600 at 360  C in high
temperature water, using constant load stress corrosion cracking test. Reproduced from Yonezawa, T.; Onimura, K.; Saito, I.;
Takamatsu, H. In Materials for Nuclear Reactor Core Applications; BNES: London, 1987; pp 77–83, with permission from
British Nuclear Energy Society.

improvement to the intergranular carbide precipitation is also almost always a benefit.
The degree of stress relief is known to be a function of time and temperature. Available data from the
International Nickel Company (INCO) are given in
Figure 19.56 It can be seen that at a temperature of
482  C (900  F) for 4 h, $21% of the stress is relieved.
Also, at a temperature of 870  C (1600  F) for 4 h,
$88% of the stress is relieved. It should be noted that
essentially all the stress relief occurs within $1 h at
the temperature of interest. Residual stresses are

generally reduced from values near the room temperature yield strength to values that are near zero stress.
5.04.4.6

Irradiation

Irradiation-assisted stress corrosion cracking
(IASCC) is an age-related degradation mechanism
where materials exposed to neutron radiation become
more susceptible to SCC with increasing fluence.57
IASCC, like PWSCC, is a distinctive subset of SCC.
Despite numerous investigations and research efforts,


Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

Applied
stress
(kgf mm-2)

Material
Heat
treat
ment

Stress corrosion cracking
testing time (h)

Applied
stress
-2
Frac- (kgf mm )
ture
0.8σ0.2

Cold
working

1.0sO2

—

24

8%
tensile
strained

45

15%
tensile
strained

57

´ (2.793)
´ (2.713)

IGSCC

975 ЊC 25%
0.07 annea- tensile
ling strained

72

´ (159)
´ (3.231)

IGSCC

C (%)

20%
cold
rolled

87

2.000 4.000 6.000 8.000 10.000 12.000 14.000

´ (1.717)

Fracture

2.000 4.000 6.000 8.000 10.000 12.000 14.000

—

9.505
´ (3.704)
´ (5.772)

Stress corrosion cracking
testing time (h)

89

IGSCC
IGSCC

IGSCC

IGSCC
IGSCC

—

45

(12.8481)

IGSCC

´ (2.374)
´ (4.191)

57

IGSCC

—

No crack
´ Crack

Figure 18 Effect of applied stress to yield strength ratio on time to stress corrosion cracking for Alloy 600 steam
generator tubing. Effect of cold working on the stress corrosion cracking resistance for the same ratio of applied stress to
yield strength of mill-annealed Alloy 600 at 360  C in high temperature water. Reproduced from Yonezawa, T.; Onimura, K.;
Saito, I.; Takamatsu, H. In Materials for Nuclear Reactor Core Applications; BNES: London, 1987; pp 77–83, with permission
from British Nuclear Energy Society.

30
482 °C (900 °F)
593 °C (1100 °F)
25
538 °C (1000 °F)
Residual stress (1000 psi)

649 °C (1200 °F)
20

15
704 °C (1300 °F)
10
760 °C (1400 °F)
5
871 °C (1600 °F)

0

0

1

2
Time (h)

3

4

Figure 19 Effect of heating time and temperature on residual stress of cold-drawn, annealed Alloy 600 rod.
Reproduced from Inconel Alloy 600, International Nickel Company: Huntington, WV, 1978, with permission from Special
Metal Corporation.

details of the IASCC mechanism remain hypothetical.58–61 The current consensus is that IASCC results
from a synergistic effect of irradiation damage to the
material, water environment with possible radiolysis

effects, and a stress state. At present, interactions
between these variables have not adequately been
quantified and no primary IASCC controlling mechanism has been identified.


90

Corrosion and Stress Corrosion Cracking of Ni-Base Alloys

Nickel-base alloys, such as Alloy 600 and 690,
are not typically utilized in areas of high fluence
because of their high nickel content. Nickel will
become highly radioactive when exposed to neutron
radiation. No known IASCC-related studies have
been reported in the open literature for these alloys.
There are, however, some data available for Alloy
X-750 at low neutron fluence levels. Results obtained
from swelling capsule tests indicate that IASCC
resistance of Alloy X-750 is approximately the same
as Types 304 and 316 stainless steel.62,63 Some tensile
test data of small scale Alloy X-750 HTH Condition
bolts irradiated in a PWR to 1.4 Â 1019 n cmÀ2
(E > 1.0 MeV) performed without any failure up to
0.95% strain.64 Simulated BWR testing demonstrated
an irradiation-enhanced IGSCC susceptibility of
Alloy X-750 HTH Condition as a function of fluence
and boron content.65 Fluence at 1019 n cmÀ2 versus
1014–1018 n cmÀ2 (E > 1.0 MeV) showed an increase
in susceptibility.

Alloy 718 has an excellent record in PWR primary
water applications often at operating stresses close to or
exceeding the yield point that can be $1000–
1100 MPa (145–160 ksi). Alloy 718 is used in fuel applications where very high neutron fluxes are also experienced. The few failures that have occurred have been
attributed to a manufacturing defect that allowed components to enter service with preexisting intergranular
defects. Alloy 718 is known to be highly resistant to
crack initiation, but IGSCC will propagate rapidly in
PWR primary water from preexisting defects.
5.04.4.7

Mitigation

A number of techniques have been evaluated and are
available to delay or eliminate the occurrence of SCC
in LWRs. Many of the available and future mitigation
techniques were evaluated in a paper presented at a
conference held by the US Nuclear Regulatory Commission (NRC) in 2003.66 In general, these techniques
fall into three main categories:
1. Mechanical surface enhancement (MSE)
2. Environmental barriers or coatings
3. Chemical or electrochemical corrosion potential
(ECP) control
MSE techniques represent processes that reduce surface tensile residual stresses or induce compressive
surface stresses on a component item or weld. Examples of MSE techniques include shot peening and
electropolishing. Environmental barrier or coating
techniques represent processes that protect the

material surface from an aggressive environment.
Coating examples include nickel plating and weld
deposit overlays or inlays. Chemical or ECP control

techniques represent changes to the environment
that alter the corrosion process or produce corrosion
potentials outside the critical range for SCC. Examples of chemical or ECP control include zinc additions
to the reactor coolant and modified water chemistry
(e.g., hydrogen water chemistry and noble metal chemical additions to BWR coolant; variations in dissolved
hydrogen levels, lithium concentrations, and boron
concentrations for PWR coolant).

5.04.5 Outlook
Wrought nickel-base alloys and their weld metals
were originally used in LWRs due to the materials’
inherent resistance to general corrosion in a number
of aggressive environments and because of a coefficient of thermal expansion that is very close to that of
low alloy and carbon steel. Over the last 40 years,
SCC has been observed in numerous component
items and associated welds, sometimes after relatively
long incubation times. The occurrence of SCC has
been responsible for significant downtime and
replacement power costs.
Nickel-base materials will continue to be used both
for repair and replacement activities in currently
operating nuclear units and in the next generation of
LWRs. Although the continued use of Alloy 600 material, for example, has essentially been eliminated in
LWRs, acceptance and use of the higher chromium
material, Alloy 690, is widely recognized by the industry. These higher chromium materials have been
shown to be highly resistant to SCC in laboratory
experiments and component repairs. Their use has
proven to be an effective decision, since replacements
have been free from cracking in operating reactors over
periods up to about 20 years. Improvements in the

weldability of the higher chromium weld metals (i.e.,
variants of Alloys 52 and 152) have been identified
through significant research and the results have been
put into application today.

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