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

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

of Ni-Base Alloys

S Fyfitch

AREVA NP Inc., Lynchburg, VA, USA

ß 2012 Elsevier Ltd All rights reserved.

ASME American Society of Mechanical Engineers

ASTM American Society for Testing and Materials

B&PV Boiler and Pressure Vessel

BWR Boiling water reactors

CEDM Control element drive mechanism

CMTRs Certified material test reports

CRDM Control rod drive mechanism ECP Electrochemical potential EPRI Electric Power Research Institute GMAW Gas-metal-arc welding

GTAW Gas-tungsten-arc welding HAZ Heat-affected zone IASCC Irradiation-assisted stress corrosion

cracking

69

IGSCC Intergranular stress corrosion cracking

INCO International Nickel Company

LWR Light water reactor

MSE Mechanical surface enhancement

NRC Nuclear Regulatory Commission

PWR Pressurized water reactors

PWSCC Primary water stress corrosion cracking

RCS Reactor coolant system

RUBs Reverse U-bends

SAW Submerged-arc welding

SCC Stress corrosion cracking

SEM Scanning electron microscope

SMAW Shielded-metal-arc welding

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,1especially 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

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

5.04.2 Ni-Base Alloy Use in PWRS/ BWRS

5.04.2.1 Wrought Ni–Cr–Fe AlloysThe wrought nickel-base alloys that are typically usedfor nuclear applications are Alloy 600 and, morerecently, Alloy 690, which contain approximatelytwice the chromium content These materials areused primarily for their inherent resistance to generalcorrosion (i.e., oxidation resistance), strength at ele-vated temperatures, and a coefficient of thermal expan-sion very close to carbon and low-alloy steels Thetypical chemical composition and mechanical proper-ties of these alloys are summarized inTables 1 and 2,respectively

Both Alloy 600 and Alloy 690 are hardenable, austenitic solid-solution strengthenedmaterials No precipitation reaction is possible witheither alloy to increase strength; however, strengthcan be increased by cold-working the material Theyare normally used in the annealed condition; how-ever, a low-temperature heat treatment, or ‘thermaltreatment’ (TT), is also generally used with thesealloys, which tends to improve the resistance toSCC in primary water chemistry conditions, which

non-age-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 forAlloy 600 material, in crack initiation testing.3

Table 1 Chemical composition of wrought nickel-base alloys used in nuclear applications

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

mech-anism (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

Figures 1–4 show typical applications of Alloy 600material as used in the reactor coolant systems of thefour 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 AmericanSociety for Testing and Materials (ASTM)4or Amer-ican Society of Mechanical Engineers (ASME) Boilerand Pressure Vessel (B&PV) Code5 specifications(e.g., ASTM B 166 and B 167 or ASME SB-166 andSB-167).Table 3lists the various industry specifica-tions that are used to procure these materials

Feedwater recirculation inlet/outlet welds RPV attachments/

brackets

Shroud support structure

CRD in-core housing instrumentation penetrations Courtesy GE nuclear

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.

Table 2 Typical room temperature mechanical properties of wrought nickel-base alloys used in nuclear applications

aAlloy X-750 HTH: Solution annealing at 1093C (2000F) and age-hardening at 704–718C (1300–1324F).

bAlloy 718: Solution annealing at 1100–1400C (1832–2000F) and age-hardening at 720 and 620C (1328 and 1148F).

cAlloy 800: Solution annealing at 1038–1066C (1900–1950F) and age-hardening at 760C (1400F) for 10 h, furnace cool to 649C

(1200F), hold for 20 h.

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

appli-cations 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

com-position to Alloy 600, but contains additions of

tita-nium and aluminum, which combine with nickel to

form the g0precipitates, Ni3Al and Ni3(Al, Ti), for

strengthening.6

Alloy 718 is another age-hardenable austenitic

nickel-base alloy, originally developed for the

aero-space 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

reac-tions from solid solution involving g0and g00(NiNb)

secondary phases within the austenitic matrix.8 Theaddition of niobium sets this alloy apart from otherhigh-strength nickel-base alloys (e.g., Alloy X-750)that are strengthened by g0alone Both the g0 and g00precipitates are quite small and can only be resolvedwith an SEM (scanning electron microscope) unlessgross over-aging has taken place The microstructures

of the solution-annealed and age-hardened conditionsare indistinguishable with light microscopy (LM).Alloy 718 has also been used extensively in PWRprimary coolant systems, predominantly for fuel assem-bly hardware.9Alloy 718 is utilized for fuel assemblyhold-down springs, bolts, and spacer grids It has beenshown to possess superior SCC initiation resistancecompared to Alloy X-750 Although Alloy 718 hasexperienced some isolated failures in PWRs due tofatigue/fretting cracking, it is considered highly resis-tant to intergranular SCC (IGSCC) initiation andother forms of corrosion

Core guide lugs (ID)

RV bottom head instrument penetrations Instrument nozzles and drain penetrations (all cold legs)

SG nozzle dam rings (both SGs)

Primary drain nozzles (both SGs)

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

Hot surge nozzle weld

leg-PZR surge nozzle weld

PZR heater sleeves and diaphragm plates

PZR steam and water instrument penetrations

PZR vent, spray, and relief line welds

Instrument and vent penetrations (both hot legs)

Decay heat line weld CRDM nozzles CRDM motor housings Leak-off monitor lines

HPI/MU

nozzle

welds

(all cold legs)

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.

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 inTables 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

gas-tungsten-arc welding (GTAW), shielded-metal-arc

welding (SMAW), and gas-metal-arc welding

(GMAW).7Submerged-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 joinAlloy 690 component items to carbon or low-alloysteel vessels in the reactor coolant system

Occasionally, there is a need for welded AlloyX-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 andfiller metal Alloy 718 is the currently recommendedmaterial for welding

The typical chemical compositions and industryprocurement specifications of these alloys are sum-marized inTables 3 and 4

One of the main reasons that nickel-base alloys werechosen for LWR applications is that they have theability to withstand a wide variety of severe operatingconditions involving corrosive environments, hightemperatures, high stresses, and combinations of thesefactors General corrosion can be defined as uniformdeterioration of a metal surface by chemical or

(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)

Spray nozzles (2 cold legs)

Shutdown cooling inlet nozzle (all cold legs) Guide lugs flow skirt ICI nozzles-ICI guide

and discharge (all cold legs)

Bottom channel head drain tube and welds (both SGs) Shutdown cooling outlet nozzle (1 hot leg)

Primary nozzle closure rings and welds (both SGs)

PZR and RC pipe-surge line connections CEDM motor housing CEDM/ICI nozzles to RPV head welds

RPV top head vent nozzle RPV head leak monitor tubes (2)

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.

electrochemical reaction with the environment Nickel

has good resistance to corrosion in the normal

atmo-sphere, 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 additionsprovide resistance to sulfur compounds and also pro-vide resistance to oxidizing conditions at high tem-peratures or in corrosive solutions Details of thecorrosion resistance in these types of environmentscan be found elsewhere (i.e., see also Chapter 2.08,Nickel Alloys: Properties and Characteristics).2,10

Core support

and welds (all SGs)

Primary nozzle closure rings and welds (all SGs)

Thermowells (all hot and cold legs)

Surge pipe welds

nozzle-CRDM nozzles

CRDM motor housings

Spray pipe weld

nozzle-Safety and relief nozzle-pipe welds

RPV head leak monitor tube

Head vent pipe

Bottom-mounted instrument nozzles 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

5.04.3.1 Water Chemistry

The water chemistry of LWRs is discussed in detail

inChapter5.02, 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

corro-sion They are able to withstand very high flow rates,

on the order of 18.3 m s1(60 ft s1), without concern

Corrosion rates in such flowing conditions for

nickel-base materials are expected to be <2.5 mm year1

(<0.1 mil year1).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

environ-ments, mitigation is not really necessary As noted

in this section on LWR Structural Materials, PWRenvironments have reducing conditions and generalcorrosion is not of concern However, mitigation ofcorrosion concerns in the more oxidizing environ-ment of BWRs has been through the use of hydrogenwater chemistry (i.e., to make the environment morereducing, similar to a PWR) and noble metal chemicaladditions.12,13

5.04.4 Stress Corrosion Cracking

SCC of nickel-base alloys is an important age-relatedphenomenon affecting LWRs This type of failuremechanism for nickel-base alloys typically occursintergranularly and is generally termed intergranularstress corrosion cracking (IGSCC) In PWRs, IGSCC

is typically termed primary water stress corrosioncracking (PWSCC) The occurrence of SCC of nickel-base alloys has been extensively studied since the firstreported observation of cracking in laboratory testsusing Alloy 600 in high-purity water by Coriouet al.14

in 1959 Over the last three decades, IGSCC hasbeen observed numerous times in LWRs and it hasaffected both the safe and economic operation ofthe reactors In BWRs, cracking of nickel-basecomponents such as safe ends, shroud bolts, and accesshole covers has occurred; however, the predominantfailures have been identified in Alloy 182 welds InPWRs, PWSCC of Alloy 600 component items hasbeen observed in steam generators, pressurizers, and

Table 4 Chemical composition of nickel-base welding alloys used in nuclear applications

Alloy 72 filler metal

Alloy 82 filler metal

Alloy 132 electrode

Alloy 152 a

electrode

Alloy 182 electrode

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

CRDM nozzles, and most recently in Alloy 182 and

Alloy 82 welds The mechanism of this cracking

phenomenon is not completely understood, and

pre-diction of crack initiation time has proven to be

extremely difficult, if not impossible, due to the

uncer-tainty 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

mate-rials occurs as a result of the following three factors:

 susceptibility of the material

 a tensile stress (including both operating and

resid-ual 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

treat-ment 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

precipi-tate from the solution at both intragranular and

intergranular locations If the cooldown from the

anneal is sufficiently slow, a greater number of carbideswill precipitate at the grain boundaries (i.e., intergra-nularly) and the resistance to PWSCC will beimproved Well-decorated grain boundaries are anindication that an Alloy 600 material has receivedproper heat treatment and that sufficient carbonwas available in the solution to combine with chro-mium If adequate amounts of carbon and chromiumexist, but the anneal is not at a high enough tempera-ture or sufficient time is not allowed to solutionize thecarbon, an adequate amount of carbon will not beavailable to precipitate intergranularly as chromiumcarbides, leading to minimal grain boundary decora-tion Most precipitation occurs during cooldownfollowing annealing; however, stress relief treatmentscan 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 arealso reduced given the extended time at the tempera-ture Thus, if the anneal has not adequately solutio-nized carbon for chromium carbide precipitation atthe grain boundaries, stress relief treatment will notreduce susceptibility to PWSCC

Tensile stresses, resulting from both residual andoperating stresses, can be significant for some Alloy

600 component items Operating stresses are duced from mechanical and thermal loading, whileresidual stresses are generated as a result of fabrica-tion, installation, and welding processes Residualstresses are more difficult to quantify than operatingstresses and, in many instances, are of a higher mag-nitude than operating stresses

pro-PWSCC is a thermally activated degradationmechanism, that is, as the temperature increases, therate of PWSCC increases exponentially Thus, thehot leg temperature of the RCS creates a moreaggressive environment in which the Alloy 600 com-ponents must operate

The cracking observed in PWRs to date is cally axially oriented (although circumferentially ori-ented cracks have been observed) and occurs in

typi-an area, such as a weld heat-affected zone (HAZ),that has high residual tensile stresses In a cylindricallyshaped component (e.g., piping, vessel, and nozzles),the circumferential stresses are inherently higher thanaxial stresses Thus, in a homogeneous material with noinitial flaws, cracking would be expected to occuraxially because of the higher circumferential stresses.PWSCC has been the subject of much researchand analyses in recent years as a result of the manyfailures that have been attributed to it However,

Mechanical Operational tensile stresses

residual tensile stress

Corrosive Electrochemical corrosion potential temperature pH-value

Susceptible material

Chemical composition

microstructure

SCC

Figure 5 Synergistic factors affecting stress corrosion

cracking of nickel-base materials.

a reliable crack initiation model has yet to be

devel-oped 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,

hydro-gen, lithium, boron, and sulfur content), and

electro-chemical potential (ECP) Each of these factors is

evaluated as follows

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 28–39C

(50–70F) temperature differences between hot andcold legs are enough to significantly influence thetime to initiation and subsequent crack growth rate.Temperature is generally believed to affect therate of SCC attack in accordance with an activationmodel for thermally controlled processes (Arrheniusequation), exp(Q/RT), where Q is the activationenergy, R is the ideal gas constant, and T is theabsolute temperature The current consensus is thatthe activation energy for crack initiation falls in therange of 188–230 kJ mol1 (45–55 kcal mol1) andmany predictions are based on 210 kJ mol1(50 kcalmol1) There is also evidence that the activationenergy varies with material carbon content.155.04.4.1.2 Water chemistry

The water chemistry of LWRs can generally bedescribed as essentially pure water PWRs primarilyinclude hydrogen, boron, and lithium to produce reduc-ing conditions BWRs primarily operate with low levels

of oxygen, but in recent times, hydrogen additions havebeen introduced to limit the oxidizing potential ofthe environment Additional details are included inChapter5.02, Water Chemistry Control in LWRs

– – Hydrogen

– – –

500 ppm B 1.0 ppm Li hydrogen

1100 ppm B 2.0 ppm Li hydrogen

Figure 6 Stress corrosion cracking initiation times for Alloy 600 steam generator tubing at 360C (680F) 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.

5.04.4.1.2.1 Hydrogen

The effect of dissolved hydrogen on SCC

suscepti-bility 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

tempera-tures of 360C (680F) and above The authors

con-cluded that the susceptibility of Alloy 600 increased

when the amount of dissolved hydrogen increased

Airey17has shown that dissolved hydrogen increases

the rate of PWSCC of steam generator tubing in

autoclave tests at 360C (680F) An example of his

data, shown inFigure 6, shows that the SCC tion time for pure water is decreased dramaticallywhen hydrogen is added However, in the primarywater of PWRs, the effect of hydrogen appears to be afunction of the boron and lithium content Bandy andVan Rooyen18have shown a similar effect with boronalone versus pure water and primary water (Figure 7).They showed that 83% of the specimens cracked inpure water with hydrogen versus only 2% in purewater without hydrogen

initia-More recently, Norring has reported on tests todetermine the effect of hydrogen overpressure in

330C (626F) water.19Results of these tests, shown

in Figure 8(a), suggest that the rate of PWSCCincreases with increasing hydrogen overpressure.The most recent update on the influence ofhydrogen on PWSCC was prepared by Cassagne

et al.20

in 1997 All the data seem to indicate that thesusceptibility of Alloy 600 decreases drastically forlow hydrogen values (<10 kPa (1.45 psi)) regardless oftemperature For hydrogen partial pressure above

100 kPa (14.5 psi), a more progressive decrease insusceptibility seems to occur between 360 and

400C (680 and 752F) Data are not available inthis range for lower temperatures Between 10 and

100 kPa (1.45 and 14.5 psi), it appears that PWSCCinitiation is not greatly affected for all temperaturesbetween 400 and 310C (752 and 590F) At 290C(554F), the influence of hydrogen cannot beassessed because of a lack of data

5.04.4.1.2.2 Boron and lithium

Evaluations of boron and lithium on PWSCC of Alloy

600 have been performed by numerous investigators.Norringet al.19

concluded that increasing the lithiumcontent from 2.4 to 3.5 ppm significantly decreasedthe time to crack initiation (seeFigure 8(b)).The most complete evaluation was performed byOgawaet al.21

In these tests, hydrogen overpressure waskept at a constant level of 30 cm3kg1H2O The results

of this work are shown inFigures 9 and 10 It appearsthat maintaining a constant pH of 7.1–7.3 (at 285C(545F)) will produce a range of crack initiation times(Figure 9) and that increasing the boron contentgreater than 1200 ppm (at any lithium level) decreasesthe crack initiation times (Figure 10) Therefore,the beginning of cycle boron concentrations appears

to be the worst condition for PWSCC initiation taining a pH level of 7.3 (at 285C (545F)) alsoappears to be better than a pH level of 7.1

Main-Follow-up tests were performed at high boronconcentrations with varying lithium concentrations

Flattened specs, I.E.

Cold worked, pure H2O

As received, pure H2O + H 2

As received, pure H2O + H 3 BO3

As received, pure H2O

As received, (0.03 % C), pure H2O

Primary H2O

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.

100

99 90 50 25 10 5

99 90 50 25 10 5

7.0 kPa 7.354

329.6 ⬚C

1441 ppm 2.38 ppm 25.2 ml kg -1

13.6 kPa 7.351

13.6 kPa 7.351

99 90 50 25 10 5

1

328.9 ⬚C

1241 ppm 3.54 ppm 24.7 ml kg -1

13.6 kPa 7.575

to simulate beginning of fuel cycle conditions Ogawa

et al.22

report that there is little effect of lithium

con-tent from 2 to 10 ppm at boron concentrations greater

than 1200 ppm (seeFigure 11) and PWSCC

suscep-tibility at 1600 ppm boron (2–10 ppm lithium) was

higher than that at 500 or 280 ppm boron

concentra-tions (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 ofattack, although all nickel-base materials will beattacked by sulfate

Andresen23 has shown that the time to failuredecreased by 2–3 orders of magnitude in constantload tests between pure water and sulfate impurities(conductivities ranging from 5 to 55 mS cm1due tosulfuric acid additions)

Bandy et al.24

have shown that sulfates are verypotent cracking agents for Alloy 600 materials Tem-perature significantly accelerates the cracking andmost likely decreases the threshold stress for cracking

to occur

0 1 2 3 4 5

1.1 1.0

6.9

Figure 9 Isosusceptibility diagram for primary water stress corrosion cracking of Alloy 600 as a function of lithium and pH at

285C (545F) 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.

1.7 1.6