Comprehensive nuclear materials 5 02 water chemistry control in LWRs

Comprehensive nuclear materials 5 02 water chemistry control in LWRs Comprehensive nuclear materials 5 02 water chemistry control in LWRs Comprehensive nuclear materials 5 02 water chemistry control in LWRs Comprehensive nuclear materials 5 02 water chemistry control in LWRs Comprehensive nuclear materials 5 02 water chemistry control in LWRs Comprehensive nuclear materials 5 02 water chemistry control in LWRs Comprehensive nuclear materials 5 02 water chemistry control in LWRs

C J Wood

Electric Power Research Institute, Palo Alto, CA, USA

ß 2012 Elsevier Ltd All rights reserved.

5.02.2.2 Mitigating Effects of Water Chemistry on Degradation of Reactor Materials 20

Abbreviations

flux depression in reactor core caused by buildup of boron- containing deposits Originally called AOA for axial offset anomaly.

ammonia for pH control in steam generators

referring to designated standard points in BWR reactors for radiation field measurements

fuel element surfaces

potential

BWRs in the United States and some other countries

hydrogen

hydrogen

cracking

secondary side)

steam generator tubes

17

OLNC On-line noble chemistry

cracking

Other chapters of this comprehensive describe the

vari-ous degradation processes affecting the structural

mate-rials used in the construction of nuclear power plants

(seeChapter5.04, Corrosion and Stress Corrosion

Cracking of Ni-Base Alloys;Chapter5.05,

Corro-sion and Stress CorroCorro-sion Cracking of Austenitic

Stainless Steels; andChapter 5.06, Corrosion and

Environmentally-Assisted Cracking of Carbon and

Low-Alloy Steels) This chapter describes the influence

of water chemistry on corrosion of the most important

materials in light water reactors (LWRs) In particular,

alloys susceptible to intergranular attack (IGA) and stress

corrosion cracking (SCC) are significantly impacted

by water chemistry, most notably, sensitized 304

stain-less steel in boiling water reactors (BWRs) and

nickel-based alloys in pressurized water reactors (PWRs)

Excellent water quality is essential if material

degradation is to be controlled In the early days of

nuclear power plant operation, impurities in the

coolant water were a major factor in causing excessive

corrosion Chlorides and sulfates are particularly

aggressive in increasing intergranular stress corrosion

cracking (IGSCC) and other corrosion processes

Transient increases of impurities in the coolant that

occur during fault conditions (e.g., condenser leaks

and ingress of oil or ion exchange resins) proved to be

particularly damaging Thus, water chemistry was

traditionally regarded as a key cause of material

deg-radation Initial efforts to improve water quality

brought about a slow but steady reduction in

impu-rities through improved design and operation of

puri-fication systems Not only were the average

concentrations of impurities reduced over time, but

the frequency and magnitude of impurity ‘spikes’

from transient fault conditions were also diminished

However, excellent water chemistry alone was notsufficient to control corrosion Hence, programs tomodify water chemistry were introduced, includingminimizing oxygen to reduce the electrochemicalcorrosion potential (ECP) in BWRs, and oxygen and

pH control in PWRs More recently, additives tofurther inhibit the corrosion process have been devel-oped and are now in widespread use As a result,water chemistry advances are now an important part

of the overall operating strategy to control materialdegradation

Primary system water chemistry also affects fuelperformance through the deposition of corrosionproducts on fuel pin surfaces, and influences radia-tion fields outside the core Core uprating throughincreased fuel duty has reduced margins for tolerat-ing corrosion products (CRUD) on BWR fuel pinsurfaces In PWRs, increasing fuel cycle durationhas increased the challenge of controlling pH withinthe optimum range At the same time, regulatorylimits on worker radiation exposure are tending to

be tightened worldwide, putting pressure on theoperators to reduce radiation dose rates Successfuloperation of PWR steam generators (SGs) and theremainder of the secondary system demand strictwater chemistry control in secondary side systems ifcorrosion problems are to be avoided

Other operating parameters also influence theoptimization process, for example, life extension (to

60 years) has emphasized the importance of controllingdegradation of circuit materials Therefore, althoughcontrol of structural material degradation remains thehighest priority, water chemistry must be optimizedbetween the sometimes-conflicting requirements affect-ing other parts of the reactor

Advances in water chemistry have enabled plantoperators to respond successfully to these technicalchallenges, and the overall performance has steadilyimproved in recent years.1 Plant-specific considera-tions sometimes influence or indeed limit the optionsfor controlling water chemistry, so we see differentchemistry specifications at different plants This is espe-cially true internationally and significant differencesbetween countries are noted The US industry starteddeveloping water chemistry guidelines 25–30 yearsago, and these now provide the technical basis forguidelines in many other countries The early editions

of these guidelines presented impurity specificationsand required action if limits were exceeded Whenadvanced water chemistries were developed and qual-ified, the guidelines evolved into a menu of optionswithin an envelope of specifications that should not be

exceeded Guidance is now provided on how to select

a plant-specific water chemistry strategy.2

The basis for water chemistry control was

dis-cussed in detail by Cohen.3 The remainder of this

chapter describes more recent water chemistry

devel-opments for BWRs, PWR primary systems, and

PWR secondary systems including SGs, with a

short section on flow-assisted corrosion (FAC) in

both BWRs and PWRs

5.02.2 BWR Chemistry Control

5.02.2.1 Evolution of BWR Chemistry

Strategies

BWR water chemistry has to be optimized between

the requirements to minimize material degradation,

avoid fuel performance issues, and control radiation

fields These factors are depicted inFigure 1,4which

also includes the main chemistry changes involved in

the optimization process

Plant-specific considerations sometimes influence

or indeed limit the options for controlling water

chem-istry, so we see different chemistry specifications at

different plants This is especially true internationally

and significant differences in chemistry strategies

between countries are noted Design features are an

important reason for these different chemistry regimes,

to which must be added the effects of different

opera-tional strategies in recent years For example, a key

issue facing BWRs in the United States concerns

IGSCC of reactor internals, as discussed in other

chapters The occurrence of IGSCC resulted in the

implementation of hydrogen water chemistry, with orwithout noble metal chemical addition (NMCA), toensure that extended plant lifetimes are achieved.German plants use 347 stainless steel, which is lesssusceptible to IGSCC than sensitized 304 stainlesssteel used originally in US-designed plants SomeSwedish and Japanese plants have replaced 304 stain-less steel reactor internals with 316 nuclear gradematerial to minimize potential problems, as this mate-rial is less susceptible to IGSCC As a result, many ofthese plants continue to use oxygenated normal waterchemistry, whereas all US plants control IGSCCthrough the use of hydrogen water chemistry (HWC)with or without normal metal chemical addition toimprove the efficiency of the hydrogen in reducingECP Second, BWRs in United States undoubtedlyhave greater cobalt sources than plants in most othercountries, despite strong efforts to replace cobaltsources This resulted in higher out-of-core radiationfields, leading all US plants to implement zinc injection

to control fields, whereas only a small number of plants

of other designs use zinc Third, the move to longer fuelcycles and increased fuel duty at US plants, whilehaving major economic benefits, has led to new con-straints on chemistry specifications in order to avoidfuel performance issues

Figure 2 depicts the changing chemistry gies over the past 30 years, showing the focus onimproving water quality in the early 1980s and themove to educing chemistry to control IGSCC, which

strate-in turn resulted strate-in strate-increased radiation fields, quently controlled by zinc injection

subse-Chemistry control issues

Materials degradation and mitigation

Fuel performance

Water chemistry guidelines

Radiation exposure

BWR internals IGSCC, IASCC:

HWC or NMC required

Clad corrosion crud deposition:

Limits on feedwater zinc

Radiation fields crud bursts:

Impurity control:

Monitoring/analysis

Figure 1 Boiling water reactor chemistry interactions.

Increasing concerns about core internals

crack-ing led to the need to increase hydrogen injection

rates, which in turn resulted in the introduction

of NMCA to reduce operating radiation fields from

N-16.Figure 3 shows the rate of implementation of

HWC, zinc and NMCA, and online noble metal

addition (OLNC) The rationale and implications of

these developments are discussed in greater detail in

subsequent sections

The goal for BWRs is therefore to specify chemistry

regimes that, together with the improved materials

used in replacement components (e.g., 316 nuclear

grade stainless steel), will ensure that the full extended

life of the plants will be achieved without the need for

further major replacements At the same time,

radia-tion dose rates, and hence worker radiaradia-tion exposure,

must be closely controlled, and fuel performance must

not be adversely affected by chemistry changes

The first requirement of plant chemistry is to tain high-purity water in all coolant systems, includingthe need to avoid impurity transients, which arebeyond the scope of this paper The performance ofall plants has improved steadily over the years, asshown by the trend for reactor water conductivity forGE-designed plants, given in Figure 4 This figureshows that conductivity now approaches the theoreti-cal minimum for pure water In fact, deliberately addedchemicals, such as zinc (discussed in the followingsection), account for much of the difference betweenmeasured values and the theoretical minimum.The conductivity data are consistent with thereactor water concentrations for sulfate and chloride

main-In fact, sulfate is the most aggressive impurity fromthe viewpoint of IGSCC, and much effort has goneinto reducing it

5.02.2.2 Mitigating Effects of WaterChemistry on Degradation of ReactorMaterials

IGSCC was first observed in small bore piping usingsensitized 304 stainless steel fairly soon after BWRsstarted operation Laboratory studies showed thatimpurities increased IGSCC rates, and in fact waterquality in BWRs gradually improved in the early1980s However, the same studies found IGSCC inhigh-purity oxygenated water typical of good BWRoperations The key parameter affecting IGSCC wasfound to be ECP, as shown inFigure 5 In this labo-ratory test, the change from oxidizing conditionstypical of normal water chemistry (NWC) operation

0 5 10 15 20 25 30 35 40

Core internals cracking

control with lower fields

2006–2008:

Online Noblechem Promising new option

Figure 2 Evolution of Boiling water reactor chemistry

options from 1977 to 2008.

to reducing conditions greatly reduced the rate of

crack growth

Furthermore, hydrogen injection was effective at

reducing the ECP in BWRs, as shown inFigure 6

In this figure, it can be seen that crack growth rates

(CGR) for Alloy 182 were low in hydrogen water

chemistry (HWC), but increased greatly when theplant reverted to normal water chemistry (NWC).These results indicated that continuous hydrogeninjection was required to fully mitigate cracking.Examination of extensive inspection data from severalplants indicated that no IGSCC was observed with an

Outlet cond: 0.30 μS cm –1 Inlet cond: 0.27 μS cm –1 Na2SO4Dissolved O2

Figure 5 Laboratory results showing the effect of reducing oxygen concentration on crack growth of 304 stainless steel.

Theoretical conductivity limit, 25 ºC

Figure 4 Boiling water reactor mean reactor water conductivity at US boiling water reactor.

ECP of
230 mV or lower, using a standard hydrogen

electrode (SHE) This is the basis for the
230 mV

requirement used by US plants for IGSCC control

In BWRs, the radiation field in the core decomposes

water to hydrogen and oxygen species, most of which

immediately recombine back to water But some

remain as oxygen or hydrogen peroxide, because some

hydrogen is stripped into the steam phase before it can

recombine These same radiolysis reactions cause

hydrogen to react with oxygen or peroxide to reduce

ECP These reactions occur mainly in the downcomer,

and relatively low hydrogen concentrations are

effec-tive at lowering ECP in out-of-core regions of the

system More than half the BWRs in the United States

adopted low hydrogen injection rates of 0.2–0.5 ppm

(called HWC-L), which, coupled with the replacement

of recirculation piping using 316 stainless steel,

miti-gated IGSCC of recirculation piping

In the 1990s, concerns about the cracking of core

internals increased, but the low concentrations of

hydrogen used to protect out-of-core regions were

not sufficient to reduce ECP enough to mitigate

IGSCC of in-core materials, because of the radiolysis

of water occurring in the core As a result, it was

necessary to increase hydrogen concentrations to

1.6–2.0 ppm to lower the in-core ECP sufficiently to

provide protection in the reactor vessel (termed

HWC-M for moderate concentrations of hydrogen)

Although this approach was effective in protecting

core internals, it also increased radiation fields in the

steam side of the circuit, including the turbines, as a

result of carryover of nitrogen-16 under reducing

chemistry (Under the oxidizing conditions of NWC,

most of the N-16 remains in the water as soluble

species such as nitrate, and only a small percent istransported with the steam.) In some plants, localshielding of turbine components has reduced theimpact of the gamma radiation to acceptable levels,but the projected 4–6-fold increase did in fact curtailplans for increased hydrogen injection rates at manyplants Note that these N-16 radiation fields are aproblem only when the plant is at power, as rapiddecay occurs at shutdown because of the short half-life of N-16 (By contrast, out-of-core radiation fieldsfrom Cobalt-60 persist after shutdown and impact onmaintenance work during outages.)

NMCA was developed to increase the efficiency

of hydrogen in BWR cores, to avoid high N-16 fields

In this process, a nanolayer of platinumþ rhodium isdeposited on the wetted surfaces of the reactor Thesetreated surfaces catalyze the hydrogen redox reac-tion, converting oxygen back to water When theaddition of hydrogen to the feedwater raises themolar ratio of H2 to O2 to 2 or higher, the ECP ofthe treated surfaces drops to the hydrogen/oxygenredox potential, which is about
450 mV (SHE) Thiscan be achieved with hydrogen concentrations ofonly about 0.2 ppm, and under these conditions, themain steam radiation level is not increased to anunacceptable level The first plant used NMCA suc-cessfully in 1997, and over 25 plants have alreadyfollowed, with excellent results Field measurementsshow that NMCA has been effective in providingmitigation against IGSCC by lowering the ECPbelow the
230 mV (SHE) threshold with relativelylow hydrogen injection rates

The NMCA process is typically applied at ing outage, before new fuel is inserted into the core,

refuel-Time (h)

HWC ECP = −510 mV (SHE) ECP = +110 mV (SHE)NWC

and is effective for about three fuel cycles, before

reapplication is necessary The regions of the reactor

vessel internals that are protected by HWC-M or

NMCA are shown inFigure 7 While both

techni-ques offer significant areas of mitigation, there is an

additional benefit with NMCA on the upper, outershroud regions, as indicated by the additional shading

in the left-hand side of the figure5 It is estimated thatnoble metals protect slightly more of the outer coreregion than does moderate HWC (HWC-M), but thedifference is not significant

Figure 8 shows the dramatic benefit of noblemetals in reducing the rate of stub tube cracking atNine Mile Point 1 since the application in 2000.Before 2000, several stub tubes had to be repaired

or replaced at each outage, but since the application,only one tube leaked, and this was believed to havealready cracked before NMCA

Recently, attention has been focused on the onlineapplication of noble metals, with the first application

at the KKM plant in Switzerland By April 2008,there were four applications in the United States.This is discussed in a later section

5.02.2.3 Radiation Field ControlCorrosion products deposited on the fuel becomeactivated, are released back into the coolant, andmay be deposited on out-of-core surfaces Both solu-ble and insoluble species may be involved, the lattertending to deposit in stagnate areas (‘crud traps’) Thechemistry changes to control IGSCC resulted inincreased out-of-core radiation fields, and the imple-mentation by most plants of depleted zinc injection to

Noble metal applied mid cycle

control dose rates, as discussed later in this section.

During shutdowns, the major radiation source for

personnel exposure is activated corrosion products,

deposited on primary system surfaces Exposures are

generally accumulated at high-radiation field locations

where maintenance work is frequently needed

Although improvement of maintenance equipment

and procedures, reduction of maintenance

require-ments, increased hot-spot shielding, and control of

contamination dispersion have significantly reduced

total exposure, further reduction of radiation fields is

a major goal in programs for minimizing occupational

radiation exposure

The primary source of radiation field buildup on

out-of-core surfaces in BWRs is 60Co, which in

mature plants usually accounts for 80–90% of the

total dose 60Co has a relatively long half-life of

5.27 years The higher the soluble60Co concentration

in the coolant, the more 60Co is incorporated and

deposited on out-of-core systems and components,

resulting in higher dose rates on recirculation piping,

the reactor water cleanup system, dead legs, and

other crud traps in the system Other activated

tran-sition metals such as 54Mn, 58Co, 59Fe, and 65Zn

contribute the remainder of the dose.51Cr also

con-tributes significantly to the piping dose in some

NMCA plants The radiation fields commonly

measured in a BWR at the straight vertical section

of recirculation pipes are considered to be more

representative for the purposes of radiation buildup

trending and comparison with other plants These

measurements are done in a prescribed manner

developed under the EPRI BWR Radiation and

Con-trol program and are called BRAC point

measure-ments These measurements represent primarily the

incorporation of soluble60Co into the corrosion film

on the piping surfaces and tend to be a fairly good

predictor of drywell dose rates The deposition of

particulate oxides that contain 60Co and other

acti-vated species can also contribute significantly to

out-of-core radiation levels in BWRs, especially in hot

spots The particulate oxides, which vary in size,

originate primarily from corrosion of the

steam/con-densate system and are introduced via the feedwater

The sole precursor of the gamma-emitting 60Co

isotope is59Co.59Co is present as an impurity in the

nickel in structural alloys used in BWRs (e.g., Type

304 stainless steel) and is the main constituent of

wear-resistant alloys (e.g., Stellite), used as hard

fac-ing in valves and other applications requirfac-ing

out-standing wear resistance Corrosion and wear lead to

release of59Co into the coolant from these sources,

which is transported to the core and incorporatedinto the crud that deposits on the fuel rods The

59

Co is activated to 60Co by neutron activation,released back into the coolant, and incorporated as

a minor constituent into the passive films that form

on components that are inspected, repaired, andreplaced by maintenance personnel Components inthe neutron flux (e.g., the control blades) directlyrelease60Co Cobalt source removal is clearly impor-tant if radiation fields are to be minimized Anothergamma-emitting isotope,58Co, is formed by the acti-vation of nickel from stainless steel and nickel-basedalloys.58Co has a shorter half-life and is not as major

a contributor to radiation fields as60Co in BWRs, but

is much more significant in PWRs

Shutdown drywell dose rates increase when ant chemistry is changed for the first time fromoxidizing (NWC) to reducing (HWC) conditions.This results from a partial restructuring of the oxidesformed under the oxidizing conditions of NWC(Fe2O3 type) to a more reducing spinel type oxidecompound (Fe3O4type) The oxides affected are thefuel deposits, the corrosion films on stainless steelpiping, and out-of core deposits This results in anincrease in the chemical cobalt (and 60Co) concen-tration in the oxide because of the higher solid-statesolubility of transition metals in the spinel structure.The presence of a higher soluble reactor 60Co con-centration released from fuel crud while this conver-sion is occurring only aggravates the situation Theprocesses are depicted inFigure 9 The net result atmost plants is a temporary increase in reactor water

• Corrosion films

• Vessel crud

• Fuel crud

Small insoluble particles containing

60 Co, 54 Mn, etc.

Restructuring under HWC conditions

Fe2O3(containing 60 Co,

Soluble 60 Co, etc released during restructure

Figure 9 Boiling water reactor oxide behavior under reducing conditions.

As mentioned earlier, zinc addition reduces

radia-tion field buildup The mechanism of the zinc ion effect

is complex, as release of60Co from fuel crud is reduced,

and deposition out-core is also reduced Overall,

reac-tor water 60Co is decreased significantly after zinc

addition, as shown by plant data inFigure 10

Aqueous zinc ion promotes the formation of a

more protective spinel-structured corrosion film on

stainless steel, especially when reducing conditions

are present Second, both cobalt and zinc favor

tetra-hedral sites in the spinel structure, but the site

pref-erence energy favors zinc incorporation Thus, the

available sites have a higher probability of being filled

with a zinc ion than a cobalt ion (or60Co ion), and

hence the uptake of60Co into the film will be

signi-ficantly less if zinc ion is present in the water The

60

Co remains longer in the water and is eventually

removed by the cleanup system

The zinc was originally added to the feedwater as

ZnO, but it was quickly found that the65Zn that was

created by activation of the naturally occurring64Zn

isotope in natural zinc created problems With the

use of zinc oxide depleted in the64Zn isotope, called

depleted zinc oxide (DZO), this drawback was

elimi-nated Because of the high cost of DZO, feedwater

zinc injection was not implemented widely until

HWC shutdown dose issues emerged

For the case of plants treated with NMCA and

injecting hydrogen, the oxidant concentration on the

surface of the stainless steel is zero (due to the Pt

and Rh catalyzing the reaction of any oxidant withthe surplus hydrogen) The net result is that the ECP

is at or very near the hydrogen redox potential,typically about –490 mV (SHE) for neutral BWRwater This low potential causes a much more thor-ough restructuring of the oxides to the spinel statethan observed under moderate hydrogen waterchemistry (HWC-M)

Feedwater iron ingress has a significant influence

on the effectiveness of zinc injection As discussed inthe next section, deposits on fuel cladding surfaces(called ‘CRUD’) are mainly composed of iron oxides,with other constituents Therefore, reducing ironingress from the feedwater has the benefit of mini-mizing crud buildup, which is important for fuelreliability (next section) For these reasons, extensiveefforts have been made to reduce iron ingress, withsignificant success Furthermore, fuel crud has a largecapacity for incorporating zinc and is in fact wheremost of the zinc ends up The lower the amount ofcrud on the fuel, the greater the proportion of zincthat remains in solution and can subsequently beincorporated in out-of-core surfaces Therefore, atplants with low feedwater iron, less zinc is captured

by the crud on the fuel, so a relatively greater amountremains in solution and is available to control out-ofcore radiation fields This is very important, as zincinjection rates are limited by fuel performance con-cerns, and hence lowering feedwater iron is essentialfor maintaining lower radiation fields

0 0.2

Figure 10 Hydrogen water chemistry plant RxW60Co response to zinc addition.

5.02.2.4 Fuel Performance Issues

Fuel durability has improved over the years, and

failures have declined, helped by improvements in

water purity In operation, zircaloy fuel cladding

develops a thin oxide layer (ZrO2), which typically

does not adversely affect performance However, an

increase of deposition of corrosion product deposits

(‘crud’) on this oxide film is undesirable because it

can reduce heat transfer and increase fuel pin

tem-peratures, with resultant increased corrosion of the

fuel cladding, ultimately increasing the risk of fuel

failure Moreover, the addition of additives to control

corrosion may increase the risk of crud buildup on

the fuel For example, zinc and noble metals in BWRs

tend to increase the adherence of crud deposits on

the fuel, which can result in undesirable oxide

spal-ling in higher-rated cores In fact, corrosion-related

fuel failures occurred at four plants in the United

States between 1999 and 2003 Although the precise

root cause of fuel failures is often difficult to

deter-mine, it is clear that excessive crud buildup played a

role in these failures With progressive uprating of

fuel duty in both PWRs (and BWRs), the margin to

tolerate crud has been reduced and additional care

has to be taken in specifying the water chemistry to

avoid undesirable fuel performance issues Despite

these more demanding conditions, fuel failures have

decreased in recent years

Concern about the possibility of adverse effects of

NMCA on fuel has prompted imposition of a strict

limit on the amount of noble metal that can end up on

the fuel and guidance on the injection of zinc Plant

data indicate that spalling of the corrosion layer from

fuel cladding, which is often regarded as a precursor

to cladding failure, is prevented if the cycle averagefeedwater zinc is maintained below 0.4 ppb in NMCAplants (0.6 ppb for non-NMCA plants) More recentdata indicate that quarterly averages may be as high

as 0.5 ppb for NMCA plants, without occurrence ofspalling.5

These feedwater zinc data are the basis for limits

in the water chemistry guidelines The 2008 try guidelines7 retain the cycle average feedwaterzinc limit of 0.4 ppb (0.6 ppb for non-NMCA plants)but enable a slight increase in the quarterly average

chemis-to 0.5 ppb, which may allow flexibility in controllingradiation buildup in parts of the cycle

The tighter control of water chemistry in recentyears has been successful in controlling crud forma-tion on fuel cladding, andFigure 118shows failuresfrom pellet–clad interaction causing SCC, fabricationdefects, debris, and crud/corrosion Note that therehave been zero crud/cladding related fuel failures in

US BWRs since 2004 (although assessment of 2007failures is not yet complete, crud/corrosion is notbelieved to be a factor here)

Analysis of recent plant data confirms that control

of feedwater iron ingress has the positive benefit ofreducing the amount of crud on the fuel Control ofcopper, which generally originates from admiraltybrass alloys, is also beneficial; not only can copperhave detrimental effects on the fuel, but it also limitsthe ability of hydrogen to reduce the ECP, and it alsoleads to higher radiation fields As a result, most USplants have replaced condensers containing brasstubing

0 5 10 15 20 25 30

EOC year

PCI-SCC Unknown Fabrication Debris Crud/corrosion

Figure 11 US boiling water reactor fuel failures by mechanism for each end-of-cycle (EOC) year.

5.02.2.5 Online Addition of Noble Metals

As discussed earlier, the classic NMCA process is

generally applied during refueling outages before

the new fuel is loaded into the core Reapplication

after about three cycles of operation takes

approxi-mately 2 days, while the plant maintains 107–154C

as it enters the refueling outage To reduce this outage

time, GE-Hitachi developed OLNC, first

demon-strated at KKM (a GE design of plant in Switzerland)

in 2005, with several more additions subsequently

Preliminary results indicate that there have been no

unexpected chemistry effects during the first OLNC

applications, and shutdown radiation fields actually

decreased at KKM after OLNC.5Subsequently, CGR

of susceptible welds decreased significantly, as shown

by the decrease in slopes inFigure 12 after OLNC

initiation for two welds that have been monitored for

several years

The effects of OLNC on fuel have been

exten-sively studied in fuel removed from KKM, and no

adverse effects have been observed The jury is still

out on this concern, but the general assessment is that

OLNC will have no more impact than the classic

application, and may well prove to be of less concern

More IGSCC and fuel measurements are planned,

but with no issues emerging to date, it appears that

OLNC applications about every 12 months would be

effective and economical, avoiding the critical path

time necessitated for the classic NMCA application

during refueling outages Initial OLNC applications

have been carried out at plants that had previously

applied noble metals in the classic off-power manner

However, the first OLNC application at a plant thathas not used noble meals previously occurred in late

2008, but no results are available

Chemistry Control5.02.3.1 Evolution of PWR PrimaryChemistry Strategies

In the very early days of PWR operation, heavy crudbuildup on fuel cladding surfaces was caused by thetransport of corrosion products from the SGs into thereactor core As a result, activated corrosion productscaused high-radiation fields on out-of-core surfaces(Figure 13), fuel performance was compromised, andeven coolant flow issues were observed in someplants

These problems were initially mitigated by imposing

a hydrogen overpressure on the primary system, toreduce the ECP, and raising the primary chemistry

pH Materials degradation in primary systems wasthen not a major concern, with most of the emphasisfocused on secondary side corrosion issues in theSGs Commercial PWR power plants use a steadilydecreasing concentration of boric acid as a chemicalshim (for reactor control) throughout the fuel cycle,which results in the use of lithium hydroxide tocontrol pH Some 30 years ago, the concept of

‘coordinated boron and lithium’ was developed,whereby the concentration of LiOH was graduallyreduced in line with the boric acid reduction tomaintain a constant pH

1997 1998

0 50

98 g OLNC

198 g OLNC

199 g

Not inspected in 00, 01,04

Indications 9,10 may be seeing mitigation by OLNC

Corrosion products released from the steam

gen-erator tubes are transported, dissolved, or deposited

by the coolant on the basis of solubility differences

The solubility of nickel and iron depend on pH,

temperature, and redox potential, all of which vary

with location around the nonisothermal system

Originally, a constant at-temperature pH of 6.9 was

recommended, based on the minimum temperature

coefficient of solubility of magnetite In fact, it was

determined that heavy fuel crud buildup was avoided

if a constant pH of at least 6.9 was maintained This

was possible with 12-month fuel cycles, but fuel

clad-ding corrosion concerns limited the maximum LiOH

concentration to 2.2 ppm Consequently, plants often

started the fuel cycle with pH below 6.9, which

resulted in deposition of corrosion products on the

fuel, activation of cobalt and nickel, and subsequent

transport to out-of-core surfaces, resulting in

radia-tion fields remaining relatively high

Even though detailed studies of fuel crud showed

that the prime constituent of the crud was nickel

ferrite (for which the optimum pH is 7.4), this

coor-dinated chemistry had remained the standard for

many years, until higher pHs became the norm in

the 1990s Although research and plant

demonstra-tions showed that the 2.2 ppm limit was excessively

conservative, the move to higher Li concentrations

has been slow However, detailed fuel examinations

from a recent plant demonstration (that will be

dis-cussed later) have indicated that Li can be raised to as

high as 6 ppm

About 25 years ago, primary water stress sion cracking (PWSCC) of Alloy 600 SG tubes wasobserved in a few plants, leading to studies on miti-gating this effect Following successful demonstra-tion of zinc injection in BWRs, initial field tests atPWRs showed that radiation fields were reduced,and laboratory studies indicating that PWSCC wasreduced were eventually confirmed As a result, zincinjection is being implemented at an increasing rate,although concerns about fuel performance at high-duty plants have not been completely resolved Mostrecently, buildup of boron-containing crud in areas

corro-of subcooled nucleate boiling leading to localizedflux depression has encouraged the use of higher Liconcentrations to minimize corrosion product trans-port Concerns about the potential adverse effects ofzinc deposited in high-crud regions have resulted inseveral highly rated plants applying in situ ultrasonicfuel cleaning before implementing zinc injection.Although zinc injection was developed for radia-tion field control, laboratory studies showed that italso inhibited SCC under PWR conditions Theidentification of PWSCC in reactor vessel penetra-tions in the last 15–20 years has encouraged the use

of zinc injection, but has also focused attention onthe effects of dissolved hydrogen, for which therecommended range has remained 25–50 ml kg
1for 30 years It now appears that raising hydrogenwill reduce PWSCC rates, while lowering it maydelay initiation of PWSCC The interactions ofmaterials, radiation fields, and fuels in PWR primary

Corrosion products released from SG tubing

Primary loop Figure 13 Transport and activation of corrosion products in pressurized water reactor primary systems.

chemistry and optimization issues covered in the

water chemistry guidelines, which are discussed

later, are depicted inFigure 14

The evolution of water chemistry control in PWR

primary systems in the United States over the last

30 years is shown inFigure 15

The following sections address the three main

factors – pH control, zinc injection, and dissolved

hydrogen control – that have dominated PWR

pri-mary chemistry strategies in the past and continue to

do so today.9Each of these factors is considered from

the viewpoint of materials degradation, radiation

field control, and fuel performance concerns

5.02.3.2 Materials DegradationMaterials degradation has been covered in detail inChapter 5.04, Corrosion and Stress CorrosionCracking of Ni-Base AlloysandChapter5.05, Cor-rosion and Stress Corrosion Cracking of AusteniticStainless Steels, and here only the specific effects ofwater chemistry variables on materials in PWR pri-mary systems will be reviewed, particularly those thatmay affect the chemistry of optimization process.Recent papers by Andresen et al.10,11 providedetailed results of a comprehensive study of the effects

of PWR primary water chemistry on PWSCC ofnickel-based alloys Extensive studies have been car-ried out to determine the effect of lithium, boron, and

pH on PWSCC, and the generally held conclusion isthat any effects are minimal, especially compared tomaterial susceptibility, stress state and temperature,and other operational issues Crack initiation testsusing the most reliable types of reverse U-bend speci-mens indicate that pH has a relatively small effect oncrack initiation (generally less than a factor of 2).Although the most rapid crack initiation occurred at

pH310C7.25, with slower rates at higher or lower pHs,CGR tests generally confirm that pH has minimaleffect

The effect of lithium is even smaller than the pHeffect, and the influence of boron is minor or nonexis-tent Andresen et al concluded that the effects of rele-vant variations in PWR primary water chemistry (B, Li,

Elevated constant pH (7.3/7.4) Ultrasonic fuel cleaning Elevated constant pH (7.1/7.2) Zinc injection Modified elevated lithium program

Elevated lithium program

PWSCC:

pH (Li, B) minimal effect

Zn beneficial dissolved H2 effect

Radiation fields:

pH (Li, B), Zn beneficial

PWR chemistry control Plant

dose rates

Materials degradation

Fuel performance

Plant operations

Dissolved H2control range

Figure 14 Pressurized water reactor primary chemistry optimization Reproduced from Fruzzetti, K.; Perkins, D PWR chemistry: EPRI perspective on technical issues and industry research In VGB NPC’08 Water Chemistry Conference, Berlin, Sept 14–18, 2008.

and pH) have little effect on the SCC growth rate in

Alloy 600, and thus provide little opportunity for

miti-gation of PWSCC Plant data have found no adverse

effects from increasing lithium and pH in primary

systems As a result, it is considered that adjusting

pH, lithium, or boron to minimize crack initiation

may be of minimal value The 2007 edition of the

PWR Primary Water Chemistry Guidelines12reviewed

the most recent data and concluded that pH strategy

changes based on PWSCC considerations are not

war-ranted This means that plants have the flexibility to

pursue B/Li/pHt chemistry adjustments to minimize

crud transport and radiation buildup without concern

for negative effects on PWSCC susceptibility of

nickel-based alloys, although of course chloride and sulfate

impurities should continue to be minimized

Following good experience in BWRs, zinc

injec-tion has been implemented in the primary systems of

PWRs, both to reduce primary side cracking of

nickel-based alloys and to control dose rates The qualification

work for BWRs showed that zinc inhibited SCC, but

the benefit was not sufficient to avoid the need for

hydrogen water chemistry to mitigate IGSCC Thus,

the motivation for BWR zinc injection was exclusively

radiation field control

The situation in PWRs is different, as laboratory

work13showed that initiation of PWSCC was

signifi-cantly delayed by zinc injection, and hence the

moti-vation for the initial applications of zinc in most US

PWRs at the 10–30 ppb level was to control PWSCC

of SG tubing Additionally, German-designed PWRs

and a few US plants used5 ppb depleted zinc for

to give ZnCr2O4, which is very stable The benefits

of PWR zinc injection have been clearly strated in reducing PWSCC degradation (especiallygrowth rate) of Alloy 600 SG tubes, and in controllingradiation fields Evaluation of currently availablelaboratory data2 indicates that PWSCC initiationwill be reduced, and PWSCC CGR may be reduced

demon-in thicker cross-section components, dependdemon-ing uponother factors such as the stress intensity factor of thespecimen Andresen et al.11 concluded that crackgrowth mitigation by adding Zn requires furtherstudy, although two of four tests show a decrease

in growth rate of >3 Molander et al.14

also foundthat the effect of zinc on CGR was minor Hence,more work is needed before making definitive con-clusions from laboratory studies regarding the benefit

The largest effect of zinc appears to be on tion of cracking, with a smaller effect on CGR, withthe data indicating a factor of 2–10 reduction forinitiation and about a factor of 1.5 reduction in CGR,consistent with the extensive laboratory work,11

initia-Application of zinc in world PWRs

0

1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006 2007 5

10 15 20 25 30 35 40 45 50

Year

Percent of PWRs injecting Number of units injecting

Figure 16 Application of zinc injection in pressurized water reactors worldwide.

indicating that zinc inhibits mainly by delaying the

initiation of PWSCC However, the SG NDE data

also showed that zinc reduced the rate of crack

prop-agation (depth) by 17–60%

These results are consistent with initial laboratory

test data indicating that zinc reduced crack

propaga-tion by a factor of approximately 3 at low stress

intensities, but had no effect at higher stress

intensi-ties In addition, the lack of cracking in the Farley

PWR pressure vessel head penetrations (exposed to

zinc for over 12 years), compared to PWSCC

indica-tions in similar pressure vessel heads in other plants,

suggests that zinc addition is beneficial for Alloy 600

(and possibly Alloy 82/182) thick-section

compo-nents under PWR primary service conditions

Recent work has studied the influence of dissolved

hydrogen on PWSCC In the early days of PWR

operation, the lower limit on hydrogen was set at

25 ml kg
1, to provide adequate margin against

radi-olysis and heavy crud formation Plant tests in France

showed that this limit was excessively conservative

and that less than 10 ml kg
1 would be satisfactory,

provided good control of oxygen was maintained in

makeup water Several workers have found that the

maximum in PWSCC CGR occurs close to the ECP

corresponding to the Ni/NiO thermodynamic

equi-librium condition.15 Although this potential is

unaf-fected by lithium/boron/pH (consistent with the fact

that these do not greatly influence PWSCC over the

range of practical relevance), the equilibrium

poten-tial is significantly affected by the dissolved hydrogen

concentration Andresen et al found that the peak inSCC growth rate versus H2fugacity was temperaturedependent, but generally fell within the hydrogenconcentration range used in PWRs This provides

an opportunity for mitigation, by perhaps a factor of

2 in Alloy 600 and a factor of 5 in Alloys 182, 82,and X750, as the median value of the dissolvedhydrogen concentration for US plants is approxi-mately 35 ml kg
1

US PWRs currently operate within dissolvedhydrogen within the recommended 25–50 ml kg
1range, with the majority in the 30–40 ml kg
1range,but none with more than 44 ml kg
1 (Figure 18).The lower limit is set conservatively to provide anoperating margin over the level of hydrogen required

0 20 40 60 80 100 120 140

Unit 1 Unit 2

Refueling outage

X indicates last refueling outage

before start of zinc injection

Figure 17 Effect of zinc on steam generator tube degradation at a US pressurized water reactor.

0 10 20 30 40 50 60 70