A B i f O i f FACA Brief Overview of FAC Investigations, Experiences and g , p

Lessons Learned

Douglas MunsonWANO SeminarWANO Seminar

Effective Monitoring and Control of FACHaiyan, Zhejiang, China

January 15-17, 2008

ENT000035 Submitted: March 28, 2012

Historical PerspectiveHistorical Perspective

� Flow-accelerated corrosion (FAC) is not a new phenomenon

� Historically it was called erosion corrosion� Historically it was called erosion-corrosion� As used historically, erosion-corrosion

described two different mechanisms� A chemical dissolution of the protective oxide layer in

a moving stream of water or wet steam� A mechanical wearing away of the protective oxide� A mechanical wearing away of the protective oxide

layer

� E-C is identified as one of the 8 forms of corrosion in Fontana & Greene’s classical textcorrosion in Fontana & Greene s classical text book “Corrosion Engineering” published in 1967

Erosion-Corrosion and Flow-Accelerated Corrosion

Film Free

Breakawayte

Film Breakdown

Velocity

rosi

on R

a

Erosion-CorrosionCor

r

FAC

VelocityBreakaway velocity for carbon steel straight pipe in water ~10 m/sec, wet stream ~30 m/sec

1940s1940s

� Most early researchers were concerned about turbines, heat exchanger tubes, and feedwater i (FAC i th i f f d t i )iron (FAC is the primary source of feedwater iron)

� Failures of power plant piping attributed to FAC were identified in the 1940swere identified in the 1940s� Unfortunately they were not well documented� Fossil and industrial plants typically replaced

the piping and restarted

1940s (continued)

� In the late 1940s, Mars Fontana at Ohio State ,University (and others) began to investigate E-C. Studies were conducted on many different alloys with different fluidswith different fluids

� Studies included carbon steel and distilled water. Found:� Rate f(velocity)� Rate f(pH); rate was negligible for pH > 10.0� Rate f(temperature)� Rate f(alloy and fluid)

1950s

D i f i l l l t i th id� Design of commercial nuclear power plants in the mid 1950s necessitated a higher level of safety and design

� Shippingport (PWR) went operational 12/1957

� Dresden 1 (BWR) went operational 7/1960

� In the late 1950s, the US Atomic Energy Commission sponsored a series of experiments on PWR type reactorssponsored a series of experiments on PWR-type reactors

� Concluded carbon steel could be used in the primary loop if oxygen was low and pH > 10.5 to 11.5

� Also in the late 1950s, Oak Ridge National Laboratory tested carbon steel at 250°C, a water velocity of 8 m/s and varying oxygen, and noted variations in oxide film and corrosion rates as a function of oxygen

1960s - GE1960s GE

� In the late 1950s, GE began a program to study corrosion in� In the late 1950s, GE began a program to study corrosion in steam, wet steam, saturated water and subcooled water conditions

� Included was the release of corrosion products in typical BWR� Included was the release of corrosion products in typical BWR feedwater conditions (then O2 < 15 ppb)

� Prior data was taken at ~25°C and 288°C, with little to no data in betweenin between

� One study used a side stream test loop at the Humboldt Bay Nuclear Power Plant located in California (operational 8/1963)

� Quantified the dependence of O on FAC rates� Quantified the dependence of O2 on FAC rates� Also concluded that:

� Carbon steel can be used only if O2 > 15 ppbS i l b d f ll l l f O� Stainless can be used for all levels of O2

� Lead investigators were Brush and Pearl

1974Keller Develops the First Predictive ModelKeller Develops the First Predictive Model

� In 1974, H. Keller of Siemens/KWU developed the first predictive model� Valid for 2-phase lines with a steam quality of 70-100%� Valid for 2-phase lines with a steam quality of 70-100%

FAC Rate = [f(T) · f(X) · V · Kc] – Ks (mm/104 hours)

Wheref(T) = dimensionless coefficientf(X) = dimensionless coefficient related to steam wetnessV = fluid velocity (m/s)Kc = factor to account for local geometryKs = a constant that must be exceeded before FAC is observed

1974Keller Develops the First Predictive Model

( ti d)(continued)

1974Keller Develops the First Predictive Model

(continued)(continued)

Keller’s GeometryKeller s Geometry Factors

1970s – CEGB (UK)( )

� In the late 1970s and early 1980s G Bignold I� In the late 1970s and early 1980s, G. Bignold, I. Woolsey and others at the Central Electricity Generating Board (UK) were performing systematic studies of FAC� Quantified wall loss as f(temperature, pH,

velocity alloy composition)velocity, alloy composition)� CEGB researchers also applied thin layer surface

activation to laboratory experiments to accurately measure wall loss rates in real time

� Developed a predictive modelFAC 4k3 (Co )3 (H+)2/B2FAC = 4k3 · (Co

eq)3 · (H+)2/B2

1970s – EDF (France)( )

� Starting in the early 1970s, P. Berge, M. g y , g ,Bouchacourt, F. Remy and others at Electricite de France were also performing systematic studies of FAC

� Included laboratory tests to investigate oxide layer, influence of steam qualify, surface roughness, mass transfer alloy composition pH amine etctransfer, alloy composition, pH, amine, etc

� EDF also started the development of a mechanistic model to predict the rate of FAC

FAC = (Ceq - C�)(1/2K + 1/k)

1982Ducreux Material InvestigationsDucreux Material Investigations

� In 1982, J. Ducreux of EDF (France) published laboratory data� In 1982, J. Ducreux of EDF (France) published laboratory data on the effect of alloy composition on FAC rates. His model:

FAC Rate/FAC Ratemax = 1/(83 · Cr%0.89 · Cu%0.25 · Mo%0.20)

� The Ducreux relationship is the basis for most of the predictive technology in use today

Early 1980s – Siemens/KWU (Germany)y ( y)

� Starting in the late 1970s, H. Heitmann, W. Kastner and others at KWU (now Siemens) were also ( )performing systematic studies of FAC� Included were rate studies versus pH

� In the mid 1980s, the Keller model was further developed

FAC = Kc · F1(V, T, h) · F2(pH) · F3 (O2) · F4(t) · F5(x)

1984Huijbregts Materials InvestigationsHuijbregts Materials Investigations

� In 1984, W. M. M. Huijbreghts of KEMA (Netherlands) published laboratory data on the effectlaboratory data on the effect of alloy composition on FAC rates. His model:

� FAC Rate/FAC Ratemax = 1/(0.61 + 2.43Cr(%) +

1.64Cu(%) + 0.3Mo(%))

June 1978Oyster Creek (US)Oyster Creek (US)

� General Electric BWR� General Electric BWR� Failure occurred in a 200 x 350 mm reducer

downstream of a feedwater pump � Failure was attributed to cavitation

� Significant and reoccurring damage was found in several feedwater control valvesseveral feedwater control valves� Damage was attributed to flashing

� Significant wall thinning was found in several piping a eas mostl do nst eam of cont olpiping areas, mostly downstream of control valves

� Event and findings had little effect on nuclear industry

November 1982 Navajo Fossil Plant (US) (continued)Navajo Fossil Plant (US) (continued)

� Feedwater just downstream of feed pump� Original thickness = 9.3 mm� Thickness @ failure = 0.7 mm

� There was a backing ring at the upstream weldL H i h d i t h i t� Low pH ammonia + hydrazine water chemistry

� Unusual because it was a fossil failure that was publicized and analyzedpublicized and analyzed

� Little change in the way fossil and nuclear plants approached FAC.

November 1982 Navajo Fossil Plant (US)Navajo Fossil Plant (US)

Note direction of flow

December 1986Surry Unit 2 (US)Surry Unit 2 (US)

Flow

Condensate system just before the feed pump

December 1986Surry Unit 2 (US) ( i d)Surry Unit 2 (US) (continued)

� Westinghouse PWR� Westinghouse PWR� Four workmen were killed. Four others were injured� Ammonia water chemistry with a low pH y p� Many replacements (~190) were made in both units � Similar conditions as Navajo� This failure showed:

� Seriousness of FACS ibili f i l h C� Susceptibility of single-phase systems to FAC

� Need for an inspection program� Maximum damage may not be at extrados of� Maximum damage may not be at extrados of

elbows

1987US Industry Responds to SurryUS Industry Responds to Surry

� NUMARC assembled a working group and issued guidelines for utilities to implement an inspection program for single-phase systems

NRC d INPO b i t t d i th i� NRC and INPO became interested in the issue

� US nuclear plants start to implement inspection programs of single phase pipingprograms of single-phase piping

1987 CurrentIndustry Improves its Predictive

TechnologyTechnology

� EPRI develops a predictive model and software for� EPRI develops a predictive model and software for utility use� Best estimate model developed by Bindi Chexal and Jeff

Horowitz using a regression analysis of laboratory data� Released CHEC in 1987: 1-phase lines only� Released CHECMATE in 1989: 1 and 2-phase lines� Released CHECMATE in 1989: 1 and 2-phase lines� Released CHECWORKS in 1993 (current version is 2.2)

� Component-by-component predictions of rate of wall thinning, total wall loss to date, remaining service life

� Water chemistry and network flow analysis� Storage and evaluation of component inspection datag� Management of related outage activities

1987 - CurrentIndustry Improves its Predictive Technology

(continued)(continued)

� EDF develops the BRT-CICERO software based on� EDF develops the BRT-CICERO software based on the Bignold/Berge/Bouchacourt model:

FAC = f(Cr) · f(�) · (Ceq - C�)[0.5 · (1/k + �/D)]

� Results include:� Wear and wear rate� Residual thickness � Range of validity of thickness taking uncertainties

into accountinto account

1987 - CurrentIndustry Improves its Predictive Technology

( i d)(continued)

� Siemens/KWU develops the WATHEC and DASY programs:

FAC = Kc · F1(V, T, h) · F2(pH) · F3 (O2) · F4(t) · F5(x)

� Results include:� Wall thinning and remaining life� Designed to provide conservative predictions of� Designed to provide conservative predictions of

maximum probable thinning� Current version is called COMSYS

In l des othe me hanisms s h as st ain ind ed� Includes other mechanisms such as strain-induced cracking, material fatigue, cavitation, droplet impingement

For More Information

� Details of the various models, a theoretical treatment and laboratory data are provided in y pEPRI report TR-106611-R1. Primary authors:� Bindi Chexal and Jeff Horowitz – EPRI� Michel Bouchacourt and Francois Remy – EDF� Wolfgang Kastner – KWU/Siemens

April 1989Arkansas Nuclear One (US)Arkansas Nuclear One (US)

April 1989k l ( )Arkansas Nuclear One (US) (continued)

� Combustion Engineering PWR� Two-phase conditions

L ti d t f hi h� Location was downstream of high-pressure extraction nozzle

� This failure showed:� This failure showed:� Aggressive nature of FAC in 2-phase lines and

need to include them in the inspection program

May 1989US NRC Issues Generic Letter 89-08US NRC Issues Generic Letter 89-08

� Required the US utilities to:� Implement a long-term FAC monitoring

program� Include all susceptible high-energy carbon

steel piping systemssteel piping systems� Include both single- and two-phase lines� Utilize the NUMARC/EPRI or equally effective Ut e t e U C/ o equa y e ect e

method

December 1989 Santa Maria de Garona (Spain)Santa Maria de Garona (Spain)

� GE BWR� A small piece of the feedwater line was blown p

out� Failure was just downstream of a flowmeter.� Line operated with very low oxygen (~6 ppb)� Failure demonstrated:

� Need for FAC program for BWRs� Need for FAC program for BWRs� Dangers of operating with low oxygen in

neutral water

July 1989EPRI forms CHUGEPRI forms CHUG

Then (1989) Now (2008)� Purpose was to support the

CHEC and CHECMATE computer codes

� FAC only

� Issue group to deal with degradation in FAC susceptible systems

� FAC erosion weld degradation� FAC only� 2 meetings/year� Training� 10 members representing ~ 30

� FAC, erosion, weld degradation� 2 meetings/year� Web site� Technical investigationsp g

nuclear plants (US only)g

� Training� 46 members representing >

160 nuclear plantsl•Belgium

•Canada (all)•Czech Republic•France

( CO

•Romania•Slovakia•South Korea•Spain (all)

i•Japan (TEPCO only)•Mexico

•Taiwan•US (all)

May 1990 Loviisa Unit 1 (Finland)Loviisa Unit 1 (Finland)

Note orificeNote orifice.

Downstream Pipe

Upstream Flange

May 1990 Loviisa Unit 1 (Finland) (continued)Loviisa Unit 1 (Finland) (continued)

� Russian built PWR� Russian built PWR� Failure just downstream of orifice in feedwater line� Water chemistry was neutral water with low O2

� Failure through orifice flange. There was little wear in pup piece and downstream pipe

� 11 of 12 sister locations were < minimum thickness.� 11 of 12 sister locations were < minimum thickness.� This failure showed:

� The significance of Cr (there was > 0.1% Cr in d t i d 0 C i th fl )downstream pipe and ~0 Cr in the flange)

� Importance of high oxygen if using neutral water� High risk at orificesHigh risk at orifices� All types of nuclear plant designs are at risk

� Note: Unit 2 had a similar failure in 1993

December 1990 Millstone 3 (US)Millstone 3 (US)

December 1990 Millstone 3 (US) (continued)Millstone 3 (US) (continued)

� Westinghouse PWR� Westinghouse PWR� Simultaneous failures of two (of four) parallel lines

downstream of level control valves and downstream of moisture separator drain tankmoisture separator drain tank

� The lines were omitted from the CHEC© analysis.� This failure showed:

� Need for a comprehensive susceptibility analysis� High risk downstream of control valves

C t d l t i l d ll tibl t� Computer models must include all susceptible systems� Importance of good communications between central

engineering and the plantg g p� Parallel lines may wear differently

November 1991 Millstone 2 (US)Millstone 2 (US)

November 1991 Millstone 2 (US) ( d)Millstone 2 (US) (continued)

� Combustion Engineering PWR� Combustion Engineering PWR� Failure downstream of level control valve in the

reheater drain line.� Location had not been previously inspected.� In both of the Millstone accidents, personnel were

i th i i it f th b k l ti h tl b fin the vicinity of the break locations shortly before the components failed, but were not injured

� This failure showed:� A large effort was needed to improve the FAC

program at all of the owners units. This required a significant restart effortrequired a significant restart effort.

March 1993 Sequoyah Unit 2 (US)Sequoyah Unit 2 (US)

� Westinghouse PWR� Westinghouse PWR� Failure of a 275 mm OD pipe downstream of a tee

in a high-pressure extraction line � 150 x 75 mm “fish-mouth” failure

� Post-accident investigation indicated numerous programmatic deficienciesprogrammatic deficiencies

� Lengthy shutdown for both Sequoyah units was required

� This failure showed:� Need for personnel training � Need to identify a program owner� Need to identify a program owner� Dangers of excessive personnel turnover

November 1993EPRI Issues NSAC-202LEPRI Issues NSAC-202L

� In response to continuing leaks and failures, in 1992 EPRI began a series of plant visits to understand how FAC knowledge and technology were being implementedg gy g p

� Visits found a wide range of implementation details

� In 1993 EPRI and CHUG developed NSAC-202L “R d ti f Eff ti Fl A l t d“Recommendations for an Effective Flow-Accelerated Corrosion Program”

� Has been accepted by INPO, the US NRC, and regulators p y , , gin many other countries as the standard for FAC control

� Considered a living document, the latest version is revision 3 issued in 2006 (EPRI report 1011838)revision 3 issued in 2006 (EPRI report 1011838)

Content of NSAC-202L-R3Content of NSAC 202L R3

� Overview of an effective program� Procedures and documentation� Recommendations for FAC tasks

� Performing a susceptibility analysis� Performing a susceptibility analysis� Performing a FAC analysis� Selection of inspection locations

P f i i ti� Performing inspections� Evaluating inspection data� Evaluating worn components� Replacements and repairs

� Developing a long-term strategy� Recommended program for small-bore piping� Recommended program for small bore piping� Recommended program for vessels and equipment

November 1994 S h U it 1 (US)Sequoyah Unit 1 (US)

� Westinghouse PWR� Crack caused a leak in the condensate system� Crack caused a leak in the condensate system.� Flow straightener used during construction was

inadvertently left in place despite drawingsinadvertently left in place despite drawings indicating that it was removed.

� This failure showed:� Importance of knowing as-built condition of

the plant and inspecting “new” locations

February 1995 Pleasant Prairie Fossil Plant (US)Pleasant Prairie Fossil Plant (US)

February 1995 Pleasant Prairie Fossil Plant (US) (continued)( ) ( )

� Catastrophic failure of a straight, seamless pipe� Catastrophic failure of a straight, seamless pipe downstream of a tee in feedwater system

� Two plant employees were killed� Low pH ammonia and hydrazine water chemistry� The pipe had a measured Cr of 0.03% and the tee

had a measured Cr of 0 12%had a measured Cr of 0.12%.� This failure showed:

� Importance of chromium� Importance of chromium� Need for fossil plants to implement a FAC

inspection program� Importance of water chemistry for fossil plants

August 1995Millstone 2 (US)Millstone 2 (US)

Gate Valve

August 1995Mill t 2 (US)Millstone 2 (US) (continued)

� Failure downstream of gate valve in heater drain tank bypass line

� Post accident analysis indicated that water� Post accident analysis indicated that water hammer caused this rupture although the pipe was thinned by FAC. � Failure occurred even though the pipe was

above minimum wall thickness� This failure showed:� This failure showed:

� The importance of knowing operating history. The valve had apparently been used to throttle pp ythe flow

April 1997 Fort Calhoun (US)Fort Calhoun (US)

April 1997 Fort Calhoun (US) ( i d)Fort Calhoun (US) (continued)

� Combustion Engineering PWR� Combustion Engineering PWR� A 5 diameter sweep in a high pressure extraction line

failed catastrophically.� Another elbow located downstream was very thin (~

0.5 mm).� The plant had previously replaced the upstream� The plant had previously replaced the upstream

component, and inspected it instead of the sweep.� This failure showed:

� Importance of knowing replacement history� Need to fully implement CHECWORKS and NSAC-

202L (plant only had partial models, partial202L (plant only had partial models, partial implementation)

May 1999 Point Beach Unit 1 (US)Point Beach Unit 1 (US)

#2 heater , operating temperature = 175°C, steam quality = 88%

May 1999 Point Beach Unit 1 (US) ( d)Point Beach Unit 1 (US) (continued)

� Westinghouse PWR

� Nominal wall thickness = 13 mm

� Fishmouth type failure with opening size of 685 x 22� Fishmouth type failure with opening size of 685 x 22 mm. Degradation extended 1219 mm

� Failure location was where steam entering the gfeedwater heater hit the impingement plate, and deflected to the shell

Simila deg adation in pa allel t ain� Similar degradation in parallel train

� This failure showed:

� Need for the inspection program to include vessels� Need for the inspection program to include vessels

August 1999 Callaway (US)Callaway (US)

B li it d t lBeam limited travel

August 1999 Callaway (US) (continued)y ( ) ( )

� Westinghouse PWR� Westinghouse PWR� Failure was on first stage reheater drain line (170 mm

diameter) just downstream of a very long horizontal run

� A 380 x 530 mm section of pipe was flattened and ejectedejected

� Operating conditions uncertain and unusual:� Quality believed to be about 4.5% at ~ 215°C � This was because the level control valve was

located near the upstream end of the line. Usually, such valves are located near the downstream endsuch valves are located near the downstream end

August 1999 C ll (US) ( d)Callaway (US) (continued)

� The void fraction was estimated to be about 55% Most� The void fraction was estimated to be about 55%. Most two-phase lines in nuclear plants either have:� Void fractions of very near one (e.g., extraction lines),

oror � Void fractions near zero (e.g., cascading drains)

� A backing ring was in the line and contributed to the wear� The thinning did not affect all sister locations� This failure showed:

� Importance of extra inspections if uncertain or unusualImportance of extra inspections if uncertain or unusual operating conditions

� Importance of extra inspections in lines with backing ringsrings

� Importance of inspecting parallel trains

March 2000 Susquehanna Unit 1 (US)Susquehanna Unit 1 (US)

March 2000 Susquehanna Unit 1 (US) ( d)Susquehanna Unit 1 (US) (continued)

� GE BWR� Damage to #3 feedwater heaters (operating

temperature = 142°C, steam quality = 91%)� Shells

T b t� Tube supports� Tie rods

� This event showed:� This event showed:� Importance of inspecting both the shells and the

internal elements of susceptible equipment

July 2002Wagner 3 Fossil Plant (US)Wagner 3 Fossil Plant (US)

Feedwater Heater drain Line

August 2004Mihama Unit 3 (Japan)Mihama Unit 3 (Japan)

Condensate line downstream of a flow measuring orifice

August 2004Mihama Unit 3 (Japan) ( i d)Mihama Unit 3 (Japan) (continued)

� Mitsubishi PWR� 560 mm pipe had thinned from 10 mm to ~ 1.4 mm� Five workers were killed and six were injured� Five workers were killed and six were injured.� Location had never been inspected� The location was similar to the Surry & Loviisa failures� The location was similar to the Surry & Loviisa failures

� Immediately downstream of an orifice� Approximate operating conditions:

� Temperature ~ 140° C� Pressure ~ 0.93 MPa

Velocity 1 94 m/s� Velocity ~ 1.94 m/s

March 2005Edwards Fossil Plant (US)Edwards Fossil Plant (US)

Failure Located Between Main Feedwater Regulator and the Regulator Discharge Block Valve

August 2005 South Ukraine Unit 2 (Ukraine)( )

August 2005 South Ukraine Unit 2 (Ukraine) ( d)South Ukraine Unit 2 (Ukraine) (continued)

R i b il PWR� Russian built PWR� Failure was in a 45° carbon steel elbow of the

moisture separator first stage drain tank to themoisture separator first stage drain tank to the deaerator� Pipe size was 219 x 8 mm

� Pipe wall had thinned to 0.5 - 2.5 mm � Two-phase conditions 19 kgf/cm2 and 211°C

A th t d i J l 2005 th� Another rupture occurred in July 2005 on the drain line from high pressure heater 6A to the deaerator

February 2006 Kakrapar Unit 2 (India)Kakrapar Unit 2 (India)

orifice

flow

February 2006 Kakrapar Unit 2 (India)Kakrapar Unit 2 (India)

� Utility built PHWR

� Failure was in the 10% feedwater system d t f ifidownstream of an orifice.

� Pipe size was 80 mm. Material was A106 Grade B.

� This location had been planned for replacement but was not replaced.

What Do These Pictures Have in Common?

Close-up of Rupture Overall Viewp p

What can be learned from history?y

FAC f il i ll f l� FAC failures occur in all types of power plants� Equipment and equipment internals are also

susceptible to FACsusceptible to FAC� Locations downstream of orifices and control

valves are especially susceptible� It is important to know replacement history� Location of maximum thinning varies, e.g.,

� Navajo was on upstream intrados of elbow� Navajo was on upstream intrados of elbow� Surry was on upstream side of elbow� Fort Calhoun and South Ukraine were on

extrados

What can be learned from history? (continued)y ( )

� It is important to know actual operating conditions of the lines and if they are being used differently than designedthan designed

� It is important to look in new locations around the plant as conditions are not always as assumedp y

� Knowledge of chromium is important� Parallel trains can wear differently

Conclusion

“You can either inspect it all (every outage), replace p ( y g ), pit all, or run some level of risk”

David Smith, Duke Energy, Past Chairman of CHUG

� The causes and prevention of FAC are well known� An intelligent well implemented program can� An intelligent, well implemented program can

minimize risk at a reasonable cost� Particularly as regards to large-bore piping and

vessels� But there will always be some risk for current

plants particularly as regards small-bore pipingplants, particularly as regards small bore piping

The Good NewsThe Good News

� The lessons learned from history have been incorporated into the requirements of NSAC-p q202L-R3

� No plant that has fully implemented NSAC 202L� No plant that has fully implemented NSAC-202L has ever had a major failure

� Details of the process to be discussed tomorrow