ML12088A401
| ML12088A401 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 01/17/2008 |
| From: | Munson D - No Known Affiliation |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| Shared Package | |
| ML12088A381 | List: |
| References | |
| RAS 22101, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01 | |
| Download: ML12088A401 (66) | |
Text
A B i f O i
f FAC A Brief Overview of FAC Investigations, Experiences and g
p Lessons Learned Douglas Munson WANO Seminar WANO Seminar Effective Monitoring and Control of FAC Haiyan, Zhejiang, China January 15-17, 2008 ENT000035 Submitted: March 28, 2012
Historical Perspective Historical 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 & Greenes classical text corrosion in Fontana & Greene s classical text book Corrosion Engineering published in 1967
Erosion-Corrosion and Flow-Accelerated Corrosion Film Free Breakaway te Film Breakdown Velocity rosion Ra Erosion-Corrosion Corr FAC Velocity Breakaway velocity for carbon steel straight pipe in water ~10 m/sec, wet stream ~30 m/sec 1940s 1940s
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 1940s were 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 fluids with 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 reactors sponsored 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
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 between in 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 ppb S
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
1974 Keller Develops the First Predictive Model Keller 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)
Where f(T) = dimensionless coefficient f(X) = dimensionless coefficient related to steam wetness V = fluid velocity (m/s)
Kc = factor to account for local geometry Ks = a constant that must be exceeded before FAC is observed
1974 Keller Develops the First Predictive Model
(
ti d)
(continued)
1974 Keller Develops the First Predictive Model (continued)
(continued)
Kellers Geometry Keller 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 model FAC 4k3 (Co
)3 (H+)2/B2 FAC = 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 etc transfer, 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)
1982 Ducreux Material Investigations Ducreux 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)
1984 Huijbregts Materials Investigations Huijbregts Materials Investigations
In 1984, W. M. M.
Huijbreghts of KEMA (Netherlands) published laboratory data on the effect laboratory 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 1978 Oyster Creek (US)
Oyster Creek (US)
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 valves several feedwater control valves
Damage was attributed to flashing
Significant wall thinning was found in several piping a eas mostl do nst eam of cont ol piping 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 weld L
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 analyzed publicized 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 1986 Surry Unit 2 (US)
Surry Unit 2 (US)
Flow Condensate system just before the feed pump
December 1986 Surry Unit 2 (US) (
i d)
Surry Unit 2 (US) (continued)
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 FAC S
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
1987 US Industry Responds to Surry US 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 piping programs of single-phase piping
1987 Current Industry Improves its Predictive Technology Technology
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 data g
Management of related outage activities
1987 - Current Industry 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 account into account
1987 - Current Industry 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
p EPRI report TR-106611-R1. Primary authors:
Bindi Chexal and Jeff Horowitz - EPRI
Michel Bouchacourt and Francois Remy - EDF
Wolfgang Kastner - KWU/Siemens
April 1989 Arkansas Nuclear One (US)
Arkansas Nuclear One (US)
April 1989 k
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 1989 US NRC Issues Generic Letter 89-08 US 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 systems steel 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)
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:
Dangers of operating with low oxygen in neutral water
July 1989 EPRI forms CHUG EPRI 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 only
2 meetings/year
Training
10 members representing ~ 30
FAC, erosion, weld degradation
2 meetings/year
Web site
Technical investigations p
g nuclear plants (US only) g
Training
46 members representing >
160 nuclear plants l
- 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 orifice Note 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 orifices High 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)
Simultaneous failures of two (of four) parallel lines downstream of level control valves and downstream of moisture separator drain tank moisture 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 plant g
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 f in 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 effort required a significant restart effort.
March 1993 Sequoyah Unit 2 (US)
Sequoyah Unit 2 (US)
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 deficiencies programmatic 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 1993 EPRI Issues NSAC-202L EPRI 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 implemented g
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 g
in 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-R3 Content 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)
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 drawings inadvertently 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 1995 Millstone 2 (US)
Millstone 2 (US)
Gate Valve
August 1995 Mill 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 y
the 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, partial 202L (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)
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 g
feedwater 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 l
Beam limited travel
August 1999 Callaway (US) (continued) y (
) (
)
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 ejected ejected
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 end such 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),
or or
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 unusual Importance of extra inspections if uncertain or unusual operating conditions
Importance of extra inspections in lines with backing rings rings
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)
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 2002 Wagner 3 Fossil Plant (US)
Wagner 3 Fossil Plant (US)
Feedwater Heater drain Line
August 2004 Mihama Unit 3 (Japan)
Mihama Unit 3 (Japan)
Condensate line downstream of a flow measuring orifice
August 2004 Mihama 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 2005 Edwards 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 the moisture 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
ifi downstream 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 View p
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 FAC susceptible 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 designed than designed
It is important to look in new locations around the plant as conditions are not always as assumed p
y
Knowledge of chromium is important
Parallel trains can wear differently
Conclusion You can either inspect it all (every outage), replace p
(
y g ),
p it 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 piping plants, particularly as regards small bore piping
The Good News The Good News
The lessons learned from history have been incorporated into the requirements of NSAC-p q
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