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
ENT000035 Submitted: March 28, 2012 A Brief B i f Overview O i off FAC Investigations, g , Experiences p and Lessons Learned Douglas Munson WANO Seminar Effective Monitoring and Control of FAC Haiyan, Zhejiang, China January 15-17, 2008
Historical Perspective Flow-accelerated corrosion (FAC) is not a new phenomenon Historically it was called erosion erosion-corrosion 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 layer E-C is identified as one of the 8 forms of corrosion in Fontana & Greene Greenes s classical text book Corrosion Engineering published in 1967
Erosion-Corrosion and Flow-Accelerated Corrosion Film Free Breakaway Corrrosion Rate Velocity Film Breakdown Erosion-Corrosion FAC Velocity Breakaway velocity for carbon steel straight pipe in water ~10 m/sec, wet stream ~30 m/sec 1940s Most early researchers were concerned about turbines, heat exchanger tubes, and feedwater i
iron (FAC iis th the primary i source off ffeedwater d t iiron))
Failures of power plant piping attributed to FAC 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 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 Design D i off commercial i l nuclear l power plants l t iin th the mid id 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 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 - GE 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 feedwater conditions (then O2 < 15 ppb)
Prior data was taken at ~25°C and 288°C, with little to no data 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 O2 on FAC rates Also concluded that:
Carbon steel can be used only if O2 > 15 ppb Stainless S i l can b be used d ffor all ll llevels l off O2 Lead investigators were Brush and Pearl
1974 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%
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
(
(continued) ti d)
1974 Keller Develops the First Predictive Model (continued)
Keller s Geometry Kellers Factors
1970s - CEGB (UK) ( )
In the late 1970s and early 1980s, 1980s G.
G Bignold, Bignold I.
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, 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 * (Coeq)3 * (H+)2/B2
1970s - EDF (France)
( )
Startingg in the early y 1970s,, P. Berge, g , M.
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, transfer, composition pH, pH amine, 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 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 y 1980s - Siemens/KWU ((Germany) 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 In 1984, W. M. M.
Huijbreghts of KEMA (Netherlands) published 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)
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 valves Damage was attributed to flashing Significant wall thinning was found in several piping aareas, eas mostly mostl do downstream nst eam of control cont ol valves Event and findings had little effect on nuclear industry
November 1982 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
Low pH H ammonia i +h hydrazine d i water t chemistry h i t Unusual because it was a fossil failure that was publicized and analyzed Little change in the way fossil and nuclear plants approached FAC.
November 1982 Navajo Fossil Plant (US)
Note direction of flow
December 1986 Surry Unit 2 (US)
Flow Condensate system just before the feed pump
December 1986 Surry Unit 2 (US) (continued)
( i d)
Westinghouse PWR Four workmen were killed. Four others were injured Ammonia water chemistry y with a low pH p
Many replacements (~190) were made in both units Similar conditions as Navajo This failure showed:
Seriousness of FAC Susceptibility S ibili off single-phase i l h systems to FAC C
Need for an inspection program Maximum damage may not be at extrados of elbows
1987 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 and d INPO became b interested i t t d in i the th iissue US nuclear plants start to implement inspection programs of single-phase single phase piping
1987 Current Industry Improves its Predictive Technology 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 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 g and evaluation of component inspection data Management of related outage activities
1987 - Current Industry Improves its Predictive Technology (continued)
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
1987 - Current Industry Improves its Predictive Technology
(
(continued) i d)
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 maximum probable thinning Current version is called COMSYS Includes In l des other othe mechanisms me hanisms such s h as strain-induced st ain ind ed cracking, material fatigue, cavitation, droplet impingement
For More Information Details of the various models, a theoretical treatment and laboratory y data are provided p in 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)
April 1989 Arkansas k Nuclear l One (US)
( ) (continued)
Combustion Engineering PWR Two-phase conditions L
Location ti was downstream d t off high-pressure hi h extraction nozzle 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 Required the US utilities to:
Implement a long-term FAC monitoring program Include all susceptible high-energy carbon steel piping systems Include both single- and two-phase lines Ut Utilize e tthe e NUMARC/EPRI U C/ o or equa equally y effective e ect e method
December 1989 Santa Maria de Garona (Spain)
GE BWR A small ppiece of the feedwater line was blown out Failure was just downstream of a flowmeter.
Line operated with very low oxygen (~6 ppb)
Failure demonstrated:
Need for FAC program for BWRs Dangers of operating with low oxygen in neutral water
July 1989 EPRI forms CHUG Then (1989) Now (2008)
Purpose was to support the Issue group to deal with CHEC and CHECMATE computer degradation in FAC susceptible codes systems FAC only FAC erosion, FAC, erosion weld degradation 2 meetings/year 2 meetings/year Training Web site 10 members representing p g ~ 30 Technical investigations g
nuclear plants (US only) Training 46 members representing >
160 nuclear plants
- Belgium l *Romania
- Canada (all) *Slovakia
- Czech Republic *South Korea
- France *Spain (all)
- Japan (TEPCO
( CO *Taiwan i
only) *US (all)
- Mexico
May 1990 Loviisa Unit 1 (Finland)
Note orifice orifice.
Downstream Pipe Upstream Flange
May 1990 Loviisa Unit 1 (Finland) (continued)
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.
This failure showed:
The significance of Cr (there was > 0.1% Cr in d
downstream t pipe i and d ~0 0CCr iin th the fl flange))
Importance of high oxygen if using neutral water 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)
December 1990 Millstone 3 (US) (continued)
Westinghouse PWR Simultaneous failures of two (of four) parallel lines downstream of level control valves and downstream of 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 Computer C t models d l mustt iinclude l d all ll susceptible tibl systems t
Importance of good communications between central g
engineering g and the plant p
Parallel lines may wear differently
November 1991 Millstone 2 (US)
November 1991 Millstone 2 (US) (continued)
( d)
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 in the vicinity i i it off th the b break k llocations ti shortly h tl bbefore f
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.
effort
March 1993 Sequoyah Unit 2 (US)
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 deficiencies Lengthy shutdown for both Sequoyah units was required This failure showed:
Need for personnel training Need to identify a program owner Dangers of excessive personnel turnover
November 1993 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 g and technologygy were beingg implemented p
Visits found a wide range of implementation details In 1993 EPRI and CHUG developed NSAC-202L R
Recommendations d ti for f an Effective Eff ti Fl Flow-Accelerated A l t d Corrosion Program Has been accepted p by y INPO,, the US NRC,, and regulators 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)
Content of NSAC NSAC-202L-R3 202L R3 Overview of an effective program Procedures and documentation Recommendations for FAC tasks Performing a susceptibility analysis Performing a FAC analysis Selection of inspection locations Performing P f i iinspections ti Evaluating inspection data Evaluating worn components Replacements and repairs Developing a long-term strategy Recommended program for smallsmall-bore bore piping Recommended program for vessels and equipment
November 1994 S
Sequoyah hUUnit it 1 (US)
Westinghouse PWR Crack caused a leak in the condensate system.
system Flow straightener used during construction was 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)
February 1995 Pleasant Prairie Fossil Plant ((US)) ((continued))
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%.
0 12%
This failure showed:
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)
Gate Valve
August 1995 Mill t Millstone 2 (US) (continued)
Failure downstream of gate valve in heater drain tank bypass line 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:
The importance of knowing operating history.
The valve had apparently pp y been used to throttle the flow
April 1997 Fort Calhoun (US)
April 1997 Fort Calhoun (US) (continued)
( i d)
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 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 implementation)
May 1999 Point Beach Unit 1 (US)
- 2 heater , operating temperature = 175°C, steam quality = 88%
May 1999 Point Beach Unit 1 (US) (continued)
( d)
Westinghouse PWR Nominal wall thickness = 13 mm Fishmouth type failure with opening size of 685 x 22 mm. Degradation extended 1219 mm Failure location was where steam entering g the feedwater heater hit the impingement plate, and deflected to the shell Simila degradation Similar deg adation in pa parallel allel ttrain ain This failure showed:
Need for the inspection program to include vessels
August 1999 Callaway (US)
B Beam lilimited it d travel t l
August 1999 Callaway y ((US)) ((continued))
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 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
August 1999 C ll Callaway (US) (continued)
( d)
The void fraction was estimated to be about 55% 55%. Most two-phase lines in nuclear plants either have:
Void fractions of very near one (e.g., extraction lines),
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 operating conditions Importance of extra inspections in lines with backing rings Importance of inspecting parallel trains
March 2000 Susquehanna Unit 1 (US)
March 2000 Susquehanna Unit 1 (US) (continued)
( d)
GE BWR Damage to #3 feedwater heaters (operating temperature = 142°C, steam quality = 91%)
Shells Tube T b supports t Tie rods This event showed:
Importance of inspecting both the shells and the internal elements of susceptible equipment
July 2002 Wagner 3 Fossil Plant (US)
Feedwater Heater drain Line
August 2004 Mihama Unit 3 (Japan)
Condensate line downstream of a flow measuring orifice
August 2004 Mihama Unit 3 (Japan) (continued)
( i d)
Mitsubishi PWR 560 mm pipe had thinned from 10 mm to ~ 1.4 mm Five workers were killed and six were injured injured.
Location had never been inspected 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 1.94 94 m/s
March 2005 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) (continued)
( d)
Russian R i b built il PWR Failure was in a 45° carbon steel elbow of 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 rupture Another t occurredd in i July J l 2005 on the th drain line from high pressure heater 6A to the deaerator
February 2006 Kakrapar Unit 2 (India) orifice flow
February 2006 Kakrapar Unit 2 (India)
Utility built PHWR Failure was in the 10% feedwater system d
downstream t off an orifice.
ifi 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 p of Rupture p Overall View
What can be learned from history? y FAC failures f il occur in i all ll types off power plants l
Equipment and equipment internals are also 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 Surry was on upstream side of elbow Fort Calhoun and South Ukraine were on extrados
What can be learned from history? y ((continued))
It is important to know actual operating conditions of the lines and if they are being used differently than designed It is important to look in new locations around the plant as conditions are not always p y as assumed Knowledge of chromium is important Parallel trains can wear differently
Conclusion You can either inspect p it all ((every y outage),
g ), replace 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, 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 plants, small-bore bore piping
The Good News The lessons learned from history have been incorporated p into the requirements q of NSAC-202L-R3 No plant that has fully implemented NSAC NSAC-202L 202L has ever had a major failure Details of the process to be discussed tomorrow