ML14184A177
| ML14184A177 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 02/22/1980 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Jackie Jones Carolina Power & Light Co |
| References | |
| NUDOCS 8003140055 | |
| Download: ML14184A177 (7) | |
Text
FEBRUARY 2 2 1960 Docket No. 50-261 Mr. J. A. Jones Senior Executive Vice President Carolina Power and Light Company 336 Fayetteville Street Raleigh, North Carolina 27602
Dear Mr. Jones:
The cracking that was found in the feedwater system piping at your plant, is summarized in Table 1 of-the-enclosed safety analysis.
The NRC Staff has reviewed the actions you have taken and finds that the repair program, the nondestructive inspections and leakage testing per formed following the repairs are adequate to insure that the integrity of the feedwater piping will beimaintained until the recommendations of the.;wners' Group and the NRC's_ Pipe Crack Study Group have been evaluated.
Should we determine tifat further licensing.actions are required after these evaluations, you will be riotifi.ed.
Sincerely, Original Signed By A. Schwencer, Chief Operating Reactors Branch #1-.
Division of Operatiig Reactors
Enclosure:
Safety Analysis of I nterim Actions Taken to Eliminate Feedwater Piping Cracks cc:
w/enclosure See next page D
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 FEBRUARY 2 2 198O Docket No. 50-261 Mr. J. A. Jones Senior Executive Vice President Carolina Power and Light Company 336 Fayetteville Street Raleigh, North Carolina 27602
Dear Mr. Jones:
The cracking that was found in the feedwater system piping at your plant, is summarized in Table 1 of the enclosed safety analysis.
The NRC Staff has reviewed the actions you have taken and finds that the repair program, the nondestructive inspections and leakage testing per formed following the repairs are adequate to insure that the integrity of the feedwater piping will be maintained until the recommendations of the Owners' Group and the NRC's Pipe Crack Study Group have been evaluated.
Should we determine that further licensing actions are required after.
these evaluations, you will be notified.
Sincerely, A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Enclosure:
Safety Analysis of Interim Actions Taken to Eliminate Feedwater Piping Cracks cc:
w/enclosure See next page
Mr. J. A. Jones Carolina Power and Light Company FEBRUARY 2 2 cc: G. F. Trowbridge, Esquire Shaw, Pittman, Potts and Trowbridge 1800 M Street, N.W.
Washington, D. C. 20036 Hartsville Memorial Library Home and Fifth Avenues Hartsville, South Carolina 29550 Michael C. Farrar, Chairman Atomic Safety and Licensing Appeal Board Panel U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Richard S. Salzman Atomic Safety and Licensing Appeal Board Panel U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Dr. W. Reed Johnson Atomic Safety and Licensing Appeal Board Panel U. S. Nuclear Regulatory Commission Washington, D. C. 20555
SAFETY ANALYSIS OF INTERIM ACTIONS TAKEN TO ELIMINATE FEEDWATER.PIPING CRACKS On May 20, 1979, Indiana and Michigan Power Company notified the NRC of cracking in two feedwater lines at their D. C. Cook Unit 2 facility. The cracking was discovered following a shutdown on May 19 to investigate leakage inside contain ment. Leaking circumferential cracks were identified in the 16-inch diameter feed water elbows adjacent to two steam generator nozzle to elbow welds. Subsequent radiographic examinations revealed cracks in all eight steam generator feedwater lines at this location on both units 1 and 2.
On May 25, 1979, a letter was sent to all PWR licensees by the Office of Nuclear Reactor Regulation which informed licensees of the D. C. Cook failures and requested specific inforamtion on feedwater system design, fabrication, inspection and oper ating histories. To further explore the generic nature of the cracking problem, the Office of Inspection and Enforcement requested licensees of PWR plants in current outages to immediately conduct volumetric examination of certain feedwater piping welds. As a result of these actions several other licensees reported cracking in the steam generator feedwater nozzle-to-piping weld vicinity. On June 25, 1979, IE Bulletin 79-13 was issued. The Bulletin required inspection of the steam cener ator nozzle-to-pipe welds and adjacent areas within 90 days. If flaws were found in these welds, the feedwater piping welds to the first support, the feedwater piping to containment penetration and the auxiliary feedwater to main feedwater piping connection were required to be inspected.
In conformance with the Bulletin, the licensees of the plants listed in the attached Table 1 completed the radiopraphic examinations and found crackin-in the feedwater DiDing systems.'
Meetings and/or telephone conference calls were held with the respective licensees to discuss the following items regarding the feedwater piping cracks at their facilities:
- 1.
Nature and extent of the cracking.
- 2. Metallurgical evaluation of the cracking including identification of the -;ode oF failure.
- 3.
Stress analyses
- 4. Operating history
- 5.
Feedwater chemistry
- 6. Corrective actions
- 7. Safety Implications The licensees' interim reports containing the information above were submitted and reviewed by the staff prior to the units returning to power. The extent of the crackina at the facilities is summarized in Table 1. The mode of failure at all the facilities discussed in this analysis, with the exception of Yankee Rowe was i:erti fied as fatigue assisted by corrosion. The Yankee Rowe facility had gross faurication defects in its feedwater piping.
No anomolies were found in the Coce require: stress analyses at the facilities.
From the results of instrumentation installed at several plants which have e-::erienced feedwater piping cracks and other mooeling and analyses by a utility sponszre: Gwners
-2 Group, significant cyclic stresses have been identified that occur in the feedwater piping in the vicinity of the steam generator nozzle from mixing and stratification of cold auxiliary feedwater with hot water from the steami generator during low flow conditions. The Owners Group is expected to complete their investigations and make recommendations for changes in design and operating procedures in February 1980.
The licensees have repaired and/or replaced the affected piping in most cases with improved designs to minimize stress risers. In addition, the licensees have com mitted to reinspect the steam generator to feedwater piping weld vicinities at the subsequent refueling outage.
Although the piping has been repaired at the facilities listed in Table 1, the staff feels that cracking could re-occur in the future at these facilities.
The staff and Owners Group both have performed independent analyses and have deter mined that flawed feedwater piping could withstand challenges from operating and faulted loads including seismic and limited water loads without loss of piping integrity. Pipe breaks have occurred in the past in feedwater piping as the result of water hammer loads. However, design changes such as "J" tubes have been rade and operational changes have occurred to minimize the possibility of water hamer.
In the unlikely event of a feedwater pipe break frcm a severe water hammer, the consequences have been analyzed as a design base accident and acceptable measures to deal with the event have been established.
The NRC has instituted a Pipe Crack Study Group to review this and other pipe crack ing problems in PWR's. It is anticipated that the Pipe Crack Study Group will complete its work by June 1980 and provide recomrendations for review and implemen tation by licensees as new criteria for operatin; plants.
We conclude that repairs to the feedwater piping, the nondestructive inspecti:ns performed and scheduled, and the analyses perfcrme: for flawed piping ensure :hat the piping integrity will be maintainec until the recomendations of the Ownes Group and the Pipe Crack Study Group have ceen eva"uated.
Should the staff ceterrine that further actions are required after evaluaticn of the Owners Group and Pi:e Crack Study Group reommendations, the licensees wil! benotified at that time.
Table I -
Summary of PWR Feedwater Piping Cracks PLANT EXT.NT OF CRACKING (NOZZLE VICINITY)
PIPING COMPONENT PROBABLE CAUSE COMENTS max.
Locntion max.
No.
of Linen pth Depth Crack Cracked Westinghouse D. C. Cook 1/2 Thru wall TOP 8 of 8 elbow Corrosion Assisted 2 cracks thru wall Fatigue Beaver Valley 0./400" 9 O'clock 3 of 3 elbow Corrosion, Assisted 13 additional fab.
lat Fatigue Indications replaire Kawaunce 0.050" 7 O'clock 2 of 2 pipe Corrosion Assistd 3" dia. aux. feed near Fatigue SG inlet rt.
Beach 1/2 0.047" 3 O'clock 2 of 2 reducer Corrosion Assisted 3" dia. aux. feed near Fatigue SG inlet II.II.Itobinson 2 0.750" 9 O'clock 3 of 3 reducer Corrosion Assisted Shallow cracking in nozm Fatigue under thermal sleeve Salent 1 0.235" 4 of t elbow Corrosion Assisted reducer Fatigue San Onofre 1 0.1.00" lower half 3 of 3 reducer Stress Assisted Multiple branched cracks of reducer Corrosion evidence of some fa.1*1us Surry 1/2 0.080" 2 and 5 6 of 6 reducer Corrosion Assisted O'clock Fatigue Ginna 0.107" 8:30 O'clock 2 of 2 elbow Stress Assisted Cracks also at deep Corrosion/Corrosion machining marks Fatigue Zionl, 1/2 0.0ffil" 4 O'clock t of 11 elbow pipe Corrotion Assisted Fatigue Vnkee Rowe Gross fabrication defects in piping
2 Table 1. -
SumnIary of PWR Feedwater Piping Cracks PLANT EXTENT OF CRACKING (NOZZLE VICINITY)
PIPING COMPONENT PROBABLE CAUSE COMMENTS
- Max, Location max.
No. of Lines De >th De th Crack Cracked Combustion Engitcering MiIstone 2 0.250" 12 O'clock 2 of 2 pipe.
Not analyzed Ial I h
0.170" 3 and 9 O'clock 2 of 2 pipe Corrosion Assisted Cracks found also at weld Fatigue vicinity of horizontal piping