ML14170A513
| ML14170A513 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 10/17/1979 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Jackie Jones CAROLINA POWER & LIGHT CO. |
| References | |
| NUDOCS 7911070116 | |
| Download: ML14170A513 (14) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II 101 MARIETTA ST., N.W., SUITE 3100 ATLANTA, GEORGIA 30303 O CT 1 7 1979 In Reply Refer To:
RIJO Carolina Power and Light Company Attn:
J. A.. Jones Senior Executive Vice President and Chief Operating Officer 411 Fayetteville Street Raleigh, North Carolina 27602 Gentlemen:
Enclosed is IE Bulletin No. 79-13, Revision 2 which requires action by you with regard to your power reactor facility(ies).
Should you have any questions regarding this Bulletin or the actions required by you, please contact this office.
Sincerely, mes P. O'Reilly Di ector
Enclosures:
- 1.
IE Bulletin No. 79-13, Revision 2 w/Attachments
- 2.
List of IE Bulletins Issued in Last Six Months 1911070
00CT 17 1979 Carolina Power and Light Company
-2 cc w/encl:
R. B. Starkey, Jr., Plant Manager Post Office Box 790 Hartsville, South Carolina 29550
SSINS:
6830 Accession No.:
7908220135 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 October 16, 1979 IE Bulletin No. 79-13 Revision 2 CRACKING IN FEEDWATER SYSTEM PIPING Description of Circumstances:
This revision to IE Bulletin No. 79-13 is based on the results of the radio graphic examinations and ongoing investigation of the subject problem to date since the initial Bulletin was issued. The revision reduces in scope the R2 number and extent of the piping system welds required to be examined. The requirements for reporting and action time frame remain unchanged.
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 containment. Leaking circumferential cracks were identified in the 16-inch feedwater elbows adjacent to two steam generator nozzle elbow welds.
Subsequent radiographic examination revealed crack indications 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 information on feedwater system design, fabrication, inspection and operating 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 with Westinghouse steam generators reported crack indications. Southern California Edison reported on June 5, 1979, that radiographic examination revealed indications of cracking in feedwater nozzle-to-pipe welds on two of three steam generators of San Onofre Unit 1. On June 15, 1979, Carolina Power and Light reported that radiography showed crack indications in similar locations at their H. B.
Robinson Unit 2. Duquesne Power and Light confirmed on June 18, 1979, that radiography has shown cracking in their Beaver Valley Unit 1 feedwater piping to-vessel nozzle weld. Public Service Electric and Gas Company reported on June 20, 1979 that Salem Unit 1 also has crack indications. Wisconsin Public Service company decided on June 20, 1979 to cut out a feedwater nozzle-to-pipe weld which contained questionable indication, for metallurgical examination.
As of June 22, 1979 and since May 25, 1979 seven other PWR facilities have inspected the feedwater nozzle-to-pipe welds without finding cracking indications.
NOTE:
R1 and R2 indicates lines revised or added.
IE Bulletin No. 79-13 October 16, 1979 Revision 2 Page 2 of 5 The feedwater nozzle-to-pipe configurations for D. C. Cook and for San Onofre are shown on the attached figures 1 and 2. A typical feedwater nozzle-to-pipe weld joint detail showing the principal crack locations for D. C. Cook and San Onofre are shown on the attached figure 3.
On March 17, 1977, during heat-up for hot functional testing of Diablo Canyon Unit 1, a leak was discovered in the vessel nozzle-to-pipe butt weld joining the 16-inch diameter feedwater piping to steam generator 1-2. Subsequent nondestructive examination of all nozzle welds by radiography and ultrasonics revealed an approximate 6-inch circumferential crack originating in the weld root heat-affected zone of the leaking nozzle weld. The cause of this crack ing was identified as either corrosion fatigue or thermal fatigue initiating at small cracks probably induced by the welding and postweld heat treatment cycles. The system was repaired by replacing with a piping component employing greater controls on the welding including maintaining preheat temperature until postweld heat treatment.
The potential safety consequences of the cracking is an increased likelihood of a feedwater line break in the event of a seismic event or water hammer. A feedwater line break results in a loss of one of the mechanisms of heat removal from the reactor core and would result in release of stored energy from the steam generator into containment. Although a feedwater line break is an analyzed accident, the identified degradation of these joints in the absence of a routine inservice inspection requirement of these feedwater nozzle-to-pipe welds formed the basis of this Bulletin.
To date the radiographic examinations, supplemented by ultrasonic methods, have identified cracking in the steam generator nozzle to feedwater piping weldments at the following W, and C. E. plants.
D. C. Cook Units 1 & 2 Salem Unit 1 Diablo Canyon Surry Unit 1 San Onofre Unit I R. E. Ginna H. B. Robinson Unit 2 Millstone Unit 2 Beaver Valley Unit 1 Palisades Kewaunee Yankee Rowe Point Beach Unit 2 Maine Yankee Found during hot functional testing Confirmatory evaluation incomplete R2 An extensive metallurgical investigation has been conducted by Westinghouse on a substantial number of cracked weldments removed from the above plants.
Results of the metallurgical analysis lead to the conclusion that a corrosion fatigue phenomenon is the probable failure mechanism, except for the San Onofre piping which has been characteristized as stress assisted corrosion.
In parallel with the above ongoing analysis, the feedwater piping at D. C.
Cook, H. B. Robinson, R. E. Ginna, Salem 1 and other plants have been instrumented (Thermocouples, accelerometers, strain gages, and transducers) to collect data I
IE Bulletin No. 79-13 October 16, 1979 Revision 2 Page 3 of 5 on the potential forcing functions contributing to cracking under steady state and transient conditions. Preliminary unchecked results of temperature data has identified cyclic thermal gradients may exist due to stratified feedwater temperature conditions in the feedpipe weld region during zero and low power operations. This gradient tends to support the fatigue aspect of the postulated failure mechanism. No further unexpected operation loading or forcing functions have been identified by other instrumentation.
In regard to B&W plants a total of 95 welds in the main and separate auxiliary R2 feedwater piping, risers and, steam generator nozzles regions have been examined at Crystal River Unit 3 and Davis Besse. No indications of a cracking problem was found.
In view of the findings to date, the revised inspections outlined below is R2 considered acceptable to meet this intent of IE Bulletin No. 79-13.
Actions to be Taken by Licensees For all pressurized water reactor facilities with an operating license:
- 1.
Facilities which have steam generators fabricated by Westinghouse or Combustion Engineering that have not conducted volumetric examination of feedwater nozzles since May 1979 shall complete the inspection program described below at the earliest practical time but no later than 90 days after the date of Bulletin No. 79-13.
- a.
Perform radiographic examination, supplemented by ultrasonic examination as necessary to evaluate indications, of all feedwater nozzle-to-pipe welds and of adjacent pipe and nozzle areas (a distance equal to at least two wall thicknesses).
Evaluation shall be in accordance with ASME Section III, Subsection NC, Article NC-5000. Radiography shall be performed to the 2T penetrameter sensitivity level, in lieu of Table NC-5111-1, with systems void of water.
- b.
In the event cracking is identified during examination of the nozzle to-pipe weld, all feedwater line welds up to the first piping support or snubber outboard of the nozzle shall be volumetrically examined in accordance with 1.a above. All unacceptable code discontinuities shall be subject to repair unless justification for continued operation is provided.
- c.
Perform a visual inspection of feedwater system piping supports and snubbers in containment to verify operability and conformance to design.
- 2.
All pressurized water reactor facilities shall perform the inspection program described below at the next outage of sufficient duration or at the next refueling outage after the inspection required by item 1.
IE Bulletin No. 79-13 October 16, 1979 Revision 2 Page 4 of 5
- a.
For steam generator designs with a common nozzle for both main and auxiliary feedwater systems, perform volumetric examination of the feedwater nozzle-to-pipe welds, the feedwater piping welds to the first support, and the feedwater line-to-containment penetration welds in accordance with Item 1 above.
In addition, examine an area of at least one pipe diameter of the main feedwater line downstream at the auxiliary feedwater to main feedwater connection.
R2
- b.
For steam generator designs utilizing auxiliary feedwater systems connected by means of welded nozzle connections, perform volumetric examination of all auxiliary feedwater nozzle to piping welds and the first adjacent outboard pipe-to-pipe welds (risers) in accordance with item 1 above.
For designs utilizing auxiliary feedwater systems connected to the steam generator by means of bolted flange connections, perform volumetric examination of the flanged nozzle to piping and first outboard pipe-to-pipe welds (risers) in accordance with item 1 above.
The examinations specified in 2.b above are not required provided that during startup, hot standby or cold shutdown operations, the feedwater level within the steam generator is maintained essentially constant and no intermittent cold auxiliary feedwater injection is utilized; i.e., auxiliary feedwater injection where used, is preheated during the forementioned operating modes.
- c.
Perform a visual inspection of all feedwater system piping supports and snubbers in containment to verify operability and conformance to design.
- 3.
Identification of cracking indications in feedwater nozzle or piping weld areas in one unit of a multi-unit facility shall require shutdown and inspection of other similar units which have not been inspected since May 1979, unless justification for continued operation is provided.
- 4.
Any cracking or other unacceptable code discontinuities identified shall be reported to the Director of the appropriate NRC Regional Office within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of identification.
- 5.
Provide a written report to the Director of the appropriate NRC Regional Office within 20 days of the date of the orginal Bulletin (June 25, 1979) addressing the following:
- a.
Your schedule for inspection if required by item 1.
- b.
The adequacy of your operating and emergency procedures to recognize and respond to a feedwater line break accident.
- c.
The methods and sensitivity of detection of feedwater leaks in containment.
IE Bulletin No. 79-13 October 16, 1979 Revision 2 Page 5 of 5
- 6.
A written report of the results of examination, in accordance with requests by Regional Offices preceding this Bulletin and with Bulletin items 1 and 2 including any corrective measures taken, shall be submitted within 30 days of the date of the original Bulletin No. 79-13 (June 25, 1979) or within 30 days of completion of the examination, whichever is later, to the Director of the appropriate NRC Regional Office with a copy to the NRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D. C. 20555.
Actions to be Taken by Designated Applicants for Operating Licenses:
- 1.
On completion of the hot functional testing program and prior to fuel loading, perform the inspections described in item 1 above.
- 2.
During the first refueling outage, perform the inspections described in RI item 2 above.
- 3.
Submit reports as described in Items 4, 5, amd 6 above based on the date of Revision 1 to Bulletin No. 79-13 (August 30, 1979)
Approved by GAO, B180225 (R0072), clearance expires 7/31/80. Approval was given under a blanket clearance specifically for identified generic problems.
Attachments:
Figures 1, 2, and 3
e II, I
THERMAV
-SLEEVE VNLET 4k4 i~r
- 1 t
1 NOkM AL FLOW I AfoM.
ALL AT FIGURE I
--FEEDWATER PIPE ELBOW TO STEAM GENERATOR NOZZLE INSTALLATION (DC COOK)
.Fir*3 FIGURE 2 FEEDWATER PIPE REDUCER TO STEAM GENERATOR NOZZLE INSTALLATION (SAN ONOFRE)
FIGURE 3 TYPICAL FEEDWATER PIPE TO NOZZLE WELD JOINT DETAIL CW.CE )
00 DESIGNATED APPLICANTS FOR OPERATING LICENSES Salem 2 North Anna 2 Diablo Canyon 1 & 2 Sequoyah 1 RI McGuire 1 San Onofre 2 Summer Watts Bar 1 & 2
IE Bulletin No. 79-13, Revision 2 Enclosure October 17, 1979 Page 1 of 3 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Bulletin No.
Subject Date Issued Issued To 79-24 Frozen Lines 9/27/79 All Power Reactor Yacilities with an OL or a CP 79-23 Potential Failure of Emergency 9/12/79 All Power Reactor Diesel Generator Field Exciter Facilities with an Transformer OL or a CP 79-22 Possible Leakage of Tubes of 9/5/79 Each Licensee who Tritium Gas Used in Timepieces Receives Tubes of for Luminosity Tritium Gas in Time pieces for Luminosity 79-21 Temperature Effects on Level 8/13/79 All PWR's with an Measurements Operating License 79-20 Packaging Low-Level Radioactive 8/10/79 All Materials Licensees Waste for Transport and Burial who did not receive Bulletin No. 79-19 79-19 Packaging Low-Level Radioactive 8/10/79 All Power and Research Waste for Transport and Burial Reactors with OLs, Fuel Facilities except uranium mills, and certain materials licensees 79-18 Audibility Problems Encountered 8/7/79 All OLs for Action on Evacuation of Personnel from All CPs for Information High-Noise Areas 79-17 Pipe Cracks in Stagnant Borated 7/26/79 All PWRs with Water Systems at PWR Plants Operating License 79-16 Vital Area Access Controls 7/26/79 All Holders of and applicants for Power Reactor Operating Licenses who Antici page loading fuel prior to 1981 79-15 Deep Draft Pump Deficiencies 7/18/79 All Power Reactor (Supp. 1)
IE Bulletin No. 79-13, Revision 2 Enclosure October 17, 1979 Page 2 of 3 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Bulletin No.
Subject Date Issued Issued To 79-15 Deep Draft Pump Deficiencies 7/11/79 All Power Reactor
-Licensees with a CP and/or OL 79-14 Seismic Analyses for As-Built 7/27/79 All Power Reactor (Correction)
Safety-Related Piping System Facilities with an OL or a CP 79-14 Seismic Analyses for As-Built 9/7/79 All Power Reactor (Supp. 2)
Safety-Related Piping System Facilities with an OL or a CP 79-14 Seismic Analyses for As-Built 7/18/79 All Power Reactor (Rev. 1)
Safety-Related Piping System Facilities with an OL or a CP 79-14 Seismic Analyses for As-Built 7/2/79 All Power Reactor Safety-Related Piping System Facilities with an OL or a CP 79-13 Cracking in Feedwater System 10/17/79 All PWR's with an (Rev. 2)
Piping Operating License 79-13 Cracking in Feedwater System 8/30/79 All PWR's with an (Rev. 1)
Piping Operating License 79-13 Cracking in Feedwater System 6/25/79 All PWR's with an Piping OL for action. All BWRs with a CP for information 79-12 Short Period Scrams at BWR 5/31/79 All GE BWR Facilities Facilities with an OL 79-11 Faulty Overcurrent Trip Device 5/22/79 All Power Reactor in Circuit Breakers for Engin-Facilities with an eered Safety Systems OL or a CP 79-10 Requalification Training 5/11/79 All Power Reactor Program Statistics Facilities with an OL 79-06C Nuclear Incident at Three Mile 7/26/79 To all PWR Power Island - Supplement Reactor Facilities with an OL
IE Bulletin No. 79-13, Revision 2 Enclosure October 17, 1979 Page 3 of 3 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Bulletin No.
Subject Date Issued Issued To 79-05C Nuclear Incident at Three 7/26/79 To all PWR Power Mile Island - Supplement Reactor Facilities with an OL 79-02 Pipe Support Base Plate Designs 8/20/79 All Power Reactor (Rev. 1)
Using Concrete Expansion Anchor Facilities with an (Supp. 1)
Bolts OL or a CP 79-02 Pipe Support Base Plate Designs 6/21/79 All Power Reactor (Rev. 1)
Using Concrete Expansion Anchor Facilities with an Bolts OL or a CP 79-01A Environmental Qualification of 6/6/79 All Power Reactor Class 1E Equipment (Deficien-Facilities with an cies in the Environmental OL or a CP Qualification of ASCO Sole noid Valves)