ML18136A157
| ML18136A157 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna  |
| Issue date: | 10/17/1979 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Proffitt W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| References | |
| NUDOCS 7911070157 | |
| Download: ML18136A157 (14) | |
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! ~¥-.~*1 UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II p
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-.).0 101 MARIETTA.ST., N.W., SUITE 3100 ATLANTA, GEORGIA 30303 In Reply Refer To:
OCT 1 7 1979 RII:JPO 50-338 50-339
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Virginia Electric and Power Company Attn:
W. L. Proffitt Senior Vice President, Power P. 0. Box 26666 Richmond, Virginia 23261 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, D:r ector
Enclosures:
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- 1.
Revision 2 w/Attachments
- 2.
List of IE Bulletins Issued in Last Six Months
Virginia Electric and Power Company cc w/encl:
e W.R. Cartwright, Station Manager Post Office Box 402 Mineral, Virginia 23117 P. G. Perry Senior Resident Engineer Post Office Box 38 Mineral, Virginia 23117 W. L. Stewart, Manager Post Office Box 315 Surry, Virginia 23883 OCT 1 7 1979 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 October 16, 1979 CRACKING IN FEEDWATER SYSTEM PIPING Description of Circumstances:
e SSINS:
6830 Accession No.:
7908220135 IE B~lletin No. 79-13 Revision 2 This revision to IE Bulletin No. 79-13 is based on the results of the graphic examinations and ongoing investigation of the subject problem since the initial Bulletin was issued.
The revision reduces in scope number and extent of the piping system welds required to be examined.
requirements for reporting and action time frame remain unchanged.
radio-to date the The 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.
Publ{c 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 indicati~~, for metallurgical examination.
I R2 I 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:
Rl and R2 indicates lines revised or adqed.
IE Bulletin No. 79-13 Revision 2 October 16, 1979 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~, and C. E. plants.
D. C. Cook Units 1 & 2 Diablo Canyon n San Onofre Unit 1 H.B. Robinson Unit 2 Beaver Valley Unit 1 Kewaunee Point Beach Unit 2 Found during hot functional testing Confirmatory evaluation incomplete Salem Unit 1 Surry Unit 1 R. E. Ginna Millstone Unit 2 Palisades Yankee Rowe Maine Yankee 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.
R2 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 Revision 2 October 16, 1979 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 Revision 2 October 16, 1979 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.
- 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 p1p1ng supports and snubbers in containment to verify operability and conformance to design.
- 3.
Identification of cracking indications in feedwater nozzle or p1p1ng 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 containm~p.f:.
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IE Bulletin No. 79-13 Revision 2 October 16, 1979 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 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 RI
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FIGURE 2 FEEDWATER PIPE REDUCER TO STEAM GENERATOR NOZZLE INSTALL~TION {SAN ONOFRE).
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B<<ck,°':J Rl"J or.r,,.~,r H:J,./~ o/ CA?Ae K,4' -----.....i FIGURE l TYPICAL FEEDWATER PIPE TO NOZZLE WELD JOJNT DETAIL
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Salem 2 North Anna 2 Diablo Canyon 1 & 2 Sequoyah 1 McGuire 1 San Onofre 2 Summer Watts Bar 1 & 2 DESIGNATED APPLICANTS FOR OPERATING LICENSES Rl
r-IE Bulletin No. 79-13, Revision 2 October 17, 1979 Bulletin No.
79-24 79-23 79-22 79-21 79-20 79-19 79-18 79-17 79-16 79-15 (Supp. 1)
LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Subject Date Issued Frozen Lines 9/27/79 Potential Failure of Emergency 9/12/79 Diesel Generator Field Exciter Transformer Possible Leakage of Tubes of 9/5/79 Tritium Gas Used in Timepieces for Luminosity Temperature Effects on Level Measurements Packaging Low-Level Radioactive Waste for Transport and Burial Packaging Low-Level Radioactive Waste for Transport and Burial Audibility Problems Encountered on Evacuation of Personnel from High-Noise Areas Pipe Cracks in Stagnant Borated Water Systems at PWR Plants Vital Area Access Controls Deep Draft Pump Deficiencies 8/13/79 8/10/79 8/10/79 8/7 /79 7/26/79 7 /26/79 7 /18/79 Enclosure Page 1 of 3 Issued To All Power Reactor Facilities with an 01 or a CP All Power Reactor Facilities with an 01 or a CP Each Licensee who Receives Tubes of Tritium Gas in Time-pieces for Luminosity All PWR's with an Operating License All Materials Licensees who did not receive Bulletin No. 79-19 All Power and Research Reactors with OLs, Fuel Facilities except uranium mills, and certain materials licensees All OLs for Action All CPs for Information All PWRs with Operating License All Holders of and applicants for Power Reactor Operating Licenses who Antici-page loading fuel prior to 1981 All Power Reactor Licensees with a CP and/or OL
IE Bulletin No. 79-13, Revision 2 October 17, 1979 Bulletin No.
79-15 79-14 (Correction) 79-14 (Supp. 2) 79-14 (Rev. 1) 79-14 79-13 (Rev. 2) 79-13 (Rev. 1) 79-13 79-12 79-11 79-10 79-06C LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Subject Deep Draft Pump Deficiencies Seismic Analyses for As-Built Safety-Related Piping System Seismic Analyses for As-Built Safety-Related Piping System Seismic Analyses for As-Built Safety-Related Piping System Seismic Analyses for As-Built Safety-Related Piping System Cracking in Feedwater System Piping Cracking in Feedwater System Piping Cracking in Feedwater *system Piping Short Period Scrams at BWR Facilities Faulty Overcurrent Trip Device in Circuit Breakers for Engin-eered Safety Systems Requalific~tion Training Program Statistics Nuclear Incident at Three Mile Island - Supplement Date Issued 7 /11/79 7/27/79 9/7 /79 7/18/79 7 /2/79 10/17/79 8/30/79 6/25/79 5/31/79 5/22/79 5/11/79 7 /26/79 e
Enclosure Page 2 of 3 Issued To All Power Reactor
_Licensees with a CP and/or OL All Power Reactor Facilities with an OL or a CP All Power Reactor Facilities with an OL or a CP All Power Reactor Facilities with an OL or a CP All Power Reactor Facilities with an OL or a CP All PWR's with an Operating License All PWR's with an Operating License All PWR's with an OL for action.
All BWRs with a CP for information All GE BWR Facilities with an OL All Power Reactor Facilities with an OL or a CP All Power Reactor Facilities with an OL To all PWR Power Reactor Facilities with an OL
I e
IE Bulletin No. 79-13, Revision 2 October 17, 1979 Bulletin No.
79-0SC 79-02 (Rev. 1)
(Supp. 1) 79-02 (Rev. 1) 79-0lA LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Subject Date Issued Nuclear Incident at Three 7/26/79 Mile Island - Supplement Pipe Support Base Plate Designs 8/20/79 Using Concrete Expansion Anchor Bolts Pipe Support Base Plate Designs 6/21/79 Using Concrete Expansion Anchor Bolts Environmental Qualification of 6/6/79 Class lE Equipment (Deficien-cies in the Environmental Qualification of ASCO Sole-noid Valves) e Enclosure Page 3 of 3 Issued To To all PWR Power Reactor Facilities with an 01 All Power Reactor Facilities with an 01 or a CP All Power Reactor Facilities with an 01 or a CP All Power Reactor Facilities with an 01 or a CP