ML16340A620
| ML16340A620 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 10/19/1979 |
| From: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Crane P PACIFIC GAS & ELECTRIC CO. |
| References | |
| NUDOCS 7910310268 | |
| Download: ML16340A620 (26) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION REGION V i990 N. CALIFORNIABOULEVARD SUITE 202, WALNUTCREEK PLAZA WALNUTCREEK, CALIFORNIA94596 October 17, 1979 Docket No. 50-275 50-323 Pacific Gas and Electric Company 77 Beale Street San Francisco, California 94106 Attention:
t<r. Phi Iip A. Crane, Jr.
Assistant General Counsel 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 re-quired by you, please contact this office.
Sincerely~
R. H. Engelken Director
Enclosures:
1.
IE Bulletin No. 79-13, Revision 2
2.
List of IE Bulletins Issued in the Last Six tIonths cc w/enclosures:
W. Raymond, PG5E J. Worthington, PG&E R.
- Ramsay, PGIE, Diablo Canyon 6F V 910810
SSINS:
6830 Accession Ho.:
7908220135 UNITED STATES NUCLEAR REGULATORY COt1t'1ISSION OFFICE OF INSPECTION AHD ENFORCEt1EHT WASHINGTON, D.C.
20555 October 16, 1979 IE Bulletin No. 79-13 Revision 2
CRACKING IH FEEDMATER SYSTEt1 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
'umber 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 thichigan 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 l1ay 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 Hay 25,
- 1979, a letter was sent to all PWR licensees by the'gOffice of Huclear 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 generi'c 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 28.
1979, that radiography has shown cracking'in their Beaver Valley Unit 3. 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 Hay 25, 1979 seven other PWR facilities have inspected the feedwater nozzle-to-pipe welds without finding cracking indications.
NOTE:
Rl and R2 indicates lines r vised or added.
IE Bulletin No. 79-13 Revision 2
October 16, 1979 Page 2 of 5 The feedwater nozzle-to-oipe configurations for D. C. Cook and for San Onofre are shown on the attached figures I 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 I, 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 ivelds by radiograohy 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 hamaer.
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 ultr'qsonic methods, have id'entified cracking in the steam generator nozzle to feedwatler piping weldments at the following M, and C.
E. plants D. C.
Cook Unigs I 5 2 Diablo Canyon San Onofre Unit I H. B. Robinson Unit 2 Beaver Valley Unit I Kewaunee Point Beach Unit 2 Found during hot functional testing Confirmatory evaluation incomplete Salem Unit I Surry Unit I R. E. Ginna Millstone Unit 2 Pal isades Yankee Rowe Maine Yankee An extensive metallurgical investigation has been conducted by >lestinghouse 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 eedwater piping at D. C.
Cook, H. B. Robinson, R.
E. Ginna, Salem I and other plants have been instrumented (Thermocouples, accelerometers, strain
- gages, and transducers) to co1lect data
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 tend's 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 8&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 considered acceptable to meet this intent of IE Bulletin No. 79 Actions to be Taken b
Licensees R2 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 Hay 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 6y ultrasonic examination as necessary to evaluate indications",gof all feedwater nozzle-to-pipe welds and of adjacent pipe and nozzle areas (a
constance equal to at least two wall thicknesses).
Evaluation shall be in accordance with ASHE Section III., Subsection NC, Article NC-5000.
Radiography shall be performed to the 2T penetrameter sensitivi ty 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 l.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.
0
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.
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.
3.
For designs utilizing auxiliary feedwater systems connected to the steam generator by means o" 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.
'v%
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 Hay 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 l.
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'ulletin 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 item 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 Enforcemen't, Division of Reactor Operations Inspection, Mashington, D.'. 20555.
Actions to be Taken b
Desi nated A
licants for 0 eratin Licenses:
l.
On com letion of the hot functional.testin ro ram and rior to fuel loadin erform the ins ections described in item 1 above.
2.
Durin the first refue'lin outa e, erform.the ins ections described in Rl item 2 above.
Submit re orts as described in Items 4, 5, amd 6 above based on the date o
Revision 1 to Bulletin No. 79-13 Au ust 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
I
DESIGNATED APPLICANTS FOR OPERATING LICENSES Salem 2
North Anna 2
Diablo Canyon 1
& 2 Sequoyah 1
NcGuire 1
San Onofre 2
Summer Watts Bar 1
8 2
IE Bulletin No. 79-13, Revision 2
October 16; 1979 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Enclosure Page 1 of 3 Bulletin No.
'9-24 79-23 Subject Frozen Lines Potential Failure of Emergency Diesel Generator Field Exciter Transformer Date Issued 9/27/79 9/12/79 Issued To All Power Reactor facilities which have either OLs or CPs and are in late stage of construction All Power Reactor Facilities with an Operating License or a construction oermit 79-14 Seismic Analyses For (Supolement
- 2) As-Built Safety-Related Piping Systems 9/7/79 All Power Reactor Facilities with an OLor a CP 79-22 Possible Leakage of Tubes 9/5/79 o
Tritium Gas in Time-pieces for Luminosity To Each Licensee who Receives Tubes of Tritium Gas
'; Used in Timepieces
"/or Luminosity 79-13 (Pev.
1)
Cracking in Feedwater System Piping 79-02 Pipe Support Base Plate (Rev.
1)
Designs Using Concrete (Supplement
- 1) Expansion Anchor Bolts 8/30/79 8/20/79 All Designated Applicants for OLs All power Reactor Facilities with an OL or a CP 79-14 (Su ppl ement) 79-21 79-20 79-19 Seismic Analyses For As-Built Safety-Related Piping Systems Temperature Effects on Level Measurements Packaging Low-Level Radioactive Haste for Transport and Burial Packaging Low-Level Radioactive Maste for Transport and Burial 8/15/79 8/13/79 8/10/79 8/10/79 All Power Reactor Facilities with anOLor aCP All PNRs with an operating license All ?laterials Licensees who did not receive Bulletin No. 79-19 All Power and Pesearch Reactors with OLs, fuel facilities except uranium mills, and certa-n. materials licensees
IF. Bulletin No. 79-13, Revision 2
October 16, 1979 Encl osur e Page 2 of 3 Bulletin No.
Subject LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Date Issued Issued To 79-18 Audibility Probl ems Encountered on Evacuation 8/7/79 79-05CE06C Nuclear Incident at Three 7/26/79 Nile Island - Supplement All Power Reactor Facilities with an Operating License To all PVR Power Reactor Facilities with an OL 79-17 Pipe Cracks in Stagnant Borated Hater Systems at PHR Plants 7/26/79 All P!AR's with operating license 79-16 Vital Area Access Controls 7/26/79 7/18/79 79-14; Seismic Analyses For (Revision 1)
As-Built Safety-Related Piping System All Holders of'nd applicants for Power Reactor Operating Licenses who anticipate
'. loading fuel prior to
.'.1981 All Power Reactor Facilities with an OL or a CP
, 79-15 79-14 79-13 79-02 (Rev.
1)79-01A Deep Draft Pump Deficiencies Seismic Analyses for As-Built Safety-Related Piping System Cracking In Feedwater System Piping Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts Environmental (jual ificati on of Class 1E Eauipment
(".eficiencies in the ".-nvi-ronmental Oual lricat >0 or ASCO Solenoid Valves) 7/11/79 7/2/79 6/25/79 6/21/79 6/6/79 All Power Reactor Licensees with a CP and/or OL All Power Reactor faci 1 ities with an OL or a CP All PHRs with an OL for action. All B'~jRs with a CP'for information.
All Power Reactor Facilities wi h an OL or a CP All Power Reactor Facilities wi h an nr CP
0
IE Bulletin No. 79-13, Revision 2
October 16, 19?9 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Enclosure Page 3 of 3 Bulletin No.
79-12 79-11 79-10 Subject Short, Period. Scrams at B'HR Facilities.
Faulty Overcurrent Trip Device in Circuit Breakers for Engineered Safety Systems Requalification Training
, Program Statistics Date Issued 5/31/79 5/22/79 5/11/79 Issued To All GE BMR Facilities with an OL All Power Reactor Facilities with an OL or a CP All Power Reactor Facilities with an OL
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