ML19241B209
| ML19241B209 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 06/25/1979 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Linder F DAIRYLAND POWER COOPERATIVE |
| References | |
| NUDOCS 7907130117 | |
| Download: ML19241B209 (1) | |
Text
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June 25, 1979 Docket No. 5C-409 Dairylard Pawer Cooperative ATTN:
Mr. F. W. Linder Genera? Manager 2615 East Avenue - South La Crosse, WI 54601 Gentlemen:
The enclosed IE Bulletin No. 79-13 is forwarded to you for information.
No written response is required.
If you desire additional information regatding this matter, please contact this office.
Sinceraly, rb James G. Keppler Director
Enclosures:
- 2. Listing of IE Bulletins Issued in Last 12 Months cc w/ encls:
Mr. R. E.
- Shimshak, Plart Superintendent Central Files Director, NRR/ DPM Director, NRPJDOR PDR Local PDR NSIC TIC Anthony Roisman, Esq.,
Attorney 517 042 e
7907130\\G
U.S. NUCLEAR REGULAT0nY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REC 10N 111 June 23, 1979 IE Bulletin No. 79-13 CRACKING IN FEEDW TER SYSTEM PIPING Description of Circumstances:
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 Regulatic,n which informed licensees of the D. C. Cook failures and requested specific information on teedwater system design, fa~orication, inspection and operating histories. To further explore the generic nature of the cracking problem, the Office of Inspection and Enforcement requested licensees of PRR plants in current outages to immediately conduct volumetric examination of certain feedeater 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-piping wcids 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 Seaver 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 feeduater 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.
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 pipe-to-nozzle 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 51/
043 7906250348
IE Bulletin No. 79-13 June 25, 1979 revealed an approximate 6-inch circumferential crack originating in the veld root heat-affected zone of the leaking nozzlc 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.
Cracked weldments have been removed from the D. C. Cook Units 1 and 2 and San Onofre Unit 1 feedwater systems for extensive eetallurgical investigation by Westinghouse. Based on preliminary analysis, Westinghouse stated the D. C. Cook failure may be " fatigue assisted by corrosion." The San Onofre cracking was stated to be characteristic of " stress assisted corrosion."
The cracking experienced at Tiablo Canyon, D. C. Cook and San Onofre would appear to have different cau se - effect relationships which are not fully understood at this time.
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 ha==er.
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 contain=ent. 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 chese feedwater nozzle-to piping welds is the basis for this Bulletin.
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 this Bulletin.
Perform radiographic examination, supplemented by ultrasonic a.
examination as necessary to evaluate indications, of all feedwater nozzle-to-piping 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 penetra-meter sensitivity level, in lieu of Table NC-5111-1, with systems void of water.
517 044
IE Bulletin No. 79-13 June 25, 1979 b.
If cracking is identified during examination of the nozzle-to piping weld, all feedwater line welds up to the first piping support or snubber and high stress points in containment shall be volumet-rically examined in accordance with 1.a. above. All unacceptable code discontinuities, other than cracking, 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.
a.
For steam generator designs having a common nozzle for both main and auxiliary (emergency) feedwater syste=s, perform volumetric examination of all feedwater nozzle-to-pipe veld areas and all feed-water pipe weld areas inside contcinment in accordance with item 1 above.1 In additi:n, conduct an eximination of welds connecting auxiliary feedwater piping to the main feedwater line outside con-tainment. This examination should include an area of at least one pipe diameter on the main feedwater line downstream of the connection.
b.
For steam generator designs with separate nozzles fcr main feedwater and auxiliary feedwater, perform volumetric examination (in accordance with item 1 above) of all welds inside containment and upstream of the external ring header or vessel nozzle for each steam generator.
If an external ring header is employed, also inspect all welds of one inlet riser on each feed ring of each steam generator.1 Perform a visual inspection of all feedwater system piping supports c.
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.
1 Welds in the feedwater system, (other than the feedwater nozzle-to-pipe welds) that have been examined since May 1979 need not be re-examined.
517 045
IE Bulletin No. 79-13 June 25, 1979 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 this Bulletin 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.
6.
A written report of the resalts of examinations, 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 this Bulletin 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.
Approved by GAO, B180225 (R0072), clearance expires 7/31/80. Approval was given under a blanket clearance specifically for identified generic problems.
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IE Sulletin No. 79-13 June 25, 1979 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
79-02 Pipe Support Base Plate 6/21/79 All Power Reactor (Rev. 1)
Designs Using Concrete Facilities with an Expansion Anchor Bolts OL or a CP 79-12 Short Period Scrans at 5/31/79 All GE BWR Facilities BWR Facilities with an OL 79-11 Faulty overcurrent Trip 5/22/79 All Power Reactor Device in Circuit Breakers Facilities with an for Engineered Safety OL or a CP Systems 79-10 Requalification Training 5/11/79 All Power Reactor Program Statistics Facilities with an OL 79-09 Failures of GE Type AK-2 4/17/79 All Power Reactor Circuit Breaker in Safety Facilities with an Related Systems OL or CP 79-08 Events Relevant to BWR 4/14/79 All BWR Power Reactor Reactors Identified During Facilities with an OL Three Mile Island Incident 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an OL or CP 79-06B Review of Operational 4/14/79 All Combustion Engineer-Errors and System Mis-ing Designed Pressurized alignments Identified Water Power Reactor During the Three Mile Facilities with an Island Incident Operating Licensee 517 049 Enclosure Page 1 of 4
IE Bulletin No. 79-13 June 25, 1979 LISTING-OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.79-06A Review of Operational 4/18/79 All Pressurized Water (Rev 1)
Errors and System Mis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with an OL Island Incident 79-06A Review of Operational 4/14/79 All Pressurized Water Errors and System Mis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with aa OL Island Incident 79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Mis-Pcwcr Reactors with an align =ents Identified OL except B&W facilities During the Three Mile Island Incident 79-05A Nuclear Incident at 4/5/79 All B&W Power Reactor Three Mile Island Facilities with an OL 79-05 Nuclear Incident at 4/2/79 All Power Reactor Three Mile Island Facilities with an OL and CP 79-04 Incorrect Weights for 3/30/79 All Power Reactor Swing Check Valves Facilities with an Manufactured by Velan OL or CP Engineering Corporation 517 050 Enclosure Page 2 of 4
IE Bulletin No. 79-13 June 25, 1979 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.78-12B Atypical Weld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 79-03 Longitudinal Welds Defects 3/12/79 All Power Reactor In ASME SA-312 Type 304 Facilities with an Stainless Steel Pipe Spools OL or CP Manufactured by Youngstown Welding and Engineering Co.
79-02 Pipe Support Base Plate 3/2/70 All Power Reactor Designs Using Concrete Facilities with an Expansion Anchor Bolts OL or CP 79-01A Environmental Qualification 6/6/79 All Power Reactor of Class 1E Equipment Facilities with an (Deficiencies in the Envi-OL or CP ronmental Qualification of ASCO Solenoid Valves) 79-01 Lnviron= ental Qualification 2/8/79 All Power Reactor of Class IE Equipment Facilities with an OL or CP 78-14 Deterioration of Buna-N 12/19/78 All GE SWR facilities Component In ASCO with an OL or Cf Solenoids 78-13 Failures in Source Heads 10/27/78 All general and of Kay-Ray, Inc., Gauges specific licensees Models 7050, 7050B, 7051, with the subject 7051B, 7060, 7060B, 7061 Kay-Ray, Inc.
and 7061B gauges n,
m 3 \\
Enclosure Page 3 of 4
F IE Bulletin No. 79-13 June 25, 1979 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.78-12A Atygical Weld Material 11/24/78 All Power Eeactor in Reactor Pressure Facilities with an Vesse'. Welds OL or CP 78-12 Atypical Weld Material 9/29/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-11 Exa:11 nation of Mark 1 7/21/78 BWR Power Reactor Con:ainment Torus 'a' elds Facilities for action:
Peach Bottom 2 and 3, Quad Cities 1 and 2, Hatch 1, Monticello and Vermont Yankee 78-10 Bergen-Paterson Hydraulic 6/27/78 All BWR Power Reactor Shock Suppressor Accumulator Facilities with an Spring Coils OL or CP a
Enclosure Page 4 of 4