ML19246B146
| ML19246B146 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 06/25/1979 |
| From: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Groce R Maine Yankee |
| References | |
| NUDOCS 7907120504 | |
| Download: ML19246B146 (1) | |
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KING OF PRUSSIA, PENNSYLVANI A 19406 June 25, 1979 Docket No. 50-309 Maine Yankee Atomic Power Company ATTN:
Mr. Robert H. Groce Licensing Engineer 20 Turnpike Road Westborough, Massachusetts 01581 Gentlemen:
Enclosed is IE Bulletin No. 79-13 which requires action by you with regard to
,our power reactor facility (ies) with an operating license.
Should you have any questions regarding this Bulletin or the actions required by you, please contact this office.
Sincerely, d
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- Boyce H. Grier p
Director
Enclosures:
1.
IE Bulletin No. 79-13 w/ Attachments 2.
List of IE Bulletins Issued in Last Twelve Months cc w/encls:
E. Wood, Plant Superintendent E. W. Thurlow, President x\\ n -
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ENCLOSURE 1 UNITED STATES NUCLEAR REGULATORY C0ffiISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 IE Bulletin No. 79-13 Date: June E3, 1979 Page 1 of 4 CRACKING IN FEEDWATER SYSTEM PIPING Description of Circumstances:
On fiay 20, 1979, Indiana and Michigan Power Company notified the NHC 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-piping welds on two of three steam generators of San Onofre Unit 1.
On June 15, 1979, Carolina Power and Light recorted 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. Misconsin 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 15, 1979 seven other PUR 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 pioe-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 z :> o,n.
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IE Bulletin flo. 79-13 Date: June 25,1979 Page 2 of 4 revealed an approximate 6-inch circumferential crack originating in the weld root heat-affected zone of the leaking nozzle weld. The cause of this cracking 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 metallurgical investigation by Westinghouse.
Based on preliminary analysis, Westinghouse stated the D. C. Cook Tailure may be " fatigue assisted by corrosion." The San Onofre crackir.g was stated to be characteristic of " stress assisted corrosion."
The cracking experienced at Diablo Canyon, D. C. Cook and San Onofre would appear to have different cause - 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 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-oiping 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 volun tric 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 tnis 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 flC-5000.
Radiography shall be performed t.o the 2T penetra-meter sensitivity level, in lieu of Table NC-5111-i, w.th systems void of water.
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)'I IE Bulletin No. 79-13 Date: June 25, 1979 Page 3 of 4 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 insoection program described below at the next outage of sufficient duration or at the next refueling outage after the inspection reauired by item 1.
a.
For steam generator designs having a common nozzle for both main and auxiliary (emergency) feedwater systems, perform volumetric examination of all *eedwater nozzle-to-pipe weld areas and all feed-water pipe weld areas inside containment in accordance with item 1 above.* In addition, conduct an examination of welds connecting auxiliary feedwater piping to '.he main feedwater line outside con-tainment. This exaninatioi uld include an area of at least one 3
pipe diameter on the main tec ter line downstream of the connection.
b.
For steam generator designs with separate nozzles for 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 leader is employed, also inspect all welds of one inlet riser on each feed ring of each steam generator.*
Perform a visual inspection of all feedwater system piping supoorts 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.
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.
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]O IE Bulletin No. 79-13 Date:
June 25, 1979 Page 4 of 4 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.
iour 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 results of examinations, in accordance with requests by Regional Offices preceding this Bulletin and with Bulletin item 1 and 2 including any cnrrective 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.
Ati.r hments:
Figures 1, 2 and 3
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IE Bulletin No. 79-13 Date:
June 25, 1979 Page 1 of 4 ENCLOSURE 2 LISTING 0F IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
78-10 Bergen-Paterson 6/27/78 All BWR Power Reactor Hydraulic Shock Facilities with Suppressor Accumulator an OL or CP Spring Coils 78-11 Examination of Mark I 7/24/78 BWR Power Reactor Containment Torus Facilities with an OL Welds for action:
Peach Bottom 2 and 3, Quad Cities 1 and 2, Hatch 1, Monticello and Vermont Yankee.
All other BWR Power Reactor Facilities with an OL for information 78-12 Atypical Weld Aaterial 9/29/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12A Atypical Weld Material 11/24/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12B Atypical Weld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 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 70518, 7060, 70608, 7061 Kay-Ray, Inc.
and 7061B Gauges 78-14 Deterioration of Buna-N 12/19/78 All GE BWR Faci-Components In ASCO lities with an OL Solenoids (for action), and all other Power Reactor Facilities with an OL or CP (for information) zn a,
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IE Bulletin No. 79-13 Date:
June 25, 1979 Page 2 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS (CONTINUED)
Bulletin Subject Date Issued Issued To No.
79-01 Environmental Qualif-2/8/79 All Power Reactor ication of Class IE Facilities with an OL, Equipment except the 11 Systematic Evaluation Program Plants (for action), and all other Power Reactor Facilities with an OL or CP (For Information)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) 79-02 Pipe 5upport Base Plate 3/8/79 All Power Reactor Design Using Concrete Facilities with an OL Expansion Anchor Bolts or CP 79-02 Same Title as 79-02 6/21/79 Same as 79-02 (Revision No. 1) 79-03 Longitudinal Weld Detects 3/12/79 All Power Reactor in ASME SA-312 Type Facilities with 304 Stainless Steel Pipe an OL or CP Spools Manufactured by Youngstown Welding and Engineering Company 79-04 Incorrect Weights for 3/30/79 All Pover Reactor Swing Check Valves Facilities with an Manufactured by Velan OL or CP Engineering Corporation 79-05 Nuclear Incident at 4/1/79 All Babcock and Three Mile Island Wilcox Power Reactor Facilities with an OL, Except Three Mile Island 1 and 2 (For Action),
and All Other Power Reactor Facilities With an OL or CP (For Information) o
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s IE Bulletin No. 79-13 Date:
June 25, 1979 Page 3 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS (CONTINUED)
Bulletin Subject Date Issued Issued to No.79-05A Nuclear Incident at 4/5/79 Same as 79-05 Three Mile Island -
Supplement 79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Mis-Power Reactor Facil-alignments Identified ities with an OL Except During the Three Mile B&W Facili'ies (For Incident Action), All Other Power Reactor Facil-ities with an OL or CP (For Information)79-06A Same Title as 79-06 4/14/79 All Westinghouse Designed Pressurized Power Reactor Facil-ities with an OL (For Action), and All Other Power Reactor Facilities with an OL or CP (For Information)79-06A Same Title as 79-06 4/18/79 All Westinghouse (Revision 1)
Designed Pressurized Power Reactor Facil-ities with an OL (For Action), and All Other Power Reactor Facilities with an OL or CP (For Information)79-06B Same Title as 79-06 4/14/79 All Combustion Engineering Designed Pressurized Power Reactor Facilities with an OL (For Action), and All
'Other Power Reactor Facilities with an OL or CP (For Information) 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-kelated Piping Facilities with an OL or CP z,, q q- -
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IE Bulletin No. 79-13 Date:
June 25, 1979 Page 4 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS (CONTINUED)
Bulletin Subject Date Issued Issued to No.
79-08 Events Relevant to 4/14/79 All BWR Power Boiling Water Power Reactor Facilities Reactors Identified with an OL (For During Three Mile Action), All Other Island Incident Power Reactor Facil-ities with an OL or CP (For Information) 79-09 Failures of GE Type 4/17/79 All Power Reactor AK-2 Type Circuit Facilities with an Breaker in Safety OL or CP Related Systems 79-10 Requalification Training 5/11/79 All Power Reactor Program Statistics 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 CP Systems 79-12 Short Period Scrams at 5/31/79 All GE BWR Facilities BWR Facilities with an OL
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