ML19246B140
| ML19246B140 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 06/25/1979 |
| From: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| References | |
| NUDOCS 7907120479 | |
| Download: ML19246B140 (1) | |
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[ '$..s g[%h NUCLEAR REGULATORY COMMISSION M'"rA C
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June 25, 1979 Docket Nos. 50-317 50-318 Baltimore Gas and Electric Company ATTN:
Mr. A. E. Lundvall, Jr.
Vice President, Supply P. O. Box 1475 Baltimore, Maryland 21203 Gentlemen:
Enclosed is IE Bulletin rio. 79-13 which requires action by you with regard to your 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,
. / ~doyce LM/
w rier Director
Enclosures:
1.
IE Bulletin No. 79-13 w/ Attachments 2.
List of IE Bulletins Issued in Last Twelve Months cc w/encls:
R. M. Douglass, Manager, Quality Assurance L. B. Russell, Chief Engineer W. Gibson, General Supervisor, Operat'onal QA R. C. L. Olson, Senior Engineer K. H. Sebra, Principal Engineer Y
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s ENC'_05URE 1 UtlITED STATES NUCLEAR REGULATORY C0 fit 1ISSION OFFICE OF INSPECTION AND ENFORCEMEilT WASHINGTON, D.C.
20555 IE Bulletin No. 79-13 Date: June 23,1979 Page 1 of 4 CRACKING IN FEEDWATER 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 idertified in the 16-inch feedwater elbows adjacent to two steam generator nozzle ell i 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 PUR licensees by the Office of Nuclear Reactor Regulation which infor~ed 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 generatort 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 Pcwer and Light reported that radiography showed crack indicatic.'s 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.
Nisconsin 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 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 Ca yon 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 33R Uh 7906250348
IE Bulletin No. 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 failure may be " fatigue assisted by corrosion." The San Onofre cracking was stated to be characteristic of " stress assisted corrosion."
The cracking experie ced 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 poter.tial safety consequences of the cracking is an increased likelihood of a feedwater line break in the event ';f 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 feeJwater 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-piping welds is the basis foe 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 rot 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 aftur the date of this Bulletin.
a.
Perform radiographic examination, supplemented b. ultrasonic 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 MC-5000.
Radiography shall be performe ' to the 2T penetra-meter sensitivity level, in lieu of Table NC-51.1-1, with systems void of water.
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IE Bulletin No. 79-13 Date: June 25, 1979 Page 3 of 4 b.
If cracking is identified during examination of
..c 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 oubject 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 systems, perform volumetric examination of all feedwater 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 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 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 header is employed, also inspect all welds of one inlet riser on each feed rirg of each steam generator.*
c.
Perform a visual inspection of all feedwater system piping supoorts and snubbers in containment to verify operability and conformance to design.
3.
Identification of cracking indica; ions 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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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.
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 ad 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 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.
Attachments:
Figures 1, 2 and 3
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IE Bulletin No. 79-13 Date:
June 25, 1979 Page 1 of 4 ENCLOSurF._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 Material 9/23/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12A Atypical Weld Material 11/?4/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-128 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 li.ies with an OL Solenoids (for action), and all other Power Reactor Facilities with an OL or CP (for information) n-D ') aO
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IE Bulletin No. 79-13 Date:
June 25, 1979 Page 2 of 4 LJ. STING 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 Support Base Plate 3/8/79 All Power Reactor Design Using Concrete Facilities with an CL Exoansion Anchor Bolts or CP 79-02 Same Title as 79-02 o/21/79 Same as 79-02 (Revision No. 1) 79-03 Longitudinal Weld Defects 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 Power 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) 1 l) U i
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 79-0; 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 Facilities (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 Fower 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 (Ravision 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-Related Piping Facilities with an OL or CP
)U 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- 0 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 n-
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