ML18079B018
| ML18079B018 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/30/1979 |
| From: | Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Martin T Public Service Enterprise Group |
| References | |
| NUDOCS 7910020327 | |
| Download: ML18079B018 (13) | |
Text
e UNITED STATES e
NUCLEAR REGULATORY COMMISSION REGION I 631 PARK AVENUE KING OF PRUSSIA, PENNSYLVANIA 19406 Docket Nos. 50-311 Public Service Electric & Gas Company ATTN:
Mr. T. J. Martin Vice President Engineering and Construction 80 Park.Place Newark, New Jersey 07101 Gentlemen:
Enclosed is IE Bulletin 79-13, Revision 1, which requires action by you with regard to your power reactor facility as a designated applicant for an operating license.
Should you have any questions regarding this Bulletin or the actions required by you, please contact this office.
Enclosures:
Sincerely,
/J,W.~
k1Boyce H. Grier Vv v Director
- 1.
IE Bulletin No. 79-13, Revision 1
- 2.
Li st of IE Bulletins Issued in Last Six -
Months cc w/encl s:
E. N. Schwalje, Manager - Quality Assurance,
._~-Eng:i neeri ng and Construction Department
ENCLOSURE 1 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 CRACKING IN FEEDWATER SYSTEM PIPING Description of Circumstances:
IE Bulletin No. 79-13 Revision 1 Date:
August 30, 1979 Page 1 of 4 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 contain-ment.
Leaking circumferential cracks were identified in the 16-inch feedwater elbows adjacent to two steam generator nozzle elbow welds.
Subsequent radio-graphic examination revealed crack indications in all eight steam generator f.eedwater 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 Offi~e 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 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.
Public Service Electric and Gas Company reported on June 20, 197g 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 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.
e IE Bulletin No. 79-13 Revision 1
- Date: August 30, 1979 Page 2 of 4 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.
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 11fatigue assisted by corrosion.
11 The San Onofre cracking was stated to be characteristic of "stress assisted corrosion. 11 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-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.
- a.
Perform radiographic examination, supplemented by ultrasonic examin-ation 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 penetrameter sensitivity level, in lieu of Table NC-5111-1, with systems void of water.
e IE Bulletin No. 79-13 Revision 1 Date:
August 30, 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 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 examina-tion of all feedwater nozzle-to-pipe weld areas and all feedwater pipe weld areas inside containment in accordance with item 1 above.l/
In addition, conduct an examination of welds connecting auxiliary -
feedwater piping to the main feedwater line outside containment.
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 ring of each steam generator.!/
- c.
Perform a visual inspection of all *feedwater system piping supports and snubbers in containment to verify operability and conformance to design.
- 3.
Identification of cracking indications in feedwater nozzle or piping weld areas i.n 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.
1/ Welds in the feedwater system, (other than the feedwater nozzle-to-pipe welds) that have been examined since May 1979 need not be reexamined.
e IE Bulletin No. 79-13 Revision 1 Date:
August 30, 1979 Page 4 of 4
- 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 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 comple-tion of the examination, whichever is later, to the Director of the appropriate NRC Regional Office with a copy to the NRC Office of Inspec-tion 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 and 6 above.
1 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
- t
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IE Bulletin Revisio. No. 79-13 Date*.
ust 30, 1979 FEEDWATER PIP FIGURE 1 NOZZLE r~siksow*ro STEAM LLATlON {D~ GENERATOR
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Bulletin No.78-128 79-0lA 79-02 (Rev 1) 79-02 (Rev 1)
(Supplement No. 1) 79-03 79-04 79-05 ENCLOSURE 2 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Subject Atypical Weld Material in Reactor Pressure Vessel Welds Date Issued 3/19/79 Environmental Qualification 6/6/79 of Class lE Equipment (Deficiencies in the Envi-ronmental Qualification of ASCO Solenoid Valves)
Pipe Support Base Plate 6/21/79 Design Using Concrete Expansion Anchor Bolts Same Title as 79-02 8/20/79 (Rev 1)
Longitudinal Weld Defects 3/12/79 in ASME SA-312 Type 304 Stainless Steel Pipe Spools Manufactured by Youngstown Welding and Engineering Company Incorrect Weights for 3/30/79 Swing Check Valves Manufactured by Velan Engineering Corporation Nuclear Incident at 4/1/79 Three Mile Island e
IE Bulletin No. 79-13 Revision No. 1 Date:
August 30, 1979 Page 1 of 5 Issued To All Power Reactor Facilities with an OL or CP
or CP Same as 79-02 (Rev 1)
All Power Reactor Facilities with an OL or CP All Power Reactor Facilities with an OL or CP All Babcock and 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)
LIST OF IE BULLETINS ISSUED IN LAST SIX MONTHS (CONTINUED)
Bulletin Subject No.79-05A Nuclear Incident at Three Mile Island -
Supplement 79-05B Nuclear Incident at Three Mile Island -
Supplement 79-06 Review of Operational Errors and System Mis-alignments Identified During the Three Mile Incident 79-06A Same Title.as 79-06 79-06A Same Title as 79-06 (Revision 1)
Date Issued 4/5/79 5/21/79 4/11/79 4/14/79 4/18/79 e
IE Bulletin No. 79-13 Revision No. 1 Date:
August 30, 1979 Page 2 of 5 Issued To Same as 79-05 Same as 79-05 All Pressurized Water Power Reactor Facil-ities with an OL Except B&W Facilities (For Action), All Other Power Reactor Facil-ities with an OL or CP (For Information)
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)
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)
LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS (CONTINUED)
Bulletin Subject Date Issued No.79-06B Same Title as 79-06 4/14/79 79-05C&06C Nuclear Incident at 7/26/79 Three Mile Island -
Supplement 79-07 Seismic Stress Analysis 4/14/79 of Safety-Related Piping 79-08 Events Relevant to 4/14/79 Boiling Water Power Reactors Identified During Three Mile Island Incident 79-09 Fai 1 ures of GE Type 4/17 /79 AK-2 Type Circuit Breaker in Safety Related Systems 79-10 Requalification Training 5/11/79 Program Statistics e
IE Bulletin No. 79-13 Revision No. 1 Date:
August 30, 1979 Page 3 of 5 Issued To All Combustion Engineering Designed Pressurized Power Re@ctor Facilities with an OL (For Act ion), and All Other Power Reactor Facilities with an OL or CP (For Information)
All PWR Power Reactor Facilities with an OL All Power Reactor Facilities with an OL or CP All BWR Power Reactor Facilities with an OL (For Action), All Other Power Reactor Facil-ities with an OL or CP (For Information)
All Power Reactor Facilities with an OL or CP All Power Reactor Facilities with an OL
LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS (CONTINUED)
- Bulletin Subject Date Issued No.
79-11 Faulty Overcurrent Trip 5/22/79 Device in Circuit Breakers for Engineered Safety Systems 79-12 Short Period Scrams at 5/31/79 BWR F aci 1 it i es 79-13 Cracking in Feedwater 6/25/79 System Piping 79-14 Seismic Analysis for 7/2/79 As-Built Safety Related Piping Systems 79-14 Same Title as 79-14 7/18/79 (Revision 1) 79-14 Same Title as 79-14 8/15/79 (Supplement) 79-15 Deep Draft Pump Defi-7/11/79 ciencies 79-16 Vital Area Access Con-7/30/79 trols 79-17 Pipe Cracks in Stagnant 7/26/79 Borated Water Systems at PWR Plants 79-18 Audibility Problems 8/7/79 Encountered on Evacu-ation e
IE Bulletin No.* 79-13 Revision No. 1 Date:
August 30, 1979 Page 4 of 5 Issued To All Power Reactor Facilities with an OL or CP All GE BWR Facilities with an OL All PWRs wi~h an OL (for Action),
All Other Power Reactor Facilities with an OL or CP (For Information)
All Power Reactor Facilities with an OL or CP Same as 79-14 Same as 79-14 All Power Reactor Facilities with an OL or CP All Holders of and Applicants for Reactor Operating Licenses All PWR Power Reactor Facilities with an OL All Power Reactor Faci-lities with an OL
Bulletin No.
79-19 79-20 79-21 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS (CONTINUED)
Subject Packaging Low-Level Radioactive Waste for Transport and Burial Same Title as 79-19 Temperature Effects on Level Measurements Date Issued 8/10/79 8/10/79 8/13/79 IE l11etin No. 79-13 Revision No. 1 Date:
August 30, 1979 Page 5 of 5 rssued To All Power and Re-search Reactors with OL, all Fuel Faci-1 i ti es (except Uranium Mills), and certain Materials Licensees Certain Materials Licensees All Power Reactor Facilities with an OL or CP