ML17053B161
| ML17053B161 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 10/17/1979 |
| From: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Schneider R NIAGARA MOHAWK POWER CORP. |
| References | |
| NUDOCS 7911070314 | |
| Download: ML17053B161 (24) | |
Text
t UNITED STATES NUCLEAR REGULATORY COMMISSION REGiON I
631 PARK AVENUE KING OF PRUSSIA, PENNSYLVANIA1940G Docket No. 50-220 October 17, 1979 Niagara Mohawk Power Corporation ATTN:
Mr.
R.
R. Schneider Vice President Electric Operations 300 Erie Boulevard West
- Syracuse, New York 13202 Gentlemen:
The enclosed IE Bulletin 79-13, Revision 2, is forwarded to you for information.
No written response is required.
If you desire additional information regarding this matter, please contact this office.
Sincerely,
~cyce H. Brier Director
Enclosures:
l.
'IE Bulletin No. 79-13 w/Attachments 2.
Listing of IE Bulletins Issued in Last 6 Months cc w/encls:
T. E. Lempges, General Superintendent, Nuclear Generation T. J. Perkins, Station Superintendent C. L. Stuart, Operations Supervisor E.
B. Thomas, Jr.,
Esquire CONTACT:
L.
E. Tripp 337-5282 79110703 /9
ENCLOSURE 1
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 SS INS:
6830 Accession No.:
7908220135 CRACKING IN FEEDWATER SYSTEM PIPING Descri ption of Circumstances:
October 17, 1979 IE Bulletin No. 79-13 Revision 2
Page 1 of 5 This revision to IE Bulletin No. 79-13 is based on the results of the radiographic examinations and ongoing investigation of the subject problem to date since the initial Bulletin was issued.
The revision reduces in scope the number 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 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 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, inspec-tion 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-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 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.
Wisconsin Public Service company decided on June 20, 1979 to cut out a feedwater nozzle-to-pipe weld which contained question-able 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.
NOTE:
Rl and R2 indicates lines revised or added
Enclosure 1
IE Bulletin No. 79-13 Revision 2
October 17, 1979 Page 2 of 5 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 nozzle-to-pipe weld joint detail showing the principal crack locations for D.
C.
Cook and San Onofre are shown on the attached fiqure 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 nondestruc-tive 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 cracking was identified as either corrosion fatigue or thermal fatique initiating at small cracks probablv 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 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-pipe welds formed the basis of this Bulletin.
To date the radiographic examinations, supplemented by ultrasonic methods, have identified cracking in the steam generator nozzle to feedwater piping weldments at the following W, and C.
E. plants.
D. C.
Cook Units 1
8 2
Diablo Canyon*
San Onofre Unit 1 H. B. Robinson Unit 2 Beaver Valley Unit 1 Kewaunee Point Beach Unit 2 Salem Unit 1 Surry Unit 1
R.
E. Ginna Millstone Unit 2 Pal isades Yankee Rowe**
Maine Yankee**
Found during hot functional testing Confirmatory evaluation incomplete An extensive metallurgical investigation has been conducted by Westinghouse 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.
R2
Enclosure 1
I'u11etin No. 79-13 Rev is ion 2
October 1 7,
1 979 Page 3 of 5 In parallel with the above ongoing analysis, the feedwater piping at D. C. Cook, H. B. Robinson, R.
E. Ginna, Salem 1 and other plants have been instrumented (Thermocouples, accelerometers, strain gages, and transducers) to collect data 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 temper-ature conditions in the feedpipe weld region during zero and low power operations.
This gradient tends 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 B8W plants a total of 95 welds in the main and separate auxiliary R
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 con-R2 sidered acceptable to meet this intent of IE Bulletin No. 79-13.
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 Bulletin No. 79-13.
a.
Perform radiographic examination, supplemented by ultrasonic examina-.
tion as necessary to evaluate indications, of all feedwater nozzle-to-pipe 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.
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 1.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.
Enclosure 1
I ulletin No. 79-13 Revision 2
October 17, 1979 Page 4 of 5 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 0 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 suppor t, 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 examinati'on of all auxiliary feedwater nozzle to piping welds and the first adjacent outboard pipe-to-pipe welds (risers) in accordance with item 1 above.
For designs utilizing auxiliary feedwater systems connected to the steam generator by means of bolted flange connections, perform volu-metric examination of the flanged nozzle to piping and first outboard pipe-to-pipe welds (risers) in accordance with item I 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.
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.
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.
4 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 1.
Enclosure 1
t IE Bulletin No. 79-13 Revision 2
October 17, 1979 Page 5 of 5 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'xamination, 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 oriqinal 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 Enforcement, Division of Reactor Operations Inspection, Washington, D.
C. 20555.
Actions to be Taken b
Desi nated A
licants for 0 eratin Licenses:
1.
On com letion of the hot functional testin ro ram and rior to fuel oa sn erform t e sns ectsons escribed in stem a ove.
2.
Durin the first refuelin outa e, erform the ins ections described in R
item 2 a ove.
3.
Submit re orts as described in Items 4, 5, amd 6 above based on the date of Revssson 1 to Bu eton o.
9-13 Au ust 3
9 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
Date:
ctobev 17, 1979 Ch
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Zk'L+7 DSrug F/6 5 FIGURE 1
FEEDMATER PIPE ELBOW TO STEAN GENERATOR NOZZLE INSTALLATION (DC COOK)
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FIGURE2 FEEDWTER PIPE REOUCH TO STBN GENERATOR NOZZLE IttSTALLATXN (SAN )HOFRE).
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Bulletin Subject No.
ENCLOSURE 2
LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Date Issued IE Bulletin No. 79-13 Revision No.
2 Date:
October 17, 1979 Page 1 of 3 Issued To 79-10 79-11 Faulty Overcurrent Trip Device in Circuit, Breakers for Engineered Safety Systems 5/22/79 Requal ifi cation Training 5/1 1/79 Program Statistics All Power Reactor Facilities with an OL All Power Reactor Facilities with an OL or CP 79-12 Short Period Scrams at BWR Facilities 5/31/79 All GE BWR Facilities with an OL 79-01A Environmental guali fication 6/6/79 of Class lE Equipment (Deficiencies in the Envi-ronmental qua 1 ification of ASCO Solenoid Yalves)
All Power Reactor Facilities with an OL or CP 79-02 (Rev 1) 79-13 79-14 Pipe Support Base Plate Design Using Concrete Expansion Anchor Bolts Cracking in Feedwater System Piping Seismic Analysis for As-Built Safety Related Piping Systems 6/21/79 6/25/79 7/2/79 All Power Reactor Facilities with an OL or CP All PWRs with an OL (for Action),
All Other Power Reactor Facilities with an OL or CP (For Information)
All Power Reactor.
LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS (CONTINUED)
IE Bulletin No. 79-13 Revision No.
2 Date:
October 17, 1979 Page 2 of 3 Bulletin No.
79-15
,79-.14 (Revision 1)
Subject Deep Draft Pump Defi-ciencies Same Title as 79-14 Date Issued 7/11/79 7/18/79 Issued To All Power Reactor Facilities with an OL or CP Same as 79-14 79-16 79-17 Vital Area Access Con-7/30/79 trol s Pipe Cracks in Stagnant 7/26/79 Borated Water Systems at PWR Plants All Holders of and Applicants for Reactor Operating Licenses All PWR Power Reactor Facilities with an OL 79-OSC506C Nuclear Incident at Three Mile Island-Supplement 7/26/79 All PWR Power Reactor Facilities with an OL 79-18 Audibi 1 ity Problems 8/7/79 Encountered on Evacuation All Power Reactor Facilities with an OL 79-19 79-20 Packaging Low-Level Radioactive Waste for Transport and Burial Same Title as 79-19 8/10/79 8/13/79 All Power and Re-search Reactors with OL, all Fuel Facilities (except Uranium Mills),
and certain Materials Licensees Certain Materials Licensees 79-21 Temperature Effects on 8/13/79 Level Measurements All Power Reactor Facilities with an OL or CP
LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS (CONTINUED)
IE Bulletin No. 79-13 Revision No.
2 Date:
October 17, 1979 Page 3 of 3 Bulletin No.
Subject Date Issued Issued To 79-02 (Rev 1)
(Supplement No.
1)
Same Title as 79-02 79-14 Same Title as 79-14 (Supplement) 8/15/79 8/20/79 Same as 79-14 Same as 79-02 (Rev 1) 79-13 (Rev 1) 79-22 Cracking in Feedwater System Piping 8/30/79 Possible Leakage of Tubes 9/5/79 of Tritium Gas Used in Timepieces for Luminosity All Designated Applicants for OLs Each Licensee who Receives Tubes of Tritium Gas in Timepieces for Luminosity 79-14 (Supplement No. 2) 79-23 79-24 Same as Title 79-14 Potential Failure of Emer gency Diesel Generator Field Exciter Transformer Frozen Lines 9/7/79 9/12/79 9/27/79=
Same as 79-14 All Power Reactor Facilities with an OL or CP All Power Reactor
'acilities which have
'either OLs or CPs and are in late stage of construction
~ ta