ML19249B648
| ML19249B648 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 07/26/1979 |
| From: | Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Koester G KANSAS GAS & ELECTRIC CO. |
| References | |
| NUDOCS 7909040632 | |
| Download: ML19249B648 (1) | |
Text
p*s ~ t c4, UfJITED STATES f,L G
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S"h.D.h / 8 REGION IV Gil HYAN PL AZA DHIVE, SulTE 1000
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,Mf LINGToN, TE XAS /6012 July 26, 1979 In Reply Refer To:
RIV Do let No.
STN 50-482/IE Eulletin No. 79-17 Kansas Gas & Electric Co.
Attn:
Mr. C1 cati L. Koester Vice President-Operations Post office. "ox 203 Wichita, Kansas 67201 Gentlemen:
The enclosed IE Eulletin 79-17 is forwarded to you for info =ation.
No written response is required. However, the potential corrosioa behavior of safety-related syste s as it regards your plant over the lon3-tern should be taken into consideration.
If you desire additional infomation concerning this catter, please contact this office.
Sincerely, O
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J girector
Enclosures:
1.
lE Bulletin No. 79-17 2.
List of IE Bulletins Issued in Last 12 Months cc:
w/ enclosures Messrs.
Nicholas A.
Petrick, SSUPPS D. T. McPhee, Kansas City Power and Light Company Gerald Charnoff, Shaw, Pittman, Potts & Trowbrid?,e E. W. Creel, Kansas G 1s and Electric Co:aaaay e.,o -r3r-J % 1' $ U u '.,.
i 7909040 $3 2 P
UNITED STATES NUCLEAR REGULATORY CO.T!ISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D. C.
20555 IE Bulletin No. 79-17 Date: July 26, 1979 Page 1 of 4 PIPE CF_. ~. S IN STAGNMT BORATED WATER SYSTEMS AT PWR PLANTS Descrip;ica of Circumstances:
During the period of November 1974 to February 1977 a number of cracking incidents have been experienced in safety-related stainless steel piping systems and portions of systems which contain oxygenated, stagnant or essentially stagnant borated water.
Metallurgical investigations revealed these cracks occurred in the weld heat affected zone of 8-inch to 10-inch type 304 material (schedule 10 and 40), initiating on the piping I.D. surface and propagating in either an intergranular or transgranular mode typical of Stress Corrosion Cracking.
Analysis indicated the probable corrodents to be chloride and oxygen contamination in the af fected s' stems.
Plants affected up to this time were Arkansas Nuclear Unit 1, R. E. Ginna, H. B. Robinson Unit 2, Crystal River Unit 3, San Onofre Unit 1, and Surry Units 1 and 2.
The NRC issued IE Circular 76-06 (copy attached) in view of the apparent generic nature of the problem.
During the refueling outage of Three Mile Island Unit I which began in February of this year, visual inspections disclosed five (5) through-wall cracks at welds in the spent fuel cooling system piping and one (1) at a weld in the decay heat removal system.
These cracks were found as a result of local boric acid build-up and later confirmed by liquid penetrant tests.
This initial identification of cracking was reported to the NRC in a Licensee Event Report (LER) dated May 16, 1979.
A preliminary metallurgical analysis was performed by the licensee on a section of cracked and leaking weld joint from the spent fuel cooling system.
The conclusion of this analysis was that cracking was due to Intcrgranular Stress Corro; ion Cracking (IGSCC) originating on the pipe I.D.
The cracking was localized to the heat af fected zone where the type 304 stainless steel is sensitized (precipitated carbides) during welding.
In addition to the main through-wall crack, incipient cracks were observed at several locations in the weld heat affected zone including the weld root fusion area where a miniscule lack of fusion had occurred.
The stresses responsible for cracking are believed to be primarily residual welding stresses inasmuch as the calculated applied stresses were found to be less than code design limits.
There is no conclusive evidence at this time to identify those aggressive chemical species which promoted this IGSCC attack.
Further analytical efforts in this area and on other system welds are being pursued.
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IE Bulletin No. 79-11 Date: July 26, 1979 Page 2 of 4 Based cc the above analysis and vit,ual leaks, the licensee initiated a broad-b ased ultrasonic examination of potentially a; fected systems utilizing s sial techniques.
The systems examined included the spent fuel, decay t rer: val, makeup and purification, and reactor building spray systems
.a c ::ain stagnaat or intermit.tently stagnant, oxygenated boric acid environ-neuts.
dese systems range from 2-1/2-inch (HPSI) to 24-inch (borated water :.::sge tauk suctica), are type 304 stainless steel, schedule 160 to sche'._12 40 thickness, respectively.
Results of these examinations were report d to the NRC on June 30, 1979, as an update to the May 16, 1979 LER.
The ultrasonic inspection as of July 10, 1979, has identified 206 welds out of 946 inspected having UT indications characteristic of cracking randomly distributed throughout the aforementioned sizes (24"-14" u."-10"-8"-2" etc.)
of the above systems.
It is important to note that si" of the crack indications were found in 2-1/2-inch diameter pipe of the high pessure injection lines inside containment.
These lines are attached to the ma 'n coolant pipe and are nonisolable from the nain coolant system except for check valves.
All of the six cracks were found in tveo high pressure injection lines containing stagnated bora: n.
eater.
No cracks were found in the high pressure injection lines which were accasionally Ilushed during makeup operations. The ultrasonic examination is continuing in order to delineate the extent of the problem.
The above information was previously provided in Information No i~ 70-19.
For All Pressurized Water Reactor Facilities with an Operating License:
1.
Conduct a review of safety-related stainless steel piping syst. ems within 30 days of the date of this Bulletin to identify systems and portions of systems which contain stagnant oxygenated borated water. These systems typically include ECCS, decay / residual heat removal, spent fuel pool cooling, containment. spray and borated water storage tank (BWST-RWST) piping.
(a)
Provide the extent and dates of the hydrotests, visual and volturetric examinations performed per 10 CFR 50.55a(g) (Re: IE Circular 76-06 enclosed) of identified systems.
Include a description of the non-destructive examination procedures, procedure qualifications and acceptance criteria, the sampling plan, results of the examinations and any related corrective actions taken.
(b)
Provide a description of water chemistry controls, surmnary of chemistry data, any design changes and/or actions taken, such as periodic flushing of recirculatica procedures _to rnintain required water chemistry with respect to pH, B, CL, F, 0 "
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IE Bulletin No. 79-17 Date:
July 26, 1979 Page 3 of 4 (c) Describe the preservice NDE performed on the weld joints of identified systems. The description is to include the applicable ASME Code sec-tions and supplements (addenda) that were followed, and the acceptance criterion.
(d?
Facilities having previously experienced cracking in identified systems, Itea 1, are requested to identify (list) the new materials utilized in repair or replacement on a system-by-system basis.
If a report of this information and that requested above has been previously submitted to the NRC, please reference the specific report (s) in response to this Bulletin.
2.
Facilities at which ISI examinations have not been performed (i.e., visual and volumetric UT) on stagnant portions of systems identified in Item 1 above, shall complete the following actions at the earliest practical date but not later than 90 days after the date of the Bulletin.
(a) Perform ASME Section XI visual examination (IWA 2210) of normally accessible
- velds of all engineered safety systems at service pressure to verif" system integrity.
(b)
Conduct ultrasonic examination and liquid penetrant surface examination or a representative number of circumferential welds in normally accessible 4 portions of systems identified by Item 1 above.
It is intended that the sample number of welds include all pipe diameters in the 2-1/2 inch to 24-inch range with no less than a 10 percent sample by system and pipe wall thickness.
It is also intended that the UT examination cover the weld fusion zone and a minimum of 1/2-inch on each side of the weld at the pipe I.D.
The examination shall be in accordance with the provisions of ASME Code Section XI-Appendix III and Supplements of the 1975 Winter Addenda, except all signal responses shall be evaluated as to the nature of the indications.
These code methods or alternative examination methods, combination of methods, or newly developed techniques may be used provided the procedures yield a demonstrated effectiveness in detecting stress corrosion cracking in austenitic stainless steel piping.
(c)
If cracking is identified during Item (a) and (b) examinations, all velds of safety-related piping systems and associated subsystems where dynamic flow conditions do not exist during normal operations (Item 1) shall be subject to volumetric examination and re
-ir, including piping in areas which are normally inaccessible.
- Normally accessible refers to those areas of the plant which can be entered during reactor operation.
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IE Bulletin No. 79-17 Date: July 26, 1979 Page 4 of 4 3.
Idectification of cracking in one unit of a multi-unit facility which caus s safety-related systems to be inoperable shall require immediate c:c.nication of accessible portions of other similar units which have not bee inspected under the ISI provisions of 10 CFR 50.55a(g) unless justifi-cc x2 for continued operation is provided.
4.
An7 : racking identified shall be reported to the Director of the appro-priate 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 followed by a 14-day written report.
5.
Provide a written report to the Director of the appropriate NRC Regional Office within 30 days of the date of th.is Bulletin addressing the results of your review required by Item 1.
6.
Complete the examination required by Item 2 within 90 days of the date of this Bulletin and provide a written report to the Director of the appro-priate NP,C Ref;ional Of fice within 120 days of the date of this Bulletin describing the results of the inspections required by Item 2 and any corrective ceasures taken.
7.
Copies of the reports required by Items 4, 5 and 6 above shall be provided to the Director, Division of Operating Reactors, Office of Inspection and Enforcement, Washington, D. C.
20553.
Approved by GAO, U180225 (R0072), clearance expires 7/31/80.
Approval was given under a blanket clearance specifically for idenified generic problems.
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IE Circular 76-06 November 24, 1976 s
STE5S CO?dtOSIOS CRACKS IN STACNANT, LON PRESSURE STAINLESS PIPING CONC?.ISING EORIC ACID SOLUTION AT PC.T s DEEC.Z? TION OF CIRCT.4STAUCES:
%:2 ; the pariod No tember 7,1974, to November 1,1975, several inci-de = cf throu;;h-wall cracking hava occurred in the 10-inch, schedule.
10 zy 304 stainless steel piping of the Reactor Building Spray and Dec 7 lieat Systens at Arkansas Nuclear Plant No. 1.
Oa October 7,1976, Virginia Electric and Power also reported through-vall cracking in the.10-inch schedule 40 type 304 stainless discharge pipi 3; of the "A" recirculation spray heat exchanger at Surry Unit No.
2.
A recent inspection of Unit No. 1 Containaent Recirculation Spray Piping revealed cracking sinilar to Unit No. 2.
Oa October 8,1976, another incident of siallar cracking in 8-inch sched-Ole 10 type 304 stainless piping of the Safety. Injection Pump Suction Line at the Ginna facility was reported by the licensee.
Inforuation received on the netallurgical analysis conducted to date indicates that, the. failures were the result of intergranular stress
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corrosion crdchin;;'ihat initiated on the inside of the piping. A
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cox caality of factors observed associated uith the corrosion nechanism were:
1.
The cracks were adjacent to and propagated along uald zoites of the thin-valled low pressure piping, not part of the reactor coolant system.
2.
Cracking occurred in piping containing relatively stagnant boric acid solution not required for normal operating conditions.
3.
Analysis of surface products at this time indicate a chloride ion interaction with oxide fornation in the relatively stagnant boric acid solution as the probable corrodant, with the state of stress probably due to welding and/or fabrication.
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. s Trn source of the chloride ion is not definitely known.
Itouever, at l';:-l the chlorides and sulfida level observed in the surface tarnish fil: near velas is believed to have been introduced into the piping du;-ian testing of the sodina thiosulfate discharge valves, or valve le.:n;;2.
Sittilarly, at Cinua the chlorides and potential oxygen avail-ch ' - ' y were assu2ed to have been present.since original construction of
..a horated =ter storage tank which is vented to atcosphere. Corro-sis: cttack at Surry is attributed to in-leakage of chlorides through retirculation spray heat exchange tubing, allowing buildup of contaminated water in an otheruise normally dry spray piping.
ACTIO.'I TO BE TAKEti BY LICENSEE:
1.
l'rovide a description of your program for ' assuring continued integrity of those safety-related piping systens which are not frequently flushed, or which contain nonflowing liquids.
This prograa should include coir sideration of hydrostatic testing in accordance with ASME Code Section' XI rules (1974 Editica) for all active systees required for safety inj ection and containment spray, including their recirculation nodes, fro:a source of water supply up to the second isolation valve of the prinary siste'a-Similar tests should be ccusidered for other safety-related piping systens.
w 2.
Your progran shou.l also consider valueetric exanination of a repre-senrative nunber of circunferential pipe velds by nondestructive examination techniques.
Such exaninations should be performed generally in accordance with Appendix I of Section XI of the ASME Code, except that the exami.ned area should cover a distance of approxi-riately six (6) tiuas the pipe vall thickness (but not less than 2 inches and need not exceed 8 inchas) on each side of the veld.
Suppleaeatary exacination techaiques, such as radiography, should be used there necessary for evaluation or confircation of ultrasonic indications resulting f roa such exauination.
3.
A report describing your prograta and schedule for these inspections should be submitted uithin 30 days af ter receipt of this Circular.
4.
The 1:RC Regional Of fice should be inforced within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, of any adverse findings resulting during nondestructive evaluation of the accessible piping velds identified above.
s 5.
A summary report of the exaninations and evaluation of results should be subaitted within 60 days from the date of completion of proposed testing and examinations.
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Tnis ste_,ary repart should also include a brief description of,
plaat conditions, operating procedures or other activitie; which provide assurance that the effluent chenistry will ruintain low le cals of potential corrodants in such relatively stagnant regions within. the piping.
e Tour responses should be subaitted. to the Director of this office, with a copy t:, the IGC Of fice of Inspection and Enforcenent, Division of Rea: _ar Inspection Prograas, Uashington, D.C. 20555.
/.ppro721 of 1 RC requirements for reports concerning possible generic problens has been obtained under 44 U.S.C. 3152 from the U.S. General Accounting Office.
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(GAO Approval B-180255 (R0052), expires 7/31/77).
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IE Bulletin No. 79-17 July 26, 1979 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
78-11 Exauination of Mark I 7/21/78 BWR Power Reactor Containment Torus Facilities for Welds action:
Peach Bottom 2 and 3, Quad Cities 1 and 2, Hatch 1, Monti-cello and Vermont Yankee 78-12 Atypical Weld Material 9/26/78 All Power Reactor in Reactor Pressure Facilities witu an Vess21 Welds Operating License (OL) or Construc-tion Permit (CP)78-12A Atypical Weld Material 11/24/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds Operating License (OL) or Construc-tion Permit (CP)78-12D Atypical Weld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds Operating License (OL) or Construc-tion Permit (CP) 78-13 Failures in Source 10/27/78 All General and Heads of Kay-Ray, Specific Licensees Inc., Gauges Models with Kay-Ray Gauges 7050, 7050B, 7051, 70518, 7060, 7060B, 7061 and 7061B Enclosure Page 1 of 4
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IE Bulletin No. 79-17 July 26, 1979 78-14 Deterioration of i2/19/78 All GE BWR facilities Buna-N Components with an Operating in ASCO Solenoids License (OL) or Construction Permit (CP) 79-01 Environmental Quali-2/8/79 All Power Reactor fication of Class IE Facilities with an Equipment Operating License (OL) or Construction Permit (CP)79-01A Environmental Qualification 6/6/79 All Power Reactor of Class IE Equipment Facilities with an Operating License (OL) or Construction Permit (CP) 79-02 Pipe Support Base 3/8/79 All Power Reactor Plate Designs Using Facilities with an Concrete Expansion Operating License Anchor Bolts (OL) or Construction Permit (CP) 79-02 Pipe Support Base 6/21/79 All Power Reactor (Rev. 1)
Plate Designs Using Facilities with an Concrete Expansion Operating License Anchor Bolts (OL) or Construction Permit (CP) 79-03 Longitudinal Weld 3/12/79 All Power Reactor Defects In ASME SA-312 Facilities with an Type 304 Stainless Operating License Steel Pipe Spools (OL) or Construction Manufactured by Permit (CP)
Youngstown Welding and Engineering Company 79-04 Incorrect Weights 3/30/79 All Power Reactor for Swing Check Facilitics with an Valves Manufactured Operating License by Velan Engineering (OL) or Construction Corporation Perrait (CP)
Enclosure Page 2 of 4
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IE Bulletin No. 79-17 July 26, 1979 79-05 Nuclear Incident at 4/1/79 All Power Reactor Three t!ile Island Facilities with an Operating License (OL) or Construction Permit (CP)79-05A Nuclear Incident at 4/5/79
/11 Power Reactor Three IIile Island Facilities with an Operating License (OL) or Construction Permit (CP)79-05B Nuclear Incident at 4/21/79 All B&W Power Reactor Three flile Island Facilities with an Operating License (OL) 79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Power Reactor Facilities Flisaliguments Identified Except B&W Facilities During The Three Flile Island Incident 79-06A Review of Operational 4/14/79 All Westinghouse PWR Errors and System Facilities with an Misalignments Identified Operating License During the Three Mile (OL)
Island Incident 79-06A Review of Operational 4/10/79 All Pressurized Uater (Rev. 1)
Errors and Systen Flis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with an Operating License (OL)
Island Incident 79-063 Review of Operational 4/14/79 All Combustion Engineer-Errors and System ing PWR Facilities with Misaligr.ments Identified an Operating License During The Three tiile (OL)
Island 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an Operating Liceuse (OL) or Construction Permit (CP)
Enclosure Page 3 of 4 05}-[CCQ
IE Bulletin No. 79-17 July 26, 1979 79-08 Events Relevant to BUR 4/14/79 All B',iR Power Reactor Reactors Identified Facilities with an During Three B!ile Island Operating License Incident (OL) or Construction Permit (CP) 79-09 Failures of GE Type AK-2 5/11/79 All Power Reactor Circuit Breaker in Safety Facilities with an Related Systeas Operating License (OL) or Construction Permit (CP) 79-10 Requalification Training 5/11/79 All Power Reactor Program Statistics Facilities with an Operating License (OL) 79-11 Faulty Overcurrent Trip 5/22/79 All Power Reactor Device in Circuit Breakers Facilities with an for Engineered Safety Operating License (OL) or Systems a Construction Permit (CP) 79-12 Short Period Scrams at 5/31/79 All Power Peactor Facilities BWR Facilities with an Ope rating License (OL) or a Construction Permit (CP) 79-13 Cracking In Feedwater All PWRs with an Operating System Piping License (OL) for action.
All EUR with a Construction Permit (CP) for information 79-14 Seismic Analyses for 7/2/79 All Power Reactor facilities As-Euilt Safety-Related with an Operting License Piping System (OL) or a Construction Permit (CP) 79-15 Deep Draft Pump 7/11/79 All Power Reactor Facilities Deficiencies with a Construction Permit and/or Operating License (OL) 79-16 Vital Area Access Controls 7/26/79 All Po.;er Reactors with an Operating License (OL) or anticipating fuel loading prior to January 1981.
Enclosure Page 4 of 4 n=> 4,r ' 7,,ctb ac.
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