ML19249B643
| ML19249B643 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 07/26/1979 |
| From: | Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Weigand J GULF STATES UTILITIES CO. |
| References | |
| NUDOCS 7909040628 | |
| Download: ML19249B643 (1) | |
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UNITED STATES 7Ec f
NUCLEAR REGULATORY CONMISSION O..bM[/ ji o., r o@
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- ARLINGTON TEXA5 75012 July 26, 1979 In Reply Refer To:
RIV Docket Nos.
50-458/IE Bulletin No. 79-17 50-459/IE Bulletin No. 79-17 Gulf States Utilities Attn:
Dr. J. G. Weigand General Manager, Nuclear Projects Post Office Box 2951 Beaumont, Texas 77704 Gentlemen:
The enclosed IE Bulletin 79-17 is forwarded to you for i f n ormation. No written response is required.
However, the potential corrosion behavior of safety-related systems as it regards your plant over the long-term should be taken into consideration.
If you desire additional information concerning this catter, please contact this office.
Sin.erely,
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frsr1 V. Seyfrit 5
U1 rector
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Enclosures:
1.
List of IE Eulletins Issued in Last 12 Months r, q ;, : >
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UNITED STATES NUCLEAR REGULATORY C031BlISSION OFFICE OF INSPECTION AND ENFORCEtIENT WASIIINGTON,
D. C.
20555 IE Bulletin No. 79-17 Date: July 26,1979 Page 1 of 4 PIPE CFZ23 IN STAGNA'iT BORATED WATER SYSTEMS AT PWR PLANTS Descriptica 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 af fected 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 affected systems.
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 1 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 crachs 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 Intergranular Stress Corrosion Cracking (IGSCC) originating on the pipe I.D.
The cracking was localized to the heat affected 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. 79-17 Date:
July 26, 1979 Page 2 of 4 Based on the above analysis and visual leaks, the licensee initiated a broad-based ultrasonic examination of potentially affected systems utilizing special techniques.
The systems examined included the spent fuel, decay heat. removal, raakeup and purification, and reactor building spray systems which certain stagnant or intermittently stagnant, oxygenated boric acid environ-ments.
Tnese systems range from 2-1/2-inch (HPSI) to 24-inch (borated water s:: rage tank suction), are type 304 stainless steel, schedule 160 to sched.tle 40 thickness, respectively.
Results of these examinations were reported 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 a fc rementioned sizes (24"-14"-12"-10"-8"-2" etc. )
of the above systems.
It is important to note that six of the crack indications were found in 2-1/2-inch diameter pipe of the high pressure injection lines -
inside containment.
These lines are attached to the main coolant pipe and are nonisolable from the main coolant system except for check valves.
All of the six cracks were found in two high pressure injection lines containing stagnated borated water.
No cracks were found in the high pressure injection lines which were occasionally flushed 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 Notice 79-19.
For All Pressurized Vater Reactor Facilities with an Operating License:
1.
Conduct a review of safety-related stainless steel piping systems within 30 days of the date of this Dulletin 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 volumetric 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, summary of chemistry data, any design changes and/or actions taken, such as periodic flushing of recirculation procedures _to maintain 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 cere followed, and the acceptance criterion.
(c) 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 af ter the date of the Bulletin.
(a) Perform ASME Section XI visual examination (IWA 2210) of normally accessible? welds of all engineered safety systems at service pressure to verify system integrity.
(b) Conduct ultrasonic examination and liquid penetrant surface examination' or a representative number of circumferential welds in normally accessible < 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 welds 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 repair, 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.
Identification of cracking in one unit of a multi-unit facility which canses safety-related systems to be inoperable shall require immediate exaaination of accessible portions of other similar units which have not been inspected under the ISI provisions of 10 CFR 50.55a(g) unless justifi-c:.tica for continued operation is provided.
4.
Any cracking 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 this 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 NRC Regional Office within 120 days of the date of this Bulletin describing the results of the inspections required by Item 2 and any corrective measures 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, k'ashington, D. C.
20555.
Approved by GAO, B180225 (R0072), clearance expires 7/31/80.
Approval was given under a blanket clearance specifically for idenified generic problems.
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November 24, 1976 STP35S CORROSION CPACKS IN STAGNANT, LON PRESSURE STAl!TLESS PIPING CONTAINING E02IC ACID SOLUTION AT PUR's DESC?l? TION OF CIRCUMSTANCES:
Dar 2; the paried November 7,1974, to November 1,1975, several inci-dents of through-wall cracking hava occurred in the 10-inch, schedule.
10 type 304 stainlass steel piping of the Reactor Building Spray and Decay Heat Systens at Arkansas Nuclear Plant No. 1.
On October 7,1976, Virginia Electric and Power also reported through-vall cracking in the.10-inch schedule 40, type 304 stainless discharge piping of the "A" recirculation spray heat exchanger at Surry Unit No. 2.
A recent inspection of Unit No. 1 Containment Recirculation Spray Piping revealed cracking sinilar to Unit No. 2.
On October 8,1976, another incident of similar cracking in 8-inch sched-nle 10 type 304 stainless piping of the Safety. Injection l aap suction Line at the Cinna f acility was reported by the licensee.
Infornation received on the netallurgical analysis conducted to date indicates that, the.. failures were the result of intergranular stress
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corrosion cracliing'ihat initiated on the inside of the piping. A
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cornonality of f actors observed associated uith the corrosion nechanism were:
1.
The cracks vere adjacent to and propagated along veld 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 foreation in the relatively stagnant boric acid co]ution as the probable corrodant, with the state of stress probably due to velding and/or fabrication.
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The source of the chloride ion is not dcEinitely knourt.
Itouever, at A5-l the chlorides and sulfide level observed in the surface tarnish fil: near velds is believed to have been introduced into the piping during testing of tha sediua thiosulfate discharge valves, -or valve
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Siallarly, at Cinna the chlorides and potential oxygen avail-ab '/ uere assu2ed to have been present.since original construction of raa horated. rater storage tank which is vented to atcosphere.
Corro-si:2 attack at Surry is attributed to in-leakage of chlorides through recirculation spray heat exchange tubing, allouing buildup of contaminated uatar in an otheruise normally dry spray piping.
ACTION TO BE TAKE:I BY LICENSES:
1.
Provide a description of your program for ' assuring continued integrity of those caf ety-related piping systens uhich are not frequently flushed, or which coatain nonflouing liquids.
This progran should include corr-sideration of hydrostatic testing in accordance with ASME Code Section' Xi rules (1974 Edition) for all active systecs required for safety inj ection and containcent spray, including their recirculation noden, f ron cource of water supply up to the second isolation valve of the prinary ??fste'6.
Similar tests should be ccnoidered for other safety-related piping systens.
2.
Your progran should also consider volumetric exuaination of a repre-sentative number of circumferential pipe velds by nonelestructive examination techniques.
Such examinations should be performed generally in accordance with Appendix I of Section XI of the ASME Code, except that the examined area should cover a distance of approxi-Iaately six (6) times the pipe vall thici ness (but not less than 2 inches and need not exceed 8 inches) on each side of the veld.
Supplea2atary exacination techniques, such as radiography, should be used uhere necessary for evaluation or confirr.ation of ultrasonie indications resulting froa such exauination.
4 3.
A report describing your prograa and schedule for these inspections should be submitted within 30 days af ter receipt of this Circular.
4.
The I!RC Regional Of fice should be inforced uithin 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.
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A summary report of the exaninations and evaluation of results should be subnitted within 60 days from the date of completion of. proposed testing and examinations.
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Tais cuenary report should also include a brief description of.
plant conditions, operating procedures or other activities which provide assurance that the effluent chenistry will t:aintain lou 12v21s of potential corrodants in such relatively stagnant regions within the piping.
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Your responses should be sub:titted. to the Director of this office, with ' -
a cop 7 to the 1:RC Of fice of Inspection and Enforecaent, Division of Reactar Inspection Progrants, Washington, D.C. 20555.
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Approval of 1;RC requiretaents for reports concerning possible generic prob 13'c has beert obtained under 44 U.S.C. 3152 frota the U.S. General Accounting Office.
(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 Examination 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 with an Vessel 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)78-12B 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, 7051B, 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 12/19/78 All GE BWR facilities Buna-N Components with an Operating in ASCO Solenoids License (OL) or Construction Permit (CP) 79-01 Environaental Quali-2/8/79 All Power Reactor fication of Class IE Facilities with an Equipment Operating License (OL) or Construction Ptrmit (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 Facilities with an Valves Manufactured Operating License by Velan Engineering (OL) or Construction Corporation Permit (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 Mile Island Facilities with an Operating License (OL) or Construction Permit (CP)79-05A Nuclear Incident at 4/5/79 All Power Reactor Three Mile 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 Mile Island Facilities with an Operating License (0L) 79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Power Reactor Facilities Hisalignments Identified Except B&W Facilities During The Three Mile Island Incident 79-06A Review of Operational 4/14/79 All Westinghouse PUR Errors and System Facilities with an Misalignments Identified Operating License During the Three Mile (OL)
Island Incident.79-06A Review of Operational 4/18/79 All Pressurized Vater (Rev. 1)
Errors and Systeu Mis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with an Operating License (OL)
Island Incident 79-06B Review of Operational 4/14/79 All Combustion Engineer-Errors and System ing PWR Facilities with Misaligranents Identified an Operating License During The Three Mile (OL)
Island 79-07 Seisric Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an Operating License (OL) or Construction Permit (CP)
Enclosure Page 3 of 4 a,.
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IE Bulletin No. 79-17 July 26, 1979 79-08 Events Relevant to BWR 4/14/79 All BWR 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 Systems Operating License (0L) 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 Eugineered Safety Operating License (OL) or Systems a Construction Permit (CP) 79
.2 Short Period Scrams at 5/31/79 All Power Reactor Facilities BWR Facilities with an Operating License (OL) or a Construction Permit (CP) 79-13 Cracking In Feedwater All PWRs with an Operating System Piping License (OL) for action.
All BWR with a Consttuction Pe rmit (CP) for information 79-14 Seismic Analyses for 7/2/79 All Power Reactor facilities As-Built Safety-Related with rn 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 Power Reactors with an Operating License (OL) or anticipating fuel loading prior to January 1981.
Enclosure Page 4 of 4
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