ML16294A372
| ML16294A372 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/26/1979 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Parker W DUKE POWER CO. |
| References | |
| NUDOCS 7908090369 | |
| Download: ML16294A372 (13) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II 101 MARIETTA ST., N.W., SUITE 3100 ATLANTA, GEORGIA 30303 JUL 2 6 1979 In Reply Refer To:
RII:JPO 50-269 50-270 50-287 Duke Power Company ATTN: W. 0. Parker, Jr.
Vice President, Steam Production P. 0. Box 33189 Charlotte, NC 28242 Gentlemen:
The enclosed Bulletin 79-17 is forwarded to you for action. A written response is required.
If you desire additional information regarding this matter, please contact this office.
Sincerely,
- q. James P. 0'Reil y Director
Enclosure:
IE Bulletin 79-17 7908090 3 67
-2 Duke Power Company cc w/encl:
- j. E. Smith, Station Manager Post Office Box 1175 Seneca, South Carolina 29678
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D. C. 20555 July 26, 1979 IE Bulletin No. 79-17 PIPE CRACKS IN STAGNANT BORATED WATER SYSTEMS AT PWR PLANTS Description 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 contami nation 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 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 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 in as much 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.
0 July 26, 1979 IE Bulletin No. 79-17 Page 2' of7 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, makeup and purification, and reactor building spray systems which contain stagnant or intermittently stagnant, oxygenated boric acid environments.
These systems range from 2 1/2-inch (HPCI) to 24-inch (borated water storage tank suction),
are type 304 stainless steel, schedule 160 to schedule 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 aforementioned 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 Water 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 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 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, 02'
July 26, 1979 IE Bulletin No. 79-17 Page 3 of 4 (c) Describe the preservice NDE performed on the weld joints of identified systems.
The description is to include the applicable ASIIE Code sec tions and supplements (addenda) that were followed, and the acceptance criterion.
(d) Facilities having previously experienced cracking in identified systems, Item 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* 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 acces sible* portions of systems identified by I 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 U.T. 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.
July 26, 1979 IE Bulletin No. 79-17 Page 4 of 4
- 3.
Identification of cracking in one unit of a multi-unit facility which causes safety-related systems to be inoperable shall require immediate examination of accessible portions of other similar units which have not been inspected under the ISI provisions of 10 CFR 5O.55a(g) unless justifi cation for continued operation is provided.
- 4.
Any cracking 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 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 appropriate 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 also be provided to the Director, Division of Operating Reactors, Office of Inspec tion and Enforcement, 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.
Enclosures:
- 1.
- 2.
List of IE Bulletins Issued in 1979
Encl re 1 November 26, 1976 IE Circular No. 76-06 STRESS CORRASIO\\ CrCS IN STANANr, LOW PRESSURE STAINLESS PIPING CONTAINING BORIC ACID SOLUTION AT PW(I's DESCRIPTION Of CIRCUMSTANCES:
During the period November 7, 1974 to November 1, 1975, several incidents of throegh-wal crackng have occurred in the 10-inch, schedule 10 type 304 stainless steel piping of the Reactor Building Spray and Decay Eeat Remova3 System.s at Arkansas Nuclear Plant No. 1.
On October 7, 1976, Virginia Electric and Power also :epoeted :thou;.
wall cracking in the 10-inch schedule 40 type 304 stainless disc r e piping of the "A" recirculation spray heat exchanger at Surry nSt No.
- 2.
A recent inspection of Unit 1 Containnent Recirculation Spray Piping revealed crackng similar to Unit 2.
On October 8, 1976, another incident of similpr cracking in 8-inch schedule 10 type 304 stainless piing of the Sfcty Injection Pu=p Suction Line at the Ginta facility wa reported by the licensee.
Infotrmation received on the metallurgical analysis conducted to date indicates that the failures were the result of intergrnul.r stress corrosionf crackiflg that initi-atod on the inside of the piping.
A commonality of factors observed assoc-iatcd with the corrosion mtchaisa were:
- 1.
The cracks were adjacent to and propagated along wead zones of rthe thin-walled lov pressure piping, not part o. the reactor coolant system.
- 2.
Cracking occurTed in piping containing re3atively stagnant boric acid solution not required for normal operating conditions.
- 3.
Analysis of surface products at this time indicate a chloride ion interaction with oide foration in the relatively stagnant boric acid solution as the probable corrodant, with the state of stress probably due to welding and/or fabrication.
The source of tha chloride ion is not dafinitcly known.
H4owever,.ct ANO-the chloiides and sulfid level observed in the surface tarnish fi1 near welds is believed to have been introduced into the piping during testrg of the sodium thiosulfate discharge vAlves, or valve leakage.
Siilarly, at Gina the chlorides and potential oxyget IE circular No. 76-06 2
N November 26, 1976 availability were assumed to have been present since originl construction of the borated water storage tank which is vented to atmosphere.
Corrosion attack at Surry is attributed to in-leakage of chlorides through recirculation spray heat exchange tubing allowing buildup of contaminated water in an otherwise normally dry spray piping.
ACTION TO BE TAE NY LICENSEE:
Provide a description of your program for assuring continued integrity of those safetyirelated piping systems which are not frequently flushed, or which contain nofflosing liquids.
This program should include consideration of hydrostatic testing in accordance with AS-ME Code Section X rules (1974 Edition) for all active systems required for safety inection and containment spray, including their recirculation nodes, from source of water supply up to thi second isolation valve of the primary system.
Similar tests should be considered for other safery-related piping systems.
- 2.
Your program should also consider voluetcic examination of a representative number of circumferential pipe welds by non destructive examination techniques.
Such cxapnatiodx should be performed generally in accoidance with Appendix I of Section XI of the ASM Code, except that the examined area should cover a distance of approximately six (6) times the pipe wall thickness (but not dess than 2 inches and need not exceed S inches) on each side of the weld. Supplementary exazination techniques, such as radiography, should be used where necessary for evaluation or confiration of ultrasonic indications resulting from such examination.
- 3.
A report describing your progTa= and schedule for these inspec tions should be submittcd withinl 30 days after receipt of this Circular.
A.
The XRC Regional Of fice should be informed withifl 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 welds identified above.
- 5.
hsummaV report of the examinations and evaluation of results should be submitted within 60 days from the date of completion of proposed testing and examinations.
closure 1 12 CirculaT NIO. 7"-6 Noavembetr 26, 1976 This summar repoti should also include a brief description of plant conditionls, operatingi procedures or other activities which provide assurance that the effluent chemistry w1ifl =aitain lov levels of potential corrodants in such relatively stagnant region1s within the piping.
'Your responlses should be submitte'd to the Director of this office, with a copy to the NRC Office of Inspectionl and Enforcmet, Division of Reactor Inspection ?rogransi Washiflgtour D.C. 20555.
Approval of NRC requirGe,--msfor reportsconerning 0
i1@~
~
problens has been obtained under 44 U.S.C 3152 from the U.S.
General Accounting Office.
(GAO Approval B-180255 (R0062).
expire.s 7/31/77.)
IE Bulletin No. 79-17 July 26, 1979 Page 1 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
79-16 Vital Area Access Controls 7/26/79 All Holders of and applicants for Reactor Operating Licenses 79-15 Deep Draft Pump 7/11/79 All Power Reactor Deficiencies Licensees with a CP and/or OL 79-14 Seismic Analyses for 6/2/79 All Power Reactor As-Built Safety-Related facilities with an Piping System OL or a CP 79-13 Cracking in Feedwater 6/25/79 All PWRs with an System Piping OL for action. All BWRs with a CP for information.
79-02 Pipe Support Base Plate 6/21/79 All Power Reactor facilities with an (Rev. 1)
Designs Using Concrete Faclite rt aaC Expansion Anchor Boltsa CP 79-12 Short Period Scrams at 5/31/79 All GE BWR Facilities BWR 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 a CP Systems 79-10 Requalification Training 5/11/79 All Power Reactor Program Statistics Facilities with an OL 79-09 Failures of GE Type AK-2 4/17/79 All Power Reactor Circuit Breaker in Safety Facilities with an Related Systems OL or CP 708Events Relevant to BWR 4/14/79 All BWR Power Reactor 79-08 Failitiescithianes Reactors Identified During 5/22/7 Three Mile Island Incident
IE Bulletin No. 79-17 July 26, 1979 Page 2 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an OL or CP 79-06B Review of Operational 4/14/79 All Combustion Engineer Errors and System Mis-ing Designed Pressurized alignments Identified Water Power Reactor During the Three Mile Facilities with an Island Incident Operating License 79-06A Review of Operational 4/18/79 All Pressurized Water (Rev 1)
Errors and System Mis-ow etor F es alignments Identified of W n
Di During the Three Mile Island Incident 79-06A Review of Operational 4/14/79 All Pressurized Water Errors and System Mis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with an OL Island Incident 79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Mis-Power Reactors with an alignments Identified OL except B&W facilities During the Three Mile Island Incident 79-OSA Nuclear Incident at 4/5/79 All B&W Power Reactor Three Mile Island Facilities with an OL 79-05 Nuclear Incident at 4/2/79 All Power Reactor Three Mile Island Facilities with an OL and CP
IE Bulletin No. 79-17 July 26, 1979 Page 3 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
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 78-12B Atypical Weld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 79-03 Longitudinal Welds Defects 3/12/79 All Power Reactor In ASME SA-312 Type 304 Facilities with an Stainless Steel Pipe Spools OL or CP Manufactured by Youngstown Welding and Engineering Co.
79-02 Pipe Support Base Plate 3/2/70 All Power Reactor Designs Using Concrete Facilities with an Expansion Anchor Bolts OL or CP 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) 79-01 Environmental Qualification 2/8/79 All Power Reactor of Class IE Equipment Facilities with an OL or CP 8-14 Deterioration of Buna-N 12/19/78 All GE BWR facilities Component In ASCO with an OL or CP Solenoids
IE Bulletin No. 79-17 July 26, 1979 Page 4 of 4 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
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 7051B, 7060, 7060B, 7061 Kay-Ray, Inc.
and 7061B gauges78-12A Atypical Weld Material 11/24/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12 Atypical Weld Material 9/29/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-11 Examination of Mark I 7/21/78 BWR Power Reactor Containment Torus Welds Facilities for action:
Peach Bottom 2 and 3, Quad Cities and 2, Hatch 1, Monticello and Vermont Yankee