ML19249B649
| ML19249B649 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 07/26/1979 |
| From: | Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Aswell D LOUISIANA POWER & LIGHT CO. |
| References | |
| NUDOCS 7909040633 | |
| Download: ML19249B649 (1) | |
Text
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NUCLEAR REGULATORY COMMISSION h'-'j c
REGION IV
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July 26, 1979 In Reply Refer To:
RIV Docket No.
50-382/IE Bulletin No. 79-17 Louisiana Power and Light Co.
ATTN:
Mr. D. L. Aswell Vice President of Power Production 142 Delaronde Street New Orleans, Louisiana 70174 Gentla. men:
The enclosed IE Eulletin 79-17 is forwarded to you for inforuation.
No written response is required. However, the potential corrosion behavior of safety-related systems as it regards your plant over the long-tera should be taken into consideration.
If you desire additional information concerning this catter, please contact this office.
Sincerely, (7
4 N q w d xisen t (ul'1 ffarlV. Seyfrit firector
Enclosures:
1.
List of IE Eulle* ins Issued in Last 12 Months m }.. p e en ne u
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UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCENENT WASHINGTON, D. C.
20555 IE Bulletin No. 79-17 Date:
July 26, 1979 Page 1 of 4 PIPE C_% _ZS IN STAGNANT E0 RATED UATEit 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 esstatially stagnant be. rated water.
Metallurgical investigations revealed these cracks occurred in the weld heat af fected zone of T 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 a f fected 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 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 cetallurgical 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 1.D.
The cracking was localized to the heat affected zone where the type 304 stainless steel is sensitized (precipitated carbides) during welCing.
In addition to the nain through-wall crack, incipient cracks were observed at several locations in the weld heat affected zone including the weld toot 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 sere 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 ICSCC 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 excmination of potentially affected systems utilizing ecia] techniques.
The systems examined included the spent fuel, decay heat rt c-al, makeup and purification, and reactor building spray systems which :: ain stagnaat or intermitteatly stagnant, oxygenated boric acid environ-ments.
3 ese systems range from 2-1/2-inch (liPSI) to 24-inch (borated water s:mge tank suction), are type 304 stainless steel, schedule 160 to scheci e 40 thickness, respectively.
Results of these examinations were reported to the NRC cn 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"-2 4"- 12"-10"-S"-2" e t c. )
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 co'laut system except for check valves. All of the six cracks were found in two high pressure injection li es 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 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 h'/ rotests, visual and volumetric d
examinations performed per 10 CFR 50.55a(g) (Pe: 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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lE 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 supplcments (addenda) that were followed, and the acceptance criterion.
(d) 72cilities 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 examinat ions 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 examinition (IWA 2210) of rormally 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 intende<i that the UT examination cover the weld fusion zone and a minimum of 1/2-incb 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 x
during reactor operation.
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IE Eulletin No. 79-17 Date:
July 26, 1979 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 ex: ination of accessible portions of other similar units which have not bet _ inspected under the ISI provisions of 10 CFR 50.55a(g) unless justifi-can n for continued operation is provided.
4.
An; : racking identified shall be reported to the Director of the appro-pri n e 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 la-day written report.
5.
Provide a written report to the Di. rector 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 exam nation required b'j 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 Of fice 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 t he Director, Division of Operating IMactors, Of fice of Inspection and Enforcement, k'ashington, D. C.
20555.
Approved by CAO, B180223 (R0072), clearance expires 7/31/S0.
Approval was given under a blanket clearance specifically for idenified generic problems.
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s 1E Circular 76-06 I!ovtaber 24, 1976 s
STRESS CORROSIOS CMCKS IN STAGNANT, LO:! PRESSURE STAINLCSS PIPING C05TlISING EORIC ACID SOLUTION isT PiTR's DESSE? TION OF CIRCRIST/'!CES:
Dri2;; the paric:' I! ave =ber 7,1974, to November 1,1975, several inci-denn of through-vall cracking have occurred in the 10-inch, schedule,
1G r 9 304 stainlass steel piping of the Reactor Eullding Spray and Dec;y licat 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 staittless discharge pipin3 of the "A" recirculation spray heat exchanger at Surry Unit No. 2.
A recent inspection of Unit I!o. 1 Containaent Recirculatica Spray Piping revealed cracking sinilar to Unit I!o. 2.
On October 8,1976, another incident of siallar cracking in 8-inch sched-nle 10 type 304 stainless piping of the Safety Injection Puap suction I.ine at the Ginna facility was reported by the licensee.
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Infort:ation received on the rietallurgical analysis conducted to date indicates that the. failures were the result of intergranul.r stress
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corrosion cracking'ihat initiated on the inside of the piping. A
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comanality of f actors observed essociated uith the corrosion rechanism 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 systea.
2.
Cracking occurred in piping containing relatively stagnant boric acid solution not required for norcal operating conditions.
3.
Analysis of surface products at this time indicate a chloride ion interaction uith oxide fornation in the relatively stagnant boric acid solution as the prohnble 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 deiinitely knoun.
Itouever, at C-'-l the chlorides and sulfide IcVel observed in the surface tarnish fil: near velds is calieved to have been introduced into the piping
(' rf.3 testin;, of tha sodlue-thiosulf ate discharge valves, or valve 7=
a.
Siuilarly, at Ginaa the chloriden and potential oxygen avail-abiry verc-assured to have beca present.since original constructiva of i= barated cater storage tank ubich is vented to atucaphere. Co rr o--
sian attack at Surry is attributed to in-leakage of chlorides through recirculation spray heat exchange tubing, allouing buildup of contaminated vatar in an othervise normally dry spray piping.
ACTICI TO BE TAKEN LY LICENSEE:
1.
Provide a description of your program for ' assuring continuad integrity of those saf ety-related piping systens ubich are not frequently flushed, or which contain nonflouing liquids.
This progran should include _ corr-sideration of hydrostatic testing in accordasce uith ASME Code Section-(1974 Editica) for all active systens required for safety XI rules injection and containment spray, including their recirculation nodes, froa source of water supply up to the second isolation valve of the
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prinary spste'a.
Similar tests should be ccusidered for other nafety-related piping systens.
2.
Your progran should also consider volu~etric c:.aaination of a repre-senrative nunb=r of circunferential pipe velds by nondestructive
.;xaaination techniques.
Such examinations should be performed generally in accordance with Appendi.x I of Section XI of the ASME Code, c:: cept thc t the exaained area should cover a distance of approxi-rntely six (6) Lines the pipe uall thickness (but not less than 2 inches and need not exceed C inches) on each side of the veld.
Supple 22atary exacination techaiques, such as radiography, should be used where necessary for evaluation or confirnation of ultranonic indications resulting froa such exauination.
3.
A report describing your program and schedule for these inspections should be submitted uithin 30 days af ter receipt of this Circular.
4.
The 1:RC Regional Office should be inEorned 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 falds identified above.
5.
A sumry rcport of the exaninations and evaluation of results should be subaitted uithin 63 days f rom the date of completion of proposed testing and examinations.
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, This atm. ary report should also include a brieE description of plant conditicas, operating procedures or other acti.vities uhicit
- ovide accurance that the effluent chenistry u111 tulatain lou levels of potential corrodants in cuch relatively stagnant regions Ichin. tite piping.
Your r2sponses shauld be sub.titted. to the Director of this office, with a copy to the 1:~'C Of fice of Inspection and Enforcenent, Division of Peac tor Inspectica Prograna, Uashington, D.C. 20555.
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Approval of 1:nC requirements for reports concerning possible generic prob 1c._s has been obtained under 44 U.S.C. 315? frora the U.S. General ' ' -
Accouating Office.
(GAO Approval B-180255 (R0052), c>:pires 7/31/77).
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IE Bulletin No. 79-17 July 26, 1979 LISTING OF IE BULLETl?;S ISSUEll IN LAST TWELVE t10N1~ilS Bulletin Subj ect Date Issued Issued To tio.
78-11 Exataination of Mark I 7/21/73 BUR Power Reactor Coataitunent Torus Facilities for Welds action:
Peach Bottora 2 and 3, Quad Cities 1 and 2, Ifatch 1,!!onti-cello and Vermont Yankee 78-12 Atypical Weld haterial 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)78-12D Atypical Weld Material 3/19/79 All Power Peactor 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 O n - : s
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IE Bulletin 1;o. 79-17 July 26, 1979 78-14 Deterioration of 12/19/78 All GE BWR facilities Buna-?! Components with an Operating in ASCO Solencids License (OL) or Construction Pertait (CP) avironmental Quali-2/8/79 All Power Reactor 79-01 r
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 Clas, IE Equipment Facilities with an Operating License (OL) or Construction Permit (CP) 79-02 Pipe Support Ease 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 Desi ;us Using Facilities with an t
Concrete Expansion Operating Li. cense Anchor Bolts (OL) or Construction Permit (CP) 79-03 Longitudinal Weld 3/12/79 Al1 Power Reactor Defects In AS11E SA-312 Facilities with an Type 304 Stainless Operating License Steel Pipe Spool.
(OL) or Construction Manufactured by Permit (CP)
Youngstown Velding 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 sa-v-sP r ' i. _z o._
IE Bulletin No. 79-17 July 26, 1979 79-05 Nuclear Incident at 4/1/79 All Power Reactor Three flile Island Facilities with aa Operating License (OL) or Construction Permit (CP)79-05A Nuclear Incident at 4/5/79 All Power Reactor Three fIile 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 flisalign;aents 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 riisalignments Identified Operating I.icense During the Three flile (OL)
Island Incident 79-06A Review of Operational 4/18/79 All Pressurized Uater (Rev. 1)
Errors and System flis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three liile with an Operating License (OL)
Island lucident 79-063 Review of Operational 4/14/79 All Combustion Engineer-Errors and Systea ing PWR Facilities with Flisaligrunents Identified an Operating License Luring The Three t!ile (OL)
Island 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an Operating License (OL) or Construction Perrait (CP)
Enclosure Page 3 of 4 01 ;pr o
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IE Bulletin No. 79-17 July 26, 1979 79-08 Events Relevant to EWR 4/14//9 All BWR Iower Reactor Reactors Id2atified Facilit.ies wit h an During Three Mile Island Operating License Incident (OL) or Construction Permit (CP) 79-09 Failures f GE Type AK-2 5/11/79 All Power Reactor Circuit. Era aker in Sa fety Facilities with au r: elated Sys ;eas Operating License (OL) or Construction Permit (CP) 79-10 Requalificztion 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 En'_,ineered S u ety Operating License (OL) or Syste as a Construction Permit (CP) 79-12 Short Period Scrams at 5/31/79 All Power Reactor Facilities BWR Facilities with an Operating License (OL) cr a Const.ruction Permit (CP) 79-13 Cracking In Feedwater All PWRs with an Operating System Piping License (OL) for action.
All EWR with a Construction Permit (CP) for information 79-14 Seismic Analyses for 7/2/79 All Power Reactor f acilit.ies As-Euilt Sa fety-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 r Construction Permit and/or Operating License (OL) 79-16 Vital Area Access Cont.rols 7/26/79 All Po..er Reactors with an Operating License (OL) or anticipat.ing fuel loading prior to January 1981.
Enclosure Page 4 of 4 a :,
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