ML18044A189
| ML18044A189 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 10/30/1979 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Dewitt R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 7911060088 | |
| Download: ML18044A189 (6) | |
Text
Docket No. 50-255 Consumers Power Co~pany ATTN:
Mr. R. B. DeWitt Vice President Nuclear Operations 212 West Michigan Avenue Jackson, MI 49201 Gentlemen:
The enclosed IE Bulletin No. 79-17, Revision 1 is forwarded to you for action.
A written response is required. If you desire additional information regarding this matter, please contact this office.
Enclosure:
IE Bulletin No. 79-17, Revision 1 cc w/encl:
Mr. J. G. Lewis, Manager Central Files Director, NRR/DPM Director, NRR/DOR PDR Local PDR NSIC TIC Ronald Callen, Michigan Public Service Commission Myron M. Cherry, Chicago RIII Heist'an/bk 10/29/79 RIII G~
Keppler
~04?
Sincerely, James G. Keppler Director
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 October 29, 1979 SSINS No. :
6820 Accession No.:
7908220117 J IE Bulletin No. 79-17 Revision 1 P1PE CRACKS IN STAGNANT BORATED WATER SYSTEMS AT PWR PLANTS Description of Circumstances:**
IE Bulletin No. 79-17, issued July 26, 1979, provided information on the cracking Rl experienced to date in safety-related stainless steel piping systems at PWR Rl plants.
Certain actions were required of all PWR facilities with an operating Rl license within a specified 90-day time frame.
Rl After several discussions with licensee owner group representatives and inspection Rl agencies it has been determined that the requirements of Item 2, particularly Rl the ultrasonic examination, may be impractical because of unavailability of Rl qualified personnel in certain cases to complete the inspections within the time Rl specified by the Bulletin.
To alleviate this situation and allow licensees the Rl resources of improved ultrasonic inspection capabilities, a time extension and Rl clarifications to the bulletin have been made.
These are referenced to the Rl affected items of the original bulletin.
Rl 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 por-tions. of systems which contain oxygenated, stagnant or essentially stagnant bor-ated 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 eithef an inter-granular 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 Circular No. 76-06 (copy enclosed) 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 buildup 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.
Rl - Identifies those additions or revision to IE Bulletin No. 79-17
IE Bulletin No. 79-17 Revision 1 October 29, 1979 Page 2 of 5 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.
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 11-14 11-12 11-10 11-8 11-2 11 etc.) of the above systems.
It is important to note that six of the crack indications were reportedly found in 2 1/2-inch diameter Rl 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 crack indications were found in two Rl high pressure injection lines containirig stagnated borated water.
No crack Rl indications were found in high pressure injection lines which were utilized for Rl makeup operations.
Recent data reported from Three Mile Island Unit 1 indicates that the extent Rl of IGSCC experienced in stainless steel piping at that facility may be more Rl limited than originally stated above.
Of the 1902 total welds originally Rl inspected 350 contained U.T. indications which required further evaluation.
Rl These 350 welds have been reinspected with a second U.T. procedure which pur-Rl portedly provides better discrimination between actual cracks and geometrical Rl reflectors.
Hence, the licensee now estimates that approximately 38 of the Rl 350 welds contain IGSCC and the remaining welds, including those in high pressure Rl injection and decay heat lines, contain only geometrical reflectors.
Further Rl metallurgical analysis of these welds is required to verify the adequacy of the Rl U.T. procedures and to determine the nature of the cracking.
Rl
IE Bulletin No. 79-17 Revision 1 October 29, 1979 Page 3 of 5 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 (July 26, 1979) 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, conta}nment spray and borated water storage tank (BWST-RWST) piping.
.,, Rl For this review, the term 11 stagnant, oxygenated borated water systems 11 refers Rl to those systems serving as engineered safeguards having no normal operating Rl functions and contain essentially air saturated borated water where dynamic Rl flow conditions do not exist oh a continuous basis.
However, these systems Rl must be maintained ready for actuation during normal power operations.
Where Rl your definition for stagnant differed from the one given above please supple-Rl ment your previous response within 30 days of this Bulletin revision.
Rl (a) Provide the extent and dates of the hydrotests, visual and volumetric examinations performed per 10 CFR 50.55a(g) (Re:
IE Circular No. 76-06 enclosed) of identified systems.
Include a description of the non-destructive examination procedures, procedure qualifications and accep-tance 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 or recirculation procedures to maintain required water chemistry with respect to pH, B, Cl-, F-, o2.
(c) Describe the.preservice NOE 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, 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.
All operating PWR facilities shall complete the following inspection on the Rl stagnant piping systems identified in Item 1 at the earliest practical date Rl not later than twelve months from the date of this bulletin revision.
Fa-Rl cilities which have been inspected in accordance with the original Bulletin, Rl Sections 2(a) and 2(b) satisfy the requirements of this Revision.
Rl
e IE Bulletin No. 79-17 Revision 1 October 29, 1979 Page 4 of 5 (a) Until the examination required by 2(b) is completed a visual examination Rl shall be made of all normally accessible welds of the engineered safety Rl systems at least monthly to verify continued systems integrity.
Sim-Rl ilarly, the normally inaccessible welds, shall be visually examined
.,,Rl during each cold shutdown.
,,'Rl The relevant provisions of Article IWA 2000 of ASME Code Section XI and Article 9 of Section V are considered appropriate and an acceptable basis for this examination.
For insulated piping, the examination may be conducted without the removal of insulation.
During the examination particular attention shall be given to both insulated and noninsulated piping for evidence of leakage and/or boric acid residues which may have accumulated during the service period preceding the examination.
Where evidence of leakage and/or boric acid residues are detected at locations, other than those normally expected, (such as valve stems, pump seals, etc.) the piping shall be cleaned (including insulation removal) to the extent necessary to permit further evaluation *of the piping condition.
In cases where piping conditions observed are not sufficiently definitive, additional inspections (i.e.; surface and/or volumetric) shall be conducted in accordance with Item 2.(b).
. (b)
An ultrasonic examination shall be performed on*a representative sample of circumferential welds in normally accessible portions of systems identified by 1 above.
It is intended that the sample *number of welds selected for examination include all pipe diameters within the 2 1/2-inch to 24-inch range with no less than a 10 percent sampling being taken.
The approach to selection of the sample shall be based on the following criteria:
(1)
Pipe Material Chemistry - As a first consideration, those welds in austenitic stainless steel piping (Types 304 and 316 ss) having 0.05 to 0.08 wt. % carbon content based on available material certification reports.
(2)
Pipe Size and Thickness - An unbiased mixture of pipe diameters and actual wall thickness distributed among both horizontal and vertical piping runs shall be included in the sample.
Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl Rl (3) System Importance - The sample welds shall focus the examination Rl primarily on those systems required to function in the emergency Rl core cooling mode and secondly, on the containment spray system.
Rl The U.T. examination sample may be focused on noninsulated piping Rl runs.
The evaluation shall cover the weld root fusion zone and a Rl minimum of 1/2 inch on the pipe I.D. (counterbore area) on each side Rl of the weld.
The procedure(s) for this examination shall be essentially Rl
- Normally accessible refers to those areas of the plant which can be entered Rl during reactor operation.
Rl
e IE Bulletin No. 79-17 Revision 1 October 29, 1979 Page 5 of 5 in accordance with ASME Code Section XI, Appendix III and Supplements Rl of the 1975 Winter Addenda, except all signal responses shall be eval-Rl uated as to the nature of the reflectors.
Other alternative examination Rl methods, combination of methods, or newly developed techniques may be Rl used provided the procedure(s) have a proven capability of detecting
- Rl stress corrosion cracking in austenitic stainless steel piping.
~ Rl For welds of systems included in the sample having pipe wall thickness Rl of 0.250 inches an~-below, visual and liquid penetrant surface examina-Rl tion may be used in lieu of ultrasonic examination.
Rl (c) If cracking is identified during Item 2(a) and 2(b) examinations, all Rl welds in the affected system, shall be subject to examination and repair Rl considerations.
In addition, the sample welds to be examined on the Rl remaining normally accessible noninsulated piping shall be increased to Rl 25 percent using the criteria outlined in paragraph 2(b).
In the event Rl that cracking is identified in other systems at this sampling *level, Rl all accessible and inaccessible welds of the systems identified in Rl item 1 shall be subject to examination.
Rl
- 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 50.55a(g) unless justification for con-tinued operation is provided.
- 4.
Any cracking identified shall be reported to the Director of the apppropriate 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 Rl Office within 30 days of the date of this bulletin revision addressing the Rl results of your review if required by Item 1.
Provide a schedule of your Rl inspection plans in response to Item 2(b) in those cases in which the Rl inspections have not been completed.
Rl
- 6.
Provide a written report to the Di recto~ of the appropriate NRC Regional Rl Office within 30 days of the date of completion of the examinations required Rl by Items 2(a), 2(b), or 2(c) describing the inspection results and any cor-Rl rective actions taken.
Rl
- 7.
Copies of the reports required by Items above shall also be provided to the Director, Division of Operating Reactors, Office of Inspection and Enforce-ment, Washington, D.C.
20555.
Approved by GAO, 8180225 (R0072), clearance expires 7/31/80.
Approval was given under a blanket clearance specifically for identified generic problems.
Enclosures:
- 1.
IE Circular No.* 76-06
- 2.
List of IE Bulletins Issued in the Last Six Months