JPN-92-036, Provides Results of Reactor Vessel Head Weld Insp Conducted During 1992 Refueling Outage
| ML20101Q462 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/09/1992 |
| From: | Ralph Beedle POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| JPN-92-036, JPN-92-36, NUDOCS 9207140290 | |
| Download: ML20101Q462 (5) | |
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03 Mon Street Whee Plains. Ne* Yora 10601 914 681 6846
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July 9,1992 JPN 92-036 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop P1 137 Washington, D.C. 20555
Subject:
James A. FitzPatrick Nuclear Power Plant Dockot No. 50 333 Reactor Pressure Vessel Head Welds Flaw Indication lasantions and Evaluation Analysla
References:
- 1. NYPA lotter, J. C. Brons to NRC (JPN 90-040) dated May 25, 1990, " Reactor Pressure Vessel Head Flaw Indication inspections and Evaluation Analysis."
- 2. NRC letter, D. E. LaBarge to J. C. Brons (TAC 76861) dated June 13,1990, " Evaluation of Reactor Vessel Hoad Flaw Indication inspection and Evaluation Submittal -
J. A. FitzPatrick Nuclear Power Plant."
- 3. NYPA letter, H. P. Salmon, Jr. to J. P. Durr (JAFP-92-0360) dated April 30,1992, "NRC Inspection Report 50-333/92 05."
Door Sir:
Indications of possible flaws in a rcactor pressure vossoi hoed weld were found during routino in-servico refueling outage inspections in 1990. Evaluations of the indications were performed in accordance with the ASME code. Reference 1 transmitted these ovaluations to the NRC. These ovaluations confirmed the existence of subsurface flaws due to original we! ding imperfecticas. They concluded that reactor operation with the existing wold flaws did not constitute a safety concern.
In order foi the reactorwssel head to be accepted for continued service, the ASME code requires a reexamination of the wold during the next three refueling outages. The NRC requested in Refeience 2 that the results of the reexaminations be incorporated into an analytic evaluation to justify operation. Reference 2 stated that the evaluations Aould be submitted to the NRC for staff review prior to resumption of reactor operation from each of the three subsequent operating cyclee r,
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The attachment to this letter pruvidos the results of the reactor vessel head weld inspections conducted during the 1992 refueling outage. The inspection data was submitted to the NRC in Reference 3. Inconsistencies between 1990 and 1992 examination data have been resolved as described in the attachment.
Based on the results of those inspections, the reactor vossel head weld flaws do not constitute a safety concern.
If you have any questions, please contact Mr. J. A. Gray, Jr.
Very truly yours, m
w...
Ralph E. Boodle Executive Vice President Nuclear Generation cc: Regional Administrator U.S. Nuclear Rogu'Jtory Commission 475 Allendale Road King of Prussia, PA 19400 Office of the Hesid;.nt inspector U.S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, New York 13093 Mr. B. C. McCabo Project Directorate 1 1 Division of Reactor Projects 1/11 U.S. Nuclear Regulatory Commission Washington D.C. 20555 1
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Attachmsnt to JPN-92-036 James A. FitzPatrick Nuclear Power Plant Reactor Vessel Head Wold Flaw Indications 1992 In service inspection Results Introduction As part of the routine in-service inspection (ISI) program, selected reactor pressure vessel (RPV) head wolds were inspected during the 1990 refueling outago. Inspectors used ultrasonic testing (UT) techniques to detect and size flaws in reactor pressure vessel head welds.
Ultrasonic Insoections - 1990 Hofuel Outaae Results UT inspections of RPV head weld number VC TH 1+2 showed soveral recordable indications. The largest indication was observed along approximately five inches of the circumferential weld between the upper dome plate (dollar plate) and the vertical dome segments. These indications were the subject of NRC Information Notice 90 32 and General Electric Company Rapid information Communication Services Information Letter (RICSIL) 051. Both documents are dated May 3,1990.
As a result of those findings, additional examinations were performed in accordance with the requirements of ASME Section XI, paragraph IWB 2430 as stated in Reference 1.
Other inspections, beyond thoso required by ASME Section XI, were conducted on wold VC-TH 1-2 to clarify the nature and extent of the flaws. These supplemental i
inspections included visual (VT), radiographic (RT), dye penetrant (PT), and magnetic particle (MT) examinations on the reactor side (underside) of the vessel head.
Additional UT exams were performed from both the outside and inside of the head.
Construction radiographs and those taken during the 1990 refueling outage were computer enhanced to better quantify the weld characteristic.
Soms of the UT exams used in sizing these flaws were hampered by the existence of numerous small reflectors located about mid-wall in the plato. Those reflectors are believod to be metallic inclusions (also known as plate segregatos), probably manganese sulfides. These inclusions are part of the steel making process and are considered acceptable by the manufacturing specification for ASME SA 533 Grade B steel. They were also observed during pre-sarvice UT inspections.
When performing sizing exams with refracted longitudinal i transducers. shear and longitudinal sound waves are generated. The UT inspectors..ntially confused the segregate response from shcar waves with a flaw response from longitudinal waves.
The shear waves reflected off the segregates generated a response near the conter of the plate on the time display. This resembled a response from the longitudinal wcves which was interpreted es a flaw. As a result, inspectors overestimated the flaw depth to be 2 inches. The length of the flaw was similarly overestimated.
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Attachm:nt to JPN 92 036 James A. FitzPatrick Nuclear Power Plant Reactor Vessel Head Wold Flaw indications 1992 In sorvice inspection Results Flaw Evaluation Two flaws were rejectable under the guidelines of NRC Regulatory Guide 1.150.
Those f!aws were conservatively estimated to be 0.5 inch doop by 5 inches long, and 0.53 inch doop by 2.3 inches long. For the purposes of the fracture mechanics evaluation, those flaws were assumed to be open to the vesselinterior (i.e. cracks),
although inspection date indicated the contrary The assumption that a crack exists is conservative since it presupposes flaw growth.
This information, and the original structural and detailed fracturo mechanics evaluations, were provided to the NRC as Attacnments I and ll to Reference 2.
This wold was re-in:pocted during the 1992 refueling outage as required by the NRC and ASME Section XI.
Ultrasonic Insoections 1992 Refuel Outaae Results The ISIinspections performod during the 1992 refueling outage included weld VC-TH 12. The inspections were performed by Ebasco Services Inc., the ISI contractor, with additionalinspections and final data review conducted by two Authority quality assurance (QA) level lit inspectors. Although not required, all inspections were conducted by personnel certified by the BWROG EPRI IGSCC program.
The inspection techniques and equipment used during the 1992 reexaminations were comparable to those employed during the 1990 inspections. When the initial 1990 examinations were performed, no permanent references existed to ensure repeatability.
To make sure the 13 examinations captured indications identified in 1990, inspections performed in 1992 included an area larger than the locations reported in the early examinations. This also enabled Authority poisonnel to develop permanent reference marks for repeatability when performing future examinations. The examination performed in 1992 on RPV head wold VC TH 1-2 confirmed the two indications previously reported as unacceptable in 1990.
Inconsistencies Between 1990 and 1992 Data The evaluation identified some differences between the reccrded data of 1990 and 1992. The 1992 recorded data included shorter length measurements and smaller through wall dimensions. These differences prompted supplemental examinations by Authority OA level 111 personnel and a complete reevaluation of all 1990 and 1992 inspection data to determine final disposition of these indications. The examinations by the Authority, reevaluaticri of all data, and subsequent discussions with GE and Ebasco personnel, resolved the differences noted in inspection data.
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o 3ttachment to JPN 93-03G James A. FitzPatrick Nuclear Power Plant Reactor Vessel Head Weld Flaw Indications 1992 In servico inspection Results The differencos betwoon 1990 and 1992 data are attributed to differences in the evaluation techniques used during the inspections. The longer longths and greater through wall dimensions reported in 1990 are from using composite data (consolidating inspection results of various examination angles and combining automated with manual inspections).
Dotormining the dimensions of indications using the 1990 evaluation technique is an extremely conservativo methodology excoeding the sizing criteria outlined in ASME Section XI and NRC Regulatory Guido 1.150, " Ultrasonic Testing of Reactor Vessel Wolds During Pre service and Inservice Examinations," Rev.1. Applying this type of conservativo sizing in the structural evaluation assures a largo safety margin betwoon the ovaluated flaw size and the actual flaw size. Duplication of this type of evaluation when applying the sizing criteria outlined in the governing codes and 'iocuments is not possible due to the amount of conservatism built into the bounding rectangle. The inspection and evaluation, parformed during the 1992 examinations, used criteria outlined in ASME section XI and NRC Regulatory Guido 1.150, Rev.1. When data from the 1990 and 1992 inspections are compared af ter analysis, no measurable change in longth or through wall dimension is discernable.
Conclusion Based upon the inspections, and manufacturing records for the RPV head, the Authority determined that tho flaws are due to original manufacturing imperfections.
The flaws are not cracks because they do not reach the surface.
Based upon the results of evaluations performed in acc.ordance with the Technical Specifications, the ISI program, and ASME section XI, continut.d operation with the existing reactor vessel head indications do not constitute a safety concern.
The Authority will re inspect the weld during the next refueling outage in accordance with ASME Section XI (1980 odition through winter 1981 addenda), IWB 2420, and will report any changes to the NRC in the inspection results.
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