IR 05000333/1994011
| ML20029D813 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 04/28/1994 |
| From: | Mcbrearty R, Modes M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20029D808 | List: |
| References | |
| 50-333-94-11, NUDOCS 9405100138 | |
| Download: ML20029D813 (6) | |
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l U.S. NUCLEAR REGULATORY COMMISSION
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REGION I
DOCKET / REPORT NO.:
50-333/94-11
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LICENSEE:
New York Power Authority FACILITY:
James A. FitzPatrick Nuclear Power Plant INSPECTION AT:
Lycoming, New York DATES:
April 11-15,1994
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INSPECTOR:
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Robert A. McBrearty,' Reactp/ Engineer Date
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Materials Section l
Division of Reactor Safety APPROVED BY:
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Micitael C.1 Modes ~, Chief Date Materials Section Division of Reactor Safety A_Icas Insoected: An announced inspection was conducted of erosion / corrosion program implementation, licensee action on previous inspection findings, and licensee actions relative i
to the reactor pressure vessel weld augmented volumetric examination, required by the Code
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of I'ederal Regulations, Title 10, Part 50.55a(g).
Results: Unresolved item no. 92-05-01, regarding the licensee's method for ultrasonically examining welds repaired with the weld overlay repair method, was closed. The erosion / corrosion program implementation was determined to be effective and capable of detecting pipe wall thinning prior to pipe failure, and the requirements related tc the augmented vessel weld examination are clearly understood by the licensee.
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PDR ADOCK 05000333 l
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DETAIIS 1.0 LICENSEE ACTION ON PREVIOUS INSPECTION FINDINGS (Closed) Unresolved item 92-05-01. Ultrasonic Examination of Weld Overlay R_epaired Welds The ability to detect defects ultrasonically is dependent on establishing sufficient examination sensitivity. For the ultrasonic examination of cracked welds that were repaired with the weld overlay repair method, the Electric Power Research Institute (EPRI) NDE Center at Charlotte, North Carolina, originally recommended that the calibration block containing weld overlay exactly match the production weld. EPRI has since revised its recommendations and recognized that the calibration method used at FitzPatrick and other nuclear facilities is a viable way of determining the acceptability of welds that have been repaired with weld overlay.
The calibration technique employed at FitzPatrick uses the repaired weld as the calibration block to establish test sensitivity. With the transducer on the production weld, the examination system gain is increased until a noise level ranging from 5% to 15% of full screen height is displayed on the cathode ray tube. That produces the greatest usable sensitivity level for the particular weld and is the level at which the examination is performed. Further, the material characteristics remain unchanged from examination to examination so that examination reproducibility is achieved.
The inspector reviewed data associated with overlay repaired 28" diameter recirculation system welds 28-02-2-33 and -53 and ascertained that the calibration method described above, and in the applicable procedure, JAF-UT-W81-6, Revision 3, dated January 28, 1992, was used for the examination of those welds.
Based on the above, this item is considered closed.
2.0 INSERVICE INSPECTION ACTIVITIES (IP73753)
The NRC has revoked all previously granted reliefs under paragraph 50.55a to licensees for the extent of volumetric examination of reactor vessel shell welds specified in Item Bl.10 of Examination Category B-A of the ASME Boiler and Pressure Vessel Code,Section XI.
Additionally, an augmented volumetric examination of those welds to the extent required by the 1989 Edition of Section XI, is required of all licensees on a one-time basis.
Licensees whose ten-year inspection interval in effect on September 8,1992, had less than 40 months remaining as of September 8,1992, can defer the performance of the augmented examination until the first period of its next interval. The inspection interval at the FitzPatrick plant that was in effect on September 8,1992, had less than 40 months remaining, and the licensee expressed its intent to defer the examination. The licensee, additionally, is considering extending its present inspection interval, as permitted by
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Section XI. The licensee is concerned that, by extending the present interval to the maximum allowed by the Code, it may preclude deferral of the augmented examination.
The inspector contacted cognizant NRC personnel at NRR regarding the licensee's concern and was advised that taking the full extension would not preclude the licensee's eligibility to defer the augmented vessel weld examination until the first period of its next inspection interval.
At the exit meeting, the inspector discussed with the licensee his conversation with NRR and advised the licensee to contact NRR prior to taking action regarding the examination deferral, along with the interval extension.
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During the ongoing 1994 maintenance outage, the licensee performed a visual inspection of valves and associated bolting to satisfy its inservice inspection commitment in that regard.
The valve internals are inspected when the valve is disassembled for some other purpose.
The inspector reviewed visual inspection data to ascertain that the inspections were properly performed and that the results were documented. Data related to the following valves were selected for review:
e 12 MOV-15 e
13 MOV-15 e
23 MOV-15 e
29 AOV-80A e
34 NRV-Ill A and -lllB The visual inspections. VT-1 and VT-3, were performed of the valve bolting and internal surfaces by properly qualified and certified visual inspectors. The inspection results were appropriately documented, and the as-found condition of each component was clearly described. Additionally, the documentation indicated the acceptability of the specific component.
Conclusions The licensee, if it wishes, can defer the performance of the reactor pressure vessel augmented volumetric examination required by the Code of Federal Regulations, Title 10, Part 50.55a(g), until the first period of its next ten-year inspection interval. Additionally, it can extend the current interval, as permitted by the ASME Code,Section XI, ac' retain its eligibility to defer the augmented examination.
The visual inservice inspection of valves and related bolting was performed by qualified visual inspectors, and the results were clearly documented.
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3.0 EROSION / CORROSION PROGRAM IMPLEMINI ATION (IP49001)
During the current 1994 maintenance outage, the licensee performed erosion / corrosion inspections to obtain more accurate data regarding the rate of pipe wall thinning that was occurring.
The inspector reviewed the applicable implementing Procedure No. JAF-UT-W81-9, Revision 5, inspection results and the calculations that were performed to determine the remaining life of specific components.
Procedure No. JAF-UT-W81-9, Revision 5, dated March 21, 1994, entitled, " Ultrasonic Thickness Measurements," performs double duty in that it controls ultrasonic thickness measurements to determine weld profiles during the performance of inservice inspection (ISI), and the wall thickness measurements obtained to detect pipe wall thinning caused by crosion/ corrosion. Two data sheets are included for reporting examination results. One form is for inservice inspections, and the other is for reporting erosion / corrosion inspection results. The erosion / corrosion (E/C) data sheet requires that wall thickness screening criteria be identified on the form; the ISI form does not require that information.
The inspector noted that the ISI data sheet was used to report E/C results of small bore piping inspections, and the E/C data sheet was used for reporting all other E/C inspection results. The procedure does not identify when each form should be used. When the discrepancy was brought to the licensee's attention, Action Item No.11141, dated April 14,1994, was issued to document the licensee's commitment to correct the discrepancy by revising the procedure. Additionally, the appropriate screening information is being entered on the various ISI data sheets that reported E/C results.
The inspector noted that indication notification report (INR) was issued for all of the components that displayed wall thinning in excess of the screening criteria established by the licensee even when that information was not included on the data sheet. For that reason, the significance of the discrepancy was considered to be of minor importance. The INR resulted in further analysis and disposition of the component and, in some cases, the performance of engirteering calculations to determine the rate of wall thinning and the remaining life of the component. That information was also used to decide when the next inspection is required of the component.
The licensee's implementation of its E/C program is effective in that pipe wall thinning is detected before failure of the pipe occurs and inspections are scheduled so that information is available to accurately determine the rate of wall thinning and the remaining life of the component.
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During the present outage, approximately 250' of small bore piping is being replaced based on plant experiences of leaks that developed in similar configurations in the past. Small bore piping has been added to the E/C program this year, and historical thickness measurement l
data are not available. The original carbon steel piping is being replaced with chromium molybdenum material, which is considered to be more resistant to E/C than the carbon steel.
Conclusions The licensee's E/C monitoring program is effective in that pipe wall thinning is detected prior to failure. The evaluation of inspection results i. performed by the staff engineer s
assigned to the E/C program; and, based on those results, future inspections are scheduled.
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Prompt action was taken by the licensee to correct a procedural discrepancy identified by the NRC although the significance was determined to be of minor importance.
l 4.0 EXIT MEETING The inspector met with licensee representatives, denoted in Attachment 1, at the conclusion of the inspection on April 15, 1994. The inspector summarized the scope and findings of the inspection, which were acknowledged by the licensee.
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ATI'ACIIMENT 1 Persons Contacted New York Power Authority M. J. Colomb General Manager, Support Services
R. Converse Senior Assessment Engineer
D. Franger Senior Quality Assurance Engineer, NDE Level III
J. Heddy Senior Licensing Engineer
D. Lindsay General Manager, Maintenance
R. Penny Maintenance Engineer, White Plains Office
H. Salmon Resident Manager
A~. Smith -
Nuclear Ops. & Maint. Engineer, White Plains Office
.i U.S. Nuclear Regulatory Commission W. Cook Senior Resident Inspector
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