JPN-95-005, Submits Results of Reactor Vessel Head Weld Inspections Conducted During 1995 Refueling Outage.Insp Data Shows No Changes from 1992 Insp Data

Submits Results of Reactor Vessel Head Weld Inspections Conducted During 1995 Refueling Outage.Insp Data Shows No Changes from 1992 Insp Data

ML20078N268
Person / Time
Site:	FitzPatrick
Issue date:	02/09/1995
From:	William Cahill POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
JPN-95-005, JPN-95-5, NUDOCS 9502150246
Download: ML20078N268 (5)

Text

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February 9,1995 JPN-95-005 United States Nuclear Regulatory Commission Document Control Desk Mail Station P1-137 Washington, D.C. 20555

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Reactor Vessel Head Weld Flaw Indication inspecions_andlvaluation_ Analysis

References:

1. NYPA letter, 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 Head Flaw Indication Inspection j

and Evaluation Submittal - J.A. FitzPatrick Nuclear Power Plant."

3. NYPA letter, R. E. Beedle to NRC (JPN-92-036) dated July 9,1992,

" Reactor Pressure Vessel Head Welds Flaw Indication Inspections and Evaluation Analysis."

4. NRC letter, B. C. McCabe to R. E. Beedle (TAC M84030) dated August 11,1992, " Evaluation of Reactor Vessel Head Flaw Indication Inspection and Evaluation Submittal - James A. FitzPatrick Nuclear Power Plant."

Dear Sir:

Indications were detected in reactor vessel head weld VC-TH-1-2 during routine L

in-service inspections in 1990. Evaluations of the indications were performed in accordance with the ASME code. Reference 1 transmitted these evaluations to the NRC. These evaluations confirmed the existence of subsurface flaws due to original I

welding or manufacturing imperfections. They concluded that reactor operation with the existing weld flaws did not constitute a safety concem.

in order for the reactor vessel head to be accepted for continued service, the NRC requested in Reference 2 that a reexamination of the weld be performed during the next three refueling outages. The NRC requested that the results of the reexaminations be incorporated into an analytic evaluation to justify operation.

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Reference 2 stated that the 3 valuations should be submitted to the NRC for staff review prior to resumption of reactor operation from each of the three subsequent

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operating cycles.

The subject vessel head weld was inspected during the 1992 refueling outage and the examination results were incorporated into an analytic evaluation. The results were transmitted to the NRC in Reference 3. Based on the results of these i

evaluations, it was concluded that continued operation with the existing vessel head indications does not constitute a safety concern. The NRC requested in Reference 4 that the vessel head weld be reexamined during the next two refueling outages, and a reevaluation of the flaw !ndications be pedomied to justifsj operation for each of the subsequent two fuel cycles.

The attachment to this letter includes the results of the reactor vessel head weld inspections conducted during the 1995 refueling outage. The inspection data i

shows no changes from the 1992 inspection data. Therefore, the original flaw

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evaluation remains valid.

j Based on the results of these inspections, the reactor vessel head weld indications do not constitute a safety concern.

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If you have any questions, please contact Ms. C. D. Faison.

I Very truly yours, gg; i

William J.

ahill, Jr.

Chief Nuclear Officer l

att:

As stated i

cc:

Regional Administrator l

U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Office of the Resident Manager U.S. Nuclear Regulatory Commission P.O. Box 13G Lycoming, New Ycrk 13093 l

i Mr. C. E. Carpenter, Project Manager i

Project Directorate 1-1 Division of Reactor Projects 1/11 U.S. Nuclear Regulatory Commission Washington, D.C. 20555

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Attachment to JPN-95-005 James A. FitzPatrick Nuclear Power Plant l

Reactor Vessel Head Weld Flaw Indications In-Service inspection Results Introduction i

o As a part of the routine in-service inspection (ISI) program, selected reactor pressure vessel (RPV) head welds were inspected during the 1990 refueling outage.

Several recordable indications were detected in RPV head weld VC-TH-1-2. Evaluations of the indications were performed in accordance with the ASME code. This report summarizes the results of initial and follow-up evaluations, including the results of the weld examination performed during the 1995 refueling outage.

t UltrasonicJospections - 1990 RefueLQutage_Be_sults UT inspections of RPV head weld VC-TH-1-2 showed several 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 these findings, additional examinations were performed in accordance with j

the requirements of ASME Section XI, paragraph IWB-2430 as stated in Reference 1.

Other inspections, beyond those required by ASME Section XI, were conducted on weld VC-TH-1-2 to clarify the nature and extent of the flaw indications. These supplemental inspections included visual (VT), radiographic (RT), dye penetrant (PT), and magnetic particle (MT) examinations on the reactor side (underside) of the vessel head. Additional l

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.

Some of the UT exams used in sizing these flaws were hampered by the existence of numerous small reflectors located about mid-wall in the plate. These reflectors are believed to be metallic inclusions (also known as plate segregates), 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-service UT inspections.

When performing sizing exams with refracted longitudinal wave transducers, shear and longitudinal sound waves are generated. The UT inspectors initially confused the segregate response from shear waves with a flaw response from longitudinal waves. The shear waves reflected off the segregates generated a response near the center of the l

plate on the time display. This resembled a response from the longitudinal waves which was interpreted as a flaw. As a result, inspectors overestimated the flaw depth to be two inches. The length of the flaw was similarly overestimated.

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Attachment to JPN-95-005 James A. FitzPatrick Nuclear Power Plant 1

Reactor Vessel Head Weld Flaw Indications In-Service Inspection Results Elaw_ Evaluation Two flaws were rejectable under the guidelines of NRC Regulatory Guide 1.150. These flaws were conservatively estimated to be 0.5 inches deep by 5 inches long. and 0.55 inches deep by 2.3 inches long. For the purposes of the fracture mechanics evaluation, these flaws were assumed to be open to the vessel interior (i.e. cracks), although the inspection data showed no crack indications on the RPV head inner diameter. The assumption that a crack exists is conservative since it presupposes flaw growth.

This information, and the original structural and detailed fracture mechanics evaluation, were provided to the NRC as Attachments I and 11 to Reference 1.

UltrasonicJnspections - 1992_ Refuel _Qutage_.Besults The ISI inspections performed during the 1992 refueling outage included weld VC-TH-12.

l The inspections were performed by Ebasco Services Inc., the ISI contractor, with additional inspections and final data review conducted by two New York Power Authority (NYPA) Quality Assurance (OA) level lll inspectors. Although not required, all inspections i

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 inspection. When the initial 1990 examinations were performed, no permanent references existed to ensure repeatability.

To ensure the reexaminations captured indications identified in 1990, inspections performed in 1992 included an area larger than the locations reported in the early examinations. This also enabled NYPA personnel to develop permanent reference marks for repeatability when performing future examinations. The examination performed in 1992 on RPV head weld VC-TH-1-2 confirmed the two indications previously reported as unacceptable in 1990.

Inconsistencies 3etweenl990_andl992Aata i

The evaluation identified some differences between the recorded data of 1990 and 1992.

The 1992 recorded data included shorter length measurements and smaller through wall dimensions. These differences prompted supplemental examinations by NYPA OA level t

lli personnel and a complete reevaluatiore of all 1990 and 1992 inspection data to detennine final disposition of these indications. The examinations by NYPA, reevaluation of all data, and subsequent discussions with GE and Ebasco personnel resolved the differences noted in the inspection data.

The differences between 1990 and 1992 data were attributed to differences in the evaluation techniques used during the inspections. The longer lengths and greater through wall dimensions reported in 1990 are from using composite data (consolidating l

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,e Attachment to JPN-95-005 James A. FitzPatrick Nuclear Power Plant Reactor Vessel Head Weld Flaw Indications in-Service Inspection Results t

inspection results of various examination angles and combining automated with manual j.

inspections).

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Determining the dimensions of indications using the 1990 evaluation technique is an extremely conservative methodology exceeding the sizing criteria outlined in ASME Section XI arkd NRC Regulatory Guide 1.150, " Ultrasonic Testing of Reactor Vessel Welds During Pre-service and Inservice Examinations," Rev.1. Applying this type of conservative sizing in the structural evaluation assures a large safety margin between the evaluated flaw size and the actual flaw size. Duplication of this type of evaluation, when applying the sizing criteria outlined in the goveming codes and documents, is not possible due to the amount of conservatism built into the bounding rectangle. The inspection and evaluation, performed during the 1992 examinations, used criteria outlined in ASME section XI and NRC Regulatory Guide 1.150, Rev.1. When data from the 1990 and 1992 inspections are compared alter analysis, no measurable change in length or through wall dimension is discemable.

UltrasonicJnspections -1995_BefueLOutage_Elesults l

Weld VC-TH-1-2 was reinspected during the 1995 refueling outage. The inspections were performed by Raytheon personnel (the new owner of Ebasco Services Inc.), certified in i

the EPRl/BWROG IGSCC Training Course. in-process surveillances and final data review was conducted by NYPA OA level lll inspectors, also certified in the EPRl/BWROG IGSCC Training Course. The inspection techniques and equipment used during the 1995 reexaminations were comparable to those employed during the 1990 and 1992 inspections. The examination on weld VC TH-1-2 confirmed the two indications previously reported as unacceptable in 1990.

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When the data is compared with the results of the 1990 and 1992 inspections, no measurable change in length or through wall dimension is discernable. Therefore, the original flaw evaluation remains valid.

Conclusion I

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Based upon the inspections, and manufacturing records for the RPV head, the Authority determined that the flaws are due to original manufacturing imperfections.

Based upon the results of evaluations performed in accordance with the Technical Specifications, the ISI program, and ASME section XI, continued operation with the existing reactor vessel head indications does not constitute a safety concern.

The Authority will re-inspect the weld during the next refueling outage in accordance with ASME Section XI (1980 edition through Winter,1981 addenda) and will report any changes to the NRC in the inspection results.

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