ML20210T960

Provides Rept on Status of Util RPV Feedwater Nozzle Insps Performed in Response to USI A-10 Re BWR Nozzle Cracking

ML20210T960
Person / Time
Site:	Monticello
Issue date:	08/12/1999
From:	Day B NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-A-10, REF-GTECI-RV, TASK-A-10, TASK-OR NUDOCS 9908190198
Download: ML20210T960 (7)

Text

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I l Northern States Power Company l

' DN '

. , Monticello Nuclear Generating Plant

. 2807 West County Road 75 l Monticello, MN 55362 August 12,1999 US Nuclear Regulatory Commission Attn: DocumenTControl Desk Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Reactor Pressure Vessel Feedwater Nozzle Inspections at Monticello

References:

i

1. Letter from Thomas A. Ippolito, NRC, to L. O. Mayer, NSP, " Implementation of l Unresolved Safety issue A-10, BWR Nozzle Cracking," August 27,1981

2. Letter from L. O. Mayer, NSP, to Director of Nuclear Reactor Regulation, NRC,

" Additional Information Related to Resolution of Unresolved Safety issue A-10, BWR Nozzle Cracking," November 3,1981

3. Letter from Thomas M. Parker, NSP, to Director, Office of Nuclear Reactor Regulation, NRC, " Refueling Outage No.13 Status and Proposed in-Vessel i Visual PT Inspection Schedule for Monticello Feedwater Nozzles" October 4, )

1989 I

4. Letter from William O. Long, NRC, to Thomas M. Parker, NSP, "Monticello - l Feedwater Nozzle PT Inspection Program (TAC No. 75101)," April 18,1990 The purpose of this letter is to report status of the Monticello reactor pressure vessel l (RPV) feedwater nozzle inspections performed in response to Unresolved Safety issue i A-10, BWR Nozzle Cracking.

In 1975, the Monticello stainless steel clad feedwater nozzles were inspected to' determine if cracking occurred as had been observed at otherindustry BWRs. This 9908190198 990812 ADOCK 05000263 PDR G PDR 8/11/99 S:S J LICENSDTopese\R P V\ST ATUS.LTRS 20 99 doc /

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USNRC NORTHERN STATES POWER COMPANY Abgust 12,1999 inspection disclosed some indications in the clad area on all four nozzles. These indications were completely removed from the stainless steel cladding at that time by ]

grinding. In 1977, the stainless steel cladding at Monticello was completely removed to mitigate further feedwater nozzle blend radius and bore cracking.

In 1989 NSP documented results of a plant specific review (Ref. 3). This letter proposed a four part, long term inspection program based on NUREG-0619 "BWR l Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking" which the NRC l

accepted (Ref. 4). The four commitments associated with References 3 and 4 are:

1. Review on-line feedwaternozzle thermal sleeve leak detection system data on a l monthly basis.

2. Perform extemal UT examinations on two of tha fourfeedwater nozzles each refueling outage.

3. Perform visualinspections of the spargers and the nozzle blend radius area of all four feedwater nozzles each refueling outage.

4. Perform PT examinations of nozzles at the next appropriate opportunity in the event that; a.) UT examinations indicate a flaw or b.) Excessive leakage (greater than 0.3 gpm) is identified by the on-line leakage monitoring systems.

The NRC Safety Evaluation Report (SER) attached to Reference 4 states that NSP wi!!

continue inspections for "9 Inspection Interval-Refueling' Cycles or 135 Start- )

up/ Shutdown Cycles" as stated in NUREG-0619. The inspection interval began with i installation of welded thermal sieeves during the 1981 refueling outage. With completion of inspections during the 1998 refueling outage, Monticello had completed the required 9 Inspection Interval-Refueling Cycles with no observed degradation of the feedwater nozzles. Commitments 1,2, and 4 have been fully implemented as identified in Attachment 2,"Feedwater Nozzle ISI Exam History."

Commitment 3 specified a visualinspection of the spargers and nozzle blend radius areas on all four feedwater nozzles each outage (Ref. 3). All four nozzles were visually inspected in 1984,1986,1987,1989,1991 and 1996; however, during the 1993,1994, and 1998 outages, only two of the nozzles were visually inspected, see Attachment 2.

This deviation from the 1990 commitment occurred due to a 1991 augmented inservice inspection (ISI) database programming error. These missed surveillance's have been evaluated under NSP's corrective action program.

i Page 2

o USNRC NORTHERN STATES POWER COMPANY August 12,1999

Visual inspections of the spargers and inner nozzle blend radius area and online leak detection monitoring were performed to aid in assuring that the cracking phenomena had been eliminated. Experience at Monticello and at BWRs elsewhere has shown that the unclad feedwater nozzle design is not susceptible to cracking. Since removal of the stainless steel cladding in 1977, no further thermal induced cracking has been detected. The missed inspections did not result.in a reduction in safety since no thermal induced cracking was found in the nozzles during any prior inspections, nor has there been any evidence of leakage identified by the online feedwater nozzle thermal sleeve leak detection system. Since the 1998 outage visualinspection did not include all four feedwater noz:les, NSP will perform one additional visual inspection beyond the original Commitment 3 of Reference 3. NSP therefore makes the following new commitment:

NSP will visually inspect reactorpressure vessel feedwater nozzles N-4A and N-4D from the vesselID during the next refueling outage currently scheduled for January 2000. l The feedwater nozzles and thermal sleeves were modified to minimize bypass flow through the thermal sleeve. The online leak detection system was installed for the primary purpose of monitoring for bypass leakage that can lead to thermal stress on the nozzle. Based on industry experience, as well as Monticello specific ultrasonic testing / visual inspection history and online feedwater nozzle thermocouple monitoring experience, the potential for significant bypass leakage has been eliminated.  !

An ultrasonic testing (UT) technique is now performed at a 70' beam angle, which can effectively evaluate fatigue cracking initiation on the inner radius nozzle surface by examination from the reactor vessel outer shell region. With the cladding now removed, the UT technique is more effective than previous techniques because the boundary is more defined (no clad to metalinterface). This technique has been analytically modeled and performance demonstrated on a nozzle mockup to detect thermal fatigue surface crack initiation and can effectively scan greater than 90% of the inner radius nozzle surface area.

Due to the excellent performance history of the Monticello feedwater nozzles with thermal sleeves and improved UT scan detection technique for identifying smaller flaws, continued monitoring in accordance with NUREG 0619, is believed to no longer be warranted. Thus, the four commitments of Reference 3 are being replaced with the above new commitment which will be completed during the 2000 refueling outage. Our ongoing ISI UT testing program based on ASME Section XI testing frequency provides an effective and reliable means for early detection of thermal fatigue crack growth l within the feedwater nozzles.

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'4m USNRC NORTHERN STATES POWER COMPANY August 12,1999 Please contact Sam Shirey, Sr. Licensing Engineer, at (612) 295-1449 if you require further information.

Byhn D. Day p PlaYit Manager Monticello Nuclear Generating Plant c: Regional Administrator-lli, NRC NRR Project Manager, NRC Sr. Resident inspector, NRC State of Minnesota, Attn: Steve Minn Attachments:

1. Status of Commitments on Feedwater Nozzles

2. Feedwater Nozzle ISI Exam History i

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