L-PI-08-011, Steam Generator Inservice Inspection Results Response to Request for Additional Information (RAI)

Steam Generator Inservice Inspection Results Response to Request for Additional Information (RAI)

ML080520187
Person / Time
Site:	Prairie Island
Issue date:	02/21/2008
From:	Wadley M Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-PI-08-011
Download: ML080520187 (6)

Text

Prairie lsland Nuclear Generating Plant Operated by Nuclear Management Company, LLC FER 3 'r 2008 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie lsland Nuclear Generating Plant Unit 2 Docket 50-306 License No. DPR-60 Prairie lsland Unit 2 Steam Generator lnservice lnspection Results Response to Request for Additional Information (RAI)

Reference:

1) 2006 Unit 2 Steam Generator lnspection Results Day Report, dated December 18,2006, (ADAMS Accession No. ML063530271)

2) 2006 Unit 2 Steam Generator Category C-3 lnspection Results 30-Day Report, dated December 26,2006 (ADAMS Accession No. ML063600314)

3) Unit 2 lnservice lnspection Summary Report, Interval 4, Period 1, Outage 2, Refueling Outage Dates: 11-15-2006 to 12-15-2006, Fuel Cycle 23: 6-1 1-2005 to 12-15-2006, dated March 14, 2007 (ADAMS Accession No. ML070780140)

4) Email from Mahesh Chawla, NRC, to Nuclear Management Company, LLC (NMC), dated November 14,2007 In response to References 1, 2, and 3, the NRC issued eight questions with the RAI of Reference 4. The NMC response to those questions is enclosed.

Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.

Michael D. Wadley Site Vice President, Prairie lsland Nuclear Generating Plant Nuclear Management Company, LLC 1717 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1121

Document Control Desk Page 2 Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC

ENCLOSURE I Nuclear Management Company, LLC (NMC) Response to Request for Additional Information on Prairie Island Unit 2 Steam Generator Inservice Inspection Results NRC Question 1: Please confirm that no crack-like indications were detected except in the hot-leg tubesheet region. If crack-like indications were detected at other locations, please discuss each indication. Please discuss whether the data quality was consistent with past inspections.

NMC Response: All of the crack like indications detected and reported were completely contained within the tubesheet region, except row 20 column 30 in Steam Generator 22 which extended above the top of the tubesheet 0.08". The single axial indication (SAI) was 0.09" in length, had a +Point@Voltage of 0.1 Volts and was initiated on the outside diameter of the tube within the sludge pile region of the steam generator. The indication length was approximately one quarter the site in-situ proof testing screening limit of 0.38" for free span axial indications and well below the minimum Voltage threshold for leak testing of 1.89 Volts for axial outside diameter stress corrosion cracking (ODSCC) in the sludge pile region.

Data quality is monitored in the current inspection per industry and site guidance. There is no attempt made at this time to directly compare data quality from inspection to inspection.

NRC Question 2: Please discuss the nature of the indications in the tubes located in row 8, column 59 and row 9, column 59 at the fourth hot-leg tube support in steam generator 21.

NMC Response: Both tubes (row 8 column 59 and row 9 column 59) were plugged during the November 1998 inspection, along with more severe indications in row 8 column 58 and row 9 column 58, to bound a loose part which exhibited wear and could not be retrieved. Row 8 column 58 and row 9 column 58 were successfully in-situ pressure tested at approximately 5300 psi with zero leakage. During the subsequent outage in May 2000 the loose part was retrieved, row 8 column 59 and row 9 column 59 were unplugged, inspected, their wear indications sized at 35 and 32 percent through wall respectively and returned to service. The indications are monitored each outage and show no change beyond normal eddy current repeatability (see table below).

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11/98 5/00 2/02 9/03 5/05 11/06 8-58 SVI 35% 33% 29% 27% 34%

9-58 SVI 32% 34% 32% 34% 34%

NRC Question 3: Please discuss the results of your foreign object search and retrieval. If any loose parts were left in the steam generator, please discuss the following regarding those loose parts:

Indications of tube damage associated with the loose parts.

The source or nature of the loose parts if known.

How tube integrity would be maintained with loose parts left in service.

NMC Response: All locations having an eddy current indication of a Potential Loose Part (PLP) were visually observed from the secondary side (see tables below). These PLP indications were left in the steam generator. There was no tube damage associated with any loose parts on the suspect tubes or any of the adjacent tubes. This was confirmed by Motorized Rotating Pancake Coil (MRPCB) examination of tubes around the part. The source of the parts is operating environment (scale and sludge rocks), parts possibly carried forward from the secondary system (wire and bar), and parts from initial installation (carpenter nail). Since there was no damage to any of the PLP tubes or adjacent tubes to the parts and the PLP's are lodged, tube integrity is not an issue.

List of Potential Loose Parts - 21 Steam Generator Page 2 of 4

List of Potential Loose Parts - 22 Steam Generator First Status Leg Row Column Location Wear Wear seen by Removed Visual Identification Rate Eddy I Lodged Current Cold 27 24 TSC + 0.14 to 0.32" None 2006 Lodged Sludge Rock and small wire NRC Question 4: Two tubes were plugged for failing the profilometry criteria. Please discuss this criteria (e.g., is it associated with re-rolling of the tubes, has the condition in these tubes changed with time (if so, why).

NMC Response: The profilometry criteria is a site imposed installation process control used immediately after the tube has been hydraulically expanded and re-rolled and is only employed during the installation of new re-rolls.

NRC Question 5: Please discuss whether any degradation was observed during your inspection of the secondary side internals in steam generator 22.

NMC Response: No degradation of secondary components was seen during 22 Steam Generator internals inspection.

NRC Question 6: Please discuss whether the growth rates for cold-leg thinning and wear at support structures was consistent with past outages.

NMC Response: From the Unit 2 Steam Generator Condition Monitoring Operational Assessment (CMOA) the 95thPercentile Degradation Growth for cold leg thinning and anti-vibration bar (AVB) wear has remained the same for the last 3 outages (4.8%

through-wall per effective full-power year (TWIEFPY) - cold leg thinning and 2.7%

TWJEFPY - AVB wear).

NRC Question 7: Please discuss whether the number or severity of the dents at the uppermost tube support plates (attributed to the thermal stress relief of the U-bend region) increased.

NMC Response: The data sorts used to develop the requested information only included dents in rows 1 through 5 (attributed to the thermal stress relief of the U-bend region) that are located within '/I1 of the centerline of the 5'" 6thor 7thtube support plate (the uppermost tube support plates).

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In steam generator 21 there were no dents reported at the 7thtube support plate, one dent reported at the 6thtube support plate and one dent reported at the 5thtube support plate in 2006. During the May 2005 inspection there were no dents reported at the 5th, 6thor 7thtube support plates. The two dents reported in 2006 were 2.09 volts and 2.1 1 volts which are just above the dent calling criteria of 2.00 volts.

In steam generator 22 there were seven dents reported at the 7thtube support plate, one dent reported at the 6thtube support plate and no dents reported at the 5thtube support plate in 2006. During the May 2005 inspection there were seven dents reported at the 7thtube support plate, one dent reported at the 6thtube support plate and no dents reported at the 5thtube support plate. One dent reported in 2006 at 3.23 volts was not reported in 2005 and one dent reported in 2005 was reported as INR (indication not reportable) in 2006. The average voltage of the seven repeatable dents was 3.14 volts in 2005 and 3.12 volts in 2006.

Based on results provided above and the knowledge that dents near the calling criteria tend to come and go from outage to outage due to the inherent variability of the bobbin coil inspection method to measure tubing inside diameter variations (complicated by tight radius u-bend geometries) there appears to be no change in the number or severity of dents.

NRC Question 8: Please confirm that tube integrity was maintained during the cycle prior to the 2006 outage.

NMC Response: The Unit 2 Steam Generator CMOA concluded: "The observed degradation at the end of cycle (EOC) 23 outage [2R24] was evaluated in a manner consistent with NEI 97-06 and EPRl guidance. Observed degradation did not present serious challenges to the deterministic [safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials] structural margin requirement at the end of the last cycle of operation or challenge required leakage integrity limits under postulated accident conditions."

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