ML043090553
| ML043090553 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 10/29/2004 |
| From: | Solymossy J Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GL-04-001, L-PI-04-109 | |
| Download: ML043090553 (6) | |
Text
Niii NMCed Commifted to Nuclear~xelc Praire Island Nuclear Generating Plant Operated by Nuclear Management Company, LLC October 29, 2004 L-PI-04-109 10 CFR 50.54(f)
U. S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, Maryland 20852 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 60-Day Response to Generic Letter 2004-01, "Requirements for Steam Generator Tube Inspections" On August 30, 2004, the Nuclear Regulatory Commission (NRC) transmitted Generic Letter (GL) 2004-01. Enclosure 1 contains the Nuclear Management Company, LLC (NMC) 60-day response to GL 2004-01 for the Prairie Island Nuclear Generating Plant.
Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.
I declare under penalty of perjury that the foregoing is true and correct. Executed on October 29, 2004.
Jo M.
lymo sy Site ice Preside Pr ie Island Nuclear Generating Plant Ngbr Manage nt Company, LLC Enclosure (1) cc:
Administrator, Region l1l, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC A115 1717 Wakonade Drive East. Welch, Minnesota 55089-9642 Telephone: 651.388.1121
ENCLOSURE I GENERIC LETTER 2004-01 PRAIRIE ISLAND NUCLEAR GENERATING PLANT 60-DAY RESPONSE Nuclear Regulatory Commission (NRC) Requested Information
- 1)
Addressees should provide a description of the SG tube inspections performed at their plant during the last inspection. In addition, if they are not using SG tube inspection methods whose capabilities are consistent with the NRC's position, addressees should provide an assessment of how the tube inspections performed at theirplant meet the inspection requirements of the TS in conjunction with Criteria IX and XI of 10 CFR Part 50, Appendix B, and corrective action taken in accordance with Appendix B, Criterion XVI. This assessment should also address whether the tube inspection practices are capable of detecting flaws of any type that may potentially be present along the length of the tube required to be inspected and that may exceed the applicable tube repair criteria.
Nuclear Management Company, LLC (NMC) Response Prairie Island, Unit I (Pl-1)
The last inspection on PI-1 was conducted during the I R22 refueling outage in November 2002'. The inspection scope is provided in Table 1 below:
TABLE I Number and Extent of Tubes Inspected SCOPE EXTENT PROBETYPE SIG 11 SIG 12 Full Length(l TEC-TEH Bobbin 100%
100%
Rows I and 2 U-Bends 07C-07H MRPC 100%
100%
Hot Leg Tubesheets TEH-TSH+3" MRPC 100%
100%
Sleeves Full Length TEH-STH MRPC N/A 25%
Sleeves Partial Length TSH-STH MRPC N/A 75%0 Hot Leg Roll Plugs UPH-LPH MRPC 25%
25%
Post In Situ Pressure Test Various MRPC 100%
100%
SupplementalOD Various MRPC 100%
100%
' Prairie Island Unit 1 steam generators are being replaced during the Fall 2004 (1 R23) refueling outage.
Page 1 of 5
TABLE I Number and Extent of Tubes Inspected SCOPE EXTENT PROBETYPE S/G 11 S/G 12 Free Span Dents Various MRPC 25%
25%
Plug Visual N/A N/A 100%
100%
Sleeve Visual N/A N/A 100%
100%
(x Except the bend portion of rows I and 2 u-bends and the sleeved portion of sleeved tubes.
0 One sleeved tube (R4C76) was not inspectable due to an obstruction and was plugged.
X ADR, CUD, DEP, DNI, DNT 2 5.0 Volts at a tube support plate, DNT 2 5.0 Volts at the top of the tube sheet, DRI, DSI, DTI, INR 2 1.5 Volts at a tube support plate, MBM, MRI, NQI, PLP (Bound MRPC PLP's), PSI, Cold Leg Thinning 2 40% or < 40% and 2 1.5 Volts.
o See Attachment I for landmark locations used for examination extents.
Acronyms/Definitions:
MRPC = Motorized Rotating Pancake Coil ADR = Absolute Drift CUD = Copper Deposit DEP = Deposit DNI = Dent with Indication DNT = Dent DRI = Distorted Roll Transition with Indication DSI = Distorted Support Signal with Indication DTI = Distorted Tube Sheet Signal with Indication INR = Indication Not Reportable MBM = Manufacturing Burnish Mark MRI = Mix Residual Indication NQI = Non-Quantifiable Indication PLP = Possible Loose Parts PSI = Possible Support Indication S/G = Steam Generator The inspection techniques employed at P1-1 are consistent with the NRC's position in that all known and potential damage mechanisms are sampled between 25 and 100 percent with techniques qualified in accordance with the Electric Power Research Institute (EPRI) Pressurized Water Reactor Steam Generator Examination Guidelines, Appendix H.
The inspection techniques employed at P1-1 are capable of detecting all potential damage mechanisms (volumetric, axial and circumferential) at the time of the inspection in regions of the tubes where they are known or postulated to occur as specified in the Degradation Assessment that may exceed the applicable repair criteria.
The detection capability of each technique employed for P1-1 is defined in the applicable EPRI or Vendor Examination Technique Specification Sheet and is accounted for in both the Condition Monitoring and Operational Assessment.
Prairie Island, Unit 2 (P1-2)
The last inspection on P1-2 was conducted during the 2R22 refueling outage in September 2003. The inspection scope is provided in Table 2 below:
Page 2 of 5
TABLE 2 Number and Extent of Tubes Inspected SCOPE EXTENTO PROBETYPE SIG 21 SIG 22 Full LengthO TEC-TEH Bobbin 100%
100%
Rows 1 through 11 U-Bends 07C-07H MRPC 100%
100%
Hot Leg Tubesheets TEH-TSH+3-6"3)
MRPC 100%
100%
Cold Leg Tubesheets TEC-TSC+1" MRPC 20%
20%
Hot Leg Roll Plugs UPH-LPH MRPC 25%
25%
Post In Situ Pressure Test Various MRPC NIA 100%
SupplementalO Various MRPC 100%
100%
Free Span Dents 2 5.0 Volts Various MRPC 25%
25%
Plug Visual NIA N/A 100%
100%
Baseline new Re-Rolls TEH-TSH Bobbin/MRPC 100%
100%
CD Except the bend portion of rows I through 4 u-bends.
O ADR, CUD, DEP, DNI, DNT 2 5.0 Volts at a tube support plate, DNT 2 5.0 Volts at the top of the tube sheet, DRI, DSI, DTI, INR 2 1.5 Volts at a tube support plate, MBM, MRI, NQI, PLP (Bound MRPC PLP's), PSI, Cold Leg Thinning 2 40% or < 40% and 2 1.5 Volts.
d See Attachment 1 for landmark locations used for examination extents.
TSH +3" except sludge pile region (inclusive of rows 12 through 22 and column 28 through 50) which will be examined TSH +6".
The inspection techniques employed at P1-2 are consistent with the NRC's position in that all known and potential damage mechanisms are sampled between 20 and 100 percent with techniques qualified in accordance with EPRI Pressurized Water Reactor Steam Generator Examination Guidelines, Appendix H.
The inspection techniques employed at PI-2 are capable of detecting all potential damage mechanisms (volumetric, axial and circumferential) at the time of the inspection in regions of the tubes where they are known or postulated to occur as specified in the Degradation Assessment that may exceed the applicable repair criteria.
The detection capability of each technique employed for P1-2 is defined in the applicable EPRI or Vendor Examination Technique Specification Sheet and is accounted for in both the Condition Monitoring and Operational Assessment.
NRC Requested Information
- 2)
If addressees conclude that full compliance with the TS in conjunction with Criteria IX, Xl and XVI of 10 CFR Part 50, Appendix B, requires corrective actions, they should discuss their proposed corrective actions (e.g.,
changing inspection practices consistent with the NRC's position or submitting a TS amendment request with the associated safety basis for Page 3 of 5
limiting the inspections) to achieve full compliance. If addressees choose to change their TS, the staff has included in the attachment suggested changes to the TS definitions for a tube inspection and for plugging limits to show what may be acceptable to the staff in cases where the tubes are expanded for the full depth of the tubesheet and where the extent of the inspection in the tubesheet region is limited.
NMC Response The SIG tube inspections at Prairie Island are in full compliance with the NRC's position in regard to Technical Specifications (TS) in conjunction with Criteria IX, XI and XVI of 10 CFR Part 50, Appendix B. Therefore, this item is not applicable and no corrective actions are required.
NRC Requested Information
- 3)
For plants where SG tube inspections have not been or are not being performed consistent with the NRC's position on the requirements in the TS in conjunction with Criteria IX, Xl, and XVI of 10 CFR Part 50, Appendix B, the licensee should submit a safety assessment (i.e., a justification for continued operation based on maintaining tube structural and leakage integrity) that addresses any differences between the licensee's inspection practices and those called for by the NRC's position. Safety assessments should be submitted for all areas of the tube required to be inspected by the TS where flaws have the potential to exist and inspection techniques capable of detecting these flaws are not being used, and should include the basis for not employing such inspection techniques. The assessment should include an evaluation of (1) whether the inspection practices rely on an acceptance standard (e.g., cracks located at least a minimum distance of x below the top of the tube sheet, even if these cracks cause complete severance of the tube) which is different from the TS acceptance standards (i.e., the tube plugging limits or repair criteria), and (2) whether the safety assessment constitutes a change to the "method of evaluation" (as defined in 10 CFR 50.59) for establishing the structural and leakage integrity of the joint. If the safety assessment constitutes a change to the method of evaluation under 10 CFR 50.59, the licensee should determine whether a license amendment is necessary pursuant to that regulation.
NMC Response The S/G tube inspections at Prairie Island are consistent with the NRC's position on the requirements in the TS in conjunction with Criteria IX, Xl, and XVI of 10 CFR Part 50, Appendix B. Therefore, this item is not applicable and no safety assessment is required.
Page 4 of 5 LANDMARK LOCATION DESCRIPTION TEH 0.01 Tube End Hot leg LPH 0.75 Lower plug roll transition UPH 2.00 Upper plug roll transition TRH 2.75 original Tube Roll Hot leg RTR 3.00 original Roll Transition Reroll 1BH l
3.75 1s additional reroll Bottom roll transition Hot leg 1TH 5.00 1st additional reroll Top roll transition Hot leg 2BH 5.75 2no additional reroll Bottom roll transition Hot leg 1 HH 6.00 Is' additional reroll top of Hydraulic transition Hot leg 2TH 7.00 2" additional reroll Top roll transition Hot leg 2HH 8.00 2 additional reroll top of Hydraulic transition Hot leg EBH 16.38 Elevated reroll Bottom roll transition Hot leg ETH 18.38 Elevated reroll Top roll transition Hot leg TSH 21.40 Tube Sheet Hot leg BUH 25.00 sleeve Bottom of Upper Hydraulic transition hot leg WCH 25.50 sleeve Weld Centerline Hot leg TUH 26.00 sleeve Top of Upper Hydraulic transition hot leg STH 27.00 Sleeve Top Hot leg 01H 71.53 1S tube support plate Hot leg 02H 122.03 2nz tube support plate Hot leg 03H 172.53 3r0 tube support plate Hot leg 04H 223.03 4tn tube support plate Hot leg 05H 273.53 5tn tube support plate Hot leg 06H 324.03 6 tube support plate Hot leg 07H 374.53 7tn tube support plate Hot leg NV1 VARIES New anti Vibration 15 NV2 VARIES New anti Vibration 2w NV3 VARIES New anti Vibration 3rt NV4 VARIES New anti Vibration 4ti 07C
-374.53 7th tube support plate Cold leg 06C
-324.03 67 tube support plate Cold leg 05C
-273.53 56 tube support plate Cold leg 04C
-223.03 4t' tube support plate Cold leg 03C
-172.53 3r4 tube support plate Cold leg 02C
-122.03 23" tube support plate Cold leg 01C
-71.53 15' tube support plate Cold leg TSC
-21.40 Tube Sheet Cold leg TRC
-2.75 original Tube Roll Cold leg UPC
-2.00 Upper Plug roll expansion Cold leg LPC
-0.75 Lower Plug roll expansion Cold leg TEC
-0.01 Tube End Cold leg Page 5 of 5