TMI-12-182, Response to Request for Additional Information Regarding 2011 Steam Generator Tube Inspection Report
| ML12355A208 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 12/19/2012 |
| From: | Libra R Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TMI-12-182 | |
| Download: ML12355A208 (10) | |
Text
7 .,
Exelon Generation.
Three Mile Island Unit 1 Route 441 South, P.O. Box 480 Middletown, PA 17057 Telephone 717-948-8000 10 CFR 50.55a TMI-12-182 December 19, 2012 u.s. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289
Subject:
Response to Request for Additional Information Regarding 2011 Steam Generator Tube Inspection Report
References:
- 1) Letter from R. W. Libra (Exelon Generation Company, LLC) to the U.S.
Nuclear Regulatory Commission, TMI-1 Steam Generator Tube Inspection Report" dated May 15, 2012
- 2) Letter from P. Bamford (U.S. Nuclear Regulatory Commission) to M. J. Pacilio (Exelon Generation Company, LLC), "Three Mile Island Nuclear Station, Unit 1 - Request for Additional Information Regarding 2011 Steam Generator Tube Inspection (TAC Nos. ME8735)," dated November 19, 2012 By letter dated May 15, 2012, Exelon Generation Company, LLC (Exelon) submitted information summarizing the results of the 2011 steam generator (SG) tube inspections performed at Three Mile Island, Unit 1 (TMI, Unit 1) during refueling outage T1 R19 (Reference 1). TMI, Unit 1 Technical Specification 6.9.6 requires a report of the inspection results be submitted within 180 days after reactor coolant temperature exceeds 200°F. The results of the examination are documented in Reference 1.
In the Reference 2 letter, the U.S. Nuclear Regulatory Commission requested information.
Attached is our response.
There are no regulatory commitments in this letter.
TMI, Unit 1 2011 Steam Generator Tube Inspections necember 19,2012 page £ If you have any questions concerning this letter, please contact Michael Fitzwater at 717-948-8228.
Respectfully, R. W. Libra Site Vice President, Three Mile Island Unit 1 Exelon Generation Company, LLC
Attachment:
Response to Request for Additional Information Regarding 2011 Steam Generator Tube Inspection Report for Three Mile Island, Unit 1 cc: Regional Administrator, Region I, USNRC D. L. Werkheiser, USNRC Senior Resident Inspector, TMI P. J. Bamford, Project Manager [TMI] USNRC
ATTACHMENT 1 Response to Request for Additional Information - 2011 Steam Generator Tube Inspection Report for Three Mile Island Nuclear Station, Unit 1
Response to Request for Information Attachment 1 2011 Steam Generator Tube Inspections Page 1 of 7 Docket No. 50 - 289 Question 1:
In order for the staff to better understand your inspection results; please provide the following general design information:
- a. Tube manufacturer.
- b. Tube pitch (and whether it is square or triangular).
- c. Please provide a diagram depicting your tube support naming convention and provide a tubesheet map showing the rows and columns of the tubes.
- d. Please discuss if any stress relief was performed on the tubing following the thermal treatment of the Alloy 690 tubing.
Response
- a. The tube manufacturer is Sumitomo.
- b. The tube pitch is triangular, 0.8750 inches center to center.
- c. See Figure 1 for tube support plate locations and Figure 2 for tube sheet layout.
- d. There was no stress relief performed after the th~rmal treatment, which was performed after final annealing, straightening and outside diameter (00) polishing.
Question 2:
Please clarify what is meant by the "best estimate structural limit" and "End of Cycle, High Probability limit." For example, is the best estimate structural limit the value determined assuming mean material properties and mean non-destructive examination uncertainties? What probability and confidence levels are used for the various input parameters used in determining the two structural limits? What were the yield and ultimate strengths used in calculating the structural limits and how were they determined?
Response
The term Best Estimate Structural Limit refers to the actual physical dimensions of tubing degradation such that the limiting structural integrity performance criteria (SIPC) is just met on a best estimate basis. Average tensile properties and nominal degraded tube strength equations are used. No uncertainty terms are included.
The term High Probability Structural Limit refers to the actual physical dimensions of tubing degradation such that the limiting SIPC is met with 0.95 probability at 50% confidence. Uncertainties in tensile properties and the appropriate degraded tube strength equation are considered. Since the SIPC must be met at the "end of cycle" (EOC) with 0.95 probability at 50% confidence, the High Probability Structural Limit is also referred to as the EOC Allowable Limit.
Per NEI-97-06 methodology, actual material properties of the installed tubing were used to determine the structural limits. There are 131 lots of tube material in the steam generators. The average tensile properties are used to calculate the Best Estimate Structural Limit. All other structural limits are calculated using a Monte Carlo procedure to combine uncertainties. For these cases a Gaussian
Response to Request for Information Attachment 1 2011 Steam Generator Tube Inspections Page 2 of 7 Docket No. 50 - 289 (normal) distribution of tensile properties is assumed with the mean value and standard deviation determined from actual measured tubing properties. The statistical analysis results for the material properties were calculated with a 0.95 probability and a 50% confidence level (95/50). Since there are no measured tensile properties at higher temperatures, the room temperature values are extrapolated using the general material behavior as provided in the ASME Code.
The material properties are as follows for the Enhanced Once Through Steam Generator (EOTSG):
TMI EOTSG Material Strengths Oy °u Oy+ou (ksi) (ksi) (ksi) measured minimum 42.8 97.0 140.8 measured maximum 49.2 107.1 154.7 measured average, m 45.64 100.52 146.16 standard deviation, s 1.3319 1.3889 2.4485 95/50 tolerance factor, k 1.645 1.645 1.645 95/50 LTL 43.45 98.24 142.13 95/50 UTL 47.83 102.80 150.18 For the Large Break Loss Of Coolant Accident (LBLOCA) analysis, the materials properties at room temperature were used. The use of room temperature values for the LBLOCA analysis are appropriate since the maximum loads occur as the tubes cool creating the maximum temperature differential between the tubes and steam generator shell. For all other Condition Monitoring and Operational Assessment (CM/OA) analysis the material properties at 650°F were used. The values used are as follows:
Extrapolated Values for TMI EOTSG Material Strengths Nominal ASME Code Values Values Based on Actual TMI Material Properties (1)
Temperature Oy °u Oy+ou Average LTL 95/50 Degrees F (ksi) (ksi) (ksi) Oy+ou Oy+ou (ksi) (ksi) 70 40.0 85.0 125.0 146.2 142.1 650 31.5 80.0 111.5 130.4 126.8 (1) Rounded values Question 3:
For wear at two lands, you indicate that the acceptance criteria is exceeded (limiting load of 3104 pounds), but that is acceptable due to the conservatism in defining the circumferential extent of the wear scar. Please discuss the conservatisms in determining the circumferential extent of the wear scars. In addition, please discuss whether this is consistent with industry guidelines for determining condition monitoring and operational assessments limits. If it isn't consistent, please discuss your plans for submitting a deviation per Nuclear Energy Institute 03-08, "Guideline for the Management of Materials Issues."
Response to Request for Information Attachment 1 2011 Steam Generator Tube Inspections Page 3 of 7 Docket No. 50 - 289
Response
The tube support plate (TSP) wear geometry is bounded by a flaw that is 1.29 inches in axial length with a circumferential extent of 56 degrees or approximately 0.3 inches. The axial length is based on the thickness of the TSP with tolerances (1.18 inches + 0.11 inches) and the circumferential extent is based on the width of a flat spot (0.3 inches) on the tube OD that just intersects the inner diameter (ID) of the tube. This circumferential extent conservatively bounds the extent of a TSP broached hole land.
This approach is conservative in that it assumes a flaw geometry that produces a uniform flat spot that intersects the tube ID. The actual tube support plate land has a flat contact surface with a nominal dimension of 22 degrees which equates to a nominal land width of approximately 0.12 inches. The potential contact area tapers from this land are back to the broached area. The combined width for two land areas is 0.24 inches, therefore the assumed flat spot of 0.3 inches is conservative.
For actual flaws in the steam generators the acceptance criteria will not be exceeded. The defined structural limits establish a baseline for hypothetic allowable tube degradation. However, as required by NEI 97-06, subsequent evaluations must be performed each inspection outage to address actual "as-found" degradation and actual plant specific growth rates. The Condition Monitoring and Operational Assessment (CM/OA) evaluations performed following each inspection are based on the actual flaw geometries and inspection technique uncertainties. The CM/OA evaluations performed following each inspection are consistent with the requirements of NEI 03-08 and 97-06 and therefore, no deviations are required.
Question 4:
In discussing the wear indications detected at the tube support plate elevations, the term "structural length" is used. Please clarify this term. Is it the length of the limiting portion of the flaw from a structural integrity standpoint? If so, was it determined for all flaws or just a subset of the flaws (Le.,
those that were inspected with an array probe)?
Response
The structural length is the burst equivalent length of a flaw taking into account the actual depth profile.
The structural length and structural depth wou'ld be the lengths/depths that have the same burst pressure as the original flaw, but the structural equivalent would be a rectangular shape flaw.
Structural lengths were obtained from line-by-line sizing the array coil data for 60 tube support plate (TSP) wear indications. Most of the indications had tapered wear shape with structural lengths less than 0.5 inches. However, one of the indications profiled in SG "8" had a flat wear shape versus a tapered wear shape. Therefore, to account for the possibility of flat wear scars, the structural length was set to a fixed value of 1.2 inches. This value is slightly longer than the thickness of the TSPs and is, therefore, very conservative.
Question 5:
Regarding the tubes with wear attributed to tube-to-tube contact, please discuss whether these tubes also have indications at the tube support plate elevations and whether there are any unique trends regarding the location of the tube support plate wear indications (e.g., all located at the upper edge of the tube support plate, all extend beyond the tube support plate, all tubes with tube-to-tube wear indications have corresponding tube support plate wear indications at supports 8 and 9, etc.). Please discuss whether the wear at the tube support plates is oriented in a specific direction (e.g., all are pointing to the center of the bundle). Please provide the mid-point of the tube (e.g., 8S + 20 inches).
Response to Request for Information Attachment 1 2011 Steam Generator Tube Inspections Page 4 of 7 Docket No. 50 - 289
Response
There are 8 tubes in SG "A" and 28 tubes in SG "B" that have both tube-to-tube wear (TT wear) and tube-to-tube support plate wear (TSP wear) indications. No unique trends were observed in this subset of tubes relative to the remainder of the tube bundle. The TSP wear indications occur at either the top or bottom edge of the tube support plate. For tubes with both TT wear and TSP wear, the TSP wear indications are as follows:
SG "A" (8 total tubes)
- 1 tube has TSP wear only at the bottom of TSP 03S The bobbin coil probe was used for the detection and sizing the TSP wear. The orientation with respect to the center of the generator cannot be determined using bobbin probe.
The overall length of the tubes is 674.65 inches. The midpoint of the tube is located at TSP 07S +
34.69 inches (5.3 inches below the center of TSP 08S).
Question 6:
Please clarify whether all 'indications of wear attributed to tube-to-tube support plate interaction were contained within the axial elevation of the tube support plate.
Response
The tube-to-tube support plate wear indications are assumed to be located within the TSP during power operation. The differences in the thermal expansion for various parts of the steam generator (tubes, tie-rods shell) results in an offset of the wear indications relative to the tube support plate edge when the generators are in the cold condition. Since the tube support plate tie rods are connected to the lower tube sheet, this effect is larger at the higher tube support plate locations. Therefore, during the eddy current examinations, the indications may appear to be outside of the support plate due to a combination of probe look ahead and the differences in the thermal expansion of the tie rods and the tubes.
Question 7:
Please discuss the current schedule for completing your assessment of the cause of the tube-to-tube wear occurring in your steam generators. Also, discuss your plans to update the NRC staff on the investigation results when they are complete, as well as any planned corrective actions.
Response to Request for Information Attachment 1 2011 Steam Generator Tube Inspections Page 5 of 7 Docket No. 50 - 289
Response
AREVA is currently performing a Root Cause Analysis (RCA) of the tube-to-tube wear in the steam generators. The schedule, based on the currently identified scope of work, is to have the RCA completed and the final documents available on 04/26/13. If information is identified that affects this expected date, Exelon will notify the NRC accordingly.
Once the RCA is complete, Exelon will schedule a meeting with the NRC staff to provide a summary of the results. The summary will include a discussion of any planned corrective actions.
Response to Request for Information Attachment 1 2011 Steam Generator Tube Inspections Page 6 of 7 Docket No. 50 - 289 Figure 1 TMI-1 EOTSG Tube Support Plate Layout Landmark - Elevation
. _ _ - - - UTE - 674.65"
~--- UTS - 650.07"
- - - 015 - 603.64"
...-- - 014-568.64
- - - 013 - 532.64"
- - 012-495.64"
" - - - - 011-457.64"
, - - - - 010-421.64"
....-- - 009 - 381 .64"
- - - 008 - 342.64"
- - - 007 - 302.64"
- - - 006 - 263.64"
- - - 005 - 226.64"
- - - 004-187.64"
- - - 003-147.64"
- - - 002 - 108.64"
- - - 001-70.64"
- - - LTS - 24.64"
..._- - - LTE - 00.00"
Response to Request for Information Attachment 1 2011 Steam Generator Tube Inspections Page 7 of 7 Docket No. 50 - 289 Figure 2 TMI -1 EOTSG Upper Tube Sheet Layout w
,Z x y
- Tubes are numbered left to right. Each row starts with tube 1. Tie rod locations are not counted.
- Rows are numbered 1 -151 with row 1 at W Axis.
- The missing tubes depict tie rod locations (52 locations).