NLS2008033, Response to Request for Additional Information for Question I.2 Regarding License Amendment Request to Revise Technical Specifications - Appendix K Measurement Uncertainty Recapture Power Uprate
| ML080780497 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 03/12/2008 |
| From: | Michael Colomb Nebraska Public Power District (NPPD) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NLS2008033, TAC MD7385 | |
| Download: ML080780497 (8) | |
Text
N Nebraska Public Power District Always there when you need us NLS2008033 March 12, 2008 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
Response to Request for Additional Information for Question 1.2 Regarding License Amendment Request to Revise Technical Specifications - Appendix K Measurement Uncertainty Recapture Power Uprate Cooper Nuclear Station,.Docket No. 50-298, DPR-46
References:
- 1. Letter from Carl F. Lyon, U.S. Nuclear Regulatory Commission, to Stewart B. Minahan, Nebraska Public Power District, dated January 23, 2008, "Cooper Nuclear Station - Request for Additional Information RE: Measurement Uncertainty Recapture Power Uprate (TAC No. MD7385)"
- 2. Letter from Stewart B. Minahan, Nebraska Public Power District, to the U.S.
Nuclear Regulatory Commission, dated March 6, 2008, "Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications - Appendix K Measurement Uncertainty Recapture Power Uprate"
- 3. Letter from Stewart B. Minahan, Nebraska Public Power District, to the U.S.
Nuclear Regulatory Commission, dated November 19, 2007, "License Amendment Request to Revise Technical Specifications - Appendix K Measurement Uncertainty Recapture Power Uprate"
Dear Sir or Madam:
The purpose of this letter is for the Nebraska Public Power District (NPPD) to submit a response to the Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI)
Question 1.2 sent on January 23, 2008 (Reference 1). The response to this question was not provided in the NPPD response, dated March 6, 2008 (Reference 2), due to the information not being available at that time. The attached information is in support of NRC review of the license amendment request (LAR) to revise the Cooper Nuclear Station (CNS) Technical Specifications for Measurement Uncertainty Recapture power uprate. This LAR was submitted by NPPD letter dated November 19, 2007 (Reference 3).
COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax:. (402) 825-5211 www.nppd.com
NLS2008033 Page 2 of 2 Attachment I contains a response to NRC RAI Question 1.2 from Reference 1. To support readability, NPPD has restated NRC Question 1.1 along with its response. This information is unchanged from that provided in Reference 2. The response to RAI Question 1.2 represents new information. This attachment does not contain information considered proprietary as defined by 10 CFR 2.390. The information submitted by this letter (including attachment) does not change the conclusion of the No Significant Hazards Consideration evaluation submitted by the Reference 3 letter.
Should you have any questions regarding this submittal, please contact David Van Der Kamp, Licensing Manager, at (402) 825-2904.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on 3- Z-- 08 Sincerely, Michael J. Colomb Acting Vice President - Nuclear and Chief Nuclear Officer
/dm Attachment cc: Regional Administrator w/ attachment USNRC - Region IV Cooper Project Manager w/ attachment USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/ attachment USNRC - CNS Nebraska Health and Human Services w/ attachment Department of Regulation and Licensure NPG Distribution w/o attachment CNS Records w/ attachment
NLS2008033 Page 1 of 5 Attachment 1 Response to Request for Additional Information for Question 1.2, Dated January 23, 2008, Regarding License Amendment Request to Revise Technical Specifications for Measurement Uncertainty Recapture Power Uprate Cooper Nuclear Station, Docket No. 50-298, DPR-46 The Nuclear Regulatory Commission (NRC) Requests for Additional Information (RAI)
Questions 1.1 and 1.2 are shown in italics. Nebraska Public Power District's (NPPD) response to each question is shown in block font. The response to RAI Question 1.2 represents new information.
NRC Request I. The following questions areprovidedfrom the Steam Generatorand Chemical EngineeringBranch (CSGB):
- 1. The flow acceleratedcorrosion (FAC) monitoringprogram includes the use of a predictive method to calculate the wall thinning of components susceptible to FAC. In order.forthe U.S. Nuclear Regulatory Commission (NRC) staff to evaluate the accuracy of these predictions,the staff requests a sample list of components for which wall thinning is predictedand measured by ultrasonic testing or other methods. Include the initialwall thickness (nominal), current (measured) wall thickness, and a comparison of the measured wall thickness to the thickness predicted by the model.
NPPD Response A sample list of components that were inspected during the most recently completed refueling outage is provided in Table 1. The list includes components from four different systems (Extraction Steam, Condensate, Condensate Drain, and Feedwater). Components are included from the Extraction Steam piping to the third Feedwater (FW) Heater, which is predicted to have the greatest increase in wear as a result of the power uprate. The list includes the nominal, predicted, and actual thickness as well as the difference between the actual and predicted thicknesses (all dimensions are in inches). As can be seen in the information provided, the model has yielded results that show the actual measured wall thicknesses were greater than those predicted by the model.
NLS2008033 Page 2 of 5 Table 1 - Comparison of Predicted versus Actual Wall Thickness*
Measured Nominal Predicted Thickness Actual -
Component ID System Size Thickness Thickness (RE23) Predicted BS-E-15-2812-2 Ex. Steam to FWH#2 24 0.500 0.365 0.428 0.063 BS-E-17-2812-2 Ex. Steam to FWH#2 24 0.500 0.438 0.451 0.013 BS-E-19-2812-2 Ex. Steam to FWH#2 24 0.500 0.365 0.456 0.091 BS-E-3-EC93877SP-1A Ex. Steam to FWH#2 24 0.375 0.269 0.367 0.098 BS-E-10-EC93877SP-IA Ex. Steam to FWH#3 20 0.375 0.224 0.312 0.088 BS-E-12-EC93877SP-1B Ex. Steam to FWH#3 20 0.375 0.251 0.287 0.036 BS-E-14-EC93877SP-IA Ex. Steam to FWH#3 20 0.375 0.247 0.289 0.042 BS-E-2-2812-1 Ex. Steam to FWH#3 20 0.375 0.258 0.294 0.036 BS-E-3-2812-2 Ex. Steam to FWH#3 20 0.375 0.238 0.312 0.074 BS-E-4-2812-1 Ex. Steam to FWH#3 20 0.375 0.238 0.281 0.043 BS-E-5-2812-1 Ex. Steam to FWH#3 20 0.375 0.186 0.311 0.125 BS-E-6-2812-2 Ex. Steam to FWH#3 20 0.375 0.288 0.304 0.016 BS-E-7-2812-1 Ex. Steam to FWH#3 20 0.375 0.251 0.252 0.001 BS-N-2-2812-2 Ex. Steam to FWH#3 20 0.375 0.166 0.312 0.146 BS-P-8-EC93877SP-IB Ex. Steam to FWH#3 20 0.375 0.232 0.287 0.055 CH-E-18-2819-3 Cond. FWH#3 to FWH#4 16 0.500 0.236 0.445 0.209 CH-R-5-2819-3 Cond. FWH#3 to FWH#4 18 X 16 0.562 0.403 0.539 0.136 CH-E-10-2819-6 Cond. FWH#4 to FWH#5 18 0.562 0.51 0.601 0.091 CH-E-4-2819-4 Cond. FWH#4 to FWH#5 16 0.500 0.355 0.434 0.079 CH-E-7-2819-4 Cond. FWH#4 to FWH#5 16 0.500 0.295 0.478 0.183 CH-E-7-2819-6 Cond. FWH#4 to FWH#5 16 0.500 0.353 0.413 0.060 CH-E-8-2819-4 Cond. FWH#4 to FWH#5 16 0.500 0.535 0.652 0.117 CH-R-2-2819-6 Cond. FWH#4 to FWH#5 18 X 16 0.562 0.425 0.484 0.059 DR-T-5-2827-2 Moisture Separator Drain 12 X 8 0.375 0.367 0.456 0.089 DR-T-5-2827-4 Moisture Separator Drain 12 X 8 0.375 0.301 0.397 0.096 RF-E-12-2849-4 Reactor Feedwater 18 1.375 1.344 1.381 0.037 RF-E-17-2849-4 Reactor Feedwater 18 1.375 1.429 1.432 0.003 RF-E-19-2849-4 Reactor Feedwater 24 1.812 1.773 1.783 0.010 RF-E-21-2849-4 Reactor Feedwater 18 1.375 1.336 1.376 0.040 RF-E-4-2509-2 Reactor Feedwater 12 1.125 0.992 1.095 0.103 RF-E-5-2509-2 Reactor Feedwater 12 1.125 0.867 0.891 0.024 RF-E-6-2849-4 Reactor Feedwater 18 1.375 1.137 1.361 0.224 RF-E-7-2849-4 Reactor Feedwater 18 1.375 1.348 1.394 0.046 RF-E-8-2509-1 Reactor Feedwater 12 1.125 0.867 1.109 0.242 RF-E-8-2509-2 Reactor Feedwater 12 1.125 0.996 1.047 0.051 RF-N-1-2849-4 Reactor Feedwater 18 X 20 1.330 1.115 1.167 0.052 RF-N-2-2849-4 Reactor Feedwater 18 X 20 1.330 1.117 1.134 0.017 RF-O-1-2849-4 Reactor Feedwater 18 1.375 1.128 1.239 0.111
NLS2008033 Page 3 of 5 Measured Nominal Predicted Thickness Actual -
Component ID System Size Thickness Thickness (RE23) Predicted RF-P-14-2849-4 Reactor Feedwater 18 1.375 1.263 1.295 0.032 RF-R-1-2509-2 Reactor Feedwater 18 X 12 1.562 1.59 1.651 0.061
- All values taken from CHECWORKS SFA predictive model for Cooper Nuclear Station (CNS).
NRC Request
- 2. The power uprate will affect several process variables that influence FAC.
Identify the systems that are expected to experience the greatest increase in wear as a result of the power uprate and discuss the effect of individualprocess variables (i.e., moisture content, temperature,oxygen, andflow velocity) on each system identified. For the most susceptible systems and components, what is the total predicted increase in wear rate due to FA C as a result ofpower uprate conditions?
NPPD Response The Erosion/Corrosion Program at CNS uses the CHECWORKS SFA predictive code and actual wall thickness measurements to model wear rate caused by Flow Accelerated Corrosion (FAC). Using this program, the uprated conditions were input and compared with the current model. The uprated conditions in the model also include several plant modifications not related to the Measurement Uncertainty Recapture (MUR) (e.g., replacement of the low pressure turbines and the fourth and fifth stage FW heaters). It is impractical to separate the effect of the plant modifications from those of the MUR.
The wear rates calculated for this response were produced using a Pass 1 analysis with the CHECWORKS SFA predictive code. Pass 1 analysis does not include component inspection data like the Pass 2 analysis used to produce the response to Question 1.1 previously supplied. The Pass 1 analysis is used to indicate the relative susceptibility of components within the piping to wear due to FAC, and the Pass 2 analysis is the method by which the Pass 1 results are calibrated so that the CHECWORKS predictions match observed conditions as closely as possible.
For determination of the effect on FAC due to a change in power, Pass 1 analyses were performed to compare the base case results to the uprated case. By performing the analyses using Pass 1, the models were not constrained by the line correction factor derived from previous inspection data. This is important since the past inspection results may not be indicative of future inspection results due to the change in power.
NLS2008033 Page 4 of 5 The average wear rate difference was calculated by determining the percent change in wear between the uprated Pass 1 and the base Pass 1 results. This percentage was multiplied with the base Pass 2 predicted wear results to determine the average wear rate difference. The uprate is predicted to result in an average wear rate increase of greater than two mils per year for:
- Extraction Steam piping to the third stage FW Heater.
- The drain piping from the fourth stage FW Heater.
- The drain piping from the fifth stage FW Heater.
None of the other piping systems are expected to have increased wear rates of greater than one mil per year (see Table 2).
The predicted increase in average wear in the Extraction Steam piping is due to the increase in wall velocity and the reduction of steam quality. The predicted increase in average wear for the drain piping is due to increased velocity for the fourth stage drains and approaching a more critical temperature for wear on the fifth stage drains. The MUR has no impact on oxygen concentration for the evaluated lines.
The estimated remaining service life for the three lines with average wear rates greater than two mils were reviewed to ensure that the expected increase in wear was acceptable for continued operation during the next fuel cycle (after the implementation of the power uprate).
NLS2008033 Page 5 of 5 Table 2: Appendix K Comparison*
102% vs 100%)
Average Wear Average Average Wall Average Average Temp Steam Rate Wear Rate Velocity Wall Temp Change Average Quality Difference Change (%) Difference (ft/s) Velocity Difference (%) Quality Change System (mils/yr) Change_(%) (F) Difference (%)
1 BS: #2 Extraction -0.628 -3.50% 7.510 71.06% -1.40 -0.67% -0.036 -5.66%
BS: #3 Extraction 2.833 22.31% 3.066 13.36% 0.60 0.23% -0.027 -3.27%
BS: #4 Extraction Excluded - FAC resistant material BS: #5 Extraction Excluded - FAC resistant material CH: Pumps to No. 1 Heaters 0.048 1.24% 0.312 2.24% -0.10 -0.10% 0.000 0.00%
CH: #1 Heaters to #2 Heaters -0.177 -4.94% 0.266 1.95% -6.60 -3.91% 0.000 0.00%
CH: #2 Heaters to #3 Heaters -0.091 -3.78% 0.279 1.64% -6.30 -3.10% 0.000 0.00%
CH: #3 Heaters to #4 Heaters -0.058 -2.66% 0.272 1.64% -5.10 -1.97% 0.000 0.00%
CH: #4 Heaters to #5 Heaters 0.241 7.69% 0.308 1.77% -2.10 -0.68% 0.000 0.00%
CH: #5 Heaters to Pumps 0.133 4.27% 0.261 1.68% -3.00 -0.82% 0.000 0.00%
DR: No. 3 Heater Drains -0.747 -13.69% -0.614 -12.67% -2.90 -1.37% -0.001 0.00%
DR: No. 4 Heater Drains 2.091 12.88% 0.683 17.77% -6.00 -2.23% 0.000 0.00%
2 DR: No. 5 Heater Drains 4.663 16.70% -2.025 -10.30% -11.33 -3.54% 0.000 0.00%
DR: MSR Drains 0.019 0.77% 0.103 2.59% 1.00 0.27% 0.000 0.00%
MS: HP Turbine to MSR 0.117 0.98% -1.780 -3.81% 5.50 1.47% -0.003 0.00%
RF: Feedwater 0.239 4.40% 0.353 1.74% -3.00 -0.82% 0.000 0.00%
- All values taken from CHECWORKS SFA predictive model for CNS.
1.
The change in average velocity is not dominate due to the lower temperature of the BS-2 piping (206'F).
2.
The increase in wear is due to approaching a more critical temperature for wear.
ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS@
ATTACHMENT 3 LIST OF REGULATORY COMMITMENTSO Correspondence Number: NLS2008033 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.
COMMITMENT COMMITTED DATE COMMITMENT NUMBER OR OUTAGE None N/A N/A i i I PROCEDURE 0.42 REVISION 22 PAGE 18 OF 25 ]