ML042040158
| ML042040158 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 07/19/2004 |
| From: | Pace P Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MC1048 | |
| Download: ML042040158 (10) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000
'JUL 1 9 200i 10 CFR 50.4 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:
In the Matter of ) Docket No.50-390 Tennessee Valley Authority )
WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - STEAM GENERATOR TUBES INSERVICE INSPECTION - RESPONSE TO NRC'S REQUEST FOR ADDITIONAL INFORMATION FOR THE CYCLE 5 REFUELING OUTAGE (TAC NO. MC 1048)
The purpose of this letter is to provide the response to NRC's request for additional information concerning the subject inspections performed during the Unit 1 Cycle 5 refueling outage. The enclosed questions were received from NRC by electronic mail (e-mail) on May 6, 2004 and June 23, 2004.
There are no regulatory commitments identified in this letter.
If you have any questions about the response to the NRC's questions, please contact me at (423) 365-1824.
Sincerely, P . Pace Manager, Site Licensing and Industry Affairs Enclosure cc: See page 2 ff f
U.S. Nuclear Regulatory Commission Page 2 JUL 19 2004 cc (Enclosure):
NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Ms. Margaret H. Chernoff, Project Manager U.S. Nuclear Regulatory Commission MS 08G9 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 MR. M. M. Comar, Project Manager U.S. Nuclear Regulatory Commission MS 08G9 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303
ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 STEAM GENERATOR TUBE INSERVICE INSPECTION REQUEST FOR ADDITIONAL INFORMATION RESPONSE The following provides TVA's response to NRC's request for additional information received by electronic mail on May 6, 2004 and June 23, 2004 concerning the WBN Unit 1 Cycle 5 Inservice Inspection of the steam generator tubes.
QUESTION 1 At Diablo Canyon Unit 2, several large bobbin voltage indications were detected during its steam generator tube inspections during its 2003 refueling outage. As a result of these findings, the licensee performed more extensive rotating probe inspections than previously performed to help determine whether certain axial outside diameter stress corrosion cracking (ODSCC) indications are prone to significant bobbin voltage growth during the course of the next cycle (refer to Information Notice 2003-13).
Question 1A Given the findings at Diablo Canyon Unit 2, discuss the +Point TM results for the larger voltage indications left in service during the Cycle 5 outage.
RESPONSE
Details of the bobbin indications that were confirmed by the rotating pancake coil (RPC) and left in service are as follows:
BOBBIN RPC BOBBIN RPC SG ROW COL VOLTS VOLTS INDICATION INDICATION LOCATION 1 8 46 0.95 0.33 DSI SAI H03 1 8 52 0.63 0.36 DSI SAI H03
= 0.3 _ SAI H03 2 4 20 0.58 0.16 DSI SAI H03 2 5 105 0.99 0.75 DSI SAI H03 2 8 42 0.38 0.39 DSI SAI H03 2 9 38 0.6 0.24 DSI SAI H02 0.2 SAI H02 2 9 41 0.59 0.51 DSI SAI H04 2 10 36 0.76 0.25 DSI SAI H03 2 10 43 0.88 0.2 DSI SAI H02 2 12 52 0.97 0.25 DSI SAI H02 Steam Generators 3 and 4 had no bobbin indications that were confirmed by the RPC and left in service.
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ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 STEAM GENERATOR TUBE INSERVICE INSPECTION REQUEST FOR ADDITIONAL INFORMATION RESPONSE Question 1B Also, discuss whether the indications with the largest growth rates during Cycle 5 were inspected with a rotating probe during both the Cycle 5 and the Cycle 4 outages.
RESPONSE
Yes, the indications with the largest growth rates (greater than 1 volt) are tested with RPC each outage due to 1 volt repair limit and requirement to test 1 volt and greater with RPC.
However, there were none that required RPC testing in both Cycle 4 and Cycle 5 outages.
Question 1C Discuss any observations from the rotating probe profiling of these indications, the bobbin and rotating probe voltages of these indications, etc. For example, discuss whether the indications with the largest bobbin voltage growth rates have large rotating probe voltage amplitudes, when compared to the bobbin voltage amplitudes, during the Cycle 4 and Cycle 5 outages.
RESPONSE
Profiling was not performed on the ODSCC indications at support plates. The bobbin versus RPC voltages of the largest growth indications are listed below:
BOBBIN VOLTS GROWTH RPC SG ROW COL CYCLE 5 RATE VOLTS RESOLUTION 1 3 47 2.06 1.43 1.21 PLUG 1 5 67 2.08 1.39 1.07 PLUG 1 8 10 2.44 1.13 1.11 PLUG 0.55 1 8 65 1.42 1.01 0.92 PLUG 1 8 71 1.57 1.06 0.94 PLUG 1 9 42 2.47 1.84 1.42 PLUG 1 9 51 1.75 1.37 1.22 PLUG 1 9 71 1.65 1.19 1.26 PLUG 1 9 74 1.75 1.16 1.05 PLUG 1 10 48 2.44 0.96 1.14 PLUG 1 10 49 1.65 1.22 0.64 PLUG E-2
ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 STEAM GENERATOR TUBE INSERVICE INSPECTION REQUEST FOR ADDITIONAL INFORMATION RESPONSE BOBBIN VOLTS GROWTH RPC SG ROW COL CYCLE 5 RATE VOLTS RESOLUTION 1 11 72 2.14 1.65 1.47 PLUG 2 5 42 2.02 1.23 1.47 PLUG 2 5 45 3.27 2.3 1.5 PLUG 2 7 44 1.8 1.01 1.28 PLUG Did Not 2 9 7 1.13 1.05 Confirm Inservice 2 9 42 1.48 1.16 0.97 PLUG 2 10 39 4.51 4.05 3.82 PULL/PLUG 2 10 50 2.1 1.38 0.26 PLUG 2 11 41 2.45 2.19 1.58 PULL/PLUG 2 11 42 1.51 1.04 0.24 PLUG 0.4 0.85 3 5 51 2.6 1.51 1.76 PLUG 3 5 69 3.53 2.69 1.88 PLUG 3 28 87 1.8 0.94 0.17 PLUG Did Not 3 29 87 1.79 1.42 Confirm Inservice Did Not 4 48 61 1.52 1.24 Confirm Inservice In the work that has been done to support probability of prior cycle detection (POPCD), voltage dependent growth, and guidelines for preventive repair of large +Point indications, it is noted that the occurrence of +Point amplitudes greater than the bobbin voltage is infrequent for ODSCC. These inspection results do not contradict this statement. None of the +Point voltages from the Cycle 5 inspection are greater than the Bobbin voltages. Also, this work indicates that preventive repair is not necessary until
+Point indications are 1.9 volts or greater. None of the indications from the Cycle 5 inspection was left in service with
+Point voltages anywhere near 1.9 volts, as the repair limit is 1.0 Bobbin volts.
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ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 STEAM GENERATOR TUBE INSERVICE INSPECTION REQUEST FOR ADDITIONAL INFORMATION RESPONSE Questions from the Tube Inspection F* Report dated October 13, 2003:
QUESTION 2 For each of the 22 designated F* tubes identified in your report, you referenced the location for each indication of degradation as either from the hot leg top of the tubesheet (HTS) or from the hot tube end (HTE). In Watts Bar Technical Specification 5.6.2.12.j, it is stated that the F* distance (1.4 inches) is measured from the bottom of the steam generator tube roll transition or the top of the tubesheet, whichever is lower in elevation (i.e., further into the tubesheet).
For each F* tube identified during the outage, designate whether the bottom of the roll transition or the HTS is lower in elevation, and provide the location of the flaw relative to the bottom of the roll transition or the HTS, whichever is lower in elevation.
RESPONSE
See the following table for Cycle 5 refueling outage report. The table indicates the flaw location with respect to the top of the tube sheet and the location of the bottom of roll transition with respect to the tube sheet.
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ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 STEAM GENERATOR TUBE INSERVICE INSPECTION REQUEST FOR ADDITIONAL INFORMATION RESPONSE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 ALTERNATE REPAIR CRITERIA F-STAR (F*)
CYCLE 5 REFUELING OUTAGE REPORT SO ROW COL TEST LENGTH VOLT IND LOCATION BRT FLAW BELOW CHARACTERIZATION LOCATION BRT OR HTS 1 28 57 ZPSN 0.83 3.04 SAI HTS-5.59 HTS-0.26 5.33 PWSCC TTS AXIAL 1 40 59 ZPSN 0.48 1.7 SAI HTS-3.57 HTS-0.34 3.23 PWSCC TTS AXIAL TOTAL THIS GENERATOR 2 2 3 108 ZPSN 0.29 1.39 SAI HTS-4.01 HTS-0.20 3.81 PWSCC TTS AXIAL TOTAL THIS GENERATOR 1 3 19 23 ZPSN 0.29 1.62 SAI HTS-4.06 HTS-0.24 3.82 PWSCC TTS AXIAL 3 19 75 ZPSN 0.24 1.75 SAI HTS-4.17 HTS-0.00 4.17 PWSCC TTS AXIAL 3 22 67 ZPSN 0.18 0.63 SAI HTs-3.96 HTS-0.20 3.76 PWSCC TTS AXIAL 3 35 58 ZPSN 0.31 0.35 SVI HTs-2.55 HTS-0.16 2.39 VOLUMETRIC 3 35 84 ZPSN 0.18 1.03 SAI HTs-4.37 HTS-0.20 4.17 PWSCC TTS AXIAL 3 38 18 ZPSN 0.27 0.18 SVI HTS-1.89 HTS-0.13 1.76 VOLUMETRIC TOTAL THIS GENERATOR 6 4 16 18 ZPSN 0.2 0.53 SAI HTS-5.97 HTS-0.15 5.82 PWSCC TTS AXIAL 4 17 32 ZPSN 0.2 0.65 SAI HTs-2.16 HTS-0.14 2.02 PWSCC TTS AXIAL 4 18 35 ZPSN 0.42 2.74 SAI HTE+.45 HTS-0.15 >20.0 PWSCC TTS AXIAL 4 19 32 ZPSN 0.43 3.16 SAI HTE+.48 HTS-0.12 >20.0 PWSCC TTS AXIAL 4 22 31 ZPSN 0.46 2.12 SAI HTE+.49 HTS-0.15 >20.0 PWSCC TTS AXIAL 4 22 37 ZPSN 0.26 2.69 SAI HTE+.52 HTS-0.21 >20.0 PWSCC TTS AXIAL 4 27 87 ZPSN 0.27 1.03 SAI HTS-4.55 HTS-0.29 4.26 PWSCC TTS AXIAL 4 35 14 ZPSN 0.13 0.53 SAI HTS-4.86 HTS-0.18 4.68 PWSCC TTS AXIAL 4 35 15 ZPSN 0.23 0.85 SAI HTS-7.08 HTS-0.23 6.85 PWSCC TTS AXIAL 4 36 36 ZPSN 0.35 2.22 SAI HTs-5.87 HTS-0.12 5.75 PWSCC TTS AXIAL 4 38 20 ZPSN 0.41 2.56 SAI HTs-4.81 HTS-0.33 4.48 PWSCC TTS AXIAL 4 39 19 ZPSN 0.27 1.11 SAI HTS-5.00 HTS-0.00 5.00 PWSCC TTS AXIAL 4 42 55 ZPSN 0.29 0.21 SVI HTE+9.50 HTS-0.16 >11.0 VOLUMETRIC TOTAL THIS GENERATOR 13 ACRONYMS:
PWSCC Primary Water Stress Corrosion Cracking ODSCC Outside Diameter Stress Corrosion Cracking BRT Bottom of Roll Transition HTS Hot Leg Top of Tube Sheet HTE Hot Tube End TTS Top of Tube Sheet SAI Single Axial Indication SVI Single Volumetric Indication ZPSN Top of Tube Sheet +Point Probe E-5
ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 STEAM GENERATOR TUBE INSERVICE INSPECTION REQUEST FOR ADDITIONAL INFORMATION RESPONSE QUESTION 3 The three tubes (3-35-58, 3-38-18, and 4-42-55) contained a volumetric indication.
Question 3A Discuss the nature of these volumetric indications. If these indications initiated from the outside diameter of the tube, discuss the driving force for these indications, given that these indications are located below the TTS by at least 1.5 inches.
RESPONSE
Tube 3-35-58: This is a small outer diameter indication in the tubesheet approximately 2 inches below the top. This area was inspected with the rotating +Point probe during Cycle 1, and upon re-review of the Cycle 1 data, the indication is present and has not changed. The fact the indication has not changed in at least four cycles suggests that this is not service induced and most probably is a buff mark.
Tube 3-38-18: This is also a small outer diameter indication in the tubesheet approximately 3 inches below the top. This area was inspected with the rotating +Point probe during Cycle 2.
Upon re-review of the Cycle 2 data, the indication is the same in Cycle 2 as it is currently. This tube was also included on the F* Report Cycle 4 dated March 12, 2002. The indication in Tube 3-38-18 is not considered to be service induced for the same reason as above for tube 3-35-88.
Tube 4-42-55: There is a skip roll of small magnitude in the same location as the indication. Within this skip roll is the outer diameter indication in question. This area had no previous cycle enhanced inspection such as rotating +Point. The indication is most likely a result of the manufacturing process due to the distance from the top of the tubesheet (approximately 11 inches). During the manufacturing process, there are circumstances which can produce outer diameter indications. If foreign material is left in the tubesheet and the tube is expanded, an outer diameter indication may be created. If the tubesheet is gouged during the boring of the hole, an indication may be produced. If the tube metal is smeared during the rolling process, an indication may be produced. Any one and/or combination of these occurrences could produce this kind of indication.
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ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 STEAM GENERATOR TUBE INSERVICE INSPECTION REQUEST FOR ADDITIONAL INFORMATION RESPONSE Question 3B Similarly, address the driving force for the ODSCC axial indications detected in the tubesheet region. Discuss whether these results call into question the bases of the F* criterion.
If secondary-side water is getting into the crevice, discuss the need to assess leakage from the joints under postulated accident conditions.
RESPONSE
Both tubes with "ODSCC" indications in the tubesheet were re-reviewed. These indications are inner diameter in nature and should have been characterized as primary water stress corrosion cracking (PWSCC). These errors were apparent upon re-review of the indications by TVA's Eddy Current Level III. The phase of the flaw indicates an inner diameter flaw. There is no way to determine if the analyst that characterized these indications thought the indications were outside diameter or if the analyst made a typographical error to indicate the indication was outside diameter. Although outside diameter degradation is not expected and questionable inside the tubesheet, TVA engineering did not question the outside diameter flaws because the F* report does not limit the degradation that can be left in service to PWSCC.
Question 3C Discuss whether any tubes have been pulled to characterize this mechanism (flaw initiation on the outside diameter of the tube in a "sealed crevice").
RESPONSE
No tubes were pulled to characterize these indications.
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ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 STEAM GENERATOR TUBE INSERVICE INSPECTION REQUEST FOR ADDITIONAL INFORMATION RESPONSE Question from the Alternate Repair Criteria 90 Day Report dated January 15, 2004:
QUESTION 4 Referring to several figures (e.g., Figures 3.1a, 3.2a, etc.)
that show both the measured and predicted number of indications as a function of the EOC-5 voltage distribution, you show that the measured number of indications was consistently larger than the predicted number of indications. The staff recognizes that the probability of tube burst and the leakage estimates associated with your inspection findings are well within acceptance limits; however, these results indicate that the probability of detection value, set at 0.6, may not adequately account for missed flaws or newly-initiated flaws. In the future, the flaws left in service may result in a higher probability of tube burst and leakage estimates, and such underpredictions may result in exceeding the acceptance limits.
Given the above, discuss what corrective actions, if any, have been undertaken or plan to be undertaken to address this underprediction.
RESPONSE
The principal contributor to the differences between projection and actual is the Generic Letter 95-05, Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking, requirement to apply a constant probability of detection (POD) of 0.6. This POD is too high for the lower voltage indications and too low for the higher voltage indications. The analyses overestimate the higher voltage population and underestimate the lower population. The appropriate methodology change to address these differences would be to apply a voltage dependent POD. Nuclear Energy Institute (NEI) has submitted voltage dependent POD, Guidelines for Preventive Repair of Large +Point Indications, and voltage dependent growth for NRC approval. If this methodology is approved, it will be used during the Unit 1 Cycle 6 inspection.
The generic letter was followed for predictions for Unit 1 Cycle 6 since there is no other approved methodology.
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