ML073480424
| ML073480424 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar, Sequoyah  |
| Issue date: | 12/14/2007 |
| From: | Wetzel B Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| EA-03-009 | |
| Download: ML073480424 (7) | |
Text
December 14, 2007 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen:
In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 50-390 SEQUOYAH NUCLEAR PLANT (SQN) UNITS 1 AND 2 AND WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) BOILER AND PRESSURE VESSEL CODE, SECTION XI -
REACTOR PRESSURE VESSEL HEAD (RPVH) PENETRATION TUBE REMOTE INNER-DIAMETER TEMPER BEAD (IDTB) REPAIR - GENERIC REQUEST FOR RELIEF G-RR RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
By letter dated December 6, 2006, TVA requested NRCs approval for the use of alternative repair methods for the repair or replacement of RPVH penetration tube welds and J-groove welds in the event that inservice examination results are determined unacceptable. Subsequent to that submittal, TVA received a request for additional information in the form of an e-mail, dated November 6, 2007, which indicated that certain safety factors used in the original TVA evaluation of the proposed repair were different than those previously approved for similar requests submitted by other utilities. In follow-up teleconferences with NRC Staff, TVA agreed to evaluate the use of the alternative repair/replacement methods using the analysis safety factors previously approved in connection with the other utility requests.
Following submittal of the December 6, 2006, request, TVA performed the required ultrasonic and visual examinations during the next SQN Unit 2 - Cycle 14
U.S. Nuclear Regulatory Commission Page 2 December 14, 2007 and Unit 1 - Cycle 15 refueling outages. These examinations were performed in compliance with the First Revised NRC Order EA-03-009. Because no recordable indications were detected during the performance of these examinations, there is no immediate need for review and approval of this request, G-RR-2, for use with the SQN reactor pressure vessel heads. Therefore, TVA is withdrawing the request for consideration of this relief with regard to the use of the alternative repair/replacement methodology for the SQN units.
However, because TVA plans to perform examinations of the WBN Unit 1 RPVH during the Spring 2008 outage, the need for the consideration of G-RR-2 and its potential for use on the WBN Unit 1 RPVH still exists. Therefore, TVA requests that G-RR-2 be considered for the WBN Unit 1 as amended by the enclosed RAI response.
Status of Inservice Inspection (ISI) Programs WBN Unit 1 is in the first period of its second 10-year ISI Program interval. WBN currently uses the 2001 Edition with addenda through the 2003 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code for the Unit 1 ISI Program Code-of-Record.
TVA requests approval of the WBN Unit 1-specific portions of G-RR-2 to support RPVH examinations to be performed in the Cycle 8 outage currently scheduled to begin on February 10, 2008. provides TVAs response to the NRC Staffs RAI on Generic Request for Relief, G-RR-2. Using the safety analysis factors as documented in Enclosure 1, the evaluation demonstrates that a repaired WBN Unit 1 RPVH Control Rod Drive Mechanism (CRDM) and Thermocouple Column nozzle would be acceptable for 3 years of operation. Enclosure 2 provides a new commitment made by response to the RAI.
If you have any questions, please contact Kevin Casey at (423) 751-8523.
Sincerely, Original signed by Fredrick C. Mashburn for Beth A. Wetzel Manager, Corporate Nuclear Licensing and Industry Affairs Enclosures cc: See page 3
U.S. Nuclear Regulatory Commission Page 3 December 14, 2007 cc (Enclosures):
U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Senior Resident Inspector U.S. Nuclear Regulatory Commission Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, TN 37379 NRC Senior Resident Inspector U.S. Nuclear Regulatory Commission Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, TN 37381 Douglas V. Pickett, Senior Project Manager U.S. Nuclear Regulatory Commission Mail Stop 08G9A One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739
ENCLOSURE 1 SEQUOYAH NUCLEAR PLANT (SQN) UNITS 1 AND 2 AND WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME)
BOILER AND PRESSURE VESSEL CODE, SECTION XI - REACTOR PRESSURE VESSEL HEAD (RPVH) PENETRATION TUBE REMOTE INNER-DIAMETER TEMPER BEAD (IDTB)
REPAIR - GENERIC REQUEST FOR RELIEF G-RR RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
The following questions are as described in the e-mail sent to TVAs Kevin E. Casey from NRCs Brendan Moroney, dated November 6, 2007. TVAs responses follow the specific questions. Note that the response for question number 2 only addresses reanalysis of the proposed alternative repair/replacement methodology for WBN Unit 1 since consideration for the use of G-RR-2 is being withdrawn for the SQN Units at this time.
Question:
- 1. TVA used the following 4 sets of safety factors for the 4 flaw evaluation cases analyzed in the flaw evaluation of the reactor vessel upper head as shown in Generic Relief Request G-RR-2 for Sequoyah and Watts Bar:
Case 1: A flaw stability evaluation based on the J-integral/tearing modulus (J-T) diagram for operating and low temperature conditions using safety factors of 3.0 on primary stresses and 1.0 on secondary stresses.
Case 2: A flaw stability evaluation based on the J-T diagram for faulted conditions using safety factors of 1.5 on primary stresses and 1.0 on secondary stresses.
Case 3: A flaw stability/crack driving force evaluation based on the J-integral/resistance (J-R) curve and a crack extension of 0.1 inch [for normal conditions]
with safety factors of 1.5 on primary stresses and 1.0 on secondary stresses.
Case 4: A flaw stability/crack driving force evaluation based on the J-R curve and a crack extension of 0.1 inch for faulted conditions with safety factors of 1.25 on primary stresses and 1.0 on secondary stresses.
Confirm the above 4 analysis cases and associated safety factors are correctly characterized.
Response
Yes, the 4 analysis cases and associated safety factors were used in Generic Relief Request G-RR-2 for Sequoyah and Watts Bar.
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Question:
- 2. TVA stated that it used draft Code Case N-749 as a technical basis to support the above safety factors. However, neither the NRC nor the ASME Code Committees have approved draft Code Case N-749. Therefore, please provide the following information:
(a) The NRC staff has approved the use of a safety factor of 3.0 for primary stresses and 1.5 for secondary (residual plus thermal) stresses for operating conditions in CRDM repair relief requests at other nuclear plants. If a safety factor of 3.0 is applied to primary stresses and 1.5 is applied to secondary stresses for the normal conditions in the flaw stability analysis for Sequoyah and Watts Bar, discuss whether the postulated flaw in the reactor vessel upper head would be within the allowable.
(b) For a crack extension of 0.1 inch, the NRC staff has approved the use of a safety factor of 1.5 on primary stresses and 1.0 on secondary stresses at other nuclear plants. If a safety factor of 1.5 is applied to primary stresses and 1.0 is applied to secondary stresses for the faulted condition for the crack extension of 0.1 inch using the J-R curve (i.e., Case 4 above) in the flaw stability/crack driving force analysis for Sequoyah and Watts Bar, discuss whether the postulated flaw in the reactor vessel upper head would be within the allowable.
Response
(a) Using the original analytical parameters (loading conditions, material properties, and service life), TVA is unable to demonstrate that the postulated flaw would be acceptable using the NRC-approved safety factors for SQN and WBN.
A supplemental elastic-plastic fracture mechanics evaluation has been performed using the Watts Bar Unit 1-specific upper shelf Charpy V-notch impact energy of 79.1 ft-lb and deleting the Turbine Roll test and 2458 psig Hydrostatic Test transients from analytical consideration. ASME Code Case N-749 was used as the basis for the elastic-plastic fracture mechanics analysis. The size of the postulated flaw that propagated in the reactor vessel upper head was 1.3 inches for the CRDM nozzle and 1.3 inches for the thermocouple column nozzle. The thickness of the WBN Unit-1 reactor pressure vessel head is 6.5 inches.
- This supplemental evaluation demonstrates that if a safety factor of 3.0 is applied to primary stresses and 1.5 is applied to secondary stresses in a flaw stability analysis for normal conditions, a repaired reactor vessel upper head CRDM nozzle would be acceptable for 3 years of operation.
- This supplemental evaluation demonstrates that safety factors of 3.0 on primary stresses and 1.33 on secondary stresses are needed to show that a repaired reactor vessel upper head Thermocouple Column nozzle would be acceptable for 3 years of operation. As discussed in the follow-up teleconference with the NRC, the safety factor of 1.33 rather than 1.5 on secondary stresses is needed to demonstrate an acceptable repaired reactor vessel head Thermocouple Column for 3 years of operation.
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(b) Using the original analytical parameters (loading conditions, material properties, and service life), TVA is unable to demonstrate that the postulated flaw would be acceptable using the NRC-approved safety factors for SQN and WBN.
A supplemental elastic-plastic fracture mechanics evaluation has been performed using the Watts Bar Unit 1-specific upper shelf Charpy V-notch impact energy of 79.1 ft-lb and deleting the Turbine Roll test and 2458 psig Hydrostatic Test transients from analytical consideration.
- This supplemental evaluation demonstrates that if a safety factor of 1.5 is applied to primary stresses and 1.0 is applied to secondary stresses in a flaw stability/crack driving force analysis for the faulted condition, a repaired reactor vessel upper head CRDM nozzle would be acceptable for 3 years of operation.
- This supplemental evaluation demonstrates that if a safety factor of 1.5 is applied to primary stresses and 1.0 is applied to secondary stresses in a flaw stability/crack driving forces analysis for the faulted condition, a repaired reactor vessel upper head Thermocouple Column nozzle would be acceptable for 3 years of operation.
For both response (a) and (b), administrative controls will prevent the future use of the Turbine Roll Test and Hydrostatic Test transients following an IDTB repair to the CRDM and/or Thermocouple Column nozzles. To accomplish this, TVA will revise WBN Unit 1 UFSAR Section 5.7.2.9 and site procedure 1-SI-0-8 to prevent the use of the Turbine Roll Test and Hydrostatic Test following an IDTB repair to the CRDM and/or Thermocouple Column nozzles. See commitment 1 in Enclosure 2.
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ENCLOSURE 2 COMMITMENT
- 1. TVA will revise WBN Unit 1 UFSAR Section 5.7.2.9 and site procedure 1-SI-0-8 to prevent the use of the Turbine Roll Test and Hydrostatic Test following an IDTB repair to the CRDM and/or Thermocouple Column nozzles.
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