CP-200901227, Revision to Request for Relief to Extend the Unit 1 and 2 in Service Inspection Interval for the Reactor Vessel Weld Examination and Withdrawal of License Amendment Request 09-004 to Add License Condition for Submittal of ISI Informatio

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Revision to Request for Relief to Extend the Unit 1 and 2 in Service Inspection Interval for the Reactor Vessel Weld Examination and Withdrawal of License Amendment Request 09-004 to Add License Condition for Submittal of ISI Information an
ML092650286
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 09/14/2009
From: Flores R
Luminant Generation Co, Luminant Power
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
CP-200901227, TAC ME0781, TAC ME0782, TXX-09106
Download: ML092650286 (14)
v • d • e


Text

Rafael Flores Luminant Power Senior Vice President P 0 Box 1002

& Chief Nuclear Officer 6322 North FM 56 rafael.flores@Luminant.com Glen Rose, TX 76043 Ll T 254 897 5550 C 817 559 0403 F 254 897 6652 CP-200901227 Ref. # 10 CFR 50.55a(a)(3)(i)

Log # TXX-09106 September 14, 2009 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION DOCKET NOS. 50-445 AND 50-446 REVISION TO REQUEST FOR RELIEF TO EXTEND THE UNIT 1 AND 2 INSERVICE INSPECTION INTERVAL FOR THE REACTOR VESSEL WELD EXAMINATION AND WITHDRAWAL OF LICENSE AMENDMENT REQUEST 09-004 TO ADD LICENSE CONDITION FOR SUBMITTAL OF ISI INFORMATION AND ANALYSES (TAC NOS. ME0781 AND ME0782)

REFERENCE:

1. Ho K. Nieh, NRC, NRR Letter to Gordon Bischoff, WOG regarding Final Safety Evaluation for PWROG Topical Report WCAP-16168-NP, Revision 2, (TAC NO.

MC9768), dated May 8, 2008

2. Letter logged TXX-09004 dated March 4, 2009 from Rafael Flores of Luminant Power to the NRC regarding Request for Relief to Extend the Unit 1 and 2 Inservice Inspection Interval for the Reactor Vessel Weld Examination and License Amendment Request 09-004 to add License Condition for Submittal of ISI Information and Analyses

Dear Sir or Madam:

Luminant Generation Company, LLC (Luminant Power) submitted Relief Request B-9 (Attachment 1) for Comanche Peak Unit 1 and Relief Request No. B-8 (Attachment 2) for Comanche Peak Unit 2 via Reference 2. These relief requests apply to the second 10-year Inservice Inspection (ISI) Intervals for Units 1 and 2, with the reactor vessel weld examinations occurring during the third 10-year ISI intervals for both units.

The NRC approved WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of The Reactor Vessel In-Service Inspection Interval," per Reference 1. This WCAP provides for extension of the inservice inspection interval for certain pressure retaining welds in the reactor vessel from 10 to 20 years.

Luminant Power proposes to implement this extended inservice inspection interval for Comanche Peak Units 1 and 2. The plant specific information identified by the above letter as needed to support this request is provided in Attachments 1 and 2. Luminant Power concluded that the proposed alternative provides an acceptable level of quality and safety. The relief is requested under the provisions of 10 CFR 50.55a(a)(3)(i).

On August 20, 2009, a teleconference was held between NRC Staff and Luminant Power. The NRC informed us that a license condition was no longer needed. Therefore, Luminant Power is withdrawing A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak - Diablo Canyon
  • Palo Verde
  • San Onofre
  • Wolf Creek

U. S. Nuclear Regulatory Commission TXX-09106 Page 2 09/14/2009 License Amendment Request 09-004.

The proposed schedule for the Unit 2 second reactor vessel inservice inspection was discussed. Moving the second inservice inspection to the refueling outage in 2021 would improve the distribution of plants performing reactor vessel inspections. This schedule change is reflected in Attachment 2.

The relief requests in Attachments 1 and 2 are also being revised for the second reactor vessel inservice inspection only. Extension of the third reactor vessel inservice inspections for Units 1 and 2 will require the submittal of new relief requests after extension of the Unit 1 and 2 operating licenses. Additionally, weld numbers in Section 1 of Attachments 1 and 2 have been revised from the previous submittal.

Luminant Power requests approval of the relief requests by December 31, 2009, to support the Comanche Peak Unit 1 Spring 2010 Refueling Outage.

This communication contains no new licensing basis commitments regarding Comanche Peak Units 1 and 2. Should you have any questions, please contact Mr. Jack Hicks at (254)897-6725.

I state under penalty of perjury that the foregoing is true and correct. Executed on the 14th of September, 2009.

Sincerely, Luminant Generation Company LLC Rafael Flores By: K422w, Fred W. Madden Director, Oversight & Regulatory Affairs Attachments - 1. Relief Request B-9 for Unit 1 Request For Relief to Extend the 10-year Reactor Vessel Inservice Inspection Interval

2. Relief Request B-8 for Unit 2 Request For Relief to Extend the 10-year Reactor Vessel Inservice Inspection Interval c- E. E. Collins, Region IV B. K. Singal, NRR Resident Inspectors, Comanche Peak Brian Welch, ANII, Comanche Peak Anthony Jones, TDLR

ATTACHMENT 1 TO TXX-09106 LUMINANT POWER COMANCHE PEAK NUCLEAR POWER PLANT RELIEF REQUEST B-9 FOR UNIT 1 REQUEST FOR RELIEF TO EXTEND THE 10-YEAR REACTOR VESSEL INSERVICE INSPECTION INTERVAL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

ALTERNATIVE PROVIDES ACCEPTABLE LEVEL OF QUALITY AND SAFETY

Attachment I to TXX-09106 Relief Request B-9 for Comanche Peak Unit I Page 2 of 6

1. ASME Code Component(s) Affected The affected component is the Comanche Peak Unit 1 reactor vessel, specifically the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code Section XI (Reference 1) examination categories and item numbers covering examinations of the reactor vessel (RV). These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code Section XI.

Weld Examination Number Category Item No. Description TBX-1-1100-1 B-A B1.30 Flange to Shell Circumferential Weld TBX-1-1100-2 B-A 131.11 Upper/Intermediate Circumferential Weld TBX-1-1100-3 B-A B1.11 Intermediate/Lower Circumferential Weld TBX-1-1100-4 B-A B1.11 Lower Shell to Lower Head Circumferential Weld TBX-l-1100-5 B-A B1.21 Lower Head Circumferential Weld TBX-1-1100-6 B-A B1.12 Upper Shell Longitudinal Weld (420)

TBX-1-1100-7 B-A B1.12 Upper Shell Longitudinal Weld (162°)

TBX-1-1100-8 B-A B1.12 Upper Shell Longitudinal Weld (282°)

TBX-1-1100-9 B-A B1.12 Intermediate Shell Longitudinal Weld (0°)

TBX-1-1100-10 B-A B1.12 Intermediate Shell Longitudinal Weld (120')

TBX-1-1100-11 B-A B1.12 Intermediate Shell Longitudinal Weld (2400)

TBX-1-1100-12 B-A B1.12 Lower Shell Longitudinal Weld (900)

TBX-1-1100-13 B-A B1.12 Lower Shell Longitudinal Weld (210')

TBX-1-1100-14 B-A B1.12 Lower Shell Longitudinal Weld (330')

TBX-1-1100-15 B-A B1.22 Lower Head Meridional Weld (0°)

TBX-1-1100-16 B-A B1.22 Lower Head Meridional Weld (900)

TBX-1-1100-17 B-A B1.22 Lower Head Meridional Weld (1800)

TBX-1-1100-18 B-A B1.22 Lower Head Meridional Weld (270')

TBX-1-110OA-19 B-D B3.90 Outlet Nozzle to Shell Weld (Loop 4) (220)

TBX-1-1100A-19IR B-D B3.100 Outlet Nozzle to Shell Weld IR (Loop 4) (22')

TBX-1-110OA-20 B-D B3.90 Inlet Nozzle to Shell Weld (Loop 4) (670)

TBX-1-1100A-20R B-D B3.100 Inlet Nozzle to Shell Weld (Loop 4) (670)

TBX-1-1100A-21 B-D B3.90 Inlet Nozzle to Shell Weld (Loop 3) (113')

TBX-1-1100A-21IR B-D B3.100 Inlet Nozzle to Shell Weld (Loop 3) (1130)

TBX-1-1100A-22 B-D B3.90 Outlet Nozzle to Shell Weld (Loop 3) (1580)

TBX-1-1100A-22IR B-D B3.100 Outlet Nozzle to Shell Weld IR (Loop 3) (1580)

TBX-1-1100A-23 B-D B3.90 Outlet Nozzle to Shell Weld (Loop 2) (2020)

TBX-1-110OA-23IR B-D B3.100 Outlet Nozzle to Shell Weld (Loop 2) (202')

TBX-1-1100A-24 B-D B3.90 Inlet Nozzle to Shell Weld (Loop 2) (247°)

TBX-1-110OA-24IR B-D B3.100 Inlet Nozzle to Shell Weld IR (Loop 2) (247°)

TBX-1-110OA-25 B-D B3.90 Inlet Nozzle to Shell Weld (Loop 1) (2930)

TBX-1-1100A-25IR B-D B3.100 Inlet Nozzle to Shell Weld IR (Loop 1) (2930)

TBX-1-110OA-26 B-D B3.90 Outlet Nozzle to Shell Weld (Loop 1) (3380)

TBX-1-1100A-26IR B-D B3.100 Outlet Nozzle to Shell Weld IR (Loop 1) (338°)

(Throughout this request the above examination categories are referred to as "the subject examinations" and the ASME BPV Code,Section XI, is referred to as "the Code.")

to TXX-09106 Relief Request B-9 for Comanche Peak Unit I Page 3 of 6

2. Applicable Code Edition and Addenda

ASME Code Section XI, "Rules and Inservice Inspection of Nuclear Power Plant Components, " Code 1998 Edition with the 2000 Addenda.

3. Applicable Code Requirement

IWB-2412, Inspection Program B, requires volumetric examination of essentially 100% of reactor vessel pressure retaining welds identified in Table IWB-2500-1 once each ten year interval. The Comanche Peak Unit 1 second 10-year inservice inspection interval is scheduled to end in August 2010.

4. Reason for Request

An alternative is requested from the requirement of IWA-2412, Inspection Program B, that volumetric examination of essentially 100% of reactor pressure vessel pressure retaining welds, Examination Categories B-A and B-D welds, be performed once each ten-year interval. Extension of the inspection interval for Examination Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in man-rem exposure and examination costs.

5. Proposed Alternative and Basis for Use Luminant Power proposes to defer the ASME Code required volumetric examination of the Comanche Peak Unit 1 reactor vessel full penetration pressure retaining Category B-A and B-D welds for the second inservice inspection from refueling outage 1RF14 in 2010 until refueling outage 1RF20 in 2019. This date is within one refueling cycle relative to the information provided to the Staff in PWR Owners Group letter OG-06-356 (Reference 2). Extension of the third inservice inspection will require the submittal of a new relief request after extension of the Unit 1 operating license.

In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on the basis that the current inspection interval can be extended based on a negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 3).

The methodology used to demonstrate the acceptability of extending the second and third inspection intervals for Category B-A and B-D welds based on a negligible change in risk is contained in WCAP-16168-NP-A, Revision 2 (Reference 4). This methodology was used to develop a pilot plant analysis for Westinghouse, Combustion Engineering, and Babcock and Wilcox reactor vessel designs and is an extension of the work that was performed as part of the NRC PTS Risk Re-Evaluation (Reference 5).

The critical parameters for demonstrating that this pilot plant analysis is applicable on a plant specific basis, as identified in WCAP-16168-NP-A, Revision 2, are identified in Table 1. By demonstrating that each plant specific parameter is bounded by the corresponding pilot plant parameter, the application of the methodology to the Comanche Peak Unit 1 reactor vessel is acceptable as shown in Table 1 below.

to TXX-09106 Relief Request B-9 for Comanche Peak Unit 1 Page 4 of 6 Table 1 Critical Parameters for Application of Bounding Analysis Additional Evaluation Parameter Pilot Plant Basis Plant Specific Basis Required?

Dominant Pressurized Thermal NRC PTS Risk Study PTS Generalization No Shock (PTS) Transients in the NRC (Reference 5) Study (Reference 6)

PTS Risk Study are applicable Through Wall Cracking Frequency 1.76E-08 Events per year 7.51E-15 Events per No (Reference 4) year Frequency and Severity of Design 7 heatup/cooldowns per year Bounded by 7 No Basis Transients (Reference 4) heatup/cooldowns per year Cladding Layers (Single/Multiple) Single Layer (Reference 4) Single Layer No Additional information relative to the Comanche Peak Units 1 reactor vessel inspection is provided in Table 2. This information confirms that satisfactory examinations have been performed on the Comanche Peak Unit 1 reactor vessel.

Table 2 Additional Information Pertaining to Reactor Vessel Inspection Inspection methodology: ASME Section XI and Regulatory Guide 1.150 (Reference 7).

Number of past inspections: 1 inspection has been performed to date on each Category B-A and B-D weld.

Number of indications found: All 3 potential beltline indications in the reactor vessel are acceptable per Section XI IWB-3500. Based on the weld length and volume of plate inspected in the beltline region for Comanche Peak Unit 1, only 2 flaws are in the required ISI volume and they are both acceptable per the proposed PTS Rule in Reference 8.

Proposed inspection schedule for The second inservice inspection is currently scheduled for 2010 (1RF14).

balance of plant life: The second inservice inspection is proposed to be performed in 2019 (1RF20). (This date is within one refueling outage of the dates in PWROG Letter OG-06-356, as discussed in Section 5).

Attachment i to TXX-09106 Relief Request B-9 for Comanche Peak Unit 1 Page 5 of 6 Table 3 provides additional information relative to the calculation of the Through Wall Cracking Frequency (TWCF) for Comanche Peak Unit 1. As noted in this table, the calculation of AT30 is based upon the embrittlement trend curve equations from Revision 2 of Regulatory Guide 1.99 (Reference 9).

The values of chemistry factor (CF) and equations for fluence function (FF) were taken directly from WCAP-16346-NP (Reference 10), which was previously provided to NRC for the Unit 1 heat-up and cool-down limit curves in Reference 11.

Table 3 Details of TWCF Calculation at 60 EFPY Inputs Reactor Coolant System Temperature, TRcs[°F]: N/A Twa,, [inches]: 8.63 Region/Component C6 Ni R.G. CF Un- Fluence [1019 R Degcrion/Conent Material Cu Ni CF Irradiated Neutron/cm 2, Description [wt%/o] [wt%] Pos. [F] RTNDT(u) [°F] E>1 MeV]

I Inter. Shell Plate A 533B 0.060 0.650 1.1 37.0 10 3.70 2 Inter. Shell Plate A 533B 0.070 0.670 1.1 44.0 -10 3.70 3 Inter. Shell Plate A 533B 0.070 0.620 1.1 44.0 10 3.70 4 Lower Shell Plate A 533B 0.080 0.650 1.1 51.0 0 3.70 5 Lower Shell Plate A 533B 0.080 0.650 1.1 51.0 0 3.70 6 Lower Shell Plate A 533B 0.060 0.600 2.1 16.1 20 3.70 7 Int. Shell Axial Weld Linde 0091 0.045 0.200 2.1 11.5 -70 3.17 8 Int. Shell Axial Weld Linde 0091 0.045 0.200 2.1 11.5 -70 2.51 9 Int. Shell Axial Weld Linde 0091 0.045 0.200 2.1 11.5 -70 3.17 10 Low. Shell Axial Weld Linde 0091 0.045 0.200 2.1 11.5 -70 2.51 11 Low. Shell Axial Weld Linde 0091 0.045 0.200 2.1 11.5 -70 3.17 12 Low. Shell Axial Weld Linde 0091 0.045 0.200 2.1 11.5 -70 3.17 13 Int./Low. Circ Weld Linde 124 0.045 0.200 2.1 11.5 -70 3.70 Outputs Methodology Used to Calculate AT30: Regulatory Guide 1.99 Rev. 2 Controlling Material RTMAx Fluence [1019 FF Region # RMxx Neutron/cm 2 (fluence AT30 [°F] TWCF95-xx (From [R] , E>1 MeV] factor)

Above)

Axial Weld - AW 3 527.04 3.17 1.30 57.35 2.47E-18 Circumferential Weld - CW 3 528.61 3.70 1.34 58.92 5.52E-29 Plate - PL 3 528.61 3.70 1.34 58.92 3.OOE-15 TWCF95-TOTAL (OCAwTWCF95-AW + (XPLTWCF95-PL + (xcwTWCF95-cw + aFoTWCF95-FO): 7.51E-15 to TXX-09106 Relief Request B-9 for Comanche Peak Unit I Page 6 of 6

6. Duration of Proposed Alternative This request is applicable to the Comanche Peak Unit I inservice inspection program for the current Unit 1 Operating License.
7. References
1. ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition with the 1989 Addenda up to and including the 2004 Edition with the 2005 Addenda, American Society of Mechanical Engineers, New York.
2. OG-06-356, "Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval." MUHP 5097-99, Task 2059," October 31, 2006.
3. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

November 2002.

4. WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," June 2008.
5. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock," March, 2007.
6. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004.
7. NRC Regulatory Guide 1.150, Revision 1, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations," February 1983.
8. SECY-07-0104, "Proposed Rulemaking - Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock," June 25, 2007 (ADAMS Accession Number ML070570141).
9. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
10. WCAP-16346-NP, Rev. 0, "Comanche Peak Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," November 2004.
11. TXU Power letter TXX-06146 from Mike Blevins to the U. S. Nuclear Regulatory Commission, dated August 31, 2006 (ADAMS Accession No. ML062490287).

ATTACHMENT 2 TO TXX-09106 LUMINANT POWER COMANCHE PEAK NUCLEAR POWER PLANT RELIEF REQUEST B-8 FOR UNIT 2 REQUEST FOR RELIEF TO EXTEND THE 10-YEAR REACTOR VESSEL INSERVICE INSPECTION INTERVAL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

ALTERNATIVE PROVIDES ACCEPTABLE LEVEL OF QUALITY AND SAFETY to TXX-09106 Relief Request B-8 for Comanche Peak Unit 2 Page 2 of 6

1. ASME Code Component(s) Affected The affected component is the Comanche Peak Unit 2 reactor vessel, specifically the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code Section XI (Reference 1) examination categories and item numbers covering examinations of the reactor vessel (RV). These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code Section XI.

Weld Examination 1%inmhpr r~to 1 nr Da*eri umber rate o Ifprn N~n_

Item No Descri nl'i nn tion TCX-1-1100-1 B-A B1.30 Flange to Shell Circumferential Weld TCX-1-1100-2 B-A B1.11 Upper/Intermediate Circumferential Weld TCX-1-1100-3 B-A B1.11 Intermediate/Lower Circumferential Weld TCX-1-1100-4 B-A B1.11 Lower Shell to Lower Head Circumferential Weld TCX-1-1100-5 B-A B1.21 Lower Head Circumferential Weld TCX-1-1100-6 B-A B1.12 Upper Shell Longitudinal Weld (560)

TCX-1-1100-7 B-A B1.12 Upper Shell Longitudinal Weld (1760)

TCX-1-1100-8 B-A B1.12 Upper Shell Longitudinal Weld (2960)

TCX-1-1100-9 B-A B1.12 Intermediate Shell Longitudinal Weld (0°)

TCX-1-1100-10 B-A B1.12 Intermediate Shell Longitudinal Weld (120')

TCX-1-1100-11 B-A B1.12 Intermediate Shell Longitudinal Weld (2400)

TCX-1-1100-12 B-A B1.12 Lower Shell Longitudinal Weld (900)

TCX-1-1100-13 B-A B1.12 Lower Shell Longitudinal Weld (210')

TCX-1-1100-14 B-A B1.12 Lower Shell Longitudinal Weld (330')

TCX-1-1100-15 B-A B1.22 Lower Head Meridional Weld (0')

TCX-1-1100-16 B-A B1.22 Lower Head Meridional Weld (90')

TCX-1-1100-17 B-A B1.22 Lower Head Meridional Weld (180')

TCX-1-1100-18 B-A B1.22 Lower Head Meridional Weld (2700)

TCX-1-1100A-19 B-D B3.90 Outlet Nozzle to Shell Weld (Loop 3) (220)

TCX-1-1100A-19IR B-D B3.100 Outlet Nozzle to Shell Weld IR (Loop 3) (22')

TCX-1-1100A-20 B-D B3.90 Inlet Nozzle to Shell Weld (Loop 3) (670)

TCX-1-1100A-20R B-D B3.100 Inlet Nozzle to Shell Weld (Loop 3) (670)

TCX-1-1100A-21 B-D B3.90 Inlet Nozzle to Shell Weld (Loop 4) (1130)

TCX-1-1100A-21IR B-D B3.100 Inlet Nozzle to Shell Weld (Loop 4) (1130)

TCX-1-1100A-22 B-D B3.90 Outlet Nozzle to Shell Weld (Loop 4) (1580)

TCX-1-1100A-22IR B-D B3.100 Outlet Nozzle to Shell Weld IR (Loop 4) (1580)

TCX-1-1100A-23' B-D B3.90 Outlet Nozzle to Shell Weld (Loop 1) (2020)

TCX-1-1100A-23IR B-D B3.100 Outlet Nozzle to Shell Weld (Loop 1) (2020)

TCX-1-1100A-24 B-D B3.90 Inlet Nozzle to Shell Weld (Loop 1) (2470)

TCX-1-1100A-24IR B-D B3.100 Inlet Nozzle to Shell Weld IR (Loop 1) (2470)

TCX-1-1100A-25 B-D B3.90 Inlet Nozzle to Shell Weld (Loop 2) (2930)

TCX-1-1100A-25IR B-D B3.100 Inlet Nozzle to Shell Weld IR (Loop 2) (293')

TCX-1-1100A-26 B-D B3.90 Outlet Nozzle to Shell Weld (Loop 2) (338')

TCX-1-1100A-26IR B-D B3.100 Outlet Nozzle to Shell Weld IR (Loop 2) (3380)

(Throughout this request the above examination categories are referred to as "the subject examinations" and the ASME BPV Code,Section XI, is referred to as "the Code.")

to TXX-09106 Relief Request B-8 for Comanche Peak Unit 2 Page 3 of 6

2. Applicable Code Edition and Addenda

ASME Code Section XI, "Rules and Inservice Inspection of Nuclear Power Plant Components," Code 1998 Edition to the 2000 Addenda.

3. Applicable Code Requirement

IWB-2412, Inspection Program B, requires volumetric examination of essentially 100% of reactor vessel pressure retaining welds identified in Table IWB-2500-1 once each ten year interval. The Comanche Peak Unit 2 second 10-year inservice inspection interval is scheduled to end in August 2014.

4. Reason for Request

An alternative is requested from the requirement of IWA-2412, Inspection Program B, that volumetric examination of essentially 100% of reactor pressure vessel pressure retaining welds, Examination Categories B-A and B-D welds, be performed once each ten-year interval. Extension of the inspection interval for Examination Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in man-rem exposure and examination costs.

5. Proposed Alternative and Basis for Use Luminant Power proposes to defer the ASME Code required volumetric examination of the Comanche Peak Unit 2 reactor vessel full penetration pressure retaining Category B-A and B-D welds for the second inservice inspection from refueling outage 2RF13 in 2012 until refueling outage 2RF19 in 2021. This date is different than that provided to the Staff in PWR Owners Group letter OG-06-356 (Reference 2). This change is required because the initial date used in the PWROG Schedule is not consistent with the Comanche Peak long range outage plan. This change will also improve the distribution of plants performing reactor vessel inspections per year with the implementation of the 20 year inspection interval and would still meet the intent of the PWROG vessel inspection schedule in OG-06-356. There are currently four other plants scheduled to perform reactor vessel inservice inspections in 2012 and only one plant is scheduled for inspection in 2021. Extension of the third inservice inspection will require the submittal of a new relief request after extension of the Unit 2 operating license.

In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on the basis that the current inspection interval can be extended based on a negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 3).

The methodology used to demonstrate the acceptability of extending the second and third inspection intervals for Category B-A and B-D welds based on a negligible change in risk is contained in WCAP-16168-NP-A, Revision 2 (Reference 4). This methodology was used to develop a pilot plant analysis for Westinghouse, Combustion Engineering, and Babcock and Wilcox reactor vessel designs and is an extension of the work that was performed as part of the NRC PTS Risk Re-Evaluation (Reference 5).

The critical parameters for demonstrating that this pilot plant analysis is applicable on a plant specific basis, as identified in WCAP-16168-NP-A, Revision 2, are identified in Table 1. By demonstrating that each plant specific parameter is bounded by the corresponding pilot plant parameter, the application of the methodology to the Comanche Peak Unit 2 reactor vessel is acceptable as shown in Table 1 below.

to TXX-09106 Relief Request B-8 for Comanche Peak Unit 2 Page 4 of 6 J Table 1 Critical Parameters for Application of Bounding Analysis Additional Evaluation Parameter Pilot Plant Basis Plant Specific Basis Required?

Dominant Pressurized Thermal NRC PTS Risk Study PTS Generalization No Shock (PTS) Transients in the NRC (Reference 5) Study (Reference 6)

PTS Risk Study are applicable _

Through Wall Cracking Frequency 1.76E-08 Events per year 2.73E-16 Events per No (Reference 4) year Frequency and Severity of Design 7 heatup/cool downs per year Bounded by 7 No Basis Transients (Reference 4) heatup/cool downs per year Cladding Layers (Single/Multiple) Single Layer (Reference 4) Single Layer No Additional information relative to the Comanche Peak Unit 2 reactor vessel inspection is provided in Table 2. This information confirms that satisfactory examinations have been performed on the Comanche Peak Unit 2 reactor vessel.

Table 2 Additional Information Pertaining to Reactor Vessel Inspection Inspection methodology: ASME Section XI and Regulatory Guide 1.150 (Ref. 7).

Number of past inspections: 1 inspection has been performed to date on each Category B-A and B-D weld.

Number of indications found: No recordable indications were found in the latest reactor vessel inspection. Therefore, the results are acceptable per the proposed PTS Rule (Reference 8).

Proposed inspection schedule for The second inservice inspection is currently scheduled for 2012 (2RF13) balance of plant life: and will be moved to 2021 (2RF19

Attachment 2 to TXX-09106 Relief Request B-8 for Comanche Peak Unit 2.;

Page 5 of 6 Table 3 provides additional information relative to the calculation of the Through Wall Cracking Frequency (TWCF) for Comanche Peak Unit 2. As 'noted in this table, the calculation of AT30 is based upon.the embrittlement trend curve equations from Revision 2 of Regulatory Guide 1.99 (Reference 9).

The values of chemistry factor (CF) and equations for fluence function (FF) were taken directly from WCAP-16346-NP (Reference 10), which was previously provided to NRC for the Unit 2 heat-up and cool-down limit curves in Reference 11.

Table 3 Details of TWCF Calculation at 60 EFPY Inputs Reactor Coolant System Temperature, TRcs[°F]: N/A Twaji [inches]: 8.63 Region/Component Cu Ni R.G. CF Un- Fluence [1019 R Deg crion/Conent Material 2 Description Cu

[wt%] Ni

[wt] os. CF

[PF] Irradiated RTNDT(u) ['F] Neutron/cm E>1 MeV] ,

1 Inter. Shell Plate A 533B 0.060 0.640 1.1 37.0 -20 3.72 2 Inter. Shell Plate A 533B 0.060 0.640 2.1 21.6 10 3.72 3 Inter. Shell Plate A 533B 0.050 0.660 1.1 31.0 -20 3.72 4 Lower Shell Plate A 533B 0.040 0.630 1.1 26.0 -40 3.72 5 Lower Shell Plate A 533B 0.050 0.590 1.1 31.0 -30 3.72 6 Lower Shell Plate A 533B 0.030 0.650 1.1 20.0 0 3.72 7 Int. Shell Axial Weld Linde 0091 0.046 0.059 2.1 32.8 -50 3.16 8 Int. Shell Axial Weld Linde 0091 0.046 0.059 2.1 32.8 -50 2.57 9 Int. Shell Axial Weld Linde 0091 0.046 0.059 2.1 32.8 -50 3.16 10 Low. Shell Axial Weld Linde 0091 0.046 0.059 2.1 32.8 -50 2.57 11 Low. Shell Axial Weld Linde 0091 0.046 0.059 2.1 32.8 -50 3.16 12 Low. Shell Axial Weld Linde 0091 0.046 0.059 2.1 32.8 -50 3.16 13 Int./Low. Circ Weld Linde 124 0.046 0.059 2.1 32.8 -60 3.72 Outputs Methodology Used to Calculate AT30: Regulatory Guide 1.99 Rev. 2 Controlling A Material RTMAxXX Fluence [1019 FF Region # [R] Neutron/cm 2 (fluence AT30 ['F] TWCF95-xx (From , E>1 MeV] factor)

Above)

Axial Weld - AW 2 497.83 3.16 1.30 28.14 2.47E-18 Circumferential Weld - CW 2 498.64 3.72 1.34 28.95 5.52E-29 Plate - PL 2 498.64 3.72 1.34 28.95 1.07E-16 TWCF95-TOTAL ((XAwTWCF95-AW + OCPLTWCF95-PL + ccwTWCF95-cw + oaFoTWCF 9 5-FO): 2.73E-16 to TXX-09106 ReUef Request B-8 for Comanche Peak Unit 2 Page 6 of 6

6. Duration of Proposed Alternative This request is applicable to the Comanche Peak Unit 2 inservice inspection program for-the current Unit 2 Operating License.
7. References
1. ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition with the 1989 Addenda up to and including the 2004 Edition with the 2005 Addenda, American Society of Mechanical Engineers, New York.
2. OG-06-356, "Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval." MUHP 5097-99, Task 2059," October 31, 2006.
3. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

November 2002.

4. WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," June 2008.
5. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock," March, 2007.
6. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004.
7. NRC Regulatory Guide 1.150, Revision 1, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations," February 1983.
8. SECY-07-0104, "Proposed Rulemaking - Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock," June 25, 2007 (ADAMS Accession Number ML070570141).
9. NRC Regulatory Guide 1.99, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," November 2002.
10. WCAP-16346-NP, Rev. 0, "Comanche Peak Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," November 2004.
11. TXU Power letter TXX-06146 from Mike Blevins to the U. S. Nuclear Regulatory Commission, dated August 31, 2006 (ADAMS Accession No. ML062490287).
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