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{{#Wiki_filter:DUKE Kelvin Henderson Vice President ENERGY. Catawba Nuclear Station 803-701-4251 Duke Energy CNO1VP 1 4800 Concord Rd.York, SC 29745 May 20, 2013 10 CFR 50.55a U.S. Nuclear Regulatory Commission Attention:
{{#Wiki_filter:DUKE ENERGY.                                                                                        Kelvin ViceHenderson President Catawba Nuclear Station 803-701-4251 Duke Energy CNO1VP 1 4800 Concord Rd.
Document Control Desk Washington, DC 20555-0001
York, SC 29745 May 20, 2013                                                                     10 CFR 50.55a U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001


==Subject:==
==Subject:==
Duke Energy Carolinas, LLC (Duke Energy)Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Relief Request Serial Number 13-CN-003, Request for Alternative to the Requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D (Inspection Program B) for Reactor Pressure Vessel (RPV) Welds Pursuant to 10 CFR 50.55a(a)(3)(i), Duke Energy hereby submits Relief Request 13-CN-003 requesting an alternative to the requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D (Inspection Program B) to perform a volumetric examination of specified RPV welds once during each ten-year interval.
Duke Energy Carolinas, LLC (Duke Energy)
Extending the examination frequency to twenty years will reduce the radiological exposure associated with these examinations.
Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Relief Request Serial Number 13-CN-003, Request for Alternative to the Requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D (Inspection Program B) for Reactor Pressure Vessel (RPV) Welds Pursuant to 10 CFR 50.55a(a)(3)(i), Duke Energy hereby submits Relief Request 13-CN-003 requesting an alternative to the requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D (Inspection Program B) to perform a volumetric examination of specified RPV welds once during each ten-year interval. Extending the examination frequency to twenty years will reduce the radiological exposure associated with these examinations.
The basis for the proposed alternative is provided in the enclosures to this letter. Duke Energy requests approval of this alternative no later than March 31, 2014. The subject examinations would need to be performed in the Unit 1 End-of-Cycle 21 Refueling Outage, which is scheduled to begin in the Spring of 2014 and is the last refueling outage in the third inservice inspection interval.There are no regulatory commitments contained in this letter or its enclosures.
The basis for the proposed alternative is provided in the enclosures to this letter. Duke Energy requests approval of this alternative no later than March 31, 2014. The subject examinations would need to be performed in the Unit 1 End-of-Cycle 21 Refueling Outage, which is scheduled to begin in the Spring of 2014 and is the last refueling outage in the third inservice inspection interval.
If you have any questions or require additional information, please contact L. J. Rudy at (803)701-3084.Very truly yours, Kelvin Henderson LJR/s E nclosures U 4 _7 www.duke-energy.com U.S. Nuclear Regulatory Commission May 20, 2013 Page 2 xc (with enclosures):
There are no regulatory commitments contained in this letter or its enclosures.
V.M. McCree Regional Administrator U.S. Nuclear Regulatory Commission  
If you have any questions or require additional information, please contact L. J. Rudy at (803) 701-3084.
-Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 G.A. Hutto, III Senior Resident Inspector U.S. Nuclear Regulatory Commission Catawba Nuclear Station J.S. Kim (addressee only)NRC Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 8 C2 11555 Rockville Pike Rockville, MD 20852-2738 Enclosure 1 Request for Alternative 13-CN-003 Catawba Nuclear Station Unit 1 Request for Alternative 13-CN-003 Enclosure 1 Page 2 of 7 Catawba Nuclear Station Unit I Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)-Alternative Provides Acceptable Level of Quality and Safety-1. ASME Code Component Affected Catawba Unit 1 Reactor Pressure Vessel (RPV).2. Applicable Code Edition and Addenda The ASME Code Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components," Code 1998 Edition with the 2000 Addenda, is applicable for the Unit 1 Third Inservice Inspection Interval, which started on June 29, 2005 and is scheduled to end on July 15, 20141.3. Applicable Code Requirement The proposed alternative is requested in lieu of the requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D (Inspection Program B) to perform a volumetric examination of the specific items listed in the table below once during each 10-year interval.Examination Category Item No. Description B-A B1.11 Circumferential Shell Welds B-A 81.21 Circumferential Head Welds B-A 81.22 Meridional Head Welds B-A 81.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius (IR) Sections 4. Reason for Request The proposed alternative is requested to eliminate the radiological exposure associated with performing these examinations every ten years. Extending the examination frequency to twenty years will reduce the radiological exposure.5. Proposed Alternative and Basis for Use Duke Energy proposes to defer the ASME Code required volumetric examination of the Catawba Unit 1 Category B-A and B-D reactor vessel full penetration pressure retaining welds during the third inservice inspection interval until 2024 (plus or minus one outage). Volumetric examinations shall be performed in accordance with requirements of 10 CFR 50.55a and the I The interval end date has been adjusted in accordance with IWA-2430(d)(1).
Very truly yours, Kelvin Henderson LJR/s E nclosures                                                                                                   U 4 _7 www.duke-energy.com
 
U.S. Nuclear Regulatory Commission May 20, 2013 Page 2 xc (with enclosures):
V.M. McCree Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 G.A. Hutto, III Senior Resident Inspector U.S. Nuclear Regulatory Commission Catawba Nuclear Station J.S. Kim (addressee only)
NRC Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 8 C2 11555 Rockville Pike Rockville, MD 20852-2738
 
Enclosure 1 Request for Alternative 13-CN-003 Catawba Nuclear Station Unit 1
 
Request for Alternative 13-CN-003 Enclosure 1 Page 2 of 7 Catawba Nuclear Station Unit I Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
                    -Alternative Provides Acceptable Level of Quality and Safety-
: 1. ASME Code Component Affected Catawba Unit 1 Reactor Pressure Vessel (RPV).
 
===2. Applicable Code Edition and Addenda===
The ASME Code Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components," Code 1998 Edition with the 2000 Addenda, is applicable for the Unit 1 Third Inservice Inspection Interval, which started on June 29, 2005 and is scheduled to end on July 15, 20141.
 
===3. Applicable Code Requirement===
The proposed alternative is requested in lieu of the requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D (Inspection Program B) to perform a volumetric examination of the specific items listed in the table below once during each 10-year interval.
Examination Category                 Item No.                             Description B-A                       B1.11       Circumferential Shell Welds B-A                       81.21       Circumferential Head Welds B-A                       81.22       Meridional Head Welds B-A                       81.30       Shell-to-Flange Weld B-D                       B3.90       Nozzle-to-Vessel Welds B-D                       B3.100     Nozzle Inside Radius (IR) Sections
 
===4. Reason for Request===
The proposed alternative is requested to eliminate the radiological exposure associated with performing these examinations every ten years. Extending the examination frequency to twenty years will reduce the radiological exposure.
: 5. Proposed Alternative and Basis for Use Duke Energy proposes to defer the ASME Code required volumetric examination of the Catawba Unit 1 Category B-A and B-D reactor vessel full penetration pressure retaining welds during the third inservice inspection interval until 2024 (plus or minus one outage). Volumetric examinations shall be performed in accordance with requirements of 10 CFR 50.55a and the IThe  interval end date has been adjusted in accordance with IWA-2430(d)(1).
 
Request for Alternative 13-CN-003 Enclosure 1 Page 3 of 7 Edition and Addenda of the ASME Code, Section XI, applicable at the time of the next scheduled examination.
Request for Alternative 13-CN-003 Enclosure 1 Page 3 of 7 Edition and Addenda of the ASME Code, Section XI, applicable at the time of the next scheduled examination.
In accordance with 10 CFR 50.55a(a)(3)(i), the alternative is requested on the basis that it provides an acceptable level of quality and safety because there is a negligible change in risk, as measured by evaluating the risk criteria specified in Regulatory Guide 1.174 (Reference 8.4), and as documented herein.The methodology used to demonstrate the acceptability of extending the inspection interval for Category B-A and B-D welds is based on a negligible change in risk, and is documented in WCAP-16168-NP-A, Revision 2 (Reference 8.5). 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 8.6). 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 Catawba Unit 1 reactor vessel is acceptable as shown in Table 1 below.Table I Critical Parameters for Application of Bounding Analysis Unit 1 Bounded Unit I by Pilot Plant Parameter Pilot Plant Basis Plant-Specific Basis Basis?Dominant Pressurized NRC PTS Risk Study PTS Generalization Yes (No Further Thermal Shock (PTS) (Reference 8.6) Study (References 8.7 Evaluation Transients in the NRC PTS and 8.10) Required)Risk Study are Applicable Through-Wall Cracking 1.76E-08 Events per Year 1.11 E-14 Events per Yes (No Further Frequency (Reference 8.5) Year (References 8.5 Evaluation and 8.10) Required)Frequency and Severity of 7 Heatup/Cooldowns per Bounded by 7 Yes (No Further Design Basis Transients Year (Reference 8.5) Heatup/Cooldowns per Evaluation year Required)Cladding Layers Single Layer (Reference 8.5) Multi-Layer Yes (No Further (Single/Multiple)
In accordance with 10 CFR 50.55a(a)(3)(i), the alternative is requested on the basis that it provides an acceptable level of quality and safety because there is a negligible change in risk, as measured by evaluating the risk criteria specified in Regulatory Guide 1.174 (Reference 8.4), and as documented herein.
Evaluation Required)
The methodology used to demonstrate the acceptability of extending the inspection interval for Category B-A and B-D welds is based on a negligible change in risk, and is documented in WCAP-16168-NP-A, Revision 2 (Reference 8.5). 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 8.6). 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 Catawba Unit 1 reactor vessel is acceptable as shown in Table 1 below.
Request for Alternative 13-CN-003 Enclosure 1 Page 4 of 7 Additional information relative to the Catawba Unit 1 reactor vessel inspection is provided in Table 2. This information confirms that satisfactory examinations have been performed on the Catawba Unit 1 reactor vessel.Table 2 Additional Information Pertaining to Reactor Vessel Inspection Inspection Methodology:
Table I Critical Parameters for Application of Bounding Analysis Unit 1 Bounded Unit I           by Pilot Plant Parameter                 Pilot Plant Basis       Plant-Specific Basis         Basis?
The most recent inservice inspection of the Category B-A and B-D welds was performed in accordance with ASME Section XI Appendix VIII requirements.
Dominant Pressurized         NRC PTS Risk Study           PTS Generalization       Yes (No Further Thermal Shock (PTS)           (Reference 8.6)             Study (References 8.7       Evaluation Transients in the NRC PTS                                 and 8.10)                   Required)
Code case N-613-1 was used in lieu of the Section Xl requirements for inspection of the Category B-D nozzle to shell welds. Code case N-648-1 visual examinations were performed in lieu of the Section Xl Category B-D volumetric examinations of the inner radii.Number of Past Inspections:
Risk Study are Applicable Through-Wall Cracking         1.76E-08 Events per Year     1.11 E-14 Events per     Yes (No Further Frequency                     (Reference 8.5)             Year (References 8.5       Evaluation and 8.10)                   Required)
Two 10-Year inservice inspections have been performed.
Frequency and Severity of     7 Heatup/Cooldowns per       Bounded by 7             Yes (No Further Design Basis Transients       Year (Reference 8.5)         Heatup/Cooldowns per       Evaluation year                       Required)
Number of Indications Found: One indication was identified in the beltline region during the most recent inservice inspection.
Cladding Layers               Single Layer (Reference 8.5) Multi-Layer               Yes (No Further (Single/Multiple)                                                                     Evaluation Required)
This indication was acceptable per Table IWB-3510-1 of Section XI of the ASME Code. This indication is not within 1/1 0 th or 1" of the reactor vessel shell plate inside diameter surface. The lack of any indications near the inner surface meets the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 8.8).Proposed Inspection Schedule for The welds for which the proposed alternative is requested shall be Balance of Plant Life: examined no later than 2024 (plus or minus one refueling outage).Subsequent to 2024, these welds shall be examined in accordance with the applicable requirements of the ASME Code, Section Xl, as required by 10 CFR 50.55a.
 
Request for Alternative 13-CN-003 Enclosure 1 Page 5 of 7 Table 3 provides additional information relative to the calculation of the Through Wall Cracking Frequency (TWCF) for Catawba Unit 1.Table 3 Details of TWCF Calculation  
Request for Alternative 13-CN-003 Enclosure 1 Page 4 of 7 Additional information relative to the Catawba Unit 1 reactor vessel inspection is provided in Table 2. This information confirms that satisfactory examinations have been performed on the Catawba Unit 1 reactor vessel.
-Performed for 60 Effective Full Power Years (EFPY)Inputs Reactor Coolant System Temperature, TRcs[°F]:
Table 2 Additional Information Pertaining to Reactor Vessel Inspection Inspection Methodology:           The most recent inservice inspection of the Category B-A and B-D welds was performed in accordance with ASME Section XI Appendix VIII requirements. Code case N-613-1 was used in lieu of the Section Xl requirements for inspection of the Category B-D nozzle to shell welds. Code case N-648-1 visual examinations were performed in lieu of the Section Xl Category B-D volumetric examinations of the inner radii.
N/A Thickness of the RPV wall (including cladding), Twa 1 l [inches]:
Number of Past Inspections:       Two 10-Year inservice inspections have been performed.
8.62 Fluence [1019 No. Region/Component Material/
Number of Indications Found:     One indication was identified in the beltline region during the most recent inservice inspection. This indication was acceptable per Table IWB-3510-1 of Section XI of the ASME Code. This indication is not within 1 / 1 0 th or 1"of the reactor vessel shell plate inside diameter surface. The lack of any indications near the inner surface meets the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 8.8).
Cu Ni R.G. 1.99 C.F. Un-Irradiated Neutron/cm 2 , Description Flux Type [wt%] [wt%] Position [OF] RTNDT [OF] E > 1.0 MeV]1 Intermediate Shell Forging A 508-2 0.090 0.860 2.1 28.4 -8 3.06 2 Lower Shell Forging A 508-2 0.040 0.830 1.1 26.0 -13 3.06 3 Intermediate-to-Lower GRAU LO 0.040 0.720 2.1 23.2 -51 3.06 Shell Forging Circumferential Weld 4 Nozzle Shell Forging A 508-2 0.25 0.85 1.1 201.25 -26 0.142 5 Nozzle-to-Intermediate GRAU LO 0.03 0.75 1.1 41 0 0.142 Shell Forging Circumferential Weld Outputs Methodology Used to Calculate AT 3 0: Regulatory Guide 1.99, Revision 2 Controlling Material Fluence [1019 FF (Fluence T 3 0 Region No. RTMAx-xx Neutron /cm2, Factor) A 0  TWCF 9 5-XX (From Above) [R] E > 1.0 MeV] Factor) [OF]Circumferential Weld -CW 4 532.44 0.142 0.491 98.75 0.00 Forging- FO 4 532.44 0.142 0.491 98.75 4.45E-15 TWCF95-TOTAL (ctcwTWcF95-CW  
Proposed Inspection Schedule for The welds for which the proposed alternative is requested shall be Balance of Plant Life:             examined no later than 2024 (plus or minus one refueling outage).
+ .FoTMWCF 9 5-FO): 1.11E-14 Failure Consequences:
Subsequent to 2024, these welds shall be examined in accordance with the applicable requirements of the ASME Code, Section Xl, as required by 10 CFR 50.55a.
The failure of any of the Reactor Pressure Vessel shell, head, or nozzle welds listed in this request would result in a loss of the structural integrity of the Reactor Vessel.Conclusion:
 
Because the parameters in WCAP-16168-NP Revision 2, Appendix A bound the plant-specific parameters for Catawba Unit 1, the change in risk meets the RG 1.174 acceptance guidelines for a small change in Large Early Release Frequency (LERF).Increasing the reactor vessel inspection frequency from 10 to 20 years satisfies all the Request for Alternative 13-CN-003 Enclosure 1 Page 6 of 7 RG 1.174 criteria.
Request for Alternative 13-CN-003 Enclosure 1 Page 5 of 7 Table 3 provides additional information relative to the calculation of the Through Wall Cracking Frequency (TWCF) for Catawba Unit 1.
For these reasons, the proposed alternative provides an acceptable level of quality and safety.6. Duration of Proposed Alternative This request is applicable until 2024 (plus or minus one outage), which includes the duration of the Catawba Unit 1 Third Inservice Inspection Interval.7. Precedents 7.1 Duke Energy Carolina, LLC (Duke), McGuire Nuclear Station Unit 1, Docket No.50-369, "Relief Request Serial # 09-MN-003" dated June 29, 2009 [Safety Evaluation dated June 28, 2010 (ADAMS Accession Number ML101610306)]
Table 3 Details of TWCF Calculation - Performed for 60 Effective Full Power Years (EFPY)
7.2 Calvert Cliffs Nuclear Plant Unit No. 2, Docket No. 50-318, "Revised Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations  
Inputs Reactor Coolant System Temperature, TRcs[°F]: N/A         Thickness of the RPV wall (including cladding),
-Relief Request (ISI-020 and ISI-021)," dated October 1, 2008 (ADAMS Accession Number ML082760282).
Twa1 l [inches]:     8.62 Fluence [1019 2
7.3 Donald C. Cook Nuclear Plant Unit No. 2, Docket No. 50-316, "Request for Relief to Extend the Unit 2 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses," dated October 9, 2008 (ADAMS Accession Number ML082980354).
No.       Region/Component Description          Material/
7.4 Indian Point Nuclear Generating Units Nos. 2 and 3, Docket Nos. 50-247 and 50-286,"Request for Relief to Extend the Unit 2 and 3 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses," dated July 8, 2008 (ADAMS Accession Number ML081980058).
Flux        Cu Type [wt%]      Ni
: 8. References 8.1 ASME Boiler and Pressure Vessel Code, Section XI, 1998 Edition with the 2000 Addenda, American Society of Mechanical Engineers, New York.8.2 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.'
[wt%]    R.G. 1.99 Position    C.F.
MUHP 5097-99, Task 2059," October 31, 2006 (ADAMS Accession Number ML082210245).
[OF]  Un-Irradiated RTNDT [OF]        E > 1.0 MeV] ,
8.3 OG-10-238, "Revision to the Revised 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.'
Neutron/cm 1    Intermediate Shell Forging   A 508-2   0.090 0.860           2.1     28.4           -8               3.06 2     Lower Shell Forging         A 508-2   0.040   0.830         1.1     26.0         -13               3.06 3     Intermediate-to-Lower       GRAU LO 0.040       0.720         2.1     23.2         -51             3.06 Shell Forging Circumferential Weld 4     Nozzle Shell Forging         A 508-2   0.25     0.85         1.1     201.25         -26             0.142 5     Nozzle-to-Intermediate     GRAU LO 0.03         0.75         1.1       41             0             0.142 Shell Forging Circumferential Weld Outputs Methodology Used to Calculate AT30: Regulatory Guide 1.99, Revision 2 Controlling Material                   Fluence [1019 FF (Fluence           T30 Region No.       RTMAx-xx       Neutron /cm2,     Factor)       A0      TWCF 9 5 -XX (From Above)           [R]           E > 1.0 MeV]     Factor)       [OF]
PA-MSC-0120," July 12, 2010 8.4 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 2002.8.5 WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," June 2008.8.6 NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock," March, 2007 (ADAMS Accession Number ML070860156).
Circumferential Weld - CW           4             532.44             0.142         0.491     98.75         0.00 Forging- FO                         4             532.44             0.142         0.491     98.75       4.45E-15 TWCF95-TOTAL (ctcwTWcF95-CW +         .FoTMWCF 9 5 -FO): 1.11E-14 Failure Consequences:
Request for Alternative 13-CN-003 Enclosure 1 Page 7 of 7 8.7 NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004 (ADAMS Accession Number ML042880482).
The failure of any of the Reactor Pressure Vessel shell, head, or nozzle welds listed in this request would result in a loss of the structural integrity of the Reactor Vessel.
8.8 Code of Federal Regulations, 10 CFR Part 50.61a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S.Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 75, No.1, dated January 4, 2010, with corrections in Volume 75, No. 22, dated February 3, 2010, Volume 75, No. 44, dated March 8, 2010, and Volume 75, No. 227, dated November 26, 2010.8.9 Revised Final Safety Evaluation by the Office Of Nuclear Reactor Regulation Topical Report WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension Of The Reactor Vessel In-Service Inspection Interval", July 26, 2011 (ADAMS Accession Number ML111600303).
 
8.10 Duke Energy Corporation Calculation  
==
#DPC-1201.00-00-0010 (Catawba Calculation
Conclusion:==
#CNC-1201.01-00-0068), "Implementation of WCAP-16168-NP-A, Revision 2 for Catawba Units 1 and 2 and McGuire Units 1 and 2", Revision 3.
 
Enclosure 2 Request for Alternative 13-CN-003 Catawba Nuclear Station Unit 2 Request for Alternative 13-CN-003 Enclosure 2 Page 2 of 8 Catawba Nuclear Station Unit 2 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)-Alternative Provides Acceptable Level of Quality and Safety-1. ASME Code Component(s)
Because the parameters in WCAP-16168-NP Revision 2, Appendix A bound the plant-specific parameters for Catawba Unit 1, the change in risk meets the RG 1.174 acceptance guidelines for a small change in Large Early Release Frequency (LERF).
Affected Catawba Unit 2 Reactor Pressure Vessel.2. Applicable Code Edition and Addenda The ASME Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Code 1998 Edition with the 2000 Addenda, is applicable for the Unit 2 Third Inservice Inspection Interval, which started on October 15, 2005, and is scheduled to end on August 19, 2016.3. Applicable Code Requirement The proposed alternative is requested in lieu of the requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D (Inspection Program B) to perform a volumetric examination of the specific items listed in the table below once during each 1 0-year interval.Examination Category Item No. Description B-A B1.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B13.21 Circumferential Head Welds B-A B13.22 Meridional Head Welds B-A B13.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius (IR) Sections 4. Reason for Request The proposed alternative is requested to eliminate the radiological exposure associated with performing these examinations every ten years. Extending the examination frequency to twenty years will reduce the radiological exposure.5. Proposed Alternative and Basis for Use Duke Energy proposes to defer the ASME Code required volumetric examination of the Catawba Unit 2 Category B-A and B-D reactor vessel full penetration pressure retaining welds during the third inservice inspection interval until 2024 (plus or minus one outage). Volumetric I The interval start date was adjusted in accordance with IWA-2430(d)(1).
Increasing the reactor vessel inspection frequency from 10 to 20 years satisfies all the
 
Request for Alternative 13-CN-003 Enclosure 1 Page 6 of 7 RG 1.174 criteria. For these reasons, the proposed alternative provides an acceptable level of quality and safety.
: 6. Duration of Proposed Alternative This request is applicable until 2024 (plus or minus one outage), which includes the duration of the Catawba Unit 1 Third Inservice Inspection Interval.
: 7. Precedents 7.1 Duke Energy Carolina, LLC (Duke), McGuire Nuclear Station Unit 1, Docket No.
50-369, "Relief Request Serial # 09-MN-003" dated June 29, 2009 [Safety Evaluation dated June 28, 2010 (ADAMS Accession Number ML101610306)]
7.2 Calvert Cliffs Nuclear Plant Unit No. 2, Docket No. 50-318, "Revised Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations - Relief Request (ISI-020 and ISI-021)," dated October 1, 2008 (ADAMS Accession Number ML082760282).
7.3 Donald C. Cook Nuclear Plant Unit No. 2, Docket No. 50-316, "Request for Relief to Extend the Unit 2 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses,"
dated October 9, 2008 (ADAMS Accession Number ML082980354).
7.4 Indian Point Nuclear Generating Units Nos. 2 and 3, Docket Nos. 50-247 and 50-286, "Request for Relief to Extend the Unit 2 and 3 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses," dated July 8, 2008 (ADAMS Accession Number ML081980058).
: 8. References 8.1   ASME Boiler and Pressure Vessel Code, Section XI, 1998 Edition with the 2000 Addenda, American Society of Mechanical Engineers, New York.
8.2   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 (ADAMS Accession Number ML082210245).
8.3   OG-10-238, "Revision to the Revised 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.'
PA-MSC-0120," July 12, 2010 8.4   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.
8.5   WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," June 2008.
8.6   NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock,"
March, 2007 (ADAMS Accession Number ML070860156).
 
Request for Alternative 13-CN-003 Enclosure 1 Page 7 of 7 8.7 NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004 (ADAMS Accession Number ML042880482).
8.8 Code of Federal Regulations, 10 CFR Part 50.61a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S.
Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 75, No.
1, dated January 4, 2010, with corrections in Volume 75, No. 22, dated February 3, 2010, Volume 75, No. 44, dated March 8, 2010, and Volume 75, No. 227, dated November 26, 2010.
8.9 Revised Final Safety Evaluation by the Office Of Nuclear Reactor Regulation Topical Report WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension Of The Reactor Vessel In-Service Inspection Interval", July 26, 2011 (ADAMS Accession Number ML111600303).
8.10 Duke Energy Corporation Calculation #DPC-1201.00-00-0010 (Catawba Calculation
    #CNC-1201.01-00-0068), "Implementation of WCAP-16168-NP-A, Revision 2 for Catawba Units 1 and 2 and McGuire Units 1 and 2", Revision 3.
 
Enclosure 2 Request for Alternative 13-CN-003 Catawba Nuclear Station Unit 2
 
Request for Alternative 13-CN-003 Enclosure 2 Page 2 of 8 Catawba Nuclear Station Unit 2 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
                      -Alternative Provides Acceptable Level of Quality and Safety-
: 1. ASME Code Component(s) Affected Catawba Unit 2 Reactor Pressure Vessel.
 
===2. Applicable Code Edition and Addenda===
The ASME Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Code 1998 Edition with the 2000 Addenda, is applicable for the Unit 2 Third Inservice Inspection Interval, which started on October 15, 2005, and is scheduled to end on August 19, 2016.
 
===3. Applicable Code Requirement===
The proposed alternative is requested in lieu of the requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D (Inspection Program B) to perform a volumetric examination of the specific items listed in the table below once during each 10-year interval.
Examination Category               Item No.                                 Description B-A                 B1.11           Circumferential Shell Welds B-A                 B1.12           Longitudinal Shell Welds B-A                 B13.21         Circumferential Head Welds B-A                 B13.22         Meridional Head Welds B-A                 B13.30         Shell-to-Flange Weld B-D                 B3.90           Nozzle-to-Vessel Welds B-D                 B3.100         Nozzle Inside Radius (IR)Sections
 
===4. Reason for Request===
The proposed alternative is requested to eliminate the radiological exposure associated with performing these examinations every ten years. Extending the examination frequency to twenty years will reduce the radiological exposure.
: 5. Proposed Alternative and Basis for Use Duke Energy proposes to defer the ASME Code required volumetric examination of the Catawba Unit 2 Category B-A and B-D reactor vessel full penetration pressure retaining welds during the third inservice inspection interval until 2024 (plus or minus one outage). Volumetric I The interval start date was adjusted in accordance with IWA-2430(d)(1).
 
Request for Alternative 13-CN-003 Enclosure 2 Page 3 of 8 examinations shall be performed in accordance with requirements of 10 CFR 50.55a and the Edition and Addenda of the ASME Code, Section XI, applicable at the time of the next scheduled examination.
Request for Alternative 13-CN-003 Enclosure 2 Page 3 of 8 examinations shall be performed in accordance with requirements of 10 CFR 50.55a and the Edition and Addenda of the ASME Code, Section XI, applicable at the time of the next scheduled examination.
In accordance with 10 CFR 50.55a(a)(3)(i), the alternative is requested on the basis that it provides an acceptable level of quality and safety because there is a negligible change in risk, as measured by evaluating the risk criteria specified in Regulatory Guide 1.174 (Reference 8.4), and as documented herein.The methodology used to demonstrate the acceptability of extending the inspection interval for Category B-A and B-D welds is based on a negligible change in risk, and is documented in WCAP-16168-NP-A, Revision 2 (Reference 8.5). 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 8.7). 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 Catawba Unit 2 reactor vessel is acceptable as shown in Table 1 below.Table 1 Critical Parameters for Application of Bounding Analysis Unit 2 Bounded Unit 2 by Pilot Plant Parameter Pilot Plant Basis Plant-Specific Basis Basis?Dominant Pressurized NRC PTS Risk Study PTS Generalization Study Yes (No Further Thermal Shock (PTS) (Reference 8.6) (References 8.7 and 8.10) Evaluation Transients in the NRC PTS Required)Risk Study are Applicable Through-Wall Cracking 1.76E-08 Events per Year 1.40E-1 3 Events per Year Yes (No Further Frequency (Reference 8.5) (References 8.5 and 8.10) Evaluation Required)Frequency and Severity of 7 Heatup/Cooldowns per Bounded by 7 Yes (No Further Design Basis Transients Year (Reference 8.5) Heatup/Cooldowns per Evaluation Year Required)Cladding Layers Single Layer (Reference 8.5) Single Layer (Reference Yes (No Further (Single/Multiple) 8.10) Evaluation Required)
In accordance with 10 CFR 50.55a(a)(3)(i), the alternative is requested on the basis that it provides an acceptable level of quality and safety because there is a negligible change in risk, as measured by evaluating the risk criteria specified in Regulatory Guide 1.174 (Reference 8.4), and as documented herein.
Request for Alternative 13-CN-003 Enclosure 2 Page 4 of 8 Additional information relative to the Catawba Unit 2 reactor vessel inspection is provided in Table 2. This information confirms that satisfactory examinations have been performed on the Catawba Unit 2 reactor vessel.Table 2 Additional Information Pertaining to Reactor Vessel Inspection Inspection Methodology:
The methodology used to demonstrate the acceptability of extending the inspection interval for Category B-A and B-D welds is based on a negligible change in risk, and is documented in WCAP-16168-NP-A, Revision 2 (Reference 8.5). 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 8.7). 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 Catawba Unit 2 reactor vessel is acceptable as shown in Table 1 below.
The most recent inservice inspection of the Category B-A and B-D welds was performed in accordance with ASME Section XI Appendix VIII requirements.
Table 1 Critical Parameters for Application of Bounding Analysis Unit 2 Bounded Unit 2           by Pilot Plant Parameter                 Pilot Plant Basis       Plant-Specific Basis         Basis?
Code case N-613-1 was used in lieu of the Section Xl requirements for inspection of the Category B-D nozzle to shell welds. Code case N-648-1 visual examinations were performed in lieu of the Section XI Category B-D volumetric examinations of the inner radii.Number of Past Inspections:
Dominant Pressurized         NRC PTS Risk Study           PTS Generalization Study   Yes (No Further Thermal Shock (PTS)           (Reference 8.6)             (References 8.7 and 8.10)     Evaluation Transients in the NRC PTS                                                               Required)
Two 10-Year inservice inspections have been performed.
Risk Study are Applicable Through-Wall Cracking         1.76E-08 Events per Year     1.40E-1 3 Events per Year Yes (No Further Frequency                     (Reference 8.5)             (References 8.5 and 8.10)     Evaluation Required)
Number of Indications Six indications, which were identified in the beltline region during the most recent Found: inservice inspection, were acceptable per IWB-3510-1.
Frequency and Severity of     7 Heatup/Cooldowns per       Bounded by 7               Yes (No Further Design Basis Transients       Year (Reference 8.5)         Heatup/Cooldowns per         Evaluation Year                         Required)
Only two of these indications (both within the weld material in Lower Axial Weld #101-142A) were within the greater of 1/10th or 1" of the reactor vessel shell plate inside diameter surface and were acceptable per the requirements of the Alternate PTS Rule 10 CFR 50.61 a (Reference 8.8). A summary of these indications is provided in Table 3. The locations of these two indications are shown in Figure 1.Proposed Inspection The welds for which the proposed alternative is requested shall be examined no Schedule for Balance of later than 2024 (plus or minus one refueling outage). Subsequent to 2024, these Plant Life: welds shall be examined in accordance with the applicable requirements of the ASME Code, Section Xl, as required by 10 CFR 50.55a.Table 3 Summary of Reactor Vessel Weld Inspection Indications Through-Wall Extent, TWE (in.)Scaled Maximum TWEMIN TWEm, Number of Flaws 2  Number of Flaws 0.075 0.475 162 2 (1 Axial and 1 Circumferential) 0.125 0.475 88 2 (1 Axial and 1 Circumferential) 0.175 0.475 23 2 (1 Axial and 1 Circumferential) 0.225 0.475 9 2 (1 Axial and 1 Circumferential) 0.275 0.475 4 1 (1 Axial and 0 Circumferential) 0.325 0.475 3 1 (1 Axial and 0 Circumferential) 2 The "Scaled Maximum Number of Flaws" indicates the number allowed per the Alternate PTS rule, 10 CFR 50.61a (Reference 8.8). This number is based on the length of weld inspected in the beltline region.
Cladding Layers               Single Layer (Reference 8.5) Single Layer (Reference   Yes (No Further (Single/Multiple)                                         8.10)                         Evaluation Required)
Request for Alternative 13-CN-003 Enclosure 2 Page 5 of 8 Table 4 provides additional information relative to the calculation of the Through Wall Cracking Frequency (TWCF) for Catawba Unit 2.Table 4 Details of TWCF Calculation  
 
-Performed for 60 Effective Full Power Years (EFPY)Inputs Reactor Coolant System Temperature, TRCS[*F]:
Request for Alternative 13-CN-003 Enclosure 2 Page 4 of 8 Additional information relative to the Catawba Unit 2 reactor vessel inspection is provided in Table 2. This information confirms that satisfactory examinations have been performed on the Catawba Unit 2 reactor vessel.
N/A Thickness of the RPV wall (including cladding), Twa 1 1 [inches]:
Table 2 Additional Information Pertaining to Reactor Vessel Inspection Inspection Methodology:           The most recent inservice inspection of the Category B-A and B-D welds was performed in accordance with ASME Section XI Appendix VIII requirements.
8.78 Fluence [1019 No. Region/Component Material/
Code case N-613-1 was used in lieu of the Section Xl requirements for inspection of the Category B-D nozzle to shell welds. Code case N-648-1 visual examinations were performed in lieu of the Section XI Category B-D volumetric examinations of the inner radii.
Flu] Cu Ni R.G. 1.99 C.F. Un-Irradiated Neutron/cm 2 , Description Type [wt%] [wt%] Position [OF] RTNDT [*F] E > 1.0 MeV]1 Intermediate Plate B8616-1 A 533B 0.050 0.600 1.1 31.0 12 3.03 2 Intermediate Plate B8605-1 A 533B 0.080 0.620 2.1 44.0 15 3.03 3 Intermediate Plate B8605-2 A 533B 0.080 0.610 1.1 51.0 33 3.03 4 Lower Plate B8806-2 A 533B 0.060 0.590 1.1 37.0 -10 3.03 5 Lower Plate B8806-3 A 533B 0.060 0.590 1.1 37.0 8 3.03 6 Lower Plate B8806-1 A 533B 0.060 0.560 1.1 37.0 6 3.03 7 Intermediate Axial Weld Linde 0091 0.040 0.140 2.1 33.4 -80 1.76 101-124A 8 Intermediate Axial Weld Linde 0091 0.040 0.140 2.1 33.4 -80 2.89 101-124B 9 Intermediate Axial Weld Linde 0091 0.040 0.140 2.1 33.4 -80 2.89 101-124C 10 Lower Axial Weld 101-142A Linde 0091 0.040 0.140 2.1 33.4 -80 2.89 11 Lower Axial. Weld 101-142B Linde 0091 0.040 0.140 2.1 33.4 -80 1.76 12 Lower Axial Weld 101-142C Linde 0091 0.040 0.140 2.1 33.4 -80 2.89 13 Intermediate/Lower Linde 0091 0.040 0.140 2.1 33.4 -80 3.03 Circumferential Weld 101-171 14 Nozzle Plate B8604-1 A 533B 0.11 0.61 1.1 74.15 24 0.0670 15 Nozzle Plate B8604-2 A 533B 0.11 0.61 1.1 74.15 26 0.0670 16 Nozzle Plate B8604-3 A 533B 0.07 0.53 1.1 44 50 0.0670 17 Nozzle Axial Weld 101-122A Linde 0091 0.156 0.059 1.1 73.71 -50 0.0637 18 Nozzle Axial Weld 101-122B Linde 0091 0.156 0.059 1.1 73.71 -50 0.0670 19 Nozzle Axial Weld 101-122C Linde 0091 0.156 0.059 1.1 73.71 -50 0.0572 20 Intermediate/Nozzle Linde 0091 0.153 0.077 1.1 74.13 -40 0.0670 Circumferential Weld 103-121 Request for Alternative 13-CN-003 Enclosure 2 Page 6 of 8 Table 4 Details of TWCF Calculation  
Number of Past Inspections:       Two 10-Year inservice inspections have been performed.
-Performed for 60 Effective Full Power Years (EFPY)Outputs Methodology Used to Calculate AT 3 o: Reculatory Guide 1.99, Revision 2 Controlling Fluence [10 19 FF Material Region 2T3 No. RTMAx.xx [R] Neutron/cm2, (Fluence AT 3 0 TWCF 9 5 sxx (From Above) E > 1.0 MeV] Factor) [OF]Axial Weld -AW 3 558.04 2.89 1.282 65.37 0.00 Circumferential Weld -3 558.64 3.03 1.293 65.95 0.00 CW Plate -PL 3 558.64 3.03 1.293 65.95 5.61 E-14 TWCF95-TOTAL (aAwTWCF95-AW  
Number of Indications             Six indications, which were identified inthe beltline region during the most recent Found:                           inservice inspection, were acceptable per IWB-3510-1. Only two of these indications (both within the weld material in Lower Axial Weld #101-142A) were within the greater of 1/10th or 1"of the reactor vessel shell plate inside diameter surface and were acceptable per the requirements of the Alternate PTS Rule 10 CFR 50.61 a (Reference 8.8). A summary of these indications is provided in Table 3. The locations of these two indications are shown in Figure 1.
+ aPLTWCF95.PL  
Proposed Inspection               The welds for which the proposed alternative is requested shall be examined no Schedule for Balance of           later than 2024 (plus or minus one refueling outage). Subsequent to 2024, these Plant Life:                       welds shall be examined inaccordance with the applicable requirements of the ASME Code, Section Xl, as required by 10 CFR 50.55a.
+ acwTWCF 9 5.cw): 1.40E-13 Figure 1 Catawba Unit 2 Reactor Vessel Beltline Indication Map (Figure is not to scale. Indication location is approximate.)
Table 3 Summary of Reactor Vessel Weld Inspection Indications Through-Wall Extent, TWE (in.)
Scaled Maximum2 Number of Flaws                       Number of Flaws TWEMIN                  TWEm, 0.075                   0.475                     162                 2 (1 Axial and 1 Circumferential) 0.125                   0.475                     88                 2 (1 Axial and 1 Circumferential) 0.175                   0.475                     23                 2 (1 Axial and 1 Circumferential) 0.225                   0.475                       9                 2 (1 Axial and 1 Circumferential) 0.275                   0.475                       4                 1 (1 Axial and 0 Circumferential) 0.325                   0.475                       3                 1 (1 Axial and 0 Circumferential) 2 The "Scaled Maximum Number of Flaws" indicates the number allowed per the Alternate PTS rule, 10 CFR 50.61a (Reference 8.8). This number is based on the length of weld inspected in the beltline region.
 
Request for Alternative 13-CN-003 Enclosure 2 Page 5 of 8 Table 4 provides additional information relative to the calculation of the Through Wall Cracking Frequency (TWCF) for Catawba Unit 2.
Table 4 Details of TWCF Calculation - Performed for 60 Effective Full Power Years (EFPY)
Inputs Reactor Coolant System Temperature, TRCS[*F]: N/A   Thickness of the RPV wall (including cladding),
Twa11 [inches]: 8.78 Fluence [1019 No.       Region/Component         Material/ Flu] Cu         Ni R.G. 1.99 C.F. Un-Irradiated Neutron/cm 2, Description             Type       [wt%]     [wt%] Position [OF]       RTNDT [*F]     E > 1.0 MeV]
1   Intermediate Plate B8616-1     A 533B       0.050     0.600   1.1   31.0         12             3.03 2   Intermediate Plate B8605-1     A 533B       0.080     0.620   2.1   44.0         15             3.03 3   Intermediate Plate B8605-2     A 533B       0.080     0.610   1.1   51.0         33             3.03 4   Lower Plate B8806-2             A 533B       0.060     0.590   1.1   37.0         -10             3.03 5   Lower Plate B8806-3             A 533B       0.060     0.590   1.1   37.0           8             3.03 6   Lower Plate B8806-1             A 533B       0.060     0.560   1.1   37.0           6             3.03 7   Intermediate Axial Weld       Linde 0091   0.040     0.140   2.1   33.4         -80             1.76 101-124A 8   Intermediate Axial Weld       Linde 0091   0.040     0.140   2.1   33.4         -80             2.89 101-124B 9   Intermediate Axial Weld       Linde 0091   0.040     0.140   2.1   33.4         -80             2.89 101-124C 10   Lower Axial Weld 101-142A     Linde 0091     0.040     0.140   2.1   33.4         -80             2.89 11   Lower Axial. Weld 101-142B   Linde 0091     0.040     0.140   2.1   33.4         -80             1.76 12   Lower Axial Weld 101-142C     Linde 0091   0.040     0.140   2.1   33.4         -80             2.89 13   Intermediate/Lower             Linde 0091   0.040     0.140   2.1   33.4         -80             3.03 Circumferential Weld 101-171 14   Nozzle Plate B8604-1           A 533B       0.11     0.61   1.1   74.15         24           0.0670 15   Nozzle Plate B8604-2           A 533B       0.11     0.61   1.1   74.15         26           0.0670 16   Nozzle Plate B8604-3           A 533B       0.07     0.53   1.1     44           50           0.0670 17   Nozzle Axial Weld 101-122A     Linde 0091   0.156     0.059   1.1   73.71         -50           0.0637 18   Nozzle Axial Weld 101-122B     Linde 0091   0.156     0.059   1.1   73.71         -50           0.0670 19   Nozzle Axial Weld 101-122C     Linde 0091   0.156     0.059   1.1   73.71         -50           0.0572 20   Intermediate/Nozzle           Linde 0091   0.153     0.077   1.1   74.13         -40           0.0670 Circumferential Weld 103-121
 
Request for Alternative 13-CN-003 Enclosure 2 Page 6 of 8 Table 4 Details of TWCF Calculation - Performed for 60 Effective Full Power Years (EFPY)
Outputs Methodology Used to Calculate AT3o:       Reculatory Guide 1.99, Revision 2 Controlling                       Fluence [10 19     FF Material Region                                 2T3 No.           RTMAx.xx [R]   Neutron/cm2,     (Fluence   AT30    TWCF9 5sxx (From Above)                       E > 1.0 MeV]     Factor)     [OF]
Axial Weld - AW                 3             558.04             2.89         1.282     65.37       0.00 Circumferential Weld -         3             558.64             3.03         1.293     65.95       0.00 CW Plate - PL                     3             558.64             3.03         1.293     65.95     5.61 E-14 TWCF95-TOTAL (aAwTWCF95-AW + aPLTWCF95.PL + acwTWCF 95.cw):           1.40E-13 Figure 1 Catawba Unit 2 Reactor Vessel Beltline Indication Map (Figure is not to scale. Indication location is approximate.)
 
Request for Alternative 13-CN-003 Enclosure 2 Page 7 of 8 Failure Consequences:
Request for Alternative 13-CN-003 Enclosure 2 Page 7 of 8 Failure Consequences:
The failure of any of the Reactor Pressure Vessel shell, head, or nozzle welds listed in this request would result in a loss of the structural integrity of the Reactor Vessel.Conclusion:
The failure of any of the Reactor Pressure Vessel shell, head, or nozzle welds listed in this request would result in a loss of the structural integrity of the Reactor Vessel.
Because the parameters in WCAP-16168-NP Revision 2, Appendix A bound the plant-specific parameters for Catawba Unit 2, the change in risk meets the RG 1.174 acceptance guidelines for a small change in LERF. Increasing the reactor vessel inspection frequency from 10 to 20 years satisfies all the RG 1.174 criteria.
 
For these reasons, the proposed alternative provides an acceptable level of quality and safety.6. Duration of Proposed Alternative This request is applicable until 2024 (plus or minus one outage), which includes the duration of the Catawba Unit 2 Third Inservice Inspection Interval.7. Precedents 7.1 Duke Energy Carolina, LLC (Duke), McGuire Nuclear Station Unit 1, Docket No.50-369, "Relief Request Serial # 09-MN-003" dated June 29, 2009 [Safety Evaluation dated June 28, 2010 (ADAMS Accession Number ML101610306)]
== Conclusion:==
7.2 Calvert Cliffs Nuclear Plant Unit No. 2, Docket No. 50-318, "Revised Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations  
 
-Relief Request (ISI-020 and ISI-021)," dated October 1, 2008 (ADAMS Accession Number ML082760282).
Because the parameters in WCAP-16168-NP Revision 2, Appendix A bound the plant-specific parameters for Catawba Unit 2, the change in risk meets the RG 1.174 acceptance guidelines for a small change in LERF. Increasing the reactor vessel inspection frequency from 10 to 20 years satisfies all the RG 1.174 criteria. For these reasons, the proposed alternative provides an acceptable level of quality and safety.
7.3 Donald C. Cook Nuclear Plant Unit No. 2, Docket No. 50-316, "Request for Relief to Extend the Unit 2 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses," dated October 9, 2008 (ADAMS Accession Number ML082980354).
: 6. Duration of Proposed Alternative This request is applicable until 2024 (plus or minus one outage), which includes the duration of the Catawba Unit 2 Third Inservice Inspection Interval.
7.4 Indian Point Nuclear Generating Units Nos. 2 and 3, Docket Nos. 50-247 and 50-286,"Request for Relief to Extend the Unit 2 and 3 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses," dated July 8, 2008 (ADAMS Accession Number ML081980058).
: 7. Precedents 7.1 Duke Energy Carolina, LLC (Duke), McGuire Nuclear Station Unit 1, Docket No.
: 8. References 8.1 ASME Boiler and Pressure Vessel Code, Section XI, 1998 Edition with the 2000 Addenda, American Society of Mechanical Engineers, New York.8.2 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.'
50-369, "Relief Request Serial # 09-MN-003" dated June 29, 2009 [Safety Evaluation dated June 28, 2010 (ADAMS Accession Number ML101610306)]
MUHP 5097-99, Task 2059," October 31, 2006 (ADAMS Accession Number ML082210245).
7.2 Calvert Cliffs Nuclear Plant Unit No. 2, Docket No. 50-318, "Revised Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations - Relief Request (ISI-020 and ISI-021)," dated October 1, 2008 (ADAMS Accession Number ML082760282).
8.3 OG-10-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-1 6168-NP, Revision 1,'Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval.'
7.3 Donald C. Cook Nuclear Plant Unit No. 2, Docket No. 50-316, "Request for Relief to Extend the Unit 2 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses,"
PA-MSC-0120," July 12, 2010 Request for Alternative 13-CN-003 Enclosure 2 Page 8 of 8 8.4 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 2002.8.5 WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," June 2008.8.6 NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock," March, 2007 (ADAMS Accession Number ML070860156).
dated October 9, 2008 (ADAMS Accession Number ML082980354).
8.7 NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004 (ADAMS Accession Number ML042880482).
7.4 Indian Point Nuclear Generating Units Nos. 2 and 3, Docket Nos. 50-247 and 50-286, "Request for Relief to Extend the Unit 2 and 3 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses," dated July 8, 2008 (ADAMS Accession Number ML081980058).
8.8 Code of Federal Regulations, 10 CFR Part 50.61a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S.Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 75, No.1, dated January 4, 2010, with corrections in Volume 75, No. 22, dated February 3, 2010, Volume 75, No. 44, dated March 8, 2010, and Volume 75, No. 227, dated November 26, 2010.8.9 Revised Final Safety Evaluation by the Office Of Nuclear Reactor Regulation Topical Report WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension Of The Reactor Vessel In-Service Inspection Interval", July 26, 2011 (ADAMS Accession Number ML111600303).
: 8. References 8.1   ASME Boiler and Pressure Vessel Code, Section XI, 1998 Edition with the 2000 Addenda, American Society of Mechanical Engineers, New York.
8.10 Duke Energy Corporation Calculation  
8.2   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 (ADAMS Accession Number ML082210245).
#DPC-1201.00-00-0010 (Catawba Calculation
8.3   OG-10-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-1 6168-NP, Revision 1,
#CNC-1201.01-00-0068), "Implementation of WCAP-16168-NP-A, Revision 2 for Catawba Units 1 and 2 and McGuire Units 1 and 2", Revision 3.}}
          'Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval.'
PA-MSC-0120," July 12, 2010
 
Request for Alternative 13-CN-003 Enclosure 2 Page 8 of 8 8.4 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.
8.5 WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," June 2008.
8.6 NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock,"
March, 2007 (ADAMS Accession Number ML070860156).
8.7 NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004 (ADAMS Accession Number ML042880482).
8.8 Code of Federal Regulations, 10 CFR Part 50.61a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S.
Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 75, No.
1, dated January 4, 2010, with corrections in Volume 75, No. 22, dated February 3, 2010, Volume 75, No. 44, dated March 8, 2010, and Volume 75, No. 227, dated November 26, 2010.
8.9 Revised Final Safety Evaluation by the Office Of Nuclear Reactor Regulation Topical Report WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension Of The Reactor Vessel In-Service Inspection Interval", July 26, 2011 (ADAMS Accession Number ML111600303).
8.10 Duke Energy Corporation Calculation #DPC-1201.00-00-0010 (Catawba Calculation
    #CNC-1201.01-00-0068), "Implementation of WCAP-16168-NP-A, Revision 2 for Catawba Units 1 and 2 and McGuire Units 1 and 2", Revision 3.}}

Latest revision as of 17:36, 4 November 2019

Relief Request Serial Number 13-CN-003, Request for Alternative to the Requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D (Inspection Program B) for Reactor Pressure Vessel (RPV) Welds
ML13148A310
Person / Time
Site: Catawba  Duke Energy icon.png
Issue date: 05/20/2013
From: Henderson K
Duke Energy Carolinas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML13148A310 (17)
v • d • e


Text

DUKE ENERGY. Kelvin ViceHenderson President Catawba Nuclear Station 803-701-4251 Duke Energy CNO1VP 1 4800 Concord Rd.

York, SC 29745 May 20, 2013 10 CFR 50.55a U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

Subject:

Duke Energy Carolinas, LLC (Duke Energy)

Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Relief Request Serial Number 13-CN-003, Request for Alternative to the Requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D (Inspection Program B) for Reactor Pressure Vessel (RPV) Welds Pursuant to 10 CFR 50.55a(a)(3)(i), Duke Energy hereby submits Relief Request 13-CN-003 requesting an alternative to the requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D (Inspection Program B) to perform a volumetric examination of specified RPV welds once during each ten-year interval. Extending the examination frequency to twenty years will reduce the radiological exposure associated with these examinations.

The basis for the proposed alternative is provided in the enclosures to this letter. Duke Energy requests approval of this alternative no later than March 31, 2014. The subject examinations would need to be performed in the Unit 1 End-of-Cycle 21 Refueling Outage, which is scheduled to begin in the Spring of 2014 and is the last refueling outage in the third inservice inspection interval.

There are no regulatory commitments contained in this letter or its enclosures.

If you have any questions or require additional information, please contact L. J. Rudy at (803) 701-3084.

Very truly yours, Kelvin Henderson LJR/s E nclosures U 4 _7 www.duke-energy.com

U.S. Nuclear Regulatory Commission May 20, 2013 Page 2 xc (with enclosures):

V.M. McCree Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 G.A. Hutto, III Senior Resident Inspector U.S. Nuclear Regulatory Commission Catawba Nuclear Station J.S. Kim (addressee only)

NRC Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 8 C2 11555 Rockville Pike Rockville, MD 20852-2738

Enclosure 1 Request for Alternative 13-CN-003 Catawba Nuclear Station Unit 1

Request for Alternative 13-CN-003 Enclosure 1 Page 2 of 7 Catawba Nuclear Station Unit I Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Component Affected Catawba Unit 1 Reactor Pressure Vessel (RPV).

2. Applicable Code Edition and Addenda

The ASME Code Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components," Code 1998 Edition with the 2000 Addenda, is applicable for the Unit 1 Third Inservice Inspection Interval, which started on June 29, 2005 and is scheduled to end on July 15, 20141.

3. Applicable Code Requirement

The proposed alternative is requested in lieu of the requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D (Inspection Program B) to perform a volumetric examination of the specific items listed in the table below once during each 10-year interval.

Examination Category Item No. Description B-A B1.11 Circumferential Shell Welds B-A 81.21 Circumferential Head Welds B-A 81.22 Meridional Head Welds B-A 81.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius (IR) Sections

4. Reason for Request

The proposed alternative is requested to eliminate the radiological exposure associated with performing these examinations every ten years. Extending the examination frequency to twenty years will reduce the radiological exposure.

5. Proposed Alternative and Basis for Use Duke Energy proposes to defer the ASME Code required volumetric examination of the Catawba Unit 1 Category B-A and B-D reactor vessel full penetration pressure retaining welds during the third inservice inspection interval until 2024 (plus or minus one outage). Volumetric examinations shall be performed in accordance with requirements of 10 CFR 50.55a and the IThe interval end date has been adjusted in accordance with IWA-2430(d)(1).

Request for Alternative 13-CN-003 Enclosure 1 Page 3 of 7 Edition and Addenda of the ASME Code,Section XI, applicable at the time of the next scheduled examination.

In accordance with 10 CFR 50.55a(a)(3)(i), the alternative is requested on the basis that it provides an acceptable level of quality and safety because there is a negligible change in risk, as measured by evaluating the risk criteria specified in Regulatory Guide 1.174 (Reference 8.4), and as documented herein.

The methodology used to demonstrate the acceptability of extending the inspection interval for Category B-A and B-D welds is based on a negligible change in risk, and is documented in WCAP-16168-NP-A, Revision 2 (Reference 8.5). 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 8.6). 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 Catawba Unit 1 reactor vessel is acceptable as shown in Table 1 below.

Table I Critical Parameters for Application of Bounding Analysis Unit 1 Bounded Unit I by Pilot Plant Parameter Pilot Plant Basis Plant-Specific Basis Basis?

Dominant Pressurized NRC PTS Risk Study PTS Generalization Yes (No Further Thermal Shock (PTS) (Reference 8.6) Study (References 8.7 Evaluation Transients in the NRC PTS and 8.10) Required)

Risk Study are Applicable Through-Wall Cracking 1.76E-08 Events per Year 1.11 E-14 Events per Yes (No Further Frequency (Reference 8.5) Year (References 8.5 Evaluation and 8.10) Required)

Frequency and Severity of 7 Heatup/Cooldowns per Bounded by 7 Yes (No Further Design Basis Transients Year (Reference 8.5) Heatup/Cooldowns per Evaluation year Required)

Cladding Layers Single Layer (Reference 8.5) Multi-Layer Yes (No Further (Single/Multiple) Evaluation Required)

Request for Alternative 13-CN-003 Enclosure 1 Page 4 of 7 Additional information relative to the Catawba Unit 1 reactor vessel inspection is provided in Table 2. This information confirms that satisfactory examinations have been performed on the Catawba Unit 1 reactor vessel.

Table 2 Additional Information Pertaining to Reactor Vessel Inspection Inspection Methodology: The most recent inservice inspection of the Category B-A and B-D welds was performed in accordance with ASME Section XI Appendix VIII requirements. Code case N-613-1 was used in lieu of the Section Xl requirements for inspection of the Category B-D nozzle to shell welds. Code case N-648-1 visual examinations were performed in lieu of the Section Xl Category B-D volumetric examinations of the inner radii.

Number of Past Inspections: Two 10-Year inservice inspections have been performed.

Number of Indications Found: One indication was identified in the beltline region during the most recent inservice inspection. This indication was acceptable per Table IWB-3510-1 of Section XI of the ASME Code. This indication is not within 1 / 1 0 th or 1"of the reactor vessel shell plate inside diameter surface. The lack of any indications near the inner surface meets the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 8.8).

Proposed Inspection Schedule for The welds for which the proposed alternative is requested shall be Balance of Plant Life: examined no later than 2024 (plus or minus one refueling outage).

Subsequent to 2024, these welds shall be examined in accordance with the applicable requirements of the ASME Code, Section Xl, as required by 10 CFR 50.55a.

Request for Alternative 13-CN-003 Enclosure 1 Page 5 of 7 Table 3 provides additional information relative to the calculation of the Through Wall Cracking Frequency (TWCF) for Catawba Unit 1.

Table 3 Details of TWCF Calculation - Performed for 60 Effective Full Power Years (EFPY)

Inputs Reactor Coolant System Temperature, TRcs[°F]: N/A Thickness of the RPV wall (including cladding),

Twa1 l [inches]: 8.62 Fluence [1019 2

No. Region/Component Description Material/

Flux Cu Type [wt%] Ni

[wt%] R.G. 1.99 Position C.F.

[OF] Un-Irradiated RTNDT [OF] E > 1.0 MeV] ,

Neutron/cm 1 Intermediate Shell Forging A 508-2 0.090 0.860 2.1 28.4 -8 3.06 2 Lower Shell Forging A 508-2 0.040 0.830 1.1 26.0 -13 3.06 3 Intermediate-to-Lower GRAU LO 0.040 0.720 2.1 23.2 -51 3.06 Shell Forging Circumferential Weld 4 Nozzle Shell Forging A 508-2 0.25 0.85 1.1 201.25 -26 0.142 5 Nozzle-to-Intermediate GRAU LO 0.03 0.75 1.1 41 0 0.142 Shell Forging Circumferential Weld Outputs Methodology Used to Calculate AT30: Regulatory Guide 1.99, Revision 2 Controlling Material Fluence [1019 FF (Fluence T30 Region No. RTMAx-xx Neutron /cm2, Factor) A0 TWCF 9 5 -XX (From Above) [R] E > 1.0 MeV] Factor) [OF]

Circumferential Weld - CW 4 532.44 0.142 0.491 98.75 0.00 Forging- FO 4 532.44 0.142 0.491 98.75 4.45E-15 TWCF95-TOTAL (ctcwTWcF95-CW + .FoTMWCF 9 5 -FO): 1.11E-14 Failure Consequences:

The failure of any of the Reactor Pressure Vessel shell, head, or nozzle welds listed in this request would result in a loss of the structural integrity of the Reactor Vessel.

==

Conclusion:==

Because the parameters in WCAP-16168-NP Revision 2, Appendix A bound the plant-specific parameters for Catawba Unit 1, the change in risk meets the RG 1.174 acceptance guidelines for a small change in Large Early Release Frequency (LERF).

Increasing the reactor vessel inspection frequency from 10 to 20 years satisfies all the

Request for Alternative 13-CN-003 Enclosure 1 Page 6 of 7 RG 1.174 criteria. For these reasons, the proposed alternative provides an acceptable level of quality and safety.

6. Duration of Proposed Alternative This request is applicable until 2024 (plus or minus one outage), which includes the duration of the Catawba Unit 1 Third Inservice Inspection Interval.
7. Precedents 7.1 Duke Energy Carolina, LLC (Duke), McGuire Nuclear Station Unit 1, Docket No.

50-369, "Relief Request Serial # 09-MN-003" dated June 29, 2009 [Safety Evaluation dated June 28, 2010 (ADAMS Accession Number ML101610306)]

7.2 Calvert Cliffs Nuclear Plant Unit No. 2, Docket No. 50-318, "Revised Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations - Relief Request (ISI-020 and ISI-021)," dated October 1, 2008 (ADAMS Accession Number ML082760282).

7.3 Donald C. Cook Nuclear Plant Unit No. 2, Docket No. 50-316, "Request for Relief to Extend the Unit 2 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses,"

dated October 9, 2008 (ADAMS Accession Number ML082980354).

7.4 Indian Point Nuclear Generating Units Nos. 2 and 3, Docket Nos. 50-247 and 50-286, "Request for Relief to Extend the Unit 2 and 3 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses," dated July 8, 2008 (ADAMS Accession Number ML081980058).

8. References 8.1 ASME Boiler and Pressure Vessel Code,Section XI, 1998 Edition with the 2000 Addenda, American Society of Mechanical Engineers, New York.

8.2 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 (ADAMS Accession Number ML082210245).

8.3 OG-10-238, "Revision to the Revised 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.'

PA-MSC-0120," July 12, 2010 8.4 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.

8.5 WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," June 2008.

8.6 NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock,"

March, 2007 (ADAMS Accession Number ML070860156).

Request for Alternative 13-CN-003 Enclosure 1 Page 7 of 7 8.7 NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004 (ADAMS Accession Number ML042880482).

8.8 Code of Federal Regulations, 10 CFR Part 50.61a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S.

Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 75, No.

1, dated January 4, 2010, with corrections in Volume 75, No. 22, dated February 3, 2010, Volume 75, No. 44, dated March 8, 2010, and Volume 75, No. 227, dated November 26, 2010.

8.9 Revised Final Safety Evaluation by the Office Of Nuclear Reactor Regulation Topical Report WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension Of The Reactor Vessel In-Service Inspection Interval", July 26, 2011 (ADAMS Accession Number ML111600303).

8.10 Duke Energy Corporation Calculation #DPC-1201.00-00-0010 (Catawba Calculation

  1. CNC-1201.01-00-0068), "Implementation of WCAP-16168-NP-A, Revision 2 for Catawba Units 1 and 2 and McGuire Units 1 and 2", Revision 3.

Enclosure 2 Request for Alternative 13-CN-003 Catawba Nuclear Station Unit 2

Request for Alternative 13-CN-003 Enclosure 2 Page 2 of 8 Catawba Nuclear Station Unit 2 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Component(s) Affected Catawba Unit 2 Reactor Pressure Vessel.

2. Applicable Code Edition and Addenda

The ASME Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Code 1998 Edition with the 2000 Addenda, is applicable for the Unit 2 Third Inservice Inspection Interval, which started on October 15, 2005, and is scheduled to end on August 19, 2016.

3. Applicable Code Requirement

The proposed alternative is requested in lieu of the requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D (Inspection Program B) to perform a volumetric examination of the specific items listed in the table below once during each 10-year interval.

Examination Category Item No. Description B-A B1.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B13.21 Circumferential Head Welds B-A B13.22 Meridional Head Welds B-A B13.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius (IR)Sections

4. Reason for Request

The proposed alternative is requested to eliminate the radiological exposure associated with performing these examinations every ten years. Extending the examination frequency to twenty years will reduce the radiological exposure.

5. Proposed Alternative and Basis for Use Duke Energy proposes to defer the ASME Code required volumetric examination of the Catawba Unit 2 Category B-A and B-D reactor vessel full penetration pressure retaining welds during the third inservice inspection interval until 2024 (plus or minus one outage). Volumetric I The interval start date was adjusted in accordance with IWA-2430(d)(1).

Request for Alternative 13-CN-003 Enclosure 2 Page 3 of 8 examinations shall be performed in accordance with requirements of 10 CFR 50.55a and the Edition and Addenda of the ASME Code,Section XI, applicable at the time of the next scheduled examination.

In accordance with 10 CFR 50.55a(a)(3)(i), the alternative is requested on the basis that it provides an acceptable level of quality and safety because there is a negligible change in risk, as measured by evaluating the risk criteria specified in Regulatory Guide 1.174 (Reference 8.4), and as documented herein.

The methodology used to demonstrate the acceptability of extending the inspection interval for Category B-A and B-D welds is based on a negligible change in risk, and is documented in WCAP-16168-NP-A, Revision 2 (Reference 8.5). 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 8.7). 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 Catawba Unit 2 reactor vessel is acceptable as shown in Table 1 below.

Table 1 Critical Parameters for Application of Bounding Analysis Unit 2 Bounded Unit 2 by Pilot Plant Parameter Pilot Plant Basis Plant-Specific Basis Basis?

Dominant Pressurized NRC PTS Risk Study PTS Generalization Study Yes (No Further Thermal Shock (PTS) (Reference 8.6) (References 8.7 and 8.10) Evaluation Transients in the NRC PTS Required)

Risk Study are Applicable Through-Wall Cracking 1.76E-08 Events per Year 1.40E-1 3 Events per Year Yes (No Further Frequency (Reference 8.5) (References 8.5 and 8.10) Evaluation Required)

Frequency and Severity of 7 Heatup/Cooldowns per Bounded by 7 Yes (No Further Design Basis Transients Year (Reference 8.5) Heatup/Cooldowns per Evaluation Year Required)

Cladding Layers Single Layer (Reference 8.5) Single Layer (Reference Yes (No Further (Single/Multiple) 8.10) Evaluation Required)

Request for Alternative 13-CN-003 Enclosure 2 Page 4 of 8 Additional information relative to the Catawba Unit 2 reactor vessel inspection is provided in Table 2. This information confirms that satisfactory examinations have been performed on the Catawba Unit 2 reactor vessel.

Table 2 Additional Information Pertaining to Reactor Vessel Inspection Inspection Methodology: The most recent inservice inspection of the Category B-A and B-D welds was performed in accordance with ASME Section XI Appendix VIII requirements.

Code case N-613-1 was used in lieu of the Section Xl requirements for inspection of the Category B-D nozzle to shell welds. Code case N-648-1 visual examinations were performed in lieu of the Section XI Category B-D volumetric examinations of the inner radii.

Number of Past Inspections: Two 10-Year inservice inspections have been performed.

Number of Indications Six indications, which were identified inthe beltline region during the most recent Found: inservice inspection, were acceptable per IWB-3510-1. Only two of these indications (both within the weld material in Lower Axial Weld #101-142A) were within the greater of 1/10th or 1"of the reactor vessel shell plate inside diameter surface and were acceptable per the requirements of the Alternate PTS Rule 10 CFR 50.61 a (Reference 8.8). A summary of these indications is provided in Table 3. The locations of these two indications are shown in Figure 1.

Proposed Inspection The welds for which the proposed alternative is requested shall be examined no Schedule for Balance of later than 2024 (plus or minus one refueling outage). Subsequent to 2024, these Plant Life: welds shall be examined inaccordance with the applicable requirements of the ASME Code, Section Xl, as required by 10 CFR 50.55a.

Table 3 Summary of Reactor Vessel Weld Inspection Indications Through-Wall Extent, TWE (in.)

Scaled Maximum2 Number of Flaws Number of Flaws TWEMIN TWEm, 0.075 0.475 162 2 (1 Axial and 1 Circumferential) 0.125 0.475 88 2 (1 Axial and 1 Circumferential) 0.175 0.475 23 2 (1 Axial and 1 Circumferential) 0.225 0.475 9 2 (1 Axial and 1 Circumferential) 0.275 0.475 4 1 (1 Axial and 0 Circumferential) 0.325 0.475 3 1 (1 Axial and 0 Circumferential) 2 The "Scaled Maximum Number of Flaws" indicates the number allowed per the Alternate PTS rule, 10 CFR 50.61a (Reference 8.8). This number is based on the length of weld inspected in the beltline region.

Request for Alternative 13-CN-003 Enclosure 2 Page 5 of 8 Table 4 provides additional information relative to the calculation of the Through Wall Cracking Frequency (TWCF) for Catawba Unit 2.

Table 4 Details of TWCF Calculation - Performed for 60 Effective Full Power Years (EFPY)

Inputs Reactor Coolant System Temperature, TRCS[*F]: N/A Thickness of the RPV wall (including cladding),

Twa11 [inches]: 8.78 Fluence [1019 No. Region/Component Material/ Flu] Cu Ni R.G. 1.99 C.F. Un-Irradiated Neutron/cm 2, Description Type [wt%] [wt%] Position [OF] RTNDT [*F] E > 1.0 MeV]

1 Intermediate Plate B8616-1 A 533B 0.050 0.600 1.1 31.0 12 3.03 2 Intermediate Plate B8605-1 A 533B 0.080 0.620 2.1 44.0 15 3.03 3 Intermediate Plate B8605-2 A 533B 0.080 0.610 1.1 51.0 33 3.03 4 Lower Plate B8806-2 A 533B 0.060 0.590 1.1 37.0 -10 3.03 5 Lower Plate B8806-3 A 533B 0.060 0.590 1.1 37.0 8 3.03 6 Lower Plate B8806-1 A 533B 0.060 0.560 1.1 37.0 6 3.03 7 Intermediate Axial Weld Linde 0091 0.040 0.140 2.1 33.4 -80 1.76 101-124A 8 Intermediate Axial Weld Linde 0091 0.040 0.140 2.1 33.4 -80 2.89 101-124B 9 Intermediate Axial Weld Linde 0091 0.040 0.140 2.1 33.4 -80 2.89 101-124C 10 Lower Axial Weld 101-142A Linde 0091 0.040 0.140 2.1 33.4 -80 2.89 11 Lower Axial. Weld 101-142B Linde 0091 0.040 0.140 2.1 33.4 -80 1.76 12 Lower Axial Weld 101-142C Linde 0091 0.040 0.140 2.1 33.4 -80 2.89 13 Intermediate/Lower Linde 0091 0.040 0.140 2.1 33.4 -80 3.03 Circumferential Weld 101-171 14 Nozzle Plate B8604-1 A 533B 0.11 0.61 1.1 74.15 24 0.0670 15 Nozzle Plate B8604-2 A 533B 0.11 0.61 1.1 74.15 26 0.0670 16 Nozzle Plate B8604-3 A 533B 0.07 0.53 1.1 44 50 0.0670 17 Nozzle Axial Weld 101-122A Linde 0091 0.156 0.059 1.1 73.71 -50 0.0637 18 Nozzle Axial Weld 101-122B Linde 0091 0.156 0.059 1.1 73.71 -50 0.0670 19 Nozzle Axial Weld 101-122C Linde 0091 0.156 0.059 1.1 73.71 -50 0.0572 20 Intermediate/Nozzle Linde 0091 0.153 0.077 1.1 74.13 -40 0.0670 Circumferential Weld 103-121

Request for Alternative 13-CN-003 Enclosure 2 Page 6 of 8 Table 4 Details of TWCF Calculation - Performed for 60 Effective Full Power Years (EFPY)

Outputs Methodology Used to Calculate AT3o: Reculatory Guide 1.99, Revision 2 Controlling Fluence [10 19 FF Material Region 2T3 No. RTMAx.xx [R] Neutron/cm2, (Fluence AT30 TWCF9 5sxx (From Above) E > 1.0 MeV] Factor) [OF]

Axial Weld - AW 3 558.04 2.89 1.282 65.37 0.00 Circumferential Weld - 3 558.64 3.03 1.293 65.95 0.00 CW Plate - PL 3 558.64 3.03 1.293 65.95 5.61 E-14 TWCF95-TOTAL (aAwTWCF95-AW + aPLTWCF95.PL + acwTWCF 95.cw): 1.40E-13 Figure 1 Catawba Unit 2 Reactor Vessel Beltline Indication Map (Figure is not to scale. Indication location is approximate.)

Request for Alternative 13-CN-003 Enclosure 2 Page 7 of 8 Failure Consequences:

The failure of any of the Reactor Pressure Vessel shell, head, or nozzle welds listed in this request would result in a loss of the structural integrity of the Reactor Vessel.

Conclusion:

Because the parameters in WCAP-16168-NP Revision 2, Appendix A bound the plant-specific parameters for Catawba Unit 2, the change in risk meets the RG 1.174 acceptance guidelines for a small change in LERF. Increasing the reactor vessel inspection frequency from 10 to 20 years satisfies all the RG 1.174 criteria. For these reasons, the proposed alternative provides an acceptable level of quality and safety.

6. Duration of Proposed Alternative This request is applicable until 2024 (plus or minus one outage), which includes the duration of the Catawba Unit 2 Third Inservice Inspection Interval.
7. Precedents 7.1 Duke Energy Carolina, LLC (Duke), McGuire Nuclear Station Unit 1, Docket No.

50-369, "Relief Request Serial # 09-MN-003" dated June 29, 2009 [Safety Evaluation dated June 28, 2010 (ADAMS Accession Number ML101610306)]

7.2 Calvert Cliffs Nuclear Plant Unit No. 2, Docket No. 50-318, "Revised Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations - Relief Request (ISI-020 and ISI-021)," dated October 1, 2008 (ADAMS Accession Number ML082760282).

7.3 Donald C. Cook Nuclear Plant Unit No. 2, Docket No. 50-316, "Request for Relief to Extend the Unit 2 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses,"

dated October 9, 2008 (ADAMS Accession Number ML082980354).

7.4 Indian Point Nuclear Generating Units Nos. 2 and 3, Docket Nos. 50-247 and 50-286, "Request for Relief to Extend the Unit 2 and 3 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses," dated July 8, 2008 (ADAMS Accession Number ML081980058).

8. References 8.1 ASME Boiler and Pressure Vessel Code,Section XI, 1998 Edition with the 2000 Addenda, American Society of Mechanical Engineers, New York.

8.2 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 (ADAMS Accession Number ML082210245).

8.3 OG-10-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-1 6168-NP, Revision 1,

'Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval.'

PA-MSC-0120," July 12, 2010

Request for Alternative 13-CN-003 Enclosure 2 Page 8 of 8 8.4 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.

8.5 WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," June 2008.

8.6 NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock,"

March, 2007 (ADAMS Accession Number ML070860156).

8.7 NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004 (ADAMS Accession Number ML042880482).

8.8 Code of Federal Regulations, 10 CFR Part 50.61a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S.

Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 75, No.

1, dated January 4, 2010, with corrections in Volume 75, No. 22, dated February 3, 2010, Volume 75, No. 44, dated March 8, 2010, and Volume 75, No. 227, dated November 26, 2010.

8.9 Revised Final Safety Evaluation by the Office Of Nuclear Reactor Regulation Topical Report WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension Of The Reactor Vessel In-Service Inspection Interval", July 26, 2011 (ADAMS Accession Number ML111600303).

8.10 Duke Energy Corporation Calculation #DPC-1201.00-00-0010 (Catawba Calculation

  1. CNC-1201.01-00-0068), "Implementation of WCAP-16168-NP-A, Revision 2 for Catawba Units 1 and 2 and McGuire Units 1 and 2", Revision 3.
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