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| issue date = 02/18/2009
| issue date = 02/18/2009
| title = Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations - Relief Requests (ISI-022 and ISI-023)
| title = Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations - Relief Requests (ISI-022 and ISI-023)
| author name = Flaherty M D
| author name = Flaherty M
| author affiliation = Constellation Energy Group
| author affiliation = Constellation Energy Group
| addressee name =  
| addressee name =  
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==REFERENCES:==
==REFERENCES:==
(a) Westinghouse Owners Group Topical Report, WCAP-16168-NP-A, Revision 2, dated June 2008, Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval (b) Letter from H. K. Nieh (NRC) to G. Bischoff (PWROG), dated May 8, 2008, Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR) WCAP-16168-NP, Revision 2, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (TAC No. MC9768)(c) Letter from J. A. Spina (CCNPP) to Document Control Desk (NRC), dated October 1, 2008, License Amendment Request: Reporting of Reactor Vessel In-Service Inspection Information and Analyses in Support of Code Relief Request for Extension of Reactor Vessel In-Service Inspection Interval (d) Letter from M. D. Flaherty (CCNPP) to Document Control Desk (NRC), dated October 1, 2008, Revised Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld and Visual Examinations-Relief Requests (ISI-020 and ISI-021)In Westinghouse Topical Report WCAP-16168-NP-A, Revision 2 (Reference a), the Pressurized Water Reactor Owners Group provided the technical and regulatory basis for decreasing the frequency of inspections by extending the American Society of Mechanical Engineers (ASME) Code Section XI inservice inspection interval from the current 10 years to 20 years for ASME Code Section XI examination categories B-A and B-D reactor pressure vessel (RPV) welds. The Nuclear Regulatory Commission approved the topical report by letter dated May 8, 2008 (Reference b). To implement the change presented in Reference (a) for Unit 1, Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) hereby submits Attachment (1) (ISI-022) in accordance with the Safety Evaluation (Reference b), to request an alternative from the Code requirements pursuant to 10 CFR 50.55a(a)(3)(i) on the basis that the alternative inspection interval (20 years) provides an acceptable level of quality and safety. The plant Document Control Desk February 18, 2009 Page 2 specific information identified by Reference (a) as needed to support this proposed alternative for Unit 1 is provided in Attachment (1). The future inspection of examination categories B-A and B-D RPV welds will be performed in 2018 for Unit 1. Additionally, as required by Reference (b), a license amendment request (Reference c) (submitted for both Unit I and Unit 2) was previously submitted in support of a similar relief request for Calvert Cliffs Unit 2 (Reference d).Further, pursuant to 10 CFR 50.55a(a)(3)(ii), Calvert Cliffs is also requesting approval to delay compliance with the specified requirements of ASME Section XI, Table IWB-2500-1, for Calvert Cliffs Unit 1. Relief is requested to defer visual examinations under ASME Section XI Table IWB-2500-1 examination categories B-N-2 and B-N-3, Item Nos. B13.50, B13.60, and B13.70 for the extended inspection interval from 10 to 20 years as described in Attachment (2) (ISI-023).
(a) Westinghouse Owners Group Topical Report, WCAP-16168-NP-A, Revision 2, dated June 2008, Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval (b) Letter from H. K. Nieh (NRC) to G. Bischoff (PWROG), dated May 8, 2008, Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR) WCAP-16168-NP, Revision 2, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (TAC No. MC9768)(c) Letter from J. A. Spina (CCNPP) to Document Control Desk (NRC), dated October 1, 2008, License Amendment Request: Reporting of Reactor Vessel In-Service Inspection Information and Analyses in Support of Code Relief Request for Extension of Reactor Vessel In-Service Inspection Interval (d) Letter from M. D. Flaherty (CCNPP) to Document Control Desk (NRC), dated October 1, 2008, Revised Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld and Visual Examinations-Relief Requests (ISI-020 and ISI-021)In Westinghouse Topical Report WCAP-16168-NP-A, Revision 2 (Reference a), the Pressurized Water Reactor Owners Group provided the technical and regulatory basis for decreasing the frequency of inspections by extending the American Society of Mechanical Engineers (ASME) Code Section XI inservice inspection interval from the current 10 years to 20 years for ASME Code Section XI examination categories B-A and B-D reactor pressure vessel (RPV) welds. The Nuclear Regulatory Commission approved the topical report by letter dated May 8, 2008 (Reference b). To implement the change presented in Reference (a) for Unit 1, Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) hereby submits Attachment (1) (ISI-022) in accordance with the Safety Evaluation (Reference b), to request an alternative from the Code requirements pursuant to 10 CFR 50.55a(a)(3)(i) on the basis that the alternative inspection interval (20 years) provides an acceptable level of quality and safety. The plant Document Control Desk February 18, 2009 Page 2 specific information identified by Reference (a) as needed to support this proposed alternative for Unit 1 is provided in Attachment (1). The future inspection of examination categories B-A and B-D RPV welds will be performed in 2018 for Unit 1. Additionally, as required by Reference (b), a license amendment request (Reference c) (submitted for both Unit I and Unit 2) was previously submitted in support of a similar relief request for Calvert Cliffs Unit 2 (Reference d).Further, pursuant to 10 CFR 50.55a(a)(3)(ii), Calvert Cliffs is also requesting approval to delay compliance with the specified requirements of ASME Section XI, Table IWB-2500-1, for Calvert Cliffs Unit 1. Relief is requested to defer visual examinations under ASME Section XI Table IWB-2500-1 examination categories B-N-2 and B-N-3, Item Nos. B13.50, B13.60, and B13.70 for the extended inspection interval from 10 to 20 years as described in Attachment (2) (ISI-023).
Deferring the subject visual examinations associated with the reactor vessel core support structure removal to the extended inspection interval allows those examinations to be performed at the same time (2018 refueling outage) as the examination categories B-A and B-D RPV welds described in Attachment (1).Calvert Cliffs' proposed extension of the inservice inspection interval for these examinations will continue to provide an acceptable level of quality and safety, as described in the attached relief requests.Calvert Cliffs requests the Nuclear Regulatory Commission staff approval by October 15, 2009, in order to support preparation activities for the spring 2010 Unit 1 refueling outage.Should you have questions regarding this matter, please contact Mr. Jay S. Gaines at (410) 495-5219.Very truly yours, Mark D. Flaherty Manager-Engineering Services MDF/KLG/bjd Attachments:  
Deferring the subject visual examinations associated with the reactor vessel core support structure removal to the extended inspection interval allows those examinations to be performed at the same time (2018 refueling outage) as the examination categories B-A and B-D RPV welds described in Attachment (1).Calvert Cliffs' proposed extension of the inservice inspection interval for these examinations will continue to provide an acceptable level of quality and safety, as described in the attached relief requests.Calvert Cliffs requests the Nuclear Regulatory Commission staff approval by October 15, 2009, in order to support preparation activities for the spring 2010 Unit 1 refueling outage.Should you have questions regarding this matter, please contact Mr. Jay S. Gaines at (410) 495-5219.Very truly yours, Mark D. Flaherty Manager-Engineering Services MDF/KLG/bjd Attachments:
(1) Proposed Alternative ISI-022 (2) Proposed Alternative ISI-023 cc: D. V. Pickett, NRC S. Gray, DNR S. J. Collins, NRC Safety Program Manager, State of Maryland Resident Inspector, NRC ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Calvert Cliffs Nuclear Power Plant, Inc.February 18, 2009 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 American Society of Mechanical Engineers (ASME) Code Component(s)
(1) Proposed Alternative ISI-022 (2) Proposed Alternative ISI-023 cc: D. V. Pickett, NRC S. Gray, DNR S. J. Collins, NRC Safety Program Manager, State of Maryland Resident Inspector, NRC ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Calvert Cliffs Nuclear Power Plant, Inc.February 18, 2009 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 American Society of Mechanical Engineers (ASME) Code Component(s)
Affected The affected component is the Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) Unit 1 reactor vessel, specifically the following ASME Boiler and Pressure Vessel (BPV) Code Section XI (Reference 1)examination categories and item numbers covering examinations of the reactor vessel. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code Section XI.Examination Category Item No. Description B-A B 1.11 Circumferential Shell Welds B-A B 1.12 Longitudinal Shell Welds B-A B 1.21 Circumferential Head Welds B-A B 1.22 Meridional Shell Welds B-A B 1.30 Shell-to-Flange Weld B-A B 1.40 Head-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inner Radius Areas (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.")Applicable Code Edition and Addenda American Society of Mechanical Engineers Code Section XI, "Rules and Inservice Inspection of Nuclear Power Plant Components," 1998 Edition (no Addenda) for the third interval and 2004 Edition (no Addenda) for the fourth interval.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.
Affected The affected component is the Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) Unit 1 reactor vessel, specifically the following ASME Boiler and Pressure Vessel (BPV) Code Section XI (Reference 1)examination categories and item numbers covering examinations of the reactor vessel. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code Section XI.Examination Category Item No. Description B-A B 1.11 Circumferential Shell Welds B-A B 1.12 Longitudinal Shell Welds B-A B 1.21 Circumferential Head Welds B-A B 1.22 Meridional Shell Welds B-A B 1.30 Shell-to-Flange Weld B-A B 1.40 Head-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inner Radius Areas (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.")Applicable Code Edition and Addenda American Society of Mechanical Engineers Code Section XI, "Rules and Inservice Inspection of Nuclear Power Plant Components," 1998 Edition (no Addenda) for the third interval and 2004 Edition (no Addenda) for the fourth interval.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.
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Reason for Request An alternative is requested from the requirement of IWA-2412, Inspection Program B, that volumetric examination of essentially 100% of reactor vessel pressure retaining, Examination Category B-A and B-D welds, be performed once each ten-year interval.
Reason for Request An alternative is requested from the requirement of IWA-2412, Inspection Program B, that volumetric examination of essentially 100% of reactor vessel pressure retaining, Examination Category 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 personnel radiation exposure and examination costs.Proposed Alternative and Basis for Use Calvert Cliffs proposes to defer the ASME Code required volumetric examination of the Unit 1 reactor vessel full penetration pressure retaining Examination Category B-A and B-D welds for the third inservice inspection until 2018 and to perform the fourth inservice inspection of these welds on a 20-year inspection interval, instead of the current ten-year inspection 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 personnel radiation exposure and examination costs.Proposed Alternative and Basis for Use Calvert Cliffs proposes to defer the ASME Code required volumetric examination of the Unit 1 reactor vessel full penetration pressure retaining Examination Category B-A and B-D welds for the third inservice inspection until 2018 and to perform the fourth inservice inspection of these welds on a 20-year inspection interval, instead of the current ten-year inspection interval.
Therefore, the fourth inservice inspection is proposed to be performed in 2038 if Unit 1 requests and obtains an additional license renewal extension beyond the current license expiration date of July 31, 2034. These dates are consistent with the information provided to the Staff in Pressurized Water Reactor Owners Group letter OG-06-356 (Reference 2).I ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 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 third and fourth inspection intervals for Examination 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 Nuclear Regulatory Commission (NRC)Pressurized Thermal Shock (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 Calvert Cliffs Unit 1 reactor vessel is acceptable as shown in Table 1 below.Table 1. Critical Parameters for Application of Bounding Analysis Additional Parameter Pilot Plant Basis Plant Specific Evaluation Required?Dominant PTS Transients in the NRC PTS Risk Study PTS Generalization No NRC PTS Risk Study are (Reference  
Therefore, the fourth inservice inspection is proposed to be performed in 2038 if Unit 1 requests and obtains an additional license renewal extension beyond the current license expiration date of July 31, 2034. These dates are consistent with the information provided to the Staff in Pressurized Water Reactor Owners Group letter OG-06-356 (Reference 2).I ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 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 third and fourth inspection intervals for Examination 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 Nuclear Regulatory Commission (NRC)Pressurized Thermal Shock (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 Calvert Cliffs Unit 1 reactor vessel is acceptable as shown in Table 1 below.Table 1. Critical Parameters for Application of Bounding Analysis Additional Parameter Pilot Plant Basis Plant Specific Evaluation Required?Dominant PTS Transients in the NRC PTS Risk Study PTS Generalization No NRC PTS Risk Study are (Reference
: 5) Study (Reference 6)applicable Through-Wall Cracking 3.16E-7 Events per year 1.80E-08 Events per No Frequency (TWCF) (Reference  
: 5) Study (Reference 6)applicable Through-Wall Cracking 3.16E-7 Events per year 1.80E-08 Events per No Frequency (TWCF) (Reference
: 4) year (Calculated per Reference 5)Frequency and Severity of Design 13 heatup/cooldowns per Bounded by 13 No Basis Transients year (Reference  
: 4) year (Calculated per Reference 5)Frequency and Severity of Design 13 heatup/cooldowns per Bounded by 13 No Basis Transients year (Reference
: 4) heatup/cooldowns per year Cladding Layers Single Layer (Reference  
: 4) heatup/cooldowns per year Cladding Layers Single Layer (Reference
: 4) Single Layer No (Single/Multiple)
: 4) Single Layer No (Single/Multiple)
II Additional information relative to the Calvert Cliffs Unit 1 reactor vessel inspection is provided in Table 2. This information confirms that .satisfactory examinations have been performed on the Calvert Cliffs Unit I reactor vessel.2 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Table 2. Additional Information Pertaining to Reactor Vessel Inspection Inspection methodology:
II Additional information relative to the Calvert Cliffs Unit 1 reactor vessel inspection is provided in Table 2. This information confirms that .satisfactory examinations have been performed on the Calvert Cliffs Unit I reactor vessel.2 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Table 2. Additional Information Pertaining to Reactor Vessel Inspection Inspection methodology:
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Table 3 provides additional information relative to the calculation of the* TWCF parameter for Calvert Cliffs Unit 1.3 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Table 3. Details of TWCF Calculation at 60 EFPY Inputs Reactor Coolant System Temperature, TRCS[°F]:
Table 3 provides additional information relative to the calculation of the* TWCF parameter for Calvert Cliffs Unit 1.3 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Table 3. Details of TWCF Calculation at 60 EFPY Inputs Reactor Coolant System Temperature, TRCS[°F]:
1 548 Twal, [inches]:
1 548 Twal, [inches]:
8.625 Un- Fluence [1019 Region/Component Material Cu Ni P Mn] Irradiated Neutron/cm 2 , Description  
8.625 Un- Fluence [1019 Region/Component Material Cu Ni P Mn] Irradiated Neutron/cm 2 , Description
[wt%] [wt%] [wt%] [wt%] RTNDT(u) ['F] E>1 MeV]1 Low. Shell Axial LINDE 1092 .180 .720 .015 1.63 -56 6.36 Weld 2 Low. Shell Axial LINDE 1092 .180 .720 .015 1.63 -56 6.36 Weld 3 Low. Shell Axial LINDE 1092 .180 .720 .015 1.63 -56 6.36 Weld 4 Int. Shell Axial Weld LINDE 1092 .220 .830 .010 1.63 -50 6.36 5 Int. Shell Axial Weld LINDE 1092 .220 .830 .010 1.63 -50 6.36 6 Int. Shell Axial Weld LINDE 1092 .220 .830 .010 1.63 -50 6.36 7 Int./Low.
[wt%] [wt%] [wt%] [wt%] RTNDT(u) ['F] E>1 MeV]1 Low. Shell Axial LINDE 1092 .180 .720 .015 1.63 -56 6.36 Weld 2 Low. Shell Axial LINDE 1092 .180 .720 .015 1.63 -56 6.36 Weld 3 Low. Shell Axial LINDE 1092 .180 .720 .015 1.63 -56 6.36 Weld 4 Int. Shell Axial Weld LINDE 1092 .220 .830 .010 1.63 -50 6.36 5 Int. Shell Axial Weld LINDE 1092 .220 .830 .010 1.63 -50 6.36 6 Int. Shell Axial Weld LINDE 1092 .220 .830 .010 1.63 -50 6.36 7 Int./Low.
Circ Weld LINDE 0091 .240 .160 .014 1.63 -80 6.36 8 Lower Shell Plate A 533B .130 .540 .010 1.45 10 6.36 9 Lower Shell Plate A 533B .120 .550 .009 1.45 -10 6.36 10 Lower Shell Plate A 533B .110 .530 .008 1.45 -20 6.36 11 Inter. Shell Plate A 533B .110 .550 .011 1.45 20 6.36 12 Inter. Shell Plate A 533B .120 .640 .011 1.45 -30 6.36 13 Inter. Shell Plate A 533B .120 '.640 .011 1.45 10 6.36 Outputs Methodology Used to Calculate AT 3 0: NUREG-1874 Controlling Material Fluence [1019 Region # RTMAX-XX Neutron/cm 2 , d (flux) AT 3 0 [°F] TWCF 9 5-XX (From [R] E>1 MeV]Above)Axial Weld- AW 4,5,6 668.88 6.36 3.36E+10 259.19 8.03E-09 Circumferential Weld -CW 13 618.48 6.36 3.36E+10 148.79 2.20E-25 Plate -PL 13 618.48 6.36 3.36E+10 148.79 7.85E-12 TWCF 9 5-TOTAL (aAwTWCF 9 5-AW + CPLTWCF 9 5-PL + acwTWCF 9 5.cw): 1.80E-08 Note 1: The manganese content for the welds and plates used in the calculation of TWCF was assumed based on the "conservative estimates for chemical element weight percentages" in Table 4 of the propose PTS Rule, 10 CFR 50.61a (Reference 9). Table 4 of the proposed PTS Rule identifies values of 1.45 for plates and 1.63 for welds.The latest proposed alternate PTS Rule, 10 CFR 50.61a (Reference 10), requires the evaluation of plant specific information that could affect the calculation of the irradiation induced shift in reference temperature, AT 3 0 , for the beltline materials.
Circ Weld LINDE 0091 .240 .160 .014 1.63 -80 6.36 8 Lower Shell Plate A 533B .130 .540 .010 1.45 10 6.36 9 Lower Shell Plate A 533B .120 .550 .009 1.45 -10 6.36 10 Lower Shell Plate A 533B .110 .530 .008 1.45 -20 6.36 11 Inter. Shell Plate A 533B .110 .550 .011 1.45 20 6.36 12 Inter. Shell Plate A 533B .120 .640 .011 1.45 -30 6.36 13 Inter. Shell Plate A 533B .120 '.640 .011 1.45 10 6.36 Outputs Methodology Used to Calculate AT 3 0: NUREG-1874 Controlling Material Fluence [1019 Region # RTMAX-XX Neutron/cm 2 , d (flux) AT 3 0 [°F] TWCF 9 5-XX (From [R] E>1 MeV]Above)Axial Weld- AW 4,5,6 668.88 6.36 3.36E+10 259.19 8.03E-09 Circumferential Weld -CW 13 618.48 6.36 3.36E+10 148.79 2.20E-25 Plate -PL 13 618.48 6.36 3.36E+10 148.79 7.85E-12 TWCF 9 5-TOTAL (aAwTWCF 9 5-AW + CPLTWCF 9 5-PL + acwTWCF 9 5.cw): 1.80E-08 Note 1: The manganese content for the welds and plates used in the calculation of TWCF was assumed based on the "conservative estimates for chemical element weight percentages" in Table 4 of the propose PTS Rule, 10 CFR 50.61a (Reference 9). Table 4 of the proposed PTS Rule identifies values of 1.45 for plates and 1.63 for welds.The latest proposed alternate PTS Rule, 10 CFR 50.61a (Reference 10), requires the evaluation of plant specific information that could affect the calculation of the irradiation induced shift in reference temperature, AT 3 0 , for the beltline materials.
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The TWCFs that were calculated for this relief request were developed using the equations for irradiation induced shift from the technical basis for the proposed PTS rule (NUREG-1874, Reference 5).Therefore, these surveillance data evaluations have been performed for the Calvert Cliffs Unit I reactor vessel surveillance data.The Calvert Cliffs Unit 1 reactor vessel surveillance program includes the weld metal for the circumferential weld 9-203 (wire heat 33A277) and plate D-7206-3 (heat number C-4441-1).
The TWCFs that were calculated for this relief request were developed using the equations for irradiation induced shift from the technical basis for the proposed PTS rule (NUREG-1874, Reference 5).Therefore, these surveillance data evaluations have been performed for the Calvert Cliffs Unit I reactor vessel surveillance data.The Calvert Cliffs Unit 1 reactor vessel surveillance program includes the weld metal for the circumferential weld 9-203 (wire heat 33A277) and plate D-7206-3 (heat number C-4441-1).
While only two surveillance capsules have been withdrawn and analyzed for each of the Calvert Cliffs reactor vessels, surveillance data for materials in the Calvert Cliffs Unit 1 reactor vessel exists in the surveillance programs of other plants for which there are greater than three points of data. Weld wire heat 33A277 is also included in the Farley Unit 1 reactor vessel surveillance program. There are two points of data from the Calvert Cliffs Unit I program and six points of data from the Farley Unit 1 program. Furthermore, the Unit 1 intermediate shell weld seams, 2-203-A, B, and C are fabricated from weld wire heat 12008/20291 which is included in the McGuire Unit I reactor vessel surveillance program. There are five points of data for this material from the McGuire Unit 1 program. Therefore, there are two materials in the Calvert Cliffs Unit I reactor vessel for which there are more than three data points. These are weld wire heats 12008/20291 for the intermediate shell welds and 33A277 for the circumferential weld. The data for weld wire heat 33A277 is reported in Reference II for Calvert Cliffs Unit 1 and in Reference 12 for Farley Unit 1. The data for weld wire heat 12008/20291 is reported in Reference 13 for McGuire Unit 1.This data is summarized in Tables 4 and 5 along with the results of the data evaluations.
While only two surveillance capsules have been withdrawn and analyzed for each of the Calvert Cliffs reactor vessels, surveillance data for materials in the Calvert Cliffs Unit 1 reactor vessel exists in the surveillance programs of other plants for which there are greater than three points of data. Weld wire heat 33A277 is also included in the Farley Unit 1 reactor vessel surveillance program. There are two points of data from the Calvert Cliffs Unit I program and six points of data from the Farley Unit 1 program. Furthermore, the Unit 1 intermediate shell weld seams, 2-203-A, B, and C are fabricated from weld wire heat 12008/20291 which is included in the McGuire Unit I reactor vessel surveillance program. There are five points of data for this material from the McGuire Unit 1 program. Therefore, there are two materials in the Calvert Cliffs Unit I reactor vessel for which there are more than three data points. These are weld wire heats 12008/20291 for the intermediate shell welds and 33A277 for the circumferential weld. The data for weld wire heat 33A277 is reported in Reference II for Calvert Cliffs Unit 1 and in Reference 12 for Farley Unit 1. The data for weld wire heat 12008/20291 is reported in Reference 13 for McGuire Unit 1.This data is summarized in Tables 4 and 5 along with the results of the data evaluations.
As can be seen from these tables, the surveillance results satisfy the criteria in the proposed alternate PTS Rule for all three tests. Therefore, the use of the equations contained in NUREG-1874 for calculation of AT 3 0 is acceptable for Calvert Cliffs Unit 1.5 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Table 4. 10 CFR 50.61a Surveillance Data Evaluation for Weld Wire Heat 12008/20291 (Welds 2-203A, B, C)Capsule U X V Y W Plant McGuire McGuire McGuire McGuire McGuire Unit I Unit I Unit I Unit I Uniti Input Data Copper (Weight %) 0.210 0.210 0.210 0.210 0.210 Phosphorous (Weight %) 0.011 0.011 0.011 0.011 0.011 Nickel (Weight %) 0.880 0.880 0.880 0.880 0.880 Manganese (Weight %) 1.360 1.360 1.360 , 1.360 1.360 Fluence(xl0 1 9 n/cm 2 , E > 1.0MeV) 0.378 1.400 1.930 2.640 5.100 EFPY 1.09 4.30 7.24 10.21 19.22 Time Averaged Coolant Temperature  
As can be seen from these tables, the surveillance results satisfy the criteria in the proposed alternate PTS Rule for all three tests. Therefore, the use of the equations contained in NUREG-1874 for calculation of AT 3 0 is acceptable for Calvert Cliffs Unit 1.5 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Table 4. 10 CFR 50.61a Surveillance Data Evaluation for Weld Wire Heat 12008/20291 (Welds 2-203A, B, C)Capsule U X V Y W Plant McGuire McGuire McGuire McGuire McGuire Unit I Unit I Unit I Unit I Uniti Input Data Copper (Weight %) 0.210 0.210 0.210 0.210 0.210 Phosphorous (Weight %) 0.011 0.011 0.011 0.011 0.011 Nickel (Weight %) 0.880 0.880 0.880 0.880 0.880 Manganese (Weight %) 1.360 1.360 1.360 , 1.360 1.360 Fluence(xl0 1 9 n/cm 2 , E > 1.0MeV) 0.378 1.400 1.930 2.640 5.100 EFPY 1.09 4.30 7.24 10.21 19.22 Time Averaged Coolant Temperature
('F) 559.00 559.00 559.00 559.00 558.00 Measured AT 3 0 Transition Temperature  
('F) 559.00 559.00 559.00 559.00 558.00 Measured AT 3 0 Transition Temperature
('F) 161.10 170.60 179.80 190.20 208.00 Calculated Values Predicted AT 3 0 Transition Temperature  
('F) 161.10 170.60 179.80 190.20 208.00 Calculated Values Predicted AT 3 0 Transition Temperature
('F) 145.66 197.31 205.24 212.46 229.74 Residual (r) 15.44 -26.71 -25.44 -22.26 -21.74 Maximum Heat-Average Residual Test Slope Deviation Test Outlier Deviation Test Mean Deviation  
('F) 145.66 197.31 205.24 212.46 229.74 Residual (r) 15.44 -26.71 -25.44 -22.26 -21.74 Maximum Heat-Average Residual Test Slope Deviation Test Outlier Deviation Test Mean Deviation  
-16.14 T-Statistic  
-16.14 T-Statistic  
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Pass Pass/Fail?
Pass Pass/Fail?
Pass Second largest r* +/-1.84 -0.96 Pass/Fail?
Pass Second largest r* +/-1.84 -0.96 Pass/Fail?
Pass 6 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Table 5. 10 CFR 50.61a Surveillance Data Evaluation for Weld Wire Heat 33A277 (Weld 9-203)Capsule 2630 970 Y U X W V Z Calvert Calvert Plant Cliffs Cliffs Farley Farley Farley Farley Farley Farley Unit 1 Unit 1 Unit 1 Unit I Unit 1 Unit I Unit I Unit I Input Data Copper (Weight %) 0.240 0.240 0.140 0.140 0.140 0.140 0.140 0.140 Phosphorous (Weight %) 0.014 0.014 0.016 0.016 0.016 0.016 0.016 0.016 Nickel (Weight %) 0.180 0.180 0.190 0.190 0.190 0.190 0.190 0.190 Manganese (Weight %) 1.050 1.050 1.060 1.060 1.060 1.060 1.060 1.060 Fluence (xl09 n/cm 2 , E > 1.0MeV) 0.620 2.640 0.612 1.730 3.060 4.750 7.140 8.470 EFPY 2.98 11.07 1.15 3.08 6.11 12.43 20.16 24.26 Time Averaged Coolant 548.00 548.00 544.00 540.25 540.86 541.75 541.72 541.43 Temperature  
Pass 6 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Table 5. 10 CFR 50.61a Surveillance Data Evaluation for Weld Wire Heat 33A277 (Weld 9-203)Capsule 2630 970 Y U X W V Z Calvert Calvert Plant Cliffs Cliffs Farley Farley Farley Farley Farley Farley Unit 1 Unit 1 Unit 1 Unit I Unit 1 Unit I Unit I Unit I Input Data Copper (Weight %) 0.240 0.240 0.140 0.140 0.140 0.140 0.140 0.140 Phosphorous (Weight %) 0.014 0.014 0.016 0.016 0.016 0.016 0.016 0.016 Nickel (Weight %) 0.180 0.180 0.190 0.190 0.190 0.190 0.190 0.190 Manganese (Weight %) 1.050 1.050 1.060 1.060 1.060 1.060 1.060 1.060 Fluence (xl09 n/cm 2 , E > 1.0MeV) 0.620 2.640 0.612 1.730 3.060 4.750 7.140 8.470 EFPY 2.98 11.07 1.15 3.08 6.11 12.43 20.16 24.26 Time Averaged Coolant 548.00 548.00 544.00 540.25 540.86 541.75 541.72 541.43 Temperature
('F)MeasuredAT 3 0 Transition 59.00 93.00 66.90 75.10 87.40 98.30 117.50 113.50 Temperature  
('F)MeasuredAT 3 0 Transition 59.00 93.00 66.90 75.10 87.40 98.30 117.50 113.50 Temperature
('F) I I I IIIII Calculated Values Predicted AT 3 0 Transition 91.35 119.20 65.09 89.18 104.39 118.19 135.38 144.36 Temperature  
('F) I I I IIIII Calculated Values Predicted AT 3 0 Transition 91.35 119.20 65.09 89.18 104.39 118.19 135.38 144.36 Temperature
('F) 91.35 119.20 65.09 89.18 -1 144.36 Residual (r) -32.35 -26.20 1.81 -14.08 -16.99 -19.89 -17.88 -30.86 Maximum Heat-Average Residual Test Slope Deviation Test Outlier Deviation Test Mean Deviation  
('F) 91.35 119.20 65.09 89.18 -1 144.36 Residual (r) -32.35 -26.20 1.81 -14.08 -16.99 -19.89 -17.88 -30.86 Maximum Heat-Average Residual Test Slope Deviation Test Outlier Deviation Test Mean Deviation  
-19.56 T-Statistic  
-19.56 T-Statistic  
Line 72: Line 72:
: 10. Federal Register Notice -Supplemental Proposed Rule: Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, 10 CFR 50.61a (RIN 3150-AI01)(ADAMS Accession Number ML081440656)
: 10. Federal Register Notice -Supplemental Proposed Rule: Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, 10 CFR 50.61a (RIN 3150-AI01)(ADAMS Accession Number ML081440656)
: 11. BAW-2160, "Analysis of Capsule 970 Baltimore Gas & Electric Company Calvert Cliffs Nuclear Plant No. 1," June 1993 12. WCAP-16964-NP, "Analysis of Capsule Z from the Southern Nuclear Operating Company Joseph M.Farley Unit 1 Reactor Vessel Radiation Surveillance Program," October 2008 13. WCAP-17014-NP, "Analysis of Capsule W from the McGuire Unit No. 1 Reactor Vessel Radiation Surveillance Program for the Calvert Cliffs Unit 1 Vessel Weld," December 2008.8 ATTACHMENT (2)PROPOSED ALTERNATIVE ISI-023 Calvert Cliffs Nuclear Power Plant, Inc.February 18, 2009 ATTACHMENT (2)PROPOSED ALTERNATIVE ISI-023 American Society of Mechanical En2ineers (ASME) Code Component(s)
: 11. BAW-2160, "Analysis of Capsule 970 Baltimore Gas & Electric Company Calvert Cliffs Nuclear Plant No. 1," June 1993 12. WCAP-16964-NP, "Analysis of Capsule Z from the Southern Nuclear Operating Company Joseph M.Farley Unit 1 Reactor Vessel Radiation Surveillance Program," October 2008 13. WCAP-17014-NP, "Analysis of Capsule W from the McGuire Unit No. 1 Reactor Vessel Radiation Surveillance Program for the Calvert Cliffs Unit 1 Vessel Weld," December 2008.8 ATTACHMENT (2)PROPOSED ALTERNATIVE ISI-023 Calvert Cliffs Nuclear Power Plant, Inc.February 18, 2009 ATTACHMENT (2)PROPOSED ALTERNATIVE ISI-023 American Society of Mechanical En2ineers (ASME) Code Component(s)
Affected The affected component is the Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) Unit 1 reactor pressure vessel (RPV), specifically the following ASME Boiler and Pressure Vessel (BPV) Code Section XI (Reference  
Affected The affected component is the Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) Unit 1 reactor pressure vessel (RPV), specifically the following ASME Boiler and Pressure Vessel (BPV) Code Section XI (Reference
: 1) examination categories and item numbers covering examinations of the RPV. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code Section XI.Examination Category Item No. Description B-N-2 B 13.50 Interior Attachments Within Beltline Region B-N-2 B 13.60 Interior Attachments Beyond Beltline Region B-N-3 B 13.70 Core Support Structure (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".)Applicable Code Edition and Addenda American Society of Mechanical Engineers Code Section XI, "Rules and Inservice Inspection of Nuclear Power Plant Components," 1998 Edition (no Addenda).Applicable Code Requirement In accordance with IWA-2430(d)(1), each inspection interval may be reduced or extended by as much as one year. Adjustments shall not cause successive intervals to be altered more than one year from the original pattern of intervals.
: 1) examination categories and item numbers covering examinations of the RPV. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code Section XI.Examination Category Item No. Description B-N-2 B 13.50 Interior Attachments Within Beltline Region B-N-2 B 13.60 Interior Attachments Beyond Beltline Region B-N-3 B 13.70 Core Support Structure (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".)Applicable Code Edition and Addenda American Society of Mechanical Engineers Code Section XI, "Rules and Inservice Inspection of Nuclear Power Plant Components," 1998 Edition (no Addenda).Applicable Code Requirement In accordance with IWA-2430(d)(1), each inspection interval may be reduced or extended by as much as one year. Adjustments shall not cause successive intervals to be altered more than one year from the original pattern of intervals.
Additionally, Table IWB-2500-1, Examination Categories B-N-2 and B-N-3, Item Numbers B13.50, B13.60, and B13.70 requires a visual examination of the accessible interior attachment welds within and beyond the beltline region and a visual examination of the accessible core support structure surfaces of the RPV once each ten-year interval.
Additionally, Table IWB-2500-1, Examination Categories B-N-2 and B-N-3, Item Numbers B13.50, B13.60, and B13.70 requires a visual examination of the accessible interior attachment welds within and beyond the beltline region and a visual examination of the accessible core support structure surfaces of the RPV once each ten-year interval.
The Calvert Cliffs Unit 1 Third Ten-Year Inservice Inspection (ISI)interval is scheduled to end June 30, 2009.Reason for Request In Westinghouse Topical Report WCAP-16168-NP-A, Revision 2 (Reference 2), the Pressurized Water Reactor Owners Group provided the technical and regulatory basis for decreasing the frequency of inspections by extending the ASME Code Section XI ISI interval from the current 10 years to 20 years for ASME Code Section XI Examination Categories B-A and B-D RPV welds. The Nuclear Regulatory Commission approved the topical report by letter dated May 8, 2008 (Reference 3). To implement the change presented in Reference 2, we are submitting Attachment (1) (ISI-022), in accordance with the Safety Evaluation (Reference  
The Calvert Cliffs Unit 1 Third Ten-Year Inservice Inspection (ISI)interval is scheduled to end June 30, 2009.Reason for Request In Westinghouse Topical Report WCAP-16168-NP-A, Revision 2 (Reference 2), the Pressurized Water Reactor Owners Group provided the technical and regulatory basis for decreasing the frequency of inspections by extending the ASME Code Section XI ISI interval from the current 10 years to 20 years for ASME Code Section XI Examination Categories B-A and B-D RPV welds. The Nuclear Regulatory Commission approved the topical report by letter dated May 8, 2008 (Reference 3). To implement the change presented in Reference 2, we are submitting Attachment (1) (ISI-022), in accordance with the Safety Evaluation (Reference
: 3) to request an alternative from the Code requirements pursuant to 10 CFR 50.55a(a)(3)(i) on the basis that the alternative inspection interval (20 years) provides an acceptable level of quality and safety. In Attachment (1) we identified 2018 as the year in which future inspection of the Examination Categories B-A and B-D RPV welds will be performed.
: 3) to request an alternative from the Code requirements pursuant to 10 CFR 50.55a(a)(3)(i) on the basis that the alternative inspection interval (20 years) provides an acceptable level of quality and safety. In Attachment (1) we identified 2018 as the year in which future inspection of the Examination Categories B-A and B-D RPV welds will be performed.
The intent of this relief request (ISI-023) is to allow deferral of the subject examinations to the same time (2018 refueling outage) as the Examination Categories B-A and B-D RPV welds described in Attachment (1).During the ten-year ISI of the RPV shell, lower head, and nozzle welds in 1998, Calvert Cliffs also performed visual examinations of the RPV interior attachments and the core support structure.
The intent of this relief request (ISI-023) is to allow deferral of the subject examinations to the same time (2018 refueling outage) as the Examination Categories B-A and B-D RPV welds described in Attachment (1).During the ten-year ISI of the RPV shell, lower head, and nozzle welds in 1998, Calvert Cliffs also performed visual examinations of the RPV interior attachments and the core support structure.

Revision as of 02:09, 12 July 2019

Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations - Relief Requests (ISI-022 and ISI-023)
ML090540062
Person / Time
Site: Calvert Cliffs Constellation icon.png
Issue date: 02/18/2009
From: Flaherty M
Constellation Energy Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML090540062 (15)
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Text

-1 -1 Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 Constellation Energy Nuclear Generation Group February 18, 2009 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit No. 1; Docket No. 50-317 Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations

-Relief Requests (ISI-022 and ISI-023)

REFERENCES:

(a) Westinghouse Owners Group Topical Report, WCAP-16168-NP-A, Revision 2, dated June 2008, Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval (b) Letter from H. K. Nieh (NRC) to G. Bischoff (PWROG), dated May 8, 2008, Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR) WCAP-16168-NP, Revision 2, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (TAC No. MC9768)(c) Letter from J. A. Spina (CCNPP) to Document Control Desk (NRC), dated October 1, 2008, License Amendment Request: Reporting of Reactor Vessel In-Service Inspection Information and Analyses in Support of Code Relief Request for Extension of Reactor Vessel In-Service Inspection Interval (d) Letter from M. D. Flaherty (CCNPP) to Document Control Desk (NRC), dated October 1, 2008, Revised Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld and Visual Examinations-Relief Requests (ISI-020 and ISI-021)In Westinghouse Topical Report WCAP-16168-NP-A, Revision 2 (Reference a), the Pressurized Water Reactor Owners Group provided the technical and regulatory basis for decreasing the frequency of inspections by extending the American Society of Mechanical Engineers (ASME) Code Section XI inservice inspection interval from the current 10 years to 20 years for ASME Code Section XI examination categories B-A and B-D reactor pressure vessel (RPV) welds. The Nuclear Regulatory Commission approved the topical report by letter dated May 8, 2008 (Reference b). To implement the change presented in Reference (a) for Unit 1, Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) hereby submits Attachment (1) (ISI-022) in accordance with the Safety Evaluation (Reference b), to request an alternative from the Code requirements pursuant to 10 CFR 50.55a(a)(3)(i) on the basis that the alternative inspection interval (20 years) provides an acceptable level of quality and safety. The plant Document Control Desk February 18, 2009 Page 2 specific information identified by Reference (a) as needed to support this proposed alternative for Unit 1 is provided in Attachment (1). The future inspection of examination categories B-A and B-D RPV welds will be performed in 2018 for Unit 1. Additionally, as required by Reference (b), a license amendment request (Reference c) (submitted for both Unit I and Unit 2) was previously submitted in support of a similar relief request for Calvert Cliffs Unit 2 (Reference d).Further, pursuant to 10 CFR 50.55a(a)(3)(ii), Calvert Cliffs is also requesting approval to delay compliance with the specified requirements of ASME Section XI, Table IWB-2500-1, for Calvert Cliffs Unit 1. Relief is requested to defer visual examinations under ASME Section XI Table IWB-2500-1 examination categories B-N-2 and B-N-3, Item Nos. B13.50, B13.60, and B13.70 for the extended inspection interval from 10 to 20 years as described in Attachment (2) (ISI-023).

Deferring the subject visual examinations associated with the reactor vessel core support structure removal to the extended inspection interval allows those examinations to be performed at the same time (2018 refueling outage) as the examination categories B-A and B-D RPV welds described in Attachment (1).Calvert Cliffs' proposed extension of the inservice inspection interval for these examinations will continue to provide an acceptable level of quality and safety, as described in the attached relief requests.Calvert Cliffs requests the Nuclear Regulatory Commission staff approval by October 15, 2009, in order to support preparation activities for the spring 2010 Unit 1 refueling outage.Should you have questions regarding this matter, please contact Mr. Jay S. Gaines at (410) 495-5219.Very truly yours, Mark D. Flaherty Manager-Engineering Services MDF/KLG/bjd Attachments:

(1) Proposed Alternative ISI-022 (2) Proposed Alternative ISI-023 cc: D. V. Pickett, NRC S. Gray, DNR S. J. Collins, NRC Safety Program Manager, State of Maryland Resident Inspector, NRC ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Calvert Cliffs Nuclear Power Plant, Inc.February 18, 2009 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 American Society of Mechanical Engineers (ASME) Code Component(s)

Affected The affected component is the Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) Unit 1 reactor vessel, specifically the following ASME Boiler and Pressure Vessel (BPV) Code Section XI (Reference 1)examination categories and item numbers covering examinations of the reactor vessel. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code Section XI.Examination Category Item No. Description B-A B 1.11 Circumferential Shell Welds B-A B 1.12 Longitudinal Shell Welds B-A B 1.21 Circumferential Head Welds B-A B 1.22 Meridional Shell Welds B-A B 1.30 Shell-to-Flange Weld B-A B 1.40 Head-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inner Radius Areas (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.")Applicable Code Edition and Addenda American Society of Mechanical Engineers Code Section XI, "Rules and Inservice Inspection of Nuclear Power Plant Components," 1998 Edition (no Addenda) for the third interval and 2004 Edition (no Addenda) for the fourth interval.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 Calvert Cliffs Unit 1 third ten-year inservice inspection interval is scheduled to end June 30, 2009. American Society of Mechanical Engineers Code Section XI IWA-2430(d)(1) allows a one year extension.

Reason for Request An alternative is requested from the requirement of IWA-2412, Inspection Program B, that volumetric examination of essentially 100% of reactor vessel pressure retaining, Examination Category 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 personnel radiation exposure and examination costs.Proposed Alternative and Basis for Use Calvert Cliffs proposes to defer the ASME Code required volumetric examination of the Unit 1 reactor vessel full penetration pressure retaining Examination Category B-A and B-D welds for the third inservice inspection until 2018 and to perform the fourth inservice inspection of these welds on a 20-year inspection interval, instead of the current ten-year inspection interval.

Therefore, the fourth inservice inspection is proposed to be performed in 2038 if Unit 1 requests and obtains an additional license renewal extension beyond the current license expiration date of July 31, 2034. These dates are consistent with the information provided to the Staff in Pressurized Water Reactor Owners Group letter OG-06-356 (Reference 2).I ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 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 third and fourth inspection intervals for Examination 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 Nuclear Regulatory Commission (NRC)Pressurized Thermal Shock (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 Calvert Cliffs Unit 1 reactor vessel is acceptable as shown in Table 1 below.Table 1. Critical Parameters for Application of Bounding Analysis Additional Parameter Pilot Plant Basis Plant Specific Evaluation Required?Dominant PTS Transients in the NRC PTS Risk Study PTS Generalization No NRC PTS Risk Study are (Reference

5) Study (Reference 6)applicable Through-Wall Cracking 3.16E-7 Events per year 1.80E-08 Events per No Frequency (TWCF) (Reference
4) year (Calculated per Reference 5)Frequency and Severity of Design 13 heatup/cooldowns per Bounded by 13 No Basis Transients year (Reference
4) heatup/cooldowns per year Cladding Layers Single Layer (Reference
4) Single Layer No (Single/Multiple)

II Additional information relative to the Calvert Cliffs Unit 1 reactor vessel inspection is provided in Table 2. This information confirms that .satisfactory examinations have been performed on the Calvert Cliffs Unit I reactor vessel.2 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Table 2. Additional Information Pertaining to Reactor Vessel Inspection Inspection methodology:

ASME Section XI and Regulatory Guide 1.150 (Reference 7)inspections were Performance Demonstration Initiative qualified per Reference 8 and were therefore performed in accordance with the requirements of ASME Section XI Appendix VIII.Number of past inspections':

All reactor vessel welds have been inservice inspected at least twice with the exception of five lower head meridional welds which have beeninspected once.Number of indications found: A total of 15 indications were detected in the most recent inservice inspection which is considered to be the most accurate inspection because the latest technology was used. All 15 indications are acceptable in accordance with IWB-3500 of Section XI of the ASME Code. One of these indications is located in the reactor vessel beltline region. However, it is not within the inner 1" of the thickness.

Therefore, these indications meet the requirements for the proposed alternate PTS Rule (10 CFR 50.61a) in SECY-07-0104 (Reference 9).Proposed inspection schedule The third inservice inspection is currently scheduled for 2010. This for balance of plant life: inspection was originally proposed to be performed in 2008 but was deferred until 2010 per IWA-2430(d) of ASME Section XI. The third inservice inspection is proposed to be performed in 2018.Note 1: The Unit 1 RPV closure head was replaced in 2006. Therefore, the replacement head-to-flange weld has not received a prior inservice inspection examination.

Table 3 provides additional information relative to the calculation of the* TWCF parameter for Calvert Cliffs Unit 1.3 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Table 3. Details of TWCF Calculation at 60 EFPY Inputs Reactor Coolant System Temperature, TRCS[°F]:

1 548 Twal, [inches]:

8.625 Un- Fluence [1019 Region/Component Material Cu Ni P Mn] Irradiated Neutron/cm 2 , Description

[wt%] [wt%] [wt%] [wt%] RTNDT(u) ['F] E>1 MeV]1 Low. Shell Axial LINDE 1092 .180 .720 .015 1.63 -56 6.36 Weld 2 Low. Shell Axial LINDE 1092 .180 .720 .015 1.63 -56 6.36 Weld 3 Low. Shell Axial LINDE 1092 .180 .720 .015 1.63 -56 6.36 Weld 4 Int. Shell Axial Weld LINDE 1092 .220 .830 .010 1.63 -50 6.36 5 Int. Shell Axial Weld LINDE 1092 .220 .830 .010 1.63 -50 6.36 6 Int. Shell Axial Weld LINDE 1092 .220 .830 .010 1.63 -50 6.36 7 Int./Low.

Circ Weld LINDE 0091 .240 .160 .014 1.63 -80 6.36 8 Lower Shell Plate A 533B .130 .540 .010 1.45 10 6.36 9 Lower Shell Plate A 533B .120 .550 .009 1.45 -10 6.36 10 Lower Shell Plate A 533B .110 .530 .008 1.45 -20 6.36 11 Inter. Shell Plate A 533B .110 .550 .011 1.45 20 6.36 12 Inter. Shell Plate A 533B .120 .640 .011 1.45 -30 6.36 13 Inter. Shell Plate A 533B .120 '.640 .011 1.45 10 6.36 Outputs Methodology Used to Calculate AT 3 0: NUREG-1874 Controlling Material Fluence [1019 Region # RTMAX-XX Neutron/cm 2 , d (flux) AT 3 0 [°F] TWCF 9 5-XX (From [R] E>1 MeV]Above)Axial Weld- AW 4,5,6 668.88 6.36 3.36E+10 259.19 8.03E-09 Circumferential Weld -CW 13 618.48 6.36 3.36E+10 148.79 2.20E-25 Plate -PL 13 618.48 6.36 3.36E+10 148.79 7.85E-12 TWCF 9 5-TOTAL (aAwTWCF 9 5-AW + CPLTWCF 9 5-PL + acwTWCF 9 5.cw): 1.80E-08 Note 1: The manganese content for the welds and plates used in the calculation of TWCF was assumed based on the "conservative estimates for chemical element weight percentages" in Table 4 of the propose PTS Rule, 10 CFR 50.61a (Reference 9). Table 4 of the proposed PTS Rule identifies values of 1.45 for plates and 1.63 for welds.The latest proposed alternate PTS Rule, 10 CFR 50.61a (Reference 10), requires the evaluation of plant specific information that could affect the calculation of the irradiation induced shift in reference temperature, AT 3 0 , for the beltline materials.

In order to make this determination, the alternate PTS Rule provides requirements for evaluation of surveillance capsule data. These requirements are specified in paragraphs (f)(6)(ii), (f)(6)(iii), and (f)(6)(iv) of the proposed Rule. In summary the requirements consist of a Mean Deviation Test, a Slope Deviation Test, and an Outlier Deviation Test. The equations for performing these tests are included as Equations 8 through 12 and the acceptance criteria for these tests are provided in Tables 5, 6, and 7 of the proposed 10 CFR 50.61a. These tests must be performed when 4 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 three or more measurements of surveillance data are available for any of the reactor vessel beltline materials.

The TWCFs that were calculated for this relief request were developed using the equations for irradiation induced shift from the technical basis for the proposed PTS rule (NUREG-1874, Reference 5).Therefore, these surveillance data evaluations have been performed for the Calvert Cliffs Unit I reactor vessel surveillance data.The Calvert Cliffs Unit 1 reactor vessel surveillance program includes the weld metal for the circumferential weld 9-203 (wire heat 33A277) and plate D-7206-3 (heat number C-4441-1).

While only two surveillance capsules have been withdrawn and analyzed for each of the Calvert Cliffs reactor vessels, surveillance data for materials in the Calvert Cliffs Unit 1 reactor vessel exists in the surveillance programs of other plants for which there are greater than three points of data. Weld wire heat 33A277 is also included in the Farley Unit 1 reactor vessel surveillance program. There are two points of data from the Calvert Cliffs Unit I program and six points of data from the Farley Unit 1 program. Furthermore, the Unit 1 intermediate shell weld seams, 2-203-A, B, and C are fabricated from weld wire heat 12008/20291 which is included in the McGuire Unit I reactor vessel surveillance program. There are five points of data for this material from the McGuire Unit 1 program. Therefore, there are two materials in the Calvert Cliffs Unit I reactor vessel for which there are more than three data points. These are weld wire heats 12008/20291 for the intermediate shell welds and 33A277 for the circumferential weld. The data for weld wire heat 33A277 is reported in Reference II for Calvert Cliffs Unit 1 and in Reference 12 for Farley Unit 1. The data for weld wire heat 12008/20291 is reported in Reference 13 for McGuire Unit 1.This data is summarized in Tables 4 and 5 along with the results of the data evaluations.

As can be seen from these tables, the surveillance results satisfy the criteria in the proposed alternate PTS Rule for all three tests. Therefore, the use of the equations contained in NUREG-1874 for calculation of AT 3 0 is acceptable for Calvert Cliffs Unit 1.5 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Table 4. 10 CFR 50.61a Surveillance Data Evaluation for Weld Wire Heat 12008/20291 (Welds 2-203A, B, C)Capsule U X V Y W Plant McGuire McGuire McGuire McGuire McGuire Unit I Unit I Unit I Unit I Uniti Input Data Copper (Weight %) 0.210 0.210 0.210 0.210 0.210 Phosphorous (Weight %) 0.011 0.011 0.011 0.011 0.011 Nickel (Weight %) 0.880 0.880 0.880 0.880 0.880 Manganese (Weight %) 1.360 1.360 1.360 , 1.360 1.360 Fluence(xl0 1 9 n/cm 2 , E > 1.0MeV) 0.378 1.400 1.930 2.640 5.100 EFPY 1.09 4.30 7.24 10.21 19.22 Time Averaged Coolant Temperature

('F) 559.00 559.00 559.00 559.00 558.00 Measured AT 3 0 Transition Temperature

('F) 161.10 170.60 179.80 190.20 208.00 Calculated Values Predicted AT 3 0 Transition Temperature

('F) 145.66 197.31 205.24 212.46 229.74 Residual (r) 15.44 -26.71 -25.44 -22.26 -21.74 Maximum Heat-Average Residual Test Slope Deviation Test Outlier Deviation Test Mean Deviation

-16.14 T-Statistic

-2.39 Allowable Actual Maximum Mean Residual +/-27.51 Critical T-Statisitc

+/-4.54 Largest r* +2.88 -1.01 Pass/Fail?

Pass Pass/Fail?

Pass Second largest r* +/-1.84 -0.96 Pass/Fail?

Pass 6 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Table 5. 10 CFR 50.61a Surveillance Data Evaluation for Weld Wire Heat 33A277 (Weld 9-203)Capsule 2630 970 Y U X W V Z Calvert Calvert Plant Cliffs Cliffs Farley Farley Farley Farley Farley Farley Unit 1 Unit 1 Unit 1 Unit I Unit 1 Unit I Unit I Unit I Input Data Copper (Weight %) 0.240 0.240 0.140 0.140 0.140 0.140 0.140 0.140 Phosphorous (Weight %) 0.014 0.014 0.016 0.016 0.016 0.016 0.016 0.016 Nickel (Weight %) 0.180 0.180 0.190 0.190 0.190 0.190 0.190 0.190 Manganese (Weight %) 1.050 1.050 1.060 1.060 1.060 1.060 1.060 1.060 Fluence (xl09 n/cm 2 , E > 1.0MeV) 0.620 2.640 0.612 1.730 3.060 4.750 7.140 8.470 EFPY 2.98 11.07 1.15 3.08 6.11 12.43 20.16 24.26 Time Averaged Coolant 548.00 548.00 544.00 540.25 540.86 541.75 541.72 541.43 Temperature

('F)MeasuredAT 3 0 Transition 59.00 93.00 66.90 75.10 87.40 98.30 117.50 113.50 Temperature

('F) I I I IIIII Calculated Values Predicted AT 3 0 Transition 91.35 119.20 65.09 89.18 104.39 118.19 135.38 144.36 Temperature

('F) 91.35 119.20 65.09 89.18 -1 144.36 Residual (r) -32.35 -26.20 1.81 -14.08 -16.99 -19.89 -17.88 -30.86 Maximum Heat-Average Residual Test Slope Deviation Test Outlier Deviation Test Mean Deviation

-19.56 T-Statistic

-0.89 Allowable Actual Maximum Mean Residual +/-21.75 Critical T-Statisitc

+/-3.14 Largest r* +/-3.02 -1.23 Pass/Fail?

Pass Pass/Fail?

Pass Second largest r* +2.05 -1.17 Pass/Fail?

Pass 7 ATTACHMENT (1)PROPOSED ALTERNATIVE ISI-022 Duration of Proposed Alternative This request is applicable to the Calvert Cliffs Unit I inservice inspection program for the 60 year extended license.References

1. ASME Boiler and Pressure Vessel Code,Section XI, 2004 Edition (No 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. Bajwa, S. of U.S. NRC to Cruse, C. of Constellation, "Relief request from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) requirements to use Performance Demonstration Initiative Program -Reactor Pressure Second Ten-Year Inservice Inspection Interval -Calvert Cliffs Nuclear Power Plant Unit Nos. 1 and 2 (TAC Nos. M99819 and M99820," January 5, 1998 9. SECY-07-0104, "Proposed Rulemaking

-Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock," Enclosure 1, June 25, 2007 (ADAMS Accession Number ML070570141)

10. Federal Register Notice -Supplemental Proposed Rule: Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, 10 CFR 50.61a (RIN 3150-AI01)(ADAMS Accession Number ML081440656)
11. BAW-2160, "Analysis of Capsule 970 Baltimore Gas & Electric Company Calvert Cliffs Nuclear Plant No. 1," June 1993 12. WCAP-16964-NP, "Analysis of Capsule Z from the Southern Nuclear Operating Company Joseph M.Farley Unit 1 Reactor Vessel Radiation Surveillance Program," October 2008 13. WCAP-17014-NP, "Analysis of Capsule W from the McGuire Unit No. 1 Reactor Vessel Radiation Surveillance Program for the Calvert Cliffs Unit 1 Vessel Weld," December 2008.8 ATTACHMENT (2)PROPOSED ALTERNATIVE ISI-023 Calvert Cliffs Nuclear Power Plant, Inc.February 18, 2009 ATTACHMENT (2)PROPOSED ALTERNATIVE ISI-023 American Society of Mechanical En2ineers (ASME) Code Component(s)

Affected The affected component is the Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) Unit 1 reactor pressure vessel (RPV), specifically the following ASME Boiler and Pressure Vessel (BPV) Code Section XI (Reference

1) examination categories and item numbers covering examinations of the RPV. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code Section XI.Examination Category Item No. Description B-N-2 B 13.50 Interior Attachments Within Beltline Region B-N-2 B 13.60 Interior Attachments Beyond Beltline Region B-N-3 B 13.70 Core Support Structure (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".)Applicable Code Edition and Addenda American Society of Mechanical Engineers Code Section XI, "Rules and Inservice Inspection of Nuclear Power Plant Components," 1998 Edition (no Addenda).Applicable Code Requirement In accordance with IWA-2430(d)(1), each inspection interval may be reduced or extended by as much as one year. Adjustments shall not cause successive intervals to be altered more than one year from the original pattern of intervals.

Additionally, Table IWB-2500-1, Examination Categories B-N-2 and B-N-3, Item Numbers B13.50, B13.60, and B13.70 requires a visual examination of the accessible interior attachment welds within and beyond the beltline region and a visual examination of the accessible core support structure surfaces of the RPV once each ten-year interval.

The Calvert Cliffs Unit 1 Third Ten-Year Inservice Inspection (ISI)interval is scheduled to end June 30, 2009.Reason for Request In Westinghouse Topical Report WCAP-16168-NP-A, Revision 2 (Reference 2), the Pressurized Water Reactor Owners Group provided the technical and regulatory basis for decreasing the frequency of inspections by extending the ASME Code Section XI ISI interval from the current 10 years to 20 years for ASME Code Section XI Examination Categories B-A and B-D RPV welds. The Nuclear Regulatory Commission approved the topical report by letter dated May 8, 2008 (Reference 3). To implement the change presented in Reference 2, we are submitting Attachment (1) (ISI-022), in accordance with the Safety Evaluation (Reference

3) to request an alternative from the Code requirements pursuant to 10 CFR 50.55a(a)(3)(i) on the basis that the alternative inspection interval (20 years) provides an acceptable level of quality and safety. In Attachment (1) we identified 2018 as the year in which future inspection of the Examination Categories B-A and B-D RPV welds will be performed.

The intent of this relief request (ISI-023) is to allow deferral of the subject examinations to the same time (2018 refueling outage) as the Examination Categories B-A and B-D RPV welds described in Attachment (1).During the ten-year ISI of the RPV shell, lower head, and nozzle welds in 1998, Calvert Cliffs also performed visual examinations of the RPV interior attachments and the core support structure.

Since the core support structure (called a core barrel on Combustion Engineering designed plants) requires removal ATTACHMENT (2)PROPOSED ALTERNATIVE ISI-023 to facilitate examination of the RPV shell, lower head, and nozzle welds, the visual examinations of ASME Examination Categories B-N-2 and B-N-3 have historically been performed during the same outage at the end of the ISI interval.Calvert Cliffs has also committed to the development and implementation of a plant specific Reactor Vessel Internals (RVI) inspection program and subsequent submittal to the Nuclear Regulatory Commission two years prior to the period of extended operation.

Calvert Cliffs may elect to perform the enhanced examinations for the RVI inspection program coincident with the core barrel removal in 2018.To complete the full scope of the RVI examination it is expected to require a complete core offload and removal of all internals to facilitate implementation of the examinations.

Portions of the RVI inspection may be performed prior to this time as may be prescribed in that program.Performing all core barrel removed related examinations during the same refueling outage will result in significant savings in dose and outage duration since the same equipment and personnel used for visual and volumetric examination of the RPV shell welds and nozzle welds from the RPV interior can be used to implement the required RV1 examinations.

Additionally, removing the RPV internals only once to accommodate all the examinations discussed in this relief.request would result in significant savings in radiation exposure.Proposed Alternative and Basis for Use The third ten-year IS interval for Calvert Cliffs began on July 1, 1999 and is scheduled to conclude on June 30, 2009. An extension of one year is allowed in ASME,Section XI, IWA-2430(d)(1).

Calvert Cliffs proposes to perform the subject examinations for the third ten-year ISI interval on or before June 30, 2018. The subject examinations are currently scheduled to be performed during the spring 2010 refueling outage. The proposed alternative inspection would enable the subject examinations to be performed during the 2018 refueling outage with the risk-informed extension of the reactor vessel ISI. In accordance with 10 CFR 50.55a(a)(3)(ii), this interval extension is requested on the basis that performing the examination of the RPV interior attachments and core support structure separate in time from the RPV shell, head, and nozzle welds would result in hardship or unusual difficulty without a compensating increase in quality or safety.The full scope examination required by ASME Examination Categories B-N-2 and B-N-3 requires the removal of all the fuel and the core barrel from the RPV. An unnecessary risk is created by removal of the core barrel to perform a visual examination without a compensating increase in quality or safety.Further, the radiation exposure to establish the conditions for and perform the ASME Examination Categories B-N-2 and B-N-3 examinations would essentially double if the subject examinations were performed separate in time from the RPV shell, lower head, and nozzle weld examinations.

The visual examninations of the RPV interior attachments and the core support structure have been performed several times at Calvert Cliffs with no relevant indications noted during the examinations.

The examinations were last performed during the 1998 refueling outage with acceptable results. Additionally, review of industry surveys indicate that these examinations have been performed many times by the industry without any significant findings relevant to the Calvert Cliffs reactor vessel design.As stated in Reference 2, "... it must be recognized that all reactor coolant pressure boundary failures occurring to date have been identified as a result of leakage, and were discovered by visual examination.

The proposed RVlSI interval extension does not alter the visual examination interval.

The reactor vessel would undergo, as a minimum, the Section XI Examination Category B-P pressure tests and visual examinations conducted at the end of each refueling before plant start-up, as well as leak tests with visual 2

.ATTACHMENT (2)PROPOSED ALTERNATIVE ISI-023 examinations that precede each start-up following maintenance or repair activities." The minimum visual examinations discussed in Reference 2 are not the subject examinations (i.e., B-N-2 and B-N-3) of this relief request. During the 2009 refueling outage, Calvert Cliffs will be performing the ASME Examination Category B-N-i visual examination.

This examination will include the space that is made accessible for examination by the removal of components during normal refueling outages. This examination is required once each period and will provide reasonable assurance of structural integrity.

As discussed further in Reference 2, defenses against human errors are preserved with the increase in inspection interval.

Specifically, the increase in the inspection interval reduces the frequency for which the reactor vessel lower internals need to be removed thereby reducing the possibility for human error and damage to the core.Therefore, in accordance with 10 CFR 50.55a(a)(3)(ii), this interval change from 10 to 20 years for the subject examinations is requested on the basis that compliance, with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.Duration of Proposed Alternative This proposed alternative is applicable to the third and fourth ten-year ISI for the Examination Categories B-N-2 and B-N-3, Item Numbers B 13.50, B 13.60, and B 13.70 visual examinations.

References

1. ASME Boiler and Pressure Vessel Code,Section XI, 1998 Edition with no Addenda 2. WCAP-16168-NP-A, Revision 2, Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval, June 2008 3. Final Safety Evaluation For Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR) WCAP-16168-NP, Revision 2, "Risk-Informed Extension Of The Reactor Vessel In-Service Inspection Interval" (TAC No. MC9768), Dated May 8, 2008 3
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