ML12130A217
| ML12130A217 | |
| Person / Time | |
|---|---|
| Site: | Surry  (DPR-032, DPR-037) |
| Issue date: | 04/25/2012 |
| From: | Lane N Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 12-267 | |
| Download: ML12130A217 (17) | |
Text
VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 April 25, 2012 United States Nuclear Regulatory Commission Serial No.12-267 Attention: Document Control Desk SPS-LIC/CGL RO Washington, D.C. 20555 Docket Nos.
50-280/281 License No.
DPR-32/37 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)
SURRY POWER STATION UNITS 1 AND 2 ASME SECTION XI INSERVICE INSPECTION (ISI) PROGRAM REQUEST FOR ALTERNATIVE - IMPLEMENTATION OF EXTENDED REACTOR VESSEL INSERVICE INSPECTION INTERVAL RELIEF REQUESTS CMP-007 AND CMP-009 Dominion requests an alternative to the requirement of IWB-2412, Inspection Program B, which requires examination of identified reactor vessel (RV) pressure retaining welds once each ten year interval. Pursuant to 10 CFR 50.55a(a)(3)(i), an alternate inspection interval of 20 years is requested. The current interval can be extended based on negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174. The methodology used to conduct this analysis is based on that defined in the study WCAP-16168-NP-A, Revision 2, "Risk Informed Extension of the Reactor Vessel In-service Inspection Interval."
This study focuses on risk assessments of materials within the beltline region of the RV wall.
The results of the calculations for Surry Units 1 and 2 were compared to those obtained from the Westinghouse pilot plant evaluated in WCAP-16168-NP-A, Revision 2.
The parameters for Surry Units 1 and 2 are bounded by the results of the Westinghouse pilot plant, which qualifies Surry Units 1 and 2 for an ISI interval extension.
Relief Requests CMP-007 and CMP-009 for Surry Units 1 and 2, respectively, are attached and provide the requisite information and justification for the proposed alternative.
Dominion proposes to perform the fourth 10-year ISI interval examination of the Surry Units 1 and 2 RVs in 2023 and 2024 in lieu of 2013 and 2014, respectively. To support the alternate RV inspection interval of 20 years for Surry Units 1 and 2, Dominion requests approval of Relief Requests CMP-007 and CMP-009 by May 30, 2013.
If you have any questions or require additional information, please contact Mrs. Candee G. Lovett at (757) 365-2178.
Sincerely, N. L. Lane Site Vice President - Surry Power Station
Serial No.12-267 Docket Nos. 50-280/281 Page 2 of 2 Attachments
- 1. Relief Request CMP-007 Surry Unit 1
- 2. Relief Request CMP-009 Surry Unit 2 Commitments made by this letter: None cc:
U.S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta, Georgia 30303-1257 State Health Commissioner Virginia Department of Health James Madison Building - 7th floor 109 Governor Street, Suite 730 Richmond, Virginia 23219 NRC Senior Resident Inspector Surry Power Station Ms. K. R. Cotton, NRC Project Manager - Surry U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G9A 11555 Rockville Pike Rockville, Maryland 20852 Dr. V. Sreenivas, NRC Project Manager - North Anna U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G9A 11555 Rockville Pike Rockville, Maryland 20852 Mr. R. A. Smith Authorized Nuclear Inspector Surry Power Station
Serial No.12-267 Docket Nos. 50-280/281 Implementation of Extended Reactor Vessel Inservice Inspection Interval -
Relief Request CMP-007 Virginia Electric and Power Company (Dominion)
Surry Power Station Unit I
Serial No.12-267 Docket Nos. 50-280/281 Page 1 of 6 Implementation of Extended Reactor Vessel Inservice Inspection Interval for Surry Unit 1 Relief Request CMP-007 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 The affected component is the Surry Unit 1 reactor vessel (RV), specifically the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code, Section Xl (Reference 1) examination categories and item numbers covering examinations of the RV. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code Section Xl.
Examination Category B-A B-A B-A B-A B-A B-A B-A B-A B-D B-D Item No.
B1.11 B1.12 B1.21 B1.22 B1.30 B1.40 B1.50 B1.51 B3.90 B3. 100 Description Circumferential Shell Welds Longitudinal Shell Welds Circumferential Head Welds Meridional Shell Welds Shell-to-Flange Weld Head-to-Flange Weld Repair Weld Beltline Region Nozzle-to-Vessel Welds Nozzle Inside Radius Section (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.")
2. Applicable Code Edition and Addenda
ASME Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Code 1998 Edition with the 2000 Addenda (Reference 1)
Serial No.12-267 Docket Nos. 50-280/281 Page 2 of 6
3. Applicable Code Requirement
IWB-2412, Inspection Program B, requires volumetric examination of 100% of RV pressure retaining welds identified in Table IWB-2500-1 once each ten year interval.
The Surry Unit 1 fourth 10-year inservice inspection (ISI) interval ends in 2013.
4. Reason for Request
An alternative is requested from the requirement of IWB-2412, Inspection Program B, that volumetric examination of 100%
of RV pressure retaining, Examination Category B-A and B-D welds, be performed once each ten-year interval. Extension of the interval between examinations of Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in man-rem exposure and examination costs.
5. Proposed Alternative and Basis for Use
Dominion proposes to perform the ASME Code required volumetric examination of the Surry Unit 1 RV full penetration pressure retaining Examination Category B-A and B-D welds for the fourth inservice inspection interval in 2023 in lieu of 2013. These dates are a deviation of those provided in OG-06-356 (Reference 2) but are consistent with those in the revised implementation plan OG-09-454 (Reference 3).
In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on the basis that the current interval can be extended based on a negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 4).
The methodology used to conduct this analysis is based on that defined in the study WCAP-16168-NP-A, Revision 2, "Risk Informed Extension of the Reactor Vessel In-service Inspection Interval" (Reference 5). This study focuses on risk assessments of materials within the beltline region of the RV wall. The results of the calculations for Surry Unit 1 (Reference 6) were compared to those obtained from the Westinghouse pilot plant evaluated in WCAP-16168-NP-A, Revision 2.
The parameters for Surry Unit 1 are bounded by the results of the Westinghouse pilot plant, which qualifies Surry Unit 1 for an ISI interval extension.
Table 1 below lists the critical parameters investigated in the WCAP and compares the results of the Westinghouse pilot plant to those of Surry Unit 1. Tables 2 and 3 provide additional information that was requested by the NRC and included in Appendix A of Reference 5.
Serial No.12-267 Docket Nos. 50-280/281 Page 3 of 6 Table 1:
Critical Parameters for the Application of Bounding Analysis for Surry Unit 1 Additional Pilot Plant Plant-Specific Evaluation Parameter Basis Basis Required?
Dominant Pressurized Thermal NRC PTS Risk Study PTS Generalization No Shock (PTS) Transients in the NRC (Reference 7)
Study PTS Risk Study are Applicable (Reference 8)
Through-Wall Cracking Frequency 1.76E-8 Events per 1.41 E-8 Events per No (TWCF)
Year (Reference 5)
Year (Calculated per Reference 5)
Frequency and Severity of Design 7 Heatup/Cooldown Bounded by 7 No Basis Transients Cycles per Year Heatup/Cooldown (Reference 5)
Cycles per Year Cladding Layers (Single/Multiple)
Single Layer Single Layer No
_ (Reference 5)
SingleLayer Table 2 below provides a summary of the latest RV inspection for Surry Unit 1 and evaluation of the recorded indications (Surry Unit 1 last inspected between November 10, 2004 and November 17, 2004).
This information confirms that satisfactory examinations have been performed on the Surry Unit 1 RV.
Table 2:
Additional Information Pertaining to Reactor Vessel Inspection for Surry Unit I Inspection methodology:
The latest 10-Year RPV ISI was conducted in accordance with the ASME Code, Section Xl and Section V, 1989 Edition.
Examinations of Category B-A and B-D welds were performed to ASME Section Xl Appendix VIII 1995 Edition with the 1996 Addenda, as modified by 10 CFR 50.55a(b)(2)(xiv, xv, and xvi). Future inservice inspections will be performed to ASME Section Xl Appendix VIII requirements.
Number of past Three 10-Year inservice inspections have been performed.
inspections:
Number of indications There were six indications identified in the beltline region during the most found:
recent inservice inspection. These indications are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. None of these indications are within the inner 1/101" or 1" of the reactor vessel thickness.
Therefore, no further evaluation is required and these indications are allowable per the requirements of the Alternate PTS
- Rule, 10 CFR 50.61a (Reference 9).
Proposed inspection The fourth inservice inspection is scheduled for 2013. This inspection will schedule for balance of be performed in 2023. These dates are a deviation of those provided in plant life:
OG-06-356 (Reference 2) but are consistent with those in the revised implementation plan OG-09-454 (Reference 3).
Serial No.12-267 Docket Nos. 50-280/281 Page 4 of 6 Table 3 summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TVVCF).
Table 3: Details of TWCF Calculation for Surry Unit I at 48 Effective Full Power Years (EFPY)
Inputs Reactor Coolant System Temperature, TRcs[*F]:
N/A Twaoi (inches]:
8.209 Fluence No Region and Material Heat Cu(1) Ni(1) R.G.
CF (1 )
(1)
[1019 No Component Maera 99HeatTDTu()
etonC cmptionet No.
[wt%]
[wt%]
Pos.
[OF]
[OF]
E> 1.0 Description Ps MeV]
1 Inter. Shell Long.
SA-0.16 0.57 1.1 143.9
-5 0.897 Weld L3 1494/8T1 554 2
Inter. Shell Long.
SA-0.16 0.57 1.1 143.9
-5 0.897 Weld L4 1494/8T1554 1
3 Lower Shell Long.
SA-0.16 0.57 1.1 143.9
-5 0.897 Weld Li 1494/8T1554 4
Lower Shell Long.
SA-0.34 0.68 1.1 220.6
-5 0.897 Weld L2 1526/299L44 Inter. To Lower Shell SA-1585/72445 022 0.54 1.1 158.0
-5 4.51 Circ. Weld W05 SA-1650/72445 6
Intermediate Shell C4326-1 0.11 0.55 1.1 73.5 10 4.51 7
Intermediate Shell C4326-2 0.11 0.55 1.1 73.5 0
4.51 8
Lower Shell C4415-1 0.11 0.50 2.1 85.0 20 4.51 9
Lower Shell C4415-2 0.11 0.50 1.1 73.0 0
4.51 Outputs Methodology Used to Calculate AT 30:
Regulatory Guide 1.99, Revision 2(2)
Controlling Fluence Material RTMA.XX
[1019 FF AT Region No.
Neutron/cm2, (Fluence T
TWCF 9 5.XX (From
[OR]
E > 1.0 Factor)
[OF]
Above)
MeV]
Limiting Axial Weld - AW 4
666.55 0.897 0.970 213.88 6.26E-09 Limiting Plate - PL 8
597.10 4.51 1.382 117.43 1.51E-12 Circumferential Weld -
672.96 4.51 1.382 218.29 6.05E-13 CW TWCF95_TOTAL(aAwTWCF95.AW + aPLTWCF95.PL + acwTWCF 9 5.CW):
1.41 E-08 (1) Reference 10 (2) Reference 11
Serial No.12-267 Docket Nos. 50-280/281 Page 5 of 6
6. Duration of Proposed Alternative
This request is applicable to the Surry Unit 1 inservice inspection program for the fourth and fifth 10-year inspection intervals.
7. Precedents
Several requests for relief on this subject have been granted, including:
" Joseph M. Farley Nuclear Plant, Unit 2 (Farley Unit 2) - Relief Request for Extension of the Reactor Vessel Inservice Inspection Date to the Year 2020 (Plus or Minus One Outage) (TAC No. ME30 10), July 12, 2010 McGuire Nuclear Station, Unit 1 (McGuire 1) - Relief Request 09-MN-003 for Extension of the Reactor Vessel Inservice Inspection (ISI) Date to the Year 2020 (Plus or Minus One Outage) (TAC No. ME1822) June 28, 2010 R. E. Ginna Nuclear Power Plant: Safety Evaluation for Relief Request No. 18, Reactor Vessel Weld Examination Extension (TAC No. MD9962) July 31, 2009
- 8. References
- 1.
ASME Boiler and Pressure Vessel Code, Section Xl, 1998 Edition with the 2000 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.
OG-09-454, "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,"
December 1, 2009.
- 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.
- 5.
WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval," June 2008.
- 6.
CN-AMLRS-10-3, Revision 1, "Implementation of WCAP-16168-NP-A, Revision 2, for Surry Units 1 and 2."
Serial No.12-267 Docket Nos. 50-280/281 Page 6 of 6
- 7.
NUREG-1874, 10 CFR Part 50.61a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," 10/3/07 (ADAMS Accession Number ML070860156).
- 8.
NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004 (ADAMS Accession Number ML042880482).
- 9.
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 and No. 22 with corrections to part (g) dated February 3, 2010.
- 10. WCAP-15130, Revision 1, "Surry Units 1 and 2 WOG Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldown Limit Curves for Normal Operation,"
April 2001.
- 11.
NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", May 1988.
Serial No.12-267 Docket Nos. 50-280/281 Implementation of Extended Reactor Vessel Inservice Inspection Interval -
Relief Request CMP-009 Virginia Electric and Power Company (Dominion)
Surry Power Station Unit 2
Serial No.12-267 Docket Nos. 50-280/281 Page 1 of 7 Implementation of Extended Reactor Vessel Inservice Inspection Interval for Surry Unit 2 CMP-009 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 The affected component is the Surry Unit 2 reactor vessel (RV), specifically the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI (Reference 1) examination categories and item numbers covering examinations of the RV. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code Section XI.
Examination Category B-A B-A B-A B-A B-A B-A B-A B-A B-D B-D Item No.
B1.11 B1.12 B1.21 B1.22 B1.30 B1.40 B1.50 B1.51 B3.90 B3.100 Description Circumferential Shell Welds Longitudinal Shell Welds Circumferential Head Welds Meridional Shell Welds Shell-to-Flange Weld Head-to-Flange Weld Repair Weld Beltline Region Nozzle-to-Vessel Welds Nozzle Inside Radius Section (Throughout this request the above examination categories are referred to as "the subject examinations" and the ASME BPV Code, Section Xl, is referred to as "the Code.")
2. Applicable Code Edition and Addenda
ASME Code Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components," Code 1998 Edition with the 2000 Addenda
Serial No.12-267 Docket Nos. 50-280/281 Page 2 of 7
3. Applicable Code Requirement
IWB-2412, Inspection Program B, requires volumetric examination of 100% of RV pressure retaining welds identified in Table IWB-2500-1 once each ten year interval.
The Surry Unit 2 fourth 10-year inservice inspection (ISI) interval ends in 2014.
4. Reason for Request
An alternative is requested from the requirement of IWB-2412, Inspection Program B, that volumetric examination of 100%
of RV pressure retaining, Examination Category B-A and B-D welds, be performed once each ten-year interval. Extension of the interval between examinations of Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in man-rem exposure and examination costs.
5. Proposed Alternative and Basis for Use
Dominion proposes to perform the ASME Code required volumetric examination of the Surry Unit 2 RV full penetration pressure retaining Examination Category B-A and B-D welds for the fourth inservice inspection interval in 2024 in lieu of 2014. These dates are a deviation of those provided in OG-06-356 (Reference 2) but are consistent with those in the revised implementation plan OG-09-454 (Reference 3).
In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on the basis that the current interval can be extended based on a negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 4).
The methodology used to conduct this analysis is based on that defined in the study WCAP-16168-NP-A, Revision 2, "Risk Informed Extension of the Reactor Vessel In-service Inspection Interval" (Reference 5). This study focuses on risk assessments of materials within the beltline region of the RV wall. The results of the calculations for Surry Unit 2 (Reference 6) were compared to those obtained from the Westinghouse pilot plant evaluated in WCAP-16168-NP-A, Revision 2.
The parameters for Surry Unit 2 are bounded by the results of the Westinghouse pilot plant, which qualifies Surry Unit 2 for an ISI interval extension.
Table 1 below lists the critical parameters investigated in the WCAP and compares the results of the Westinghouse pilot plant to those of Surry Unit 2. Tables 2 and 3 provide additional information that was requested by the NRC and included in Appendix A of Reference 5.
Serial No.12-267 Docket Nos. 50-280/281 Page 3 of 7 Table 1:
Critical Parameters for the Application of Bounding Analysis for Surry Unit 2 Additional Pilot Plant Plant-Specific Evaluation Parameter Basis Basis Required?
Dominant Pressurized Thermal NRC PTS Risk Study PTS Generalization No Shock (PTS) Transients in the NRC (Reference 7)
Study (Reference 8)
PTS Risk Study are Applicable Through-Wall Cracking Frequency 1.76E-8 Events per 1.17E-12 Events per No (TWCF)
Year (Reference 5)
Year (Calculated per Reference 5)
Frequency and Severity of Design 7 Heatup/Cooldown Bounded by 7 No Basis Transients Cycles per Year Heatup/Cooldown (Reference 5)
Cycles per Year Cladding Layers (Single/Multiple)
Single Layer No (Reference 5)
Single Layer Table 2 below provides a summary of the latest RV inspection for Surry Unit 2 and evaluation of the recorded indications (Surry Unit 2 last inspected between May 4, 2005 and May 9, 2005). This information confirms that satisfactory examinations have been performed on the Surry Unit 2 RV.
Serial No.12-267 Docket Nos. 50-280/281 Page 4 of 7 Table 2:
Additional Information Pertaining to Reactor Vessel Inspection for Surry Unit 2 Inspection methodology:
The latest 10-Year RPV ISI was conducted in accordance with the ASME Code, Section Xl and Section V, 1989 Edition.
Examinations of Category B-A and B-D welds were performed to ASME Section Xl Appendix VIII 1995 Edition with the 1996 Addenda, as modified by 10 CFR 50.55a(b)(2)(xiv, xv, and xvi). Future inservice inspections will be performed to ASME Section XI Appendix VIII requirements.
Number of past Three 10-Year inservice inspections have been performed.
inspections:
Number of indications found:
There were five indications identified in the beltline region during the most recent inservice inspection. These indications are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. Four of these indications are within the inner 1 /1 0 1h or 1" of the reactor vessel thickness and are acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 9) since the number of actual flaws is less than the allowable number of flaws for each flaw size increment. A disposition of these four flaws against the limits of the Alternate PTS Rule is shown in the tables below. The following indications are located within the weld material of the RV beltline.
Through-Wall Scaled Number of Extent, TWE (in.)
maximum Flaws number of (Axial/
TWEMIN TWEMAx flaws Circ.)
0.125 1 0.475 71 1
2(0/2)
The following indications are located within the plate material of the RV beltline.
Through-Wall Scaled Number of Extent, TWE (in.)
maximum Flaws number of (Axial/
TWEMIN TWEMAx flaws Circ.)
0.125 0.375 19
]
2(0/2) 0.175 0.375 6
1 (0/1)
Proposed inspection The fourth inservice inspection is scheduled for 2014. This inspection will schedule for balance of be performed in 2024. These dates are a deviation of those provided in plant life:
OG-06-356 (Reference 2) but are consistent with those in the revised implementation plan OG-09-454 (Reference 3).
Serial No.12-267 Docket Nos. 50-280/281 Page 5 of 7 Table 3 summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF).
Table 3: Details of TWCF Calculation for Surry Unit 2 at 48 Effective Full Power Years (EFPY)
Inputs Reactor Coolant System Temperature, TRCs[°F]:
N/A Twai [inches]:
8.209 Region and Material Heat Cu'
)
Ni(1)
R.G.
CF(
R
- 1) Fluence [1019 Component 1
99 RTNDT(u)
Neutron/cm 2, Description No.
[wt%]
[wt%]
Pos.
[°F]
E[> 1.0 MeV 1
Inter. Shell Long.
SA-0.22 0.54 1.1 158.0
-5 0.940 Weld L3 1585/72445 2
Inter. Shell Long.
SA-0(2)0.940 2__
Weld L4(2) 1585/72445 0.22 0.54 1.1 158.0(.9 3
Lower Shell Long.
WF-4/8T1762 0.19 0.57 1.1 152.4
-5 0.940 Weld Li 4
Lower Shell Long.(2 LoWeld L2(2)
WF-4/8T1762 0.19 0.57 1.1 152.4(2)
-5 0.940 5
Inter. To Lower Shell R3008/0227 0.19 0.55 1.1 149.3 0
4.50 Circ. Weld W05 6
Intermediate Shell C4331-2 0.12 0.60 1.1 83.0
-10 4.50 7
Intermediate Shell C4339-2 0.11 0.54 1.1 73.4
-20 4.50 8
Lower Shell C4208-2 0.15 0.55 1.1 107.3
-30 4.50 9
Lower Shell C4339-1 0.11 0.54 1.1 73.4
-10 4.50 Outputs Methodology Used to Calculate AT 30:
Regulatory Guide 1.99, Revision 2(3)
Controlling Fluence Material
[1019 FF Region No.
Neutron/cm 2, (Fluence AT30 TWCF 95-XX (From
[OR]
E > 1.0 Factor)
[OF]
Above)
MeV]
Limiting Axial Weld - AW 1 and 2 609.93 0.940 0.983 155.26 0.OOE+00 Limiting Plate - PL 8
577.86 4.50 1.381 148.19 3.08E-13 Circumferential Weld -
665.87 4.50 1.381 206.20 1.80E-13 CW TWCF95.TOTAL(aAWTWCF95-AW + CIPLTWVCF95-PL + acwTWCF95gcw):
1.17E-12 (1) Reference 10 (2) Weld contains two different materials. The material with the most limiting properties was used for this evaluation.
(3) Reference 11
Serial No.12-267 Docket Nos. 50-280/281 Page 6 of 7
6. Duration of Proposed Alternative
This request is applicable to the Surry Unit 2 inservice inspection program for the fourth and fifth 10-year inspection intervals.
7. Precedents
Several requests for relief on this subject have been granted, including:
" Joseph M. Farley Nuclear Plant, Unit 2 (Farley Unit 2) - Relief Request for Extension of the Reactor Vessel Inservice Inspection Date to the Year 2020 (Plus or Minus One Outage) (TAC No. ME30 10), July 12, 2010
" McGuire Nuclear Station, Unit 1 (McGuire 1) - Relief Request 09-MN-003 for Extension of the Reactor Vessel Inservice Inspection (ISI) Date to the Year 2020 (Plus or Minus One Outage) (TAC No. ME1822) June 28, 2010
" R. E. Ginna Nuclear Power Plant: Safety Evaluation for Relief Request No. 18, Reactor Vessel Weld Examination Extension (TAC No. MD9962) July 31, 2009
- 8. References
- 1.
ASME Boiler and Pressure Vessel Code, Section Xl, 1998 Edition with the 2000 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.
OG-09-454, "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,"
December 1, 2009.
- 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.
- 5.
WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval," June 2008.
- 6.
CN-AMLRS-10-3, Revision 1, "Implementation of WCAP-16168-NP-A, Revision 2, for Surry Units 1 and 2."
Serial No.12-267 Docket Nos. 50-280/281 Page 7 of 7
- 7.
NUREG-1874, 10 CFR Part 50.61a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," 10/3/07 (ADAMS Accession Number ML070860156).
- 8.
NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004 (ADAMS Accession Number ML042880482).
- 9.
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 and No. 22 with corrections to part (g) dated February 3, 2010.
- 10. WCAP-15130, Revision 1, "Surry Units 1 and 2 WOG Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldown Limit Curves for Normal Operation,"
April 2001.
- 11.
NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", May 1988.