ML18152A267
| ML18152A267 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna  |
| Issue date: | 04/01/1996 |
| From: | Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-REGGD-01.099, RTR-REGGD-1.099 96-084, 96-84, NUDOCS 9604040120 | |
| Download: ML18152A267 (23) | |
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VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 April 1, 1996 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY Serial No.
NL&OS/GDM Docket Nos.
License Nos.96-084 RO 50-280, 50-281 50-338, 50-339 DPR-32, DPR-37 NPF-4, NPF-7 SURRY AND NORTH ANNA POWER STATIONS UNITS 1 AND 2 PRESSURIZED THERMAL SHOCK (PTS) SCREENING CALCULATIONS In our letter dated June 8, 1995 (Serial No.95-197), we submitted a request for revised Technical Specifications heatup and cooldown curves and Low Temperature Overpressure Protection System (L TOPS) setpoints that were valid to the end-of-license for Surry Units 1 and 2. During the NRC's review of this submittal, the staff determined that an outstanding issue existed regarding the currently docketed PTS reference temperature {RTPTS) data for limiting reactor vessel beltline materials previously provided in our letter dated December 10, 1991 (Serial No.91-328). This data had been provided in response to the revision to 1 O CFR 50.61 PTS rule.
Specifically, we indicated in our letter that:
... none of the revised RTPTS values exceed the applicable screening criterion prior to end-of-license with the exception of that of the Surry Unit 1 Lower Shell Longitudinal Weld L2.
As indicated in previous correspondence (Letters Serial Nos. 89-7 48 dated December 1, 1989,90-335 dated July 30, 1990, and 91-374 dated July 8, 1991), we are in the process of implementing a flux reduction program at Surry 1. As noted in our July 8, 1991 letter, flux suppression inserts (FSls) are planned to be installed in Surry Unit 1 during Cycle 13. The target fluence in this program is well below that which could cause RT PTS to exceed the screening criterion at EOL. We consider this plan adequate to ensure that the requirements of 10 CFR 50.61 will continue to be met throughout the operating life of the plant.
The NRC has requested that Virginia Electric and Power Company provide confirmation of Surry's compliance with the requirements of 1 O CFR 50.61 throughout the current license period. In response to this request, we have prepared revised PTS screening calculations for Surry Units 1 and 2. These calculations take credit for available plaht-specific and B&W Owners Group Master Integrated Reactor Vessel Materials Surveillance Program (MIRVSP) surveillance data, and utilize best-estimate end-of-license neutron fluence values. No credit is taken for the installation of flux suppression inserts in Unit 1 beginning with Cycle 13.
9604040120 96040_1 _________ \\
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- Because the December 10, 1991 submittal applied to North. Anna as well as Surry, revised calculations have also been prepared for North Anna Units 1 and 2.
A discussion of the development of the PTS screening calculations for Surry and North Anna is provided in Attachment 1. A summary of surveillance capsule analysis results and revised PTS screening calculations for North Anna and Surry are provided in Attachments 2 and 3, respectively, and constitute our revised licensing basis.
The revised RTPTS values are below the applicable screening criteria, and therefore ensure that the requirements of 10 CFR 50.61 will continue to be met throughout the license periods for both stations.
If you have any questions or require further information, please contact us.
~?o;Ja-L James P. OHanlon Senior Vice-President - Nuclear Attachments cc:
U.S. Nuclear Regulatory Commission Region 11 101 Marietta Street, N. W.
Suite 2900 Atlanta, Georgia 30323 Mr. M. W. Branch NRC Senior Resident Inspector Surry Power Station Mr. R. P. Mcwhorter NRC Senior Resident Inspector North Anna Power Station
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e Discussion of Pressurized Thermal Shock Screening Calculations Surry and North Anna Power Stations Units 1 and 2
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e 1."0 BACKGROUND Since our December 1991 Pressurized Thermal Shock {PTS} submittal (2), several reports have been prepared and submitted to the NRC which pertain to Surry's compliance with the requirements* of 1 O CFR 50.61. However, none of these recent submittals constitutes a revised 1 O CFR 50.61 licensing basis:
- 1.
BAW-2166 (3) provides the response to Generic Letter (GL) 92-01, Revision 1 (6).
- 2.
BAW-2222 (5) provides the response to an NRC request for additional information (RAI) on Virginia Power's response to GL 92-01.
- 3.
BAW-2257, Revision 1 (7) provides the response to GL 92-01, Supplement 1 (8) for Linde 80 weld materials.
- 4.
BAW-2260 (11) presents the response to Supplement 1 (8) for the Rotterdam weld materials in the Surry Units 1 and 2 reactor vessels.
Virginia Power docketed these reports by letters dated June 29, 1992 (27), June 30, 1994 (9) and--November 20, 1995 (10). Because it contained no reactor vessel beltline fluence estimates, the information in BAW-2166 (3) was insufficient to demonstrate compliance with the requirements of 1 O CFR 50.61. Although BAW-2222 (5) and BAW-2260 (11) included all requisite inputs for demonstration of compliance with 10 CFR 50.61, revised RTPTS values were not documented in these reports. Revised RTPTS values were determined in BAW-2257, Revision 1 (7) for Linde 80 weld materials. However, BAW-2257, Revision 1 *declared that the Regulatory Guide (RG) --
1.99, Revision 2, Position 2.1 "ratio procedure" utilized in the report is unnecessary for -
the Linde 80. class of welds, and that previously submitted reactor vessel integrity evaluations (i.e., Reference (2)) remain valid. Therefore, the values presented in BAW-2257, Revision 1 do not constitute a revised 1 O CFR 50.61 licensing basis for Surry Units 1 and 2.
Unlike the Surry responses to GL 92-01, the North Anna response to GL 92-01, Supplement 1 (11) established a revised 1 O CFR 50.61 licensing basis for both North Anna Units. Specifically, Reference (10) concluded that the RTPTS value for the limiting North Anna Units 1 and 2 beltline material increased from 227.?°F to 238.9°F.
However, like the calculations.which support the Reference (2) PTS submittal, the calculations which support this RTPTS value (10) did not take credit for available surveillance data, and utilized unnecessarily conservative neutron fluence values. presents summaries of the surveillance capsule analysis results used in chemistry factor calculations. Attachment 2 presents the results of revised PTS screening calculations for Surry and North Anna Units 1 and 2. Detailed explanations of the chemistry factor and PTS screening calculations accompany the calculation summaries.
1
l 2.0 CONSIDERATION OF AN ALTERNATE MEASURED VALUE OF INITIAL RTN DT FOR LINDE 80 WELDS A key input to the determination of RTPTS is the initial (or unirradiated) value of RTNDT* The B&W Owners Group (B&WOG) Reactor Vessel Working Group has prepared a topical report (16) which concludes that a maximum (upper-bound) unirradiated RTNDT of -2TF is justified for all Linde 80 weld materials. Although the method used to determine this unirradiated RTNDT value departs from the method currently prescribed by Paragraph NB-2331 of Section Ill of the ASME Boiler and Pressure Vessel Code (BPVC), the method was licensed for the Linde 80 weld material WF-70. In BAW-2245, Revision 1 (16), the applicability of the unirradiated RTNDT value determined for WF-70 is extended to all Linde 80 weld materials.
The method for determining the unirradiated RTNDT prescribed by ASME Section Ill, Paragraph NB-2331, is referenced in 1 O CFR 50.61. In order for Virginia* Power to utilize the alternate method, an exemption to the requirements of 1 O CFR 50.61 per 1 O CFR 50.12 is required. Because compliance with the requirements of 1 O CFR 50.61 may be demonstrated without the margins provided by BAW-2245, Revision 1 (16),
Virginia Electric and Power Company will not pursue an exemption at this time.
3.0 CONSIDERATION OF THE RG 1.99, REVISION 2, POSITION 2.1 RATIO PROCEDURE The fracture toughness of reactor pressure vessel steel is indexed to the nil-ductility transition reference temperature (RTNDT). Regulatory Guide (RG) 1.99, Revision 2 (17) provides guidance for determining RTN DT.
Because the mean chemical composition (copper and nickel) of surveillance materials may differ from the mean chemical composition of beltline materials, Paragraph 2.1 of RG 1.99, Revision 2 requires measured values of ~RTNDT (i.e., surveillance capsule analysis results) to be multiplied by the ratio of the respective chemistry factors for the surveillance and beltline materials. This "ratio procedure" adjusts observed transition temperature shifts in an effort to make surveillance capsule analysis results indicative of actual reactor vessel beltline embrittlement.
As reported in Reference (7), the B&WOG Reactor Vessel Working Group (RVWG) believes that differences between the chemical compositions of Linde 80 surveillance and beltline weld materials are fully accommodated by the Position 2.1 method without application of the ratio procedure specified therein. Specifically, the RVWG contends that the mean chemical composition of a beltline weld material (i.e., weld heat) is represented by the mean chemical composition of surveillance and original-fabrication test samples.
Therefore, when measured values of ~RTN DT for two or more surveillance capsule results are evaluated in accordance with Position 2.1 without application of the ratio procedure, the resulting calculated chemistry factor is an unbiased estimate of the chemistry factor for the beltline material. Similarly, the standard deviation of the copper and nickel concentrations of a beltline weld material are represented by the standard deviation of copper and nickel concentrations of surveillance and original-fabrication test samples. Therefore, the ratio procedure need 2
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only be applied when there is "clear evidence that the copper or nickel content of the surveillance weld differs from that of the vessel weld" (17). No such evidence exists for Linde 80 welds.
Similarly, there is no clear evidence that the copper or nickel content of the North Anna and Surry Rotterdam surveillance welds differ from those of the corresponding vessel welds. Only minimal differences in mean chemical composition, attributable to random variation, are observed in these materials.
Furthermore, as documented in the Reference (11) response to GL 92-01, Revision 1, Supplement 1, application of the ratio procedure to North Anna beltline welds or Surry Rotterdam welds would result in less limiting calculated values of RTPTS* Therefore, the ratio procedure is not applied to North Anna beltline weld materials, nor to Surry weld materials fabricated by Rotterdam.
4.0 DETERMINATION OF CHEMISTRY FACTORS USING CREDIBLE SURVEILLANCE DATA When two or more credible surveillance data sets are available, they may be used to determine the chemistry factor of the corresponding beltline materials in accordance with Position 2.1 of Regulatory Guide 1.99, Revision 2, Position 2.1. Attachment 1 presents a list of the surveillance capsule analysis results used in Position 2.1 chemistry factor calculations. Attachment 1 also presents a description of the method used to calculate chemistry factors using surveillance data, and documents the determination of credibility of the constituent surveillance data sets.
5.0 PTS SCREENING CALCULATIONS The results of PTS screening calculations for Surry Units 1 and 2 are presented in Calculations 1 through 22 of Attachment 2. The results for North Anna Units 1 and 2 are presented in Calculations 23 through 34 of Attachment 2. With the exception of Calculations 7A and 11A, all calculations are presented as proposed 10 CFR 50.61 licensing basis calculations. Calculations 7A and 11A. utilize a revised unirradiated RT NOT value based on transition range fracture toughness testing (16), and are presented for information only.
As the Attachment 2 calculations demonstrate, all Surry and North Anna reactor vessel beltline materials meet the PTS screening criteria prescribed by 1 O CFR 50.61 throughout the current license period. The limiting Surry Units 1 and 2 material (SA-1526) exhibits 8% margin to the applicable screening criterion without consideration of the revised -27°F initial RTNOT value based on transition region fracture toughness testing (16). If the revised initial RT NOT value is considered, the margin to the applicable screening criterion for this material is increased to 19%. The limiting North Anna material exhibits 18% margin to the applicable screening criterion.
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SUMMARY
AND CONCLUSIONS A revised set of PTS screening calculations has been prepared. Although the currently docketed screening calculations for Surry and North Anna Units 1 and 2 (2) took no credit for available surveillance data, the revised calculations utilize the most recently obtained plant-specific and integrated surveillance program material properties data.
Revised best-estimate chemical compositions were determined for several North Anna Units 1 and 2 beltline materials in the North Anna response to GL 92-01, Revision 1, Supplement 1 documented in BAW-2260 (11 ). The revised calculations use the best-estimate chemical compositions most recently determined (or re-confirmed) in BAW-2257, Revision 1 (7) and BAW-2260 (11).
The Reference (2) PTS submittal utilized the more conservative "design basis" fluence values from the most recentsurveillance capsule analyses for Surry Unit 1 (12), Surry Unit 2 (13), North Anna Unit 1 (14), and North Anna Unit 2 (15).
The revised calculations utilize the best-estimate fluence values from References (12), (13), (14),
and (15).
A revised -27°F initial RTNDT value based on transition region fracture toughness testing applicable to Linde 80 welds was determined in Reference (16). This value was applied to the most limiting Linde 80 weld materials to determine the reduction in the calculated RTN DT, and in the applied margin term. These calculations are submitted for information only.
.
- BAW-2257, Revision 1 (7) dismisses the need to apply the Regulatory G_uide.1.99, Revision 2, Position* 2.1 ratio procedure to Linde 80 weld materials. Reference (7) contends that differences between the chemical compositions of Linde 80 surveillance and beltline weld materials are fully accommodated by the Position 2.1 method without application of the ratio procedure specified therein. Moreover, there is no clear evidence that the copper or nickel content of the surveillance weld differs from that of the vessel weld.
Similarly, there is no clear evidence that the copper or nickel content of the North Anna and Surry Rotterdam surveillance welds differ from those of the corresponding vessel welds. Furthermore, application of the ratio procedure to North Anna beltline welds or Surry Rotterdam welds would result in less limiting calculated values of RTPTS (11 ).
Therefore, the ratio procedure is not applied to North Anna and Surry Rotterdam weld materials.
All Surry and North Anna Units.1 and 2 reactor vessel beltline materials meet the 1 O CFR 50.61 screening criteria throughout the current license periods..
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7.0 (1)
(2)
(3)
(4)
(5)
(6)
(7)
(8)
(9)
(10) e REFERENCES Letter from J. P. O'Hanlon to USNRC, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Request for Exemption - ASME Code Case N-514, Proposed Technical Specifications
- Change, Revised Pressureffemperature Limits and L TOPS Setpoint," NRC Letter 95-197, dated June 8, 1995.
Letter from W. L. Stewart to USNRC, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2, Revision to 1 O CFR 50.61 Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Serial No.91-328, dated December 10, 1991.
"B&W Owners Group Response to Generic Letter 92-01," BAW-2166, dated June, 1992.
"North Anna Units 1 and 2 Response to Closure Letter for NRC Generic Letter 92-01, Revision 1," BAW-2224, dated July, 1994.
"Reactor Vessel Working Group Response to Closure Letters to NRC Generic Letter 92-01, Revision 1," BAW-2222, dated June, 1994.
Letter from J. G. Partlow (USNRC) to All Holders of Operating Licenses or Construction Permits for-Nuclear Power Plants (Except Yankee Atomic Electric Company), "Reactor Vessel Structural Integrity, 1 O CFR 50.54(f) (Generic Letter 92-01, Revision 1 ), " dated March 6, 1992.
"B&W Owners Group Reactor Vessel Working Group Response to Generic Letter 92-01, Revision 1, Supplement 1," BAW-2257, Revision 1, dated October, 1995.
Letter from R. P. Zimmerman (USNRC) to All Holders of Operating Licenses or Construction Permits for Nuclear Power Reactors, "NRC Generic Letter 92-01, Revision 1, Supplement 1: Reactor Vessel Structural Integrity," dated May 19, 1995 (Virginia Power Serial No.95-270).
Letter from J. P. O'Hanlon to USNRC, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity, Request for Additional Information," Serial No.94-342, dated June 30, 1994.
Letter from J. P. O'Hanlon to USNRC, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2, Six-Month Response to Generic Letter 92-01, Revision 1, Supplement 1, Reactor Vessel Structural Integrity," Serial No. 95-270A, dated November 20, 1995.
5
(11)
"Response to Generic Letter 92-01, Revision 1, Supplement 1, for Virginia Power's North Anna Units 1 and 2 Beltline Materials, and Surry Units 1 and 2 Rotterdam Beltline Materials," BAW-2260, dated October, 1995.
(12)
"Analysis of Capsule V from the Virginia Electric and Power Company Surry Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-11415, Revision 0, dated February, 1987.
(13)
"Analysis of Capsule V from the Virginia Electric and Power Company Surry Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-11499, Revision 0, dated June, 1987.
(14)
"Analysis of Capsule U from the Virginia Electric and Power Company North Anna Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-11777, dated February, 1988.
(15)
"Analysis of Capsule U from the Virginia Electric and Power Company North Anna Unit 2,-Reactor Vessel Radiation Surveillance Program," WCAP-12497, dated January, 1990.
(16)-
"Initial RTN DT of Linde 80 Welds Based on Fracture Toughness in the Transition Range," BAW-2245, Revision 1, dated October 1995.
(17)
"Radiation Embrittlement of Reactor Vessel Materials," Regulatory Guide 1.99, Revision 2, dated May, 1988.
(18)
"Surry Unit 1 Reactor Vessel Radiation Surveillance Program.," WCAP-7723.
(19)
"Surry Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-8085.
(20)
"Virginia Electric and Power Company North Anna Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-8771, dated September, 1976.
(21)
"Virginia Electric and Power Company North Anna Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-8772, dated November, 1974.
(22)
"Master Integrated Reactor Vessel Surveillance Program," BAW-1543, Revision 4, dated February 1993. See also Supplement 1, dated February 1993.
(23)
"Surry Units 1 and 2 Reactor Vessel Fluence and RT{PTS} Evaluations," WCAP-11015, Revision 1, dated April, 1987.
(24}
"Surry Units 1 and 2 Reactor Vessel Fluence and RT{PTS) Evaluations for Consideration of Life Extension," WCAP-11017, Revision 1, dated April, 1987.
(25)
"North Anna Units 1 and 2 Reactor Vessel Fluence and RT{PTS} Evaluations,"
WCAP-11016, Revision 3, dated January, 1988.
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"North Anna Unit 1 Surveillance Capsule Withdrawal Schedule, dated July 1993, Virginia Power Contract ER-Ml2002, Westinghouse G.O. RM30416; Attachment to VRA-93-107.
(27)
Letter from W. L. Stewart to USNRC, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Response to Generic Letter 92-01, Reactor Vessel Structural Integrity," Serial No.92-211, dated June 29, 1992.
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Summary of Surveillance Capsule Analysis Results Used in RG 1.99, Revision 2, Position 2.1 Chemistry Factor Calculations Surry and North Anna Units 1 and 2
Surveillance I Material I Capsule ID Surry 1 / T Forging C4415-1 Surry 1 / V Forging C4415-1 TMI 2 / LG1 WF-25 TMI 2/LG1 SA-1526 CR 3 / LG1 WF-25 TMI 1 / E WF-25 TMI 1 /C WF-25 Surry 1 IT SA-1526 Surry 1 / V SA-1526 CR3-LG1 SA-1585 CR3-LG2 SA-1585 PB1-V SA-1263 PB1-S SA-1263 PB1-R SA-1263 PB1-T SA-1263 Surry 2 / X Forging C4339-1 Surry 2 / V Forging C4339-1 Surry2/X R3008 Surry 2 / V R3008 Surveillance Capsules Used in Position 2 RT(PTS) Calculations Surry Units 1 and 2 Applicable to I
Capsule I Measured I Measured - T (Beltline Material ID)
Fluence (E19)
Shift (F)
Mean RTNDT Reference "Analysis of Capsule V from the Virginia Electric and Power Forging C4415-1 (Surry 1) 0.281 50
-8 Company Surry Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-11415, Revision 0, dated February, 1987.
Forging C4415-1 (Surry 1) 1.940 110 5
WCAP-11415, Revision O (Complete reference presented above.)
"B&W Owners Group Reactor Vessel Working Group Response SA-1526 (Surry 1) 0.968 222 7
to Generic Letter 92-01, Revision 1, Supplement 1," BAW-2257, Revision 1 dated October 1995.
SA-1526 (Surry 1) 0.830 182
-24 BAW-2257, Revision 1 (Complete reference presented above)
SA-1526 (Surry 1) 0.779 214 12 BAW-2257, Revision 1 (Complete reference presented above)
SA-1526 (Surry 1) 0.107 124 31 BAW-2257, Revision 1 (Complete reference presented above)
SA-1526 (Surry 1) 0.866 203
-5 BAW-2257, Revision 1 (Complete reference presented above)
SA-1526 (Surry 1) 0.281 165 23 WCAP-11415, Revision O (Complete reference presented above.)
SA-1526 (Surry 1) 1.940 240
-16 WCAP-11415, Revision O (Complete reference presented above.)
SA-1585 (Surry 1 & 2) 0.510 148 26 BAW-2257, Revision 1 (Complete reference presented above)
SA-1585 (Surry 1 & 2) 1.670 168
-3 BAW-2257, Revision 1 (Complete reference presented above)
SA-1585 (Surry 1 & 2) 0.502 110
-11 BAW-2257, Revision 1 (Complete reference presented above)
SA-1585 (Surry 1 & 2) 0.829 165 23 BAW-2257, Revision 1 (Complete reference presented above)
SA-1585 (Surry 1 & 2) 2.380 165
-20 BAW-2257, Revision 1 (Complete reference presented above)
SA-1585 (Surry 1 & 2) 2.420 180
-5 BAW-2257, Revision 1 (Complete reference presented above)
"Analysis of Capsule V from the Virginia Electric and Power Forging C4339-1 (Surry 2) 0.302 55 9
Company Surry Unit 2 Reactor Vessel Radiation Surveillance Proqram," WCAP-11499, Revision 0, dated June, 1987.
Forging C4339-1 (Surry 2) 1.880 75
-5 WCAP-11499, Revision O (Complete reference presented above.)
R3008 (Surry 2) 0.302 95 9
WCAP-11499, Revision O (Complete reference presented above.)
R3008 (Surry 2) 1.880 145
-5 WCAP-11499, Revision O (Complete reference presented above.)
Surveillance I Material I Capsule ID Forging 03 North Anna 1 / V (Longitudinal)
Forging 03 North Anna 1 / U (Longitudinal)
North Anna 1 / V Forging 03 (Transverse)
North Anna 1 / U Forging 03 (Transverse)
North Anna 1 / V Weld 04 North Anna 1 / U Weld 04 Forging 04 North Anna 2 / V (Longitudinal)
Forging 04 North Anna 2 / U (Longitudinal)
North Anna 2 / V Forging 04 (Transverse)
North Anna 2 / U Forging 04 (Transverse)
North Anna 2 / V Weld 04 North Anna 2 / U Weld 04 Surveillance Capsules Used in Position 2 RT(PTS) Calculations North Anna Units 1 and 2 Applicable to I
Capsule I Measured I Measured - I (Beltline Material ID)
Fluence (E19)
Shift (F)
Mean RTNDT Reference "Analysis of Capsule V from the Virginia Electric and Power Forging 03 (North Anna 1) 0.249 39
-16 Company North Anna Unit 1 Reactor Vessel Radiation Surveillance Program," BAW-1638, Revision O, dated May, 1981.
"Analysis of Capsule U from the Virginia Electric and Power Forging 03 (North Anna 1) 0.828 95 11 Company North Anna Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-11777, Revision 0, dated February, 1988.
Forging 03 (North Anna 1) 0.249 21
-15 BAW-1638, Revision O (Complete reference presented above.)
Forging 03 (North Anna 1) 0.828 65 10 WCAP-11777, Revision O (Complete reference presented above.)
Weld 04 (North Anna Unit 1) 0.249 78 20 BAW-1638, Revision O (Complete reference presented above.)
Weld 04 (North Anna Unit 1) 0.828
- 75
-13 WCAP-11777, Revision O (Complete reference presented above.)
"Analysis of Capsule V from the Virginia Electric and Power Forging 04 (North Anna 2) 0.241 9
-5 Company North Anna Unit 2 Reactor Vessel Radiation Surveillance Program," BAW-1794, Revision 0, dated October, 1983.
"Analysis cif Capsule U from the Virginia Electric and Power Forging 04 (North Anna 2) 0.955 25 3
Company North Anna Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-12497, Revision o, dated January, 1990.
Forging 04 (North Anna 2) 0.241 9
-20 BAW-1794, Revision O (Complete reference presented above.)
Forging 04 (North Anna 2) 0.955 60 13 WCAP-12497, Revision O (Complete reference presented above.)
Weld 04 (North Anna Unit 2) 0.241 2
-4 BAW-1794, Revision O (Complete reference presented above.)
Weld 04 (North Anna Unit 2) 0.955 13 3
WCAP-12497, Revision O (Complete reference presented above.)
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Surveili°ance Capsule Identification The station, unit; and surveillance capsule identifier are presented in the Surveillance Capsule column. Additional information on currently docketed plant-specific and integrated surveillance program capsule withdrawal schedules is available in References (26), (15), and (22).
Surveillance Capsule Analysis Results The surveillance material identifier and the beltline material to which the surveillance capsule material is applicable are presented in the Material ID and Beltline Material ID columns, respectively. The surveillance capsule fluence (in units o.f n/cm2 x 1019) and measured transition temperature shift are presented in the Capsule Fluence and Measured Shift columns.
The Measured Shift (8RTNDT) is calculated as the difference between the temperatures at which irradiated and unirradiated Charpy specimens exhibit 30 ft-lb of absorbed energy.
Chemistry Factor Calculations According to Regulatory Guide 1.99, Revision 2, the fluence factor (FF) is defined as:
FF = f(0.28-0.1 Ologf) where f is the neutron fluence (E> 1.0 MeV) accumulated by the surveillance capsule in units of 1019 n/cm2. When surveillance data is available, the chemistry factor is calculated by dividing the sum of all 8RTNDT x FF values by the sum of all FF2 values.
The calculated CF value provides the "best fit" of available surveillance data to an equation of the form:
8RTNDT = (CF)f(0.28-0.1 Ologf) where CF is the chemistry factor, and f is the neutron fluence in units of 1019 n/cm2, E>1.0 MeV.
Assessment of Surveillance Data Credibility In consideration of the credibility of available surveillance data, RG 1.99, Revision 2, Section B (Discussion) states:
When there are two or more sets of surveillance data from one reactor, the scatter of L1RTNDT values about a best-fit line drawn as described in Regulatory Guide Position 2. 1 normally should be less than 28 °F for welds and 17°F for base metal. Even if the f/uence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for
determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E 185-82 (Ref. 1 ).
The credibility criteria established in RG 1.99, Revision 2 are 1 a values. (RG 1.99, Revision 2, Section 1.1 states "The standard deviation for L1RTNDT,<If1, is 28°F for welds and 17°F for base metal... '1 Approximately 67% of the data points can be statistically expected to lie within +/-1 a, and approximately 95% within +/-2a. All measured ~RTNDT values used in Position 2.1 chemistry factor calculations are well within +/-2a scatter bands (i.e., 2 x 28°F for welds; 2 x 17°F for base metal) around the best-fit ~RTN DT line.
Of the 31 surveillance data points considered, only two surveillance data points have ~RTNDT values which lie beyond +/-1a scatter bands (i.e., 28°F for welds; 17°F for base metal) around a best-fit ~RTNDT line. These two data points exceed the +/-1a credibility criteria by only 3°F. On this basis, all data points used in Position 2 chemistry factor calculations were determined to be credible.
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Revised Pressurized Thermal Shock Screening Calculations Surry and North Anna Units 1 and 2
Surry Unit 1 PTS Screening Calculations Calculation Material I
Heat I
I I wt% Ni I (P;:. 1) I Plant-Spec. I MIRVSP I CF Used I Inner Surf. I A>:ial Number Identification Number Weld ID wt%Cu Surv. Mtl.?
Surv. Mtl.?
in Cales Fluence (E19)
Multiplier, 1
Nozzle Shell Foroino 122V109VA1 n/a 0.09 0.74 58.0 58.0 3.96 0.12 2
Intermediate Shell C4326-1 n/a 0.11 0.55 73.5 73.5 3.96 1.00 Plate yes 3
Intermediate Shell C4326-2 n/a 0.11 0.55 73.5 73.5 3.96 1.00 Plate 4
Lower Shell Plate C4415-1 n/a 0.11 0.50 73.0 ves 89,2 3.96 1.00 5
Lower Shell Plate C4415-2 n/a 0.11 0.50 73.0 73.0 3.96 1.00 6
Nozzle to Int Shell 25017 J726 0.33 0.10 152.0 152.0 3.96 0.12 CircWeld 7
Int. to Low Sh. Gire 72445 SA-1585 0.21 0.59 162.5 149.8 3.96 1.00 Weld (ID 40%)
yes 7A Int. to Low Sh. Gire 72445 SA-1585 0.21 0.59 162.5 149.8 3.96 1.00 Weld (ID 40%)
yes 8
Int. to Low Sh. Gire 72445 SA-1650 0.21 0.59 162.5 149.8 3.96 1.00 Weld (OD 60%)
yes 9
Int Shell Long. Welds 8T1554 SA-1494 0.18 0.63 159.0 159.0 0.639 1.00 L3 & L4 10 Lower Shell Long.
8T1554 SA-1494 0.18 0.63 159.0 159.0 0.639 1.00 Weld L1 11 Lower Shell Long.
299L44 SA-1526 0.35 0.68 223.6 yes yes 217.0 0.639 1.00 Weld L2 11A Lower Shell Long.
299L44 SA-1526 0.35 0.68 223.6 yes yes 217.0 0.639 1.00 WeldL2 Calculation Adjusted I I
"Measured" I 10 CFR 50.61 I Delta I RTPTS I Screening Ratio (RT Prs Number Fluence (E19)
Initial RTNDT or "Estimate" Margin Term RTNDT Criterion
/Screen Crit.)
1 0.475 40 Measured 34 46.0 120.0 270 44%
2 3.960 10 Measured 34 99.5 143.5 270 53%
e 3
3.960 0
Measured 34 99.5 133.5 270 49%
4 3.960 20 Measured 34 120.8 174.8 270 65%
5 3.960 0
Measured 34 98.8 132.8 270 49%
6 0.475 0
Estimate 66 120.5 186.5 300 62%
7 3.960
-5 Estimate 66 202.8 263.8 300 88%
7A 3.960
-27 Measured 56 202.8 231.8 300 77%
8 3.960
-5 Estimate 66 202.8 263.8 300 88%
9 0.639
-5 Estimate 66 139.0 200.0 270 74%
10 0.639
-5 Estimate 66 139.0 200.0 270 74%
11 0.639
-7 Estimate 66 189.8 248.8 270 92%
11A 0.639
-27 Measured 56 189.8 218.8 270 81%
Surry Unit 2 PTS Screening Calculations Calculation Material I
Heat I
I I wt% Ni I CF I Plant-Spec. I MIRVSP I CF Used I Inner Surf. I Axial Number Identification Number Weld ID wt% Cu (Pos.1)
Surv. Mtl.?
Surv. Mtl.?
in Cales Fluence (E19)
Multiplier 12 Nozzle Shell Forging 123V303VA1 n/a 0.09 0.73 58.0 58.0 3.43 0.12 13 Intermediate Shell C4331-2 n/a 0.12 0.60 83.0 83.0 3.43 1.0b Plate 14 Intermediate Shell C4339-2 n/a 0.11 0.54 73.4 73.4 3.43 1.00 Plate 15 Lower Shell Plate C4208-2 n/a 0.15 0.55 107.3 107.3 3.43 1.00 16 Lower Shell Plate C4339-1 n/a 0.11 0.54 73.4 yes 68.4 3.43 1.00 17 Nozzle to Int Shell 4275 L737 0.35 0.10 160.5 160.5 3.43 0.12 CircWeld 18 Int. to Lower Shell 0227 R3008 0.19 0.55 149.3 128.0 3.43 1.00 CircWeld yes e
19 Int. Shell Long. Weld 8T1762 WF-4 0.20 0.55 152.3 152.3 0.714 1.00 L4 (ID 50%)
20 Int. Sh. Welds L3 72445 SA-1585 0.21 0.59 162.5 149.8 0.714 1.00 (100%), L4 (OD 50) yes 21 LS Welds L2 (ID 8T1762 WF-4 0.20 0.55 152.3 152.3 0.714 1.00 63%), L1 (100) 22 LS Long. Weld L2 8T1762 WF-8 0.20 0.55 152.3 152.3 0.714 1.00 (OD 37%)
Calculation Adjusted I I
"Measured" I 10 CFR 50.61 I Delta I RTPTS I Screening Ratio (RT PTs Number Fluence (E19)
Initial RTNDT or "Estimate" Margin Term RTNDT Criterion
/Screen Crit.)
12 0.412 30 Measured 34 43.7 107.7 270 40%
13 3.430
-10 Measured 34 109.7 133.7 270 50%
14 3.430
-20 Measured 34 97.0 111.0 270 41%
15 3.430
-30 Measured 34 141.8 145.8 270 54%
16 3.430
-10 Measured 34 90.4 114.4 270 42%
17 0.412 0
Estimate 66 121.0 187.0 300 62%
18 3.430 0
Estimate 66 169.2 235.2 300 78%
19 0.714
-5 Estimate 66 137.9 198.9 270 74%
20 0.714
-5 Estimate 66 135.6 196.6 270 73%
21 0.714
-5 Estimate 66 137.9 198.9 270 74%
22 0.714
-5 Estimate 66 137.9 198.9 270 74%
North Anna Unit 1 PTS Screening Calculations Calculation Material I
Heat I
I I wt% Ni I (P:: 1) I Plant-Spec. I MIRVSP I CF Used I Inner Surf. I Axial Number Identification Number Weld ID wt% Cu Surv. Mtl.?
Surv. Mtl.?
in Cales Fluence (E19)
Multiplier 23 Nozzle Shell Forqinq 990286/295213 Forqinq 05 0.16 0.74 121.5 121.5 3.95 0.07 24 Intermediate Shell 990311/298244 Forging 04 0.12 0.82 86.0 86.0 3.95 1.00 Forqinq 25 Lower Shell Forging 990400/292332 Fomino 03 0.16 0.83 123.3 yes 88.9 3.95 1.00 26 Nozzle to Int. Shell 25295 Weld 05A 0.35 0.13 162.8 162.8 3.95 0.07 Circ Weld (OD 94%)
27 Nozzle to Int. Shell 4278 Weld 058 0.12 0.11 63.0 63.0 3.95 0.07 Circ Weld (ID 6%)
28 Int. to Lower Shell 25531 Weld 04 0.11 0.13 61.4 93.1 3.95 1.00 Gire Weld yes e
Calculation Adjusted I I
"Measured" I 10 CFR 50.61 I Delta I RTPTS I Screening Ratio (RT Prs Number Fluence (E19)
Initial RTNDT or "Estimate" Margin Term RTNDT Criterion
/Screen Crit.)
23 0.277 6
Estimate 48 78.9 132.9 270 49%
24 3.950 17 Measured 34 116.4 167.4 270 62%
25 3.950 38 Measured 34 120.3 192.3 270 71%
26 0.277 0
Estimate 66 105.7 171.7 300 57%
27 0.277 0
Estimate 66 40.9 106.9 300 36%
28 3.950 19 Measured 56 126.0 201.0 300 67%
r
~
- North Anna Unit 2 PTS Screening Calculations Calculation Material I
Heat I
I I wt% Ni I (P;:. 1) I Plant-Spec. I MIRVSP I CF Used I Inner Surf. I Axial Number Identification Number Weld ID wt%Cu Surv. Mtl.?
Surv. Mtl.?
in Cales Fluence (E19)
Multiplier 29 Nozzle Shell Forgina 990598/291396 Foraina 05 0.08 0.77 51.0 51.0 4.47 0.07 30 Intermediate Shell 990496/292424 Forging 04 0.10 0.85 67.0 47.9 4.47 1.0G For!lin!l yes 31 Lower Shell Forgina 990533/297355 Forc:iinci 03 0.13 0.83 96.0 96.0 4.47 1.00 32 Nozzle to Int. Shell 4278 Weld 05A 0.12 0.11 63.0 63.0 4.47 0.07 Circ Weld (OD 94%)
33 Nozzle to Int. Shell 801 Weld 05B 0.1,8 0.11 87.8 87.8 4.47 0.07 Circ Weld (ID 6%)
34 Int. to Lower Shell 716126 Weld 04 0.07 0.05 37.8 10.4 4.47 1.00 CircWeld yes e
Calculation Adjusted I I "Measured" I 10 CFR 50.61 I Delta I RTPTS I Screening Ratio (RT Prs Number Fluence (E19)
Initial RTNDT or "Estimate" Margin Term RTNDT Criterion
/Screen Crit.)
29 Forging Forging 05 990598/291396 48 34.7 91.7 270 34%
30 Forging Forging 04 990496/292424 34 66.1 175.1 270 65%
31 Forging Forging 03 990533/297355 34 132.5 222.5 270 82%
32 Weld Weld 05A 4278 66 42.9 108.9 300 36%
33 Weld Weld 05B 801 66 59.8 125.8 300 42%
34 Weld Weld 04 716126 56 14.3 22.3 300 7%
e e
Material Identification Each reactor vessel beltline material is identified by a Material Identification, Heat Number, and Weld ID (if applicable). The information which identifies each beltline material is consistent with the descriptions in the Surry and North Anna responses to GL 92-01, Revision 1 closure letters (4),(5), and the Surry and North Anna responses to GL 92-01, Revision 1, Supplement 1 (7),(11 ).
Chemical Composition The beltline material mean chemical compositions (wt% Cu and wt% Ni) for the Surry units are documented in the responses to the NRC requests for additional information on the responses to GL 92-01, Revision 1 (5). The Linde 80 weld mean chemical compositions were re-verified in the B&W Owners Group response to GL 92-01, Revision 1, Supplement 1 (7). The mean chemical compositions for all North Anna Units 1 and 2 beltline materials, and the Surry Units 1 and 2 Rotterdam weld materials were revised (or re-verified) in the Reference (11) response to GL 92-01, Revision 1, Supplement 1. The 1 O CFR 50.61 and Position 1 of RG 1.99, Revision 2 beltline material chemistry factors (CF (Pos. 1)) were determined using these mean chemical compositions.
Chemistry Factors Used in Screening Calculations If a material is included in a plant-specific surveillance program (18),(19),(20),(21) or in the B&W Owners Group Master Integrated Reactor Vessel Surveillance Program (MIRVSP) (22), this is indicated in the Plant-Specific Surveillance Material or MIRVSP Surveillance Material columns.
If two or more credible surveillance capsule analysis results are available for a particular beltline material, *the chemistry factor calculated in accordance with Regulatory Guide 1.99, Revision 2, Paragraph 2.1
("Position 2") is used in the PTS screening calculation.
(Compare CF Used in Calculations to CF (Pos. 1 ).) The credibility assessment of surveillance capsule analysis results is described in Attachment 1.
Neutron Fluence The maximum Inner Surface Neutron Fluence values (in units of n/cm2 x 1019) were obtained from References (12), (13), (14), and (15). The Surry Unit 1 fluence estimates do not take credit for the installation of flux suppression inserts.
Correction for Axial Location of Beltline Material The Axial Multiplier corrects the maximum Inner Surface Neutron Fluence for axial elevations other than the core mid-plane. The product of the Axial Multiplier
e and the maximum Inner Surface Neutron Fluence is presented in the Adjusted Fluence column.
The Axial Multiplier for the Surry units is obtained from Figure 11.2-6 of Reference (23) or Figure 6-7 of References (12) and (13). The Axial Multiplier for the North Anna units is obtained from Figure 11.2-6 of Reference (25) or Figure 6-7 of Reference (14). The Surry Units 1 and 2 nozzle shell forging to intermediate shell circumferential weld-is located 9.0 inches above the top of the active region of the core, or 81.0 inches (206.0 cm) above the core midplane. (See Figures 111.3-1 and 111.3-2 of Reference (24).) For the North Anna Units, the nozzle shell forging to intermediate shell circumferential weld is located 13.6 inches above the top of the active region of the core, or 85.6 inches (217.0 cm) above the core midplane. (See Figures 111.1-1 and 111.1-2 of Reference (25).) These values (206.0 cm and 217.0 cm) also represent conservative estimates of the axial location of the nozzle shell forging.
Unirradiated RTN DT The Initial RTNDT column presents the unirradiated nil-ductility transition reference temperature, calculated in accordance with the requirements of the ASME Code.
Values were obtained from BAW-2222 (5) and BAW-2260 (11). Because the 10 CFR 50.61 Margin Term is dependent on the method of determining the Initial RTNDT, each Initial RTN DT value is flagged as being either a Measured or Estimated value. Supplemental calculations were performed for the limiting Surry Units 1 and 2 Linde 80 weld materials (i.e., Unit 1 Intermediate to Lower Shell Circumferential Weld, ID-40%, SA-1585; and Unit 1 Lower Shell Longitudinal Weld L2, SA-1526),-using the measured -27°F Initial RTNDT value determined in Reference (16). As previously noted, calculations using the Reference (16) Initial RTNDT value are presented for information and comment only.
1 O CFR 50.61 Margin Term The 1 O CFR 50.61 PTS Rule prescribes Margin Term values for uncertainty in the Initial RTNDT and in the Delta RTNDT due to irradiation. For welds, the Margin Term values are 56°F and 66°F for Measured and Estimated values of Initial RTNDT, respectively. For plates and forgings, the Margin Term values are 34°F and 48°F for Measured and Estimated values of Initial RTNDT, respectively.
RTN DT Shift Due to Irradiation (Delta RTN DT}
Delta RTN DT is determined in accordance with the methodology prescribed in 1 O CFR 50.61:
L\\RTNDT = (CF)f(0.28 - 0.1 Ologf)
e where CF is the Chemistry Factor described above, and f is the Neutron Fluence in units of 1019 n/cm2. RTPTS is the sum of the Initial RTNDT, the Margin Term, and Delta RTNDT* RTPTS values are compared to the applicable Screening Criterion specified by 1 O CFR 50.61. For circumferential welds, the Screening Criterion is 300°F.
For axial (longitudinal) welds, plates, and forgings, the Screening Criterion value is 270°F. The Ratio of RTPTS to the Screening Criterion permits ranking of beltline materials in terms of the relative proximity of a specific material's RTPTS to the applicable Screening Criterion. The most limiting materials have the highest ratio.