ML16161A173
| ML16161A173 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/25/1994 |
| From: | Wiens L Office of Nuclear Reactor Regulation |
| To: | Tuckman M DUKE POWER CO. |
| References | |
| GL-92-01, GL-92-1, TAC-M83735, TAC-M83736, TAC-M83834, NUDOCS 9406030070 | |
| Download: ML16161A173 (14) | |
Text
0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 25, 1994 Docket Nos. 50-269, 50-270 and 50-289 Mr. M. S. Tuckman Senior Vice President Nuclear Generation Duke Power Company P.O. Box 1006 Carlotte, North Carolina 28201
Dear Mr. Tuckman:
SUBJECT:
GENERIC LETTER (GL) 92-01, REVISION 1, "REACTOR VESSEL STRUCTURAL INTEGRITY," OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 (TAC NOS.
M83734, M83735, AND M83736)
By letter dated July 3, 1992, Duke Power Company (DPC) provided its response to GL 92-01, Revision 1. The NRC staff has completed its review of your response. Based on its review, the staff has determined that DPC has provided the information requested in GL 92-01.
The GL is part of the staff's program to evaluate reactor vessel integrity for Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The information provided in response to GL 92-01, including previously docketed information, is being used to confirm that licensees satisfy the requirements and commitments necessary to ensure reactor vessel integrity for their facilities.
A substantial amount of information was provided in response to GL 92-01, Revision 1. These data have been entered into a computerized data base designated the Reactor Vessel Integrity Database (RVID).
The RVID contains the following tables: A pressurized thermal shock (PTS) table for PWRs, a pressure-temperature limits table for BWRs and an upper-shelf energy (USE) table for PWRs and BWRs. Enclosure 1 provides the PTS tables, Enclosure 2 provides the USE tables for your facilities, and Enclosure 3 provides a key for the nomenclature used in the tables. The tables include the data necessary to perform USE and RT evaluations. These data were taken from your response to GL 92-01 and previously docketed information.
References to the specific source of the data are provided in the tables.
We request that, within 30 days, you provide confirmation of the plant specific applicability of the Topical Reports BAW-2178P and BAW-2192P and submit a request for approval of the topical reports'as the basis for demonstrating compliance with 10 CFR Part 50, Appendix G, Paragraph IV.A.1.
To demonstrate that the topical reports are applicable to Oconee Units 1, 2, and 3, you must compare the limiting material properties of the Oconee reactor vessel to the values reported in the topical reports. -This review will be a 9406030070 94025 PDR ADOCK 05600269 P
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Mr. M. S. Tuckman 2 -
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plant-specific licensing action. We further request that you verify that the information you have provided for your facilities has been accurately entered in the summary data file. If no comments are made in your response to the last request, the staff will use the information in the tables for future NRC assessments of your reactor pressure vessel.
Once your confirmation of the applicability of the topical reports and request for approval are received, the staff will consider your actions related to GL 92-01, Revision 1, to be complete.
The information requested by this letter is within the scope of the overall burden estimated in GL 92-01, Revision 1, "Reactor Vessel Structural Integrity, 10 CFR 50.54(f)."
The estimated average number of burden hours is 200 person hours for each addressee's response. This estimate pertains only to the identified response-related matters and does not include the time required to implement actions required by the regulations. This action is covered by the Office of Management and Budget Clearance Number 3150-0011, which expires June 30, 1994.
Sincerely,
/s/
L. A. Wiens, Project Manager Project Directorate 11-3 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Enclosures:
- 1. Pressurized Thermal Shock Tables
- 2. Upper-Shelf Energy Tables
- 3. Nomenclature Key cc w/enclosures:
See next page DISTRIBUTION PD23 Reading File NRC/Local PDR DMatthews LWiens LBerry GLainas OGC ACRS (10)
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Mr. M. May 25, 1994 plant-specific licensing action. We further request that you verify that the information you have provided for your facilities has been accurately entered in the summary data file. If no comments are made in your response to the last request, the staff will use the information in the tables for future NRC assessments of your reactor pressure vessel.
Once your confirmation of the applicability of the topical reports and request for approval are received, the staff will consider your actions related to GL 92-01, Revision 1, to be complete.
The information requested by this letter is within the scope of the overall burden estimated in GL 92-01, Revision 1, "Reactor Vessel Structural Integrity, 10 CFR 50.54(f)." The estimated average number of burden hours is 200 person hours for each addressee's response. This estimate pertains only to the identified response-related matters and does not include the time required to implement actions required by the regulations. This action is covered by the Office of Management and Budget Clearance Number 3150-0011, which expires June 30, 1994.
Sincerely, L. A. Wiens, Project Manager
- 1. Pressurized Thermal Shock Tables
- 2. Upper-Shelf Energy Tables
- 3. Nomenclature Key cc w/enclosures:
See next page
Mr. J. W. Hampton Duke Power Company Oconee Nuclear Station cc:
A. V. Carr, Esquire Mr. Steve Benesole Duke Power Company Compliance 422 South Church Street Duke Power Company Charlotte, North Carolina 28242-0001 Oconee Nuclear Site P. 0. Box 1439 J. Michael McGarry, III, Esquire Seneca, South Carolina 29679 Winston and Strawn 1400 L Street, NW.
Mr. Marvin Sinkule, Chief Washington, DC 20005 Project Branch #3 U. S. Nuclear Regulatory Commission Mr. Robert B. Borsum 101 Marietta Street, NW. Suite 2900 Babcock & Wilcox Atlanta, Georgia 30323 Nuclear Power Division Suite 525 Ms. Karen E. Long 1700 Rockville Pike Assistant Attorney General Rockville, Maryland 20852 North Carolina Department of Justice Manager, LIS P. 0. Box 629 NUS Corporation Raleigh, North Carolina 27602 2650.McCormick Drive, 3rd Floor Clearwater, Florida 34619-1035 Mr. G. A. Copp Licensing -
EC050 Senior Resident Inspector Duke Power Company U. S. Nuclear Regulatory Commission 526 South Church Street Route 2, Box 610 Charlotte, North Carolina 28242-0001 Seneca, South Carolina 29678 Dayne H. Brown, Director Regional Administrator, Region II Division of Radiation Protection U. S. Nuclear Regulatory Commission North Carolina Department of 101 Marietta Street, NW. Suite 2900 Environment, Health and Atlanta, Georgia 30323 Natural Resources P. 0. Box 27687 Max Batavia, Chief Raleigh, North Carolina 27611-7687 Bureau of Radiological Health South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, South Carolina 29201 County Supervisor of Oconee County Walhalla, South Carolina 29621 00 Summary File for Pressurized Thermal Shock Plant BeLtline Heat No.
ID Neut.
IRT.,
Method of Chemistry Method of
%Cu
%Ni Name Ident.
Ident.
Fluence at Determin.
Factor Determin.
EOL IRT,,,
CF Oconee 1 Lower AHR54 1.18E18 30F Plant 119.25 Table 0.16 0.65 nozzle Specific belt forging EOL:
Upper C3278-1 9.04E18 1*F Plant 83 Table 0.12 0.60 2/6/2013 shell Specific Upper C3265-1 9.04E18 1oF Plant 54.56 Calculated 0.10 0.50 shell Specific Int. shell C2197-2 7.96E18 10F Plant 104.5 Table 0.15 0.50 Specific Lower C2800-1 8.68E18 1*F Plant 74.45 Table 0.11 0.63 shill Specific Lower C2800-2 8.68E18 1F Plant 74.45 Table 0.11 0.63 shell Specific Nozzle/int 61782 1.18E18
-50F Generic 133.94 Calculated 0.25 0.54 shell circ. weld SA-1135 Int/Lower 71249 7.96E18
-5*F Generic 181.6 Table 0.26 0.61 shell circ. weld SA-1229 Upp./lower 72445 8.68E18
-5*F Generic 146.09 Calculated 0.21 0.59 shell circ. weld SA-1585 Int. shell 1PO962 6.28E18
-50F Generic 170.6 Table 0.21 0.64 axial welds SA-1073 Upper 8T1762 7.23E18
-5*F Generic 152.25 Table 0.20 0.55 shell axial welds SA-1493 Lower 8T1762 7.29E18
-5*F Generic 152.25 Table 0.20 0.55 shell axial welds SA-1430 Lower 8T1762 7.29EtS
- 5*F Generic 152.25 Table 0.20 0.55 shell axial welds SA-1426 References for Oconee 1 Fluence, chemical composition. and IR&Ls are from BAW-2166.
Chemistry Factor for S4-1585 was calcukated from Point Beach and Crystal River surveillance data that was reported in BAW 1803, Rev. 1.
The surveillance Weds were fabricated using the same weld wire heat number as SA-1585.
Chemistry Factor for SA-1135 was calculated from Ginna and Davis-Besse surveillance data that was reported in BAW-1803, Rev. 1.
The surveillance welds were fabricated with the same weld wire Heat Number as SA-1135.
Summary File for Pressurized Thermal Shock PLant Bettline Heat No.
ID Neut.
IRT.,
Method of Chemistry Method of
%Cu
%Ni Name Ident.
Ident.
Fluence at Determin.
Factor Determin.
EOL IRT.,
CF Oconee 2 Lower AMX-77 8.42E18 30F Plant 37 Table 0.06 0.76 nozzle Specific belt forging EOL:
Upper AAW-163 9.57E18 20*F Plant 12.28 Calculated 0.04 0.75 10/6/2013 shell Specific forging Lower AWG-164 9.19E18 20*F Plant 20 Table 0.02 0.80 shell Specific forging Upper 406L44 8.42E18
-5*F Generic 176.27 Calculated 0.31 0.59 circ. weld WF-154 Middle 299L44 9.19E18
-5*F Generic 221.25 Calculated 0.35 0.68 circ. weld WF-25 References for Oconee 2 Fluence, chemical composition, and IRT.,s are from BAW-2166.
Chemistry Factor for weld WF-154 was calculated from surveillance data for Oconee 1, Davis-Besse, ANO-1, Rancho-Seco and Point Beach 2 that was reported in BAW 1803, Rev. 1.
The surveillance welds were fabricated with the same weld wire Heat Number as WF-154.
Chemistry Factor for weld WF-25 was calculated from surveillance data for TMI-1, Crystal River 3 and Surry 1 that was reported in BAW-1803, Rev. 1.
The surveillance welds were fabricated with the same weld wire Heat Number as WF-25.
Summary File for Pressurized Thermal Shock Plant BeltLine Heat No.
ID Neut.
- IRTu, Method of Chemistry Method of
%Cu
%Ni Name Ident.
Ident.
Fluence at Determin.
Factor Determin.
EOL IRTa CF Oconee 3 Lower 4680 8.26E18 30F Plant 96 Table 0.13 0.91 nozzle Specific shell forging EOL:
Upper AWS-192 9.39E18 40OF Plant 47.1 Calculated 0.01 0.73 7/19/2014 sheLl Specific forging Lower ANK-191 9.01E18 40*F Plant 32.2 Calculated 0.02 0.76 shell Specific forging Nozzle 821T44 8.26E18
-5*F Generic 162.09 Calculated 0.24 0.63 belt/upper shell circ. weld WF-200 Upp./lower 72442 9.01E18
-5*F Generic 173 Table 0.24 0.60 shell circ. weld 75% ID WF-67 References for Oconee 3 Fluence, chemical composition, and IRT.,s are from BAW-2166.
The copper content for lower nozzle sheLl forging is from BAW-2128, Table 7-5, Capsule D report.
Chemistry Factor for weld WF-200 was calculated from surveillance data for Davis-Besse that was reported in BAW-1803, Rev. 1. The surveillance welds were fabricated with the same weld wire Heat Number as WF-200.
Summary File for Upper Shelf Energy Plant Name Bettline Heat No.
Material 1/4T USE 1/4T Unirrad.
Method of Ident.
Type at EOL Neutron USE Determin.
Fluence at Unirrad.
EOL USE Oconee 1 Lower AHR54 A 508-2 107 7.1E17 124 Direct nozzle belt forging EOL:
Upper C3278-1 A 3028 75 5.45E18 91 Direct 2/6/2013 shelL Upper C3265-1 A 3028 90 5.45E18 108 Direct shell Int. shell C2197-2 A 302B 73 4.80E18 91 Direct Lower C2800-1 A 302B 76 5.23E18 91 Direct shell Lower C2800-2 A 302B 99 5.23E18 119 Direct shell Nozzle/
61782 Linde 80, EMA2 7.1E18 EMA2 Generic int. shell SAW circ. weld SA-1135 Int/tower 71249 Linde 80, EMA' 4.80E18 EMA' Generic shell SAW circ. weld ID 61%
SA-1229 Upp./Lower 72445 Linde 80, EMA' 5.23E18 EMAa Generic shell SAW circ. weld SA-1585 Int. shelL 1P0962 Linde 80, EMA' 3.79E18 EMAa Generic axial SAW welds SA-1073 Upper 8T1762 Linde 80, EMA2 4.36E18 EMA' Generic shell SAW axial welds SA-1493 Lower 8T1762 Linde 80, EMA' 4.39E18 EMA' Generic shell SAW axial welds SA-1430 2Licensee must confirm applicability of Topical Reports BAW-2178P and BAW-2192P
Sunnary File for Upper Shelf Energy PLant Name Beittine Heat No.
Material 1/4T USE 1/4T Unirrad.
Method of Ident.
Type at EOL Neutron USE Determin.
Fluence at Unirred.
EOL USE Lower 8T1762 Linde 80, EMA' 4.39E18 EMA' Generic shell SAW axial welds SA-1426 References The UUSE data are from the November 22, 1993 Letter from Duke Power to USNRC,
Subject:
Oconee Units 1, 2, & 3, Response to Request for AdditionaL Information Concerning Generic Letter 92
- 01. Revision 1.
The method of determining the WSE data is from the March 24, 1994 Letter from Duke Power to NRC.
Summary File for Upper Shelf Energy Plant Name BettLine Heat No.
Material 1/4T USE 1/4T Unirrad.
Method of Ident.
Type at EOL Neutron USE Determin.
FLuence at Unirrad.
EOL USE Oconee 2 Lower ANX-77 A 508-2 104 5.08E18 124 Direct nozzle belt forging EOL:
Upper AAW-163 A 508-2 107 5.77E18 128 Direct 10/6/2013 shell forging Lower AWG-164 A 508-2 121 5.54E18 145 Direct shell forging Upper 406L44 Linde 80, EMA' 5.08E18 EMA Generic circ. weld SAW WF-154 Middle 299L44 Linde 80, EMA2 5.54E18 EMA2 Generic circ. weld SAW WF-25 References Fluence, chemical composition, and ULUSE data are from BAW-2166.
2Licensee must confirm applicability of Topical Reports BAW-2178P and BAW-2192P
Summary File for Upper Shelf Energy Plant Name Bettline Neat No.
Material 1/4T USE 1/4T Unirred.
Method of Ident.
Type at EOL Neutron USE Determin.
Fluence at Unirred.
EOL USE Oconee 3 Lower 4680 A 508-2 77 4.98E18 124 Direct nozzle shell forging EOL:
Upper AWS-192 A 508-2 75 5.66E18 90 Direct 7/19/2014 sheLL forging Lower ANK-191 A 508-2 92 5.43E18 110 Direct shell forging Nozzle 821T44 Linde 80, EMA2 4.98E18 EMA' Generic belt/upper SAW shell circ. weld WF-200 Upp./lower 72442 Linde 80, EMA' 5.43E18 EMA' Generic shell SAW circ. weld WF-67 References Fluence, chemical composition, and UUSE data are from BAW-2166.
2Licensee must confirm applicability of Topical Reports BAW-2178P and BAW-2192P PRESSURIZED THERMAL SHOCK AND USE TABLES FOR ALL PWR PLANTS NOMENCLATURE Pressurized Thermal Shock Table Column 1: Plant name and date of expiration of license.
Column 2:
Beltline material location identification.
Column 3: Beltline material heat number; for some welds that a single wire or tandem-wire process has been reported, (S) indicates single wire was used in the SAW process, (T) indicates tandem wire was used in the SAW process.
.Column 4: End-of-life (EOL) neutron fluence at vessel inner wall; cited directly from inner diameter (ID) value or calculated by using Regulatory Guide (RG) 1.99, Revision 2 neutron fluence attenuation methodology from the quarter thickness (T/4) value reported in the latest submittal (GL 92-01, PTS, or P/T limits submittals).
Column 5:
Unirradiated reference temperature.
Column 6:
Method of determining unirradiated reference temperature (IRT).
Plant-Specific This indicates that the IRT was determined from tests on material removed from the same heat of the beltline material.
MTES 5-2 This indicates that the unirradiated reference temperature was determined from following MTEB 5-2 guidelines for cases where the IRT was not determined using American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section III, NB-2331, methodology.
Generic This indicates that the unirradiated reference temperature was determined from the mean value of tests on material of similar types.
Column 7: Chemistry factor for irradiated reference temperature evaluation.
Column 8: Method of determining chemistry factor lable This indicates that the chemistry factor was determined from the chemistry factor tables in RG 1.99, Revision 2.
Calculated
. This indicates that the chemistry factor was determined from surveillance data via procedures described in RG 1.99, Revision 2.
Column 9:
Copper content; cited directly from licensee value except when more than one value was reported. (Staff used the average value in the latter case.)
No Data This indicates that no copper data has been reported and the default value in RG 1.99, Revision 2, will be used by the staff.
Column 10: Nickel content; cited directly from licensee value except when more than one value was reported. (Staff used the average value in the latter case.)
No Data This indicates that no nickel data has been reported and the default value in RG 1.99, Revision 2, will be used by the staff.
Upper Shelf Energy Table Column 1:
Plant name and date of expiration of license.
Column 2:
Beltline material location identification.
Column 3:
Beltline material heat number; for some welds that a single wire or tandem-wire process has been reported, (S) indicates single wire was used in the SAW process. (T) indicates tandem wire was used in the SAW process.
Column 4: Material type; plate types include A 533B-1, A 302B, A 302B Mod., and forging A 508-2; weld types include SAW welds using Linde 80, 0091, 124, 1092, ARCOS-B5 flux, Rotterdam welds using Graw Lo, SMIT 89, LW 320, and SAF 89 flux, and SMAW welds using no flux.
Column 5:
EOL upper-shelf energy (USE) at T/4; calculated by using the EOL fluence and either the cooper value or the surveillance data. (Both methods are described in RG 1.99, Revision 2.)
This indicates that the USE issue may be covered by either owners group or plant-specific equivalent margins analyses.
Column 6:
EOL neutron fluence at T/4 from vessel inner wall; cited directly from T/4 value or calculated by using RG 1.99, Revision 2 neutron fluence attenuation methodology from the ID value reported in the latest submittal (GL 92-01, PTS, or P/T limits submittals).
Column 7:
Unirradiated USE.
This indicates that the USE issue may be covered by either owners group or plant-specific equivalent margins analyses.
Column 8:
Method of determining unirradiated USE Direct For plates, this indicates that the unirradiated USE was from a transverse specimen. For welds, this indicates that the unirradiated USE was from test date.
65%
This indicates that the unirradiated USE was 65% of the USE from a longitudinal specimen.
Generic This indicates that the unirradiated USE was reported by the licensee from other plants with similar materials to the beltline material.
NRC generic This indicates that the unirradiated USE was derived by the staff from other plants with similar materials to the beltline material.
10, 30, 40, or 50 OF This indicates that the unirradiated USE was derived from Charpy test conducted at 10, 30, 40, or 50 OF.
Surv. Weld This indicates that the unirradiated USE was from the surveillance weld having the same weld wire heat number.
Equiv. to Surv. Weld This indicates that the unirradiated USE was from the surveillance weld having different weld wire heat number.
Sister Plant This indicates that the unirradiated USE was derived by using the reported value from other plants with the same weld wire heat number.
'0 ink indicates-that there is insufficient data to determine the unirradiated USE.