ML18153A957
| ML18153A957 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 05/24/1994 |
| From: | Buckley B Office of Nuclear Reactor Regulation |
| To: | Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| References | |
| GL-92-01, GL-92-1, TAC-M83739, TAC-M83740, NUDOCS 9406030192 | |
| Download: ML18153A957 (12) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055~001 Docket Nos. 50-280 and 50-281 Mr. J. P. O'Hanlon Senior Vice President - Nuclear Virginia Electric and Power Company 5000 Dominion Blvd.
Glen Allen, Virginia 23060
Dear Mr. O'Hanlon:
May 24, 1994
SUBJECT:
GENERIC LETTER {GL) 92-01, REVISION 1, "REACTOR VESSEL STRUCTURAL INTEGRITY," SURRY POWER STATION, UNITS 1 AND 2 {TAC NOS. M83739 AND M83740)
By letters dated June 29, 1992, and September 23, 1993, you provided your response to GL 92-01, Revision 1.
The NRC staff has completed its review of your responses.
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, Revision 1, 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. 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 an.d RT pts eva 1 uat i ans. These data were taken from your responses 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 Topical Reports BAW-2178P and BAW-2192P to Surry, Units 1 and 2, as a basis for demonstrating compliance with 10 CFR Part 50, Appendix G, Paragraph IV.A. I.
To demonstrate that the topical reports are applicable to Surry, Units 1 and 2, you must compare the limiting material properties of your reactor vessel to the values reported in the topical reports.
r)"",~ n *"°'Is
,,-----v'..L d ~: v..
9406030192*940524 PPDR ADOCK o:;;qoo2eo
.
- PDR y
e e In addition, we have determined that additional data is required to confirm that the USE at end-of-life {EOL} for the nozzle belt to intermediate shell circumferential weld in Surry, Units 1 and 2, is greater than 50 ft-lb because you have provided a generic mean value for the unirradiated USE.
These types of values are unacceptable because they do not consider material variability.
When the unirradiated USE for a particular material has not been determined, you can set the USE equal to the lower tolerance limit calculated for the group of similar materials. The unirradiated USE should be determined such that there exists 95% confidence that at least 95% of the population is greater than the lower tolerance limit. If the lower tolerance limit results in a projected USE at EOL of less than 50 ft-lb, then you must demonstrate, in accordance with Appendix G, 10 CFR Part 50, that lower values of USE will provide margins of safety against fracture equivalent to those required by Appendix G of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.
We request that you submit within 30 days a schedule for performing these analyses.
Further, we request that you verify that the information you have provided for your facility has been accurately entered in the summary file.
If no comments are made in your response to this request, the staff will use the information in the tables for future NRC assessments of your reactor pressure vessel.
Once your response is received and your schedule is determined to be satisfactory, the staff will consider your actions related to GL 92-01, Revision 1, to be complete.
When your analyses are submitted, they will be reviewed as a plant-specific licensing action.
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.
Enclosures:
- 1.
Pressurized Thermal Shock Tables
- 2.
Upper-Shelf Energy Tables
- 3.
Nomenclature Key cc w/enclosures:
See next page Distribution - See next OFC NAME Elana DATE S'/ /7/94 OFFICIAL RECORD COPY -
Sincerely, (Original Signed By)
Bart C. Buckley, Senior Project Manager Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Mr. W. L. Stewart Virginia Electric and Power Company cc:
Michael W. Maupin, Esq.
Hunton and Williams Riverfront Plaza, East Tower 951 E. Byrd Street Richmond, Virginia 23219 Mr. Michael R. Kansler, Manager Surry Power Station Post Office Box 315 Surry, Virginia 23883 Senior Resident Inspector Surry Power Station U.S. Nuclear Regulatory Commission 5850 Hog Island Road Surry, Virginia 23883 Mr. Sherlock Holmes, Chairman Board of Supervisors of Surry County Surry County Courthouse Surry, Virginia 23683 Dr. W. T. Lough Virginia State Corporation Commission Division of Energy Regulation Post Office Box 1197 Richmond, Virginia 23209 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street N.W., Suite 2900 Atlanta, Georgia 30323 Robert B. Strobe, M.D., M.P.H.
State Health Commissioner Office of the Commissioner Virginia Department of Health P.O. Box 2448 Richmond, Virginia 23218 Surry Power Station Attorney General Supreme Court Building 101 North 8th Street Richmond, Virginia 23219 Mr. M. L. Bowling, Manager Nuclear Licensing & Programs Innsbrook Technical Center Virginia Electric and Power Company 5000 Dominion Blvd.
Glen Allen, Virginia 23060
ENCLOSURE 1 Su~ry File for Pressurized Thermal4illiock Plant Bel tline Heat No.
ID Neut.
I RT""'
Method of Chemistry Method of xcu XNi Name
!dent.
!dent.
Fluence at Determin.
Factor Determin.
EOL IRT ""'
CF Surry 1 Nozzle 122V109VA1 5.27E18 40"F Plant 58 Table 0.09 0.74 belt Specific forging EOL:
Int. shell C4326-1 4.39E19 10°F Plant 73.5 Table 0.11 0.55 5/25/2012 plate Specific Int. shell C4326-2 4.39E19 O"F Plant 73.5 Table 0.11 0.55 plate Scecific Lower C4415-1 4.39E19 20°F Plant 90.294 Calculated 0.11 a.so shell Specific plate Lower C4415-2 4.39E19 0°F Plant 90.294 Calculated 0.11 0.50 shell Specific plate Int. &
8T1554 7.08E18
-5°F Generic 158.95 Table 0.18 0.63 lower axial welds, L1, L3, & L4 SA-1494 Lower 299L44 7.08E18
-5°F Generic 221.25 Calculated 0.35 0.68 shell axial weld L2 SA-1526 Int. to 72445 4.39E19
-5°F Generic 146.09 Calculated 0.21 0.59 lower shell circ.
- weld, SA-1585 Nozzle 25017 5.27E18 0°F Generic 152 Table 0.33 0.10 belt to Rotterdam int. shell circ. weld Jn6 References Chemistry Factor for SA-1585 was calculated from Point Beach 1 and Crystal River 3 surveillance data that was reported in BAW 1803, Rev. 1.
The surveillance welds were fabricated using the same heat nut*>er of weld wire as used to fabricate SA-1585.
Fluence, !RT..., and chemical" ccq,osition data are reported in BAW-2166.
Chemistry Factor for SA-1526 was calculated from Surry, Crystal River 3 and TMl-1 surveillance data that was reported in BAW-1803, Rev. 1. The surveillance welds were fabricated using the same heat nuit>er weld wire as SA-1526.
Since the IRT... data for Rotterdam welds are similar to Linde 80 welds, the IRT.... for Rotterdam weld are botrding Linde 80 generic values CIRT... = 0°F, Std. Dev.= 20°F).
Su~ry File for Pressurized Therma~hoc~
Plant Beltline Heat No.
ID Neut.
IRT....
Method of Chemistry Method of XCu Xiii Name
!dent.
I dent.
Fluence at Determin.
Factor Determin.
EOL IRT...,.
CF Surry 2 Nozzle 123V303VA1 4.45E18 30°F Plant 58 Table 0.09 0.73 belt specific forging EOL:
Lower C4208*2
- 3. 71E19
- 30°F Plant 107.25 Table
- 0. 15 0.55 1/29/2013 shell specific plate Lower C4339*1 3.71E19
- 10°F Plant 66.531 Calculated o.,,
0.54 shell specific plate Int. shell C4331*1 3.71E19
- 10°F Plant 83 Table 0.12 0.60 plate specific Int. shell C4339*2 3.71E19
- 20°F Plant 66.531 Calculated 0.11 0.54 plate specific Int. shell 72445 7,75E18
- 5°F Generic 146.09 Calculated 0.21 NA 0.59 NA axial welds SA-1585 Int. and 8T1762 7.75E18
- 5°F Generic 152.25 Table 0.20 0.55 Lower shell axial seams WF-4 Int. to 0227 3.71E19 0°F Generic 128.01 Calculated 0.19 0.56 lower Rotterdam shell circ. weld R3008 Nozzle 4275 4.45E18 0°F Generic 160.5 Table 0.35
- 0. 10 belt to Rotterdam int. circ.
weld L737 References Chemistry Factor for SA-1585 was calculated from Point Beach 1 and Crystal River 3 surveillance data that was reported in BAW 1803, Rev. 1.
The surveillance welds were fabricated using the same heat nurber of weld wire as used to fabricate SA-1585.
Since the !RT.... data for Rotterdam welds are similar to Linde 80 welds, the IRT.... for Rotterdam weld are bounding Linde 80 generic values (IRT".,. = 0°F, Std. Dev.= 20°F).
Fluence, IRT"" and chemical composition data are reported in BAW-2166.
Chemistry Factor for weld R3008 was calculated from surveillance data reported in BAW 2166.
e ENCLOSURE 2 e
Summary File for Upper Shelf En-ergy Plant Name Bel tline Heat No.
Material 1/4T USE 1/4T Unirrad; Method of I dent.
Type at EOL Neutron USE Determin.
Fluence at Unirrad.
EOL USE Surry 1 Nozzle 122V109VA1 A 508-2 71 3.31E18 83 65X belt forging EOL:
Int. shell C4326-1 A 5338-1 106 2.76E19 115 Direct 5/25/2012 Dlate Int. shell C4326-2 A 5338-1 86 2.76E19 93 65X plate Lower C4415-1 A 5338-1 95 2.76E19 103 Direct shell Dlate Lower C4415-2 A 5338-1 74 2.76E19 80 65X shell Dlate Int. &
8T1554 Linde 80, EMA' 4.45E18 EMA 2
Generic lower SAW axial welds, L 1, L3, & L4 SA-1494 Lower 299L44 Linde 80, 51 4.45E18 70 Surv. Weld shell SAW axial weld L2 SA-1526 Int. to 72445 Linde 80, 43 (EMA')
2.76E19 77 Sister lower SAW Plant shell circ. weld SA-1585 Nozzle 25017 SAW 58 3.31E18 90 7
Generic belt to Rotterdam int. shell circ. weld J726 2L i censee must confirm applicability of Topical Reports BAW-2178P and BAW-2192P 7Additional information required to confirm value
e Summary File for Upper Shelf Energy Plant Name Beltline I dent.
References Heat No.
Material Type 1/4T USE It EOL Fluence and chemical c~sftion data are in BAW-2166.
1/4T Neutron Fluence at EOL Unirrad.
USE Method of Determin.
Unirrad.
USE UUSEs for welds SA-1494, J726, SA-1585 and SA-1526 are from BAW-1803, Rev. 1. X Drop in USE for Int. and Lower Shell plates determined from surveillance data from plate C4415*1, in accordance with Section 2.2 of RG 1.99, Rev. 2. Plate C4415*1 has the same percentage copper as the other beltline plates.
UUSE for forging 122V109VA1, plates C4326*1, C4326*2, C4415*1 encl C4415*2 are reported in WCAP 11492, Rev. 1.
Percent drop in USE for Intermediate and Lower shell plates were determined from surveillance data from plate C4415*1, in accordance with Paragraph C.2.2 of RG 1.99, Rev. 2. Plate C4415*1 has the same percentage copper as the other beltline plates.
e e
Surrunary File for Upper Shelf Energy Plant Name Bel tl ine Heat No.
Material 1/4T USE 1/4T Unirracl.
Method of I dent.
Type at EOL Neutron USE Detennin.
Fluence at Unirrad.
EOL USE Surry 2 Nozzle 123V303VA1 A 508*2 89 2.80E18 103 65X belt forging EOL:
Int. shell C4208*2 A 533B*1 67 2.33E19 94 65X 1/29/2013 plate Int. shell C4339*1 A 533B*1 80 2.33E19 94 Direct plate Lower C4331 *2 A 533B*1 63 2.33E19 84 65X shell plate Lower C4339*2 A 533B*1 63 2.33E19 83 65X shel L plate Int. shell n445 Linde 80, 54 4.87E18 77 Sister axial SAW Plant welds SA-1585 Lower 8T1762 Linde 80, EMA" 4.87E18 EMA 2
Generic shell SAW axial seams WF-4 Int. to 0227 Rotterdam 58 2.33E18 90 Surv. Weld Lower SAW shel L circ. weld I
R3008 Nozzle 4275 Rotterdam 59
. 2.80E18 907 Generic belt to SAW int. circ.
weld L737 References Fluence and chemical CCll1)0Sition data are reported in BAW-2166.
UUSE for L737 and SA-1585 is reported in BAW-1803, Revision 1.
UUSE for forging 123V303VA1, plates C4208*2, C4339*1, C4331*2 and C4339-2 is reported in WCAP 11505.
UUSE for R3008 is reported in WCAP-11499.
2Licensee must confirm applicability of Topical Reports BAW-2178P and BAW-2192P 7Additional information required to confirm value Nomenclature and Tables NOMENCLATURE Pressurized Thermal Shock Table Column 1:
Column 2:
Column 3:
Column 4:
Column 5:
Column 6:
Plant name and date of expiration of license.
Beltline material location identification.
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.
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 submittal s).
Unirradiated reference temperature.
Method of determining unirradiated reference temperature
(!RT).
Plant-Specific This indicates that the !RT was determined from tests on material removed from the same heat of the beltline material.
MTEB 5-2 This indicates that the unirradiated reference temperature was determined from following MTEB 5-2 guidelines for cases where the !RT 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 Table 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 I:
Column 2:
Column 3:
Column 4:
Column 5:
Plant name and date of expiration of license.
Beltline material location identification.
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.
Material type; plate types include A 5338-I, A 302B, A 302B Mod., and forging A 508-2; weld types include SAW welds using Linde 80, 0091, 124, 1092, ARCOS-BS flux, Rotterdam welds using Graw Lo, SMIT 89, LW 320, and SAF 89 flux, and SMAW welds using no flux.
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.)
EMA 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.
EMA 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 °F This indicates that the unirradiated USE was derived from Charpy test conducted at 10, 30, 40, or 50 °F.
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.
Blank indicates that there is insufficient data to determine the unirradiated USE.
Memorandum Dated Mgy 24, 1994 Distribution
-@0*~l<'e:t'2gJ:l*e::*
NRC & Local PDRs PDII-2 Reading S. Varga, 14/E/4 G. Lainas, 14/H/l H. Berkow E. Tana B. Buckley D. McDonald OGC ACRS (10)
D. Verel l i, RII