ML20029E738
| ML20029E738 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 05/12/1994 |
| From: | Wharton L Office of Nuclear Reactor Regulation |
| To: | Schnell D UNION ELECTRIC CO. |
| References | |
| GL-92-01, GL-92-1, TAC-M83445, NUDOCS 9405200015 | |
| Download: ML20029E738 (8) | |
Text
fs wW Q g'[A UNITED STATES e
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NUCLEAR REGULATORY COMMISSION f
WASHINGTON. D.C. 20556 0001 e
May 12, 1994 Docket No. 50-483 Mr. Donald F. Schnell Senior Vice President - Nuclear Union Electric Company Post Office Box 149 St. Louis, Missouri 63166
Dear Mr. Schnell:
SUBJECT:
CALLAWAY PLANT, UNIT 1 - GENERIC LETTER (GL) 92-01, REVISION 1,
" REACTOR VESSEL STRUCTURAL INTEGRITY," (TAC N0. 83445)
By letter dated June 16, 1992, Union Electric Company 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 Union Electric Company 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 database 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 table, Enclosure 2 provides the USE table for your facility, 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 g
to GL 92-01 and previously docketed information. References to the specific source of the data are provided in the tables.
We request that you verify that the information you have provided for your facility is accurate as indicated in Enclosures 1 and 2.
No response is necessary unless an inconsistency is identified.
If no comments are received within 30 days from the date of this letter, the staff will consider your actions related to GL 92-01, Revision 1, to be complete and the staff will use the information in the enclosed tables for future NRC assessments of your reactor pressure vessel.
The information requested by this letter is within the scope of the overall burden estimated in GL 92-01, Pevision 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 i
ISOCEI
- ping Unga PP 7" "" MN P
a i
i Mr. Donald F. Schnell May 12, 1994 I
1 I
to the identified response-related matters and does not include the time required to implement actions required by the regulations.
This action is j
covered by the Office of Management and Budget Clearance Number 3150-0011, which expires June 30, 19P4.
i Sincerely, I
a ORIGINAL SIGNED BY:
L. Raynard Wharton, Project Manager Project Directorate III-3 i
Division of Reactor Projects III/IV i
Office of Nuclear Reactor Regulation
Enclosures:
1.
Pressurized Thermal Shock 1
Table i
2.
Upper-Shelf Energy Table 3.
Nomenclature Key 1
CC W/ enclosures:
See next page t
QlSTRIBUTION Docket File-Local & NRC PDRs D. Mcdonald PDIII-3 r/f J. Roe E. Hackett 2
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O Mr. D. F. Schnell Callaway Plant Union Electric Company Unit No I cc:
Cermark Fletcher Associates Mr. Neil S. Carns 18225 Flower Hill Way #A President and Chief Gaithersburg, Maryland 20879-5334 Executive Officer Wolf Creek Nuclear Operating Gerald Charnoff,.Esq.
Corporation L
Thomas A. Baxter, Esq.
P. O. Box 411 Shaw, Pittman, Potts & Trowbridge Burlington, Kansas 66839 2300 N. Street, N.W.
Washington, D.C.
20037 Mr. Dan I. Bolef, President Kay Drey, Represent &tive Mr. T. P. Sharkey Board of Directors Coalition Supervising Engineer, for the Environment Site Licensing 6267 Delmar Boulevard Union Electric Company University City, Missouri 65130 Post Office Box 620 Fulton, Missouri 65251 l
U.S. Nuclear Regulatory Commission Resident. Inspectors Office 8201 NRC Road' Steedman, Missouri 65077-1302 Mr. Alan C. Passwater, Manager Licensing and Fuels Union Electric Company Post Office Box 149 St. Louis, Missouri 63166 l
l Manager - Electric Department Missouri Public Service Commission 301 W. High Post Office Box 360 Jefferson City, Missouri 65102 Regional Administrator U.S. NRC, Region III 801 Warrenville Road j
Lisle, Illinois 60523-4351 Mr. Ronald A. Kucera, Deputy Director Department of Natural Resources P. O. Box 176 Jefferson City, Missouri 65102 1
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l ENCLOSURE 1 l'
Summary File for Pressurized Thermal Shock I
l Plant Beltline Heat No.
10 Neut.
IRT Method of Chemistry Method of ICu
%Ni I
hame Ident.
Ident.
Fluence et Detersin.
Factor
- Determin, i
EOL/EFPY I R T, CF
}
i Callaway Int. shelt 2.39E19 40'F Plant 26 Table 0.04 0.57 1
R2707 1 Specific i
EOL:
Int. Shell 2.39E19 10*F Plant 31 Table 0.05 0.59 10/18/
R2707 2 Specific 2024 t
Int. Shell 2.39E19 10'F Plant 37 Table 0.06 0.61 j
R2707 3 Specific t
Lower 2.39E19 50'F Plant 26.139 Calcuteted 0.07 0.59 Shell Specific R2708 1 Lower 2.39E19 10'F Plant 31 Table 0.05 0.57 Shell Specific
(
R2708 2 1
Lower 2.39E19 20'F Plant 44 Table 0.07 0.59 j
Shelt Specific j
R2708 3 e
i int, and 2.39E19 60*F Plant 29,7 Table 0.04 0.06 l
Lower Specific Shell AAlal Welds i
G2.03 l~
Int. to 90077 2.39E 19 60'F Plant 56.144 Calculated 0.04 0.07 Lower 2
2 Specific Shell Cire. Weld i
E3.14 t
SQ erences f
i Chemical coripesition and IRT are from Table A 3 of R.G. Lott, et.al, " Analysis of Capsule U from the union Electric Coccany Cat taway Unit i Reactor vessel Radiation Surveillance Program," WCAP 11374, Revision 1, Westinghouse Electric j
Corporation, Pittsburgh, PA 15230, Jme 1987.
i ho heat rueers were supplied for the sheLL and intermedlete and lower exial shett welds, j
fluence data are f rom Table 614 of WCAP 12946 and Decenber 18, 1991, letter from D. F. Schnell (UEco) to USNRC Docunent
)
Control Desk, subject: Revision to Technical Specification 3/4.4.9 Pressure Tenperature Limits.
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ENCLOSURE 2 l
Sumary File for Upper Shelf Energy Plant Name Bottline Meat No.
Material 1/4T USE 1/4T Unirred.
Method of Ident.
Type at EDL Woutron USE Determin.
Fluence at unfrred.
ECL USE Callaway 1 Int, shett R2707 1 A 533s 1 62 1.435E19 78 Direct EOL:
int. Shell R2707 2 A 533B 1 79 1.435E19 100 Direct 10/18/
2024 Int. Shelt R2707 3 A 5338 1 79 1.435E19 99 Direct Lower R2708 1 A 533B 1 65 1.435E19 82 Direct Shell Lower R2708 2 A 5338 1 83 1.435E19 105 Direct Shell l
Lower R2708 3 A 5338 1 80 1.435E19 101 Direct
{
Shell 1
l Int. and Not Not 113 1.435E19 143 Direct Lower provided provided Shell l
Axial
- Welds, C2.03 Int. to 90077 Lirde 124, 87 1.435E19 112 Direct Lower
$AW shell Cire.
l
- Weld, E3.14 ae4t* exes UUSE cats are from Tebte A-3 of R. G. Lott et at, " Analysis of Capsule U from the Union Electric Corpany callaway Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP 11374 Revision 1, Westinghouse Electric Corporation, Pittsburgh, PA 15230, Jme 1987 Fluence data are from Table 614 of WCAP 12946 and December 18, 1991, letter from D. F. Schnell (UEco) to USkRC Cocunent Control Desk, s@ ject: Revision to Technical Specification 3/4.4.9 Pressure Terperature Limits.
4 I
l
ENCLOSURE 3 i
l 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-l wire or tandem-wire process has been reported, (S) indicates i
single wire was used in the SAW process, (T) indicates tandem wire was used in the SAW process.
l Column 4:
End-of-life (E0L) neutron fluence at vessel inner wall; cited i
directly from inner diameter (ID) value or calculated by using l
Regulatory Guide (RG) 1.99, Revision 2, neutron fluence l
attenuation methodology from the quarter thickness (T/4) value l
reported in the latest submittal (GL 92-01, PTS, or P/T limits l
submittals).
Column 5:
Unirradiated reference temperature.
l Column 6:
Method of determining unirradiated reference temperature (IRT).
Plant-Soecific This indicates that the IRT was determined from tests on l
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 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.
l Table This indicates that the chemistry factor was determined from l
the chemistry factor tables in RG 1.99, Revision 2.
Calculated This indicates that the chemistry factor was determined from l
surveillance. data via procedures described in RG 1.99, Revision 2.
l
i
.6 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 i
single wire was used in the SAW process.
(T) indicates tandem j
wire was used in the SAW process.
Column 4: Material type; plate types include A 5338-1, 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 E0L fluence and either the cooper value or the surveillance data.
(Both methods are described in RG 1.99, Revision 2.)
i EMA i
This indicates that the USE issue may be covered by the approved equivalent margins analysis in the B&W Owners Group Topical Reports: EAW-2178P and BAW-2192-P.
Column 6:
E0L 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).
O oo Column 7: Unirradiated USE.
f.!!8 l
l This indicates that the USE issue may be covered by the approved equivalent margins analysis in the B&W Owners Group Topical Reports:
BAW-2178P and BAW-2192P.
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 l
unirradiated USE was from test date.
65%
i 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 l
licensee from other plants with similar materials to the beltline material.
NRC aeneric l
This indicates that the unirradiated USE was derived by the l
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 i
surveillance weld having the same weld wire heat number.
Eauiv. 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.
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