ML17228A588
| ML17228A588 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 05/26/1994 |
| From: | Norris J Office of Nuclear Reactor Regulation |
| To: | Goldberg J FLORIDA POWER & LIGHT CO. |
| References | |
| GL-92-01, GL-92-1, NUDOCS 9406080356 | |
| Download: ML17228A588 (13) | |
Text
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Docket Nos.
50-335 and 50-389 0
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 26, 1994 Hr. J.
H. Goldberg President Nuclear Division Florida Power and Light Company Post Office Box 14000 Juno
- Beach, Florida 33408-0420
Dear Hr. Goldberg:
SUBJECT'ENERIC LETTER (GL) 92-01, REVISION 1, "REACTOR VESSEL STRUCTURAL INTEGRITY," ST.
LUCIE PLANT, UNITS 1
AND 2, (TAC NOS.
H83505 AND H83506)
By letters dated July 1,
- 1993, and November 15, 1993, Florida Power
& Light Company (FPL) provided its 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, 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 l.
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 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 responses to GL 92-01 and previously docketed information.
References to the specific source of the data are provided in the tables.
For St. Lucie 1, we have determined that additional data is required to confirm that the USE at end-of-life (EOL) for one of your beltline materials, weld 2-203, 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 95X confidence that at least 95X of the population is greater than the lower tolerance limit. If the lower tolerance limit results in a projected USE at 9406080356 940526 gRI; lqI.E MWTM&PV
Hr. J.
H. Goldberg t 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 Hechanical Engineers Boiler and Pressure Vessel Code.
We request that you submit within 30 days a schedule for providing the required data.
- 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 the St.
Lucie 1 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 for St. Lucie 1.
When your analyses are submitted, they will be reviewed as a plant-specific licensing action.
For St. Lucie 2, we request that you verify that the information that you have provided for your facility has been accurately entered in the summary data files.
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 tables for future NRC assessments of the St. Lucie 2 pressure vessel.
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 Hanagement and Budget Clearance Number 3150-0011, which expires June 30, 1994.
Sincerely, (Original Signed By)
Jan A. Norris, Sr. Project Hanager Project Directorate II-2 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 a e Distribution sDocket...F..i.l e NRC 3 Local PDRs SVarga GLainas Verelli, RII HBerkow JNorris ETana OGC ACRS (10) 0FFIGE
-LA:PDI I-2 ETana 7
DATE 05 D5 94 D I-2 llNorri D:
D HBert. w 05 >7 94 OFFICIAL RECORD COPY DOCUHENT NAHE: S:i83505.JAN
0 II 4
i E7 l
1
Hr. J.
H. Goldberg Florida Power and Light Company CC:
Jack Shreve, Public Counsel Office of the Public Counsel c/o The Florida Legislature ill West Madison Avenue, Room 812 Tallahassee, Florida 32399-1400 Senior Resident Inspector St. Lucie Plant U.S. Nuclear Regulatory Commission 7585 S.
Kwy AIA Jensen
- Beach, Florida 34957 Hr. Joe Myers, Director Div. of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, Florida 32399-2100 Harold F. 'Reis, Esq.
Newman 8 Holtzinger 1615 L Street, N.W.
Washington, DC 20036 John T. Butler, Esq.
- Steel, Hector and Davis 4000 Southeast Financial Center Miami, Florida 33131-2398 Mr. Thomas R.L. Kindred County Administrator St. Lucie County 2300 Virginia Avenue Fort Pierce, Florida 34982 Hr. Charles B. Brinkman, Manager Washington Nuclear Operations ABB Combustion Engineering, Nuclear Power 12300 Twinbrook Par kway, Suite 330 Rockville, Maryland 20852 St. Lucie Plant Hr. Bill Passetti Office of Radiation Control Department of Health and Rehabilitative Services 1317 Winewood Blvd.
Tallahassee, Florida 32399-0700 Regional Administrator, RII U.S. Nuclear Regulatory Commission 101 Marietta Street N.W., Suite 2900 Atlanta, Georgia 30323 Hr. H. N. Paduano, Manager Licensing
& Special Projects Florida Power and Light Company
'P.O.
Box 14000 Juno
- Beach, Florida 33408-0420 D. A.'ager, Vice President St. Lucie Nuclear Plant P.O.
Box 128 Ft. Pierce, Florida 34954-0128 C. L. Burton Plant General Manager St. Lucie Nuclear Plant P.O.
Box 128 Ft. Pierce, Florida 34954-0128
Enclosure
'1 Summary File for Pressurized Thermal Shock Plant Name St.
Lucie 1
EOL:
3/1/2016 References Beltline Ident.
lnt. sheLL C.7-1 lnt. sheLL C.7.2 lnt. sheLL C-7.3 Lover shell C-8.1 Lover shell C-8-2 Lover shell C-8-3 lnt. shell axial wetds 2.203 Lover sheLL axial uetds 3-203 lnt. to Lover shell circ. Meld 9.203 Heat, No.
Ident.
A 4567-1 8 9427-1 A 4567-2 C.5935.1 C.5935.2 C.5935.3 A.8746 8 348009 305424 90136 ID Keut.
Fluence at EDL 3.372E19 3.372E19 3.372E19 3.372E19 3.372E19 3.372E19 2.131E19 2.131E19 3.372E19 O'
O' O'
200F 20'F O'
56'F
.56'F
-60'F Hethod of Determin.
IRT HTEB 5 2 HTEB 5.2 HTEB 5.2 HTEB 5 2 Plant speci fic HTEB 5.2 Generic Generic Plant speci fic Chemi s try Factor 74.6 74.6 73.8 79.418 60.853 92 192 84.79S Hethod of Determin.
CF Table Table Table Calculated Calculated Calculated Table Table Calculated 0.11 0.11 0.11 0.15 0.15 0.12 0.19 0.28 0.23 0.64 0.64 0.58 0.56 0.57 0.58 0.10 0.63 0.11
- Fluence, chemical composition, and IRT~ data are from JuLy 1, 1992, Letter from M. H. Bohlke (FPL) to USKRC Docunent Control Desk, subject:
Generic Letter 92.01, Revision 1, Response Chemical coeposttions for uetds 2-203 and 3-203 are froa the Noveater 15, 1993 Letter from D.A. Sager to USNRC, est.
Lucia Units 1 and 2 Response to Request for Additional Information, Generic Letter 92-01, Revision 1.
~,
~
Summary File for Pressurized Thermal Shock c'nclosLtre 2 (Cont.)
Plant Name St.
Lucie.2 EOL:
4/6/2023 Bettlinc Ident.
Lower shell plate H-4116.1 Lower shell plate H.4116.2 Lover sheLL plate H-4116.3 Int. shelL plate H.605 1
Neat No.
Ident.
8.8307-2 A.3131.1 A.3131.2 A.8490.2 ID Neut.
FLuence at EOL 3.07E19 3.07E19 3.07E19 3.07E19 20'F 20'F 20'F 30'F Hethod of Determin.
IRT Plant speci fic Plant speci fic Plant speci fic Plant speci fic Chemistry Factor 37 74.15 Hethod of Determin.
CF Table Table Table Table 0.06 0.07 0.07 0.11 0.57 0.60 0.60 0.61 Int. shelL plate H.605.2 8-3416-2 3.07E19 10'F Plant specific 91.5 Table 0.13 0.62 References Int. shell plate H 605-3 Int. shell axial welds 101-124 Int. shell axial wetds 101.124 Lower shell axiaL welds 101-142 Int. to lower shell circ. weld 101.171 Int. to
'lower sheLL circ. weld 101-171 A 8490-2 83637 3P7317 3.07E19 3.07E19 3.07E19 3.07E19 3.07E19 3.07E19 O'
80'F
.50'F
.50'F
-70'F
.80'F Plant speci fic P lant Speci fic Plant Speci fic Plant Speci fic Plant speci fIc Plant Speci fic 74.15 30.65 30.65 37.5 41.2 41.2 Table Table Table Table Table Table 0.11 0.04 0.04 0.05 0.07 0.07 0.61 0.07 0.07 0.10 0.08 0.08 Alt data are from the November 15, 1993 Letter from D. A. Seger to USNRC, eSt. Lucie Units 1 and 2, Response to Request for Additional Information GL 92-01, Revision 1."
Enclosure 2
Summary File for Upper Shelf Energy Plant Name St.
Lucia 1
EOL:
3/1/2016 References Belt line ident+
lnt. shell c-7-1 int. shell C.7-2 int. shell C-7-3 Lo~er shell C-8-1 Lower shell C-8-2 Lower shell C.8-3 int. shell axial welds 2-203 Lower shell axial welds 3.203 int. to lower shell circ. weld 9.203 Heat No.
A-4567-1 B.9427.1 A.4567.2 C 5935-1 C-5935-2 C.5935.3 A.8746 B
34B009 305424 Haterial Type A 5338.1 A 5338-1 A 5338.1 A 533B-1 A 533$ -1 A 5338-1 Linda 124, SAM Linda 1092~
SAM Linda
- 0091, SAll 1/4T USE at EOL 62 62 57 54 65 1/4T Neutron Fluence at EOL 2.01E19 2.01E19 2.01E19 2.01E19 2.01E19 2.01E19 1.27E19 1.27E19 2.01E19 Unirrad.
USE 76 82 103 102
'12 Hethod of Determin.
Unirrad.
USE 65K 65X Direct 65K Direct 65X Generic Sister Plant Surv. Meld WSE data for all plates except C-8-2 are from the July 1, 1992 Letter from M.H. Bohlke to USHRCg St. Lucie Units 1 and 2, GL 92-01, Revision 1 Response.4 WSE data for plate C-8-2 and the welds are from the Hoveeber 15, 1993 Letter from D.A. SaBer to USNRC, "St. Lucia Units 1 and 2, Response to Request for Additional information, GL 92-01, Revision 1.4 Additional information required to confirm value
Summary File for Upper Shelf Energy Plant Name Beltline Neat No.
Haterial Ident.
Type 1/4T USE at EOL 1/4T Neutron Fluence at EOL Unirrad.
USE Hethod of Determin..
Unirrad.
USE St.
Lucia 2 EOL!
4/6/2023 Lover shell plaCe H.4116-1 Lover shell plate H.4116.2 Lover shell plate H.4116.3 B.8307-2 A 533B-1 A-3131-1 A 5338-1 A.3131-2 A 5338.1 71 1.83E19, 1.83E19 1.83E19 91 105 100 Direct Direct Direct Int. shell A-8490-2 A 5338-1 plaCe H-605-1 Int. shell 8-3416-2 A 5338-1 plate H.605.2 Int. shell A-8490-2 A 5338-1 plate H 605.3 81 81 1.83E19 1.83E19 1.83E19 105 113 113 Direct Direct Direct Int. shell 83642 axial voids 101.124 Int. shell 83&37 axial velds 101-124 Lover shell axial velds 101-142 Int. to lover shell circ. veld 101.171 Linda 0091 90 Linda 0091 106 Linda 0091 106 Linda 124 88 1.83E19 1.83E19 1.83E19 1.83E19 116 136 136 115 Direct Direct Direct Direct Int. to lover shell circ. veld 101-171 3P7317 Linda 124 75 1.83E19 96 Direct All data are from the Neve@her 15, 1993 letter from D.A. SaBer to USNRC, "St. Lucie Units 1 and 2, Response to Request for Additional Information GL 92-01, Revision 1."
Nomenclature and Tables
PRESSURIZED THERMAL SHOCK AND USE TABLES FOR ALL PWR PLANTS NOMEN LATURE Pressurized Thermal Shock Table Column Column Column Column Column Column I
~
2:
3 t 4
~
5
~
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 submittals).
Unirradiated reference temperature.
Method of determining unirradiated reference temperature (IRT).
Ppl-Sp if'his indicates that the IRT was determined from tests on material removed from the same heat of the beltline material.
Column Column 7
~
8:
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.
Chemistry factor for irradiated reference temperature evaluation.
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 Dat 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 Dat 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 Column Column Column Column Column I
~
2:
3
~
4
~
a 5
~
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.
Material type; plate types include A 5338-1, A 302B, A 3028 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.
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.)
/MA This indicates that the USE issue may be covered by either owners group or plant-specific equivalent margins analyses.
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).
I
~ Column 7:
Unirradiated USE.
~EM 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.
65X This indicates that the unirradiated USE was 65X of the USE from a longitudinal specimen.
I Generic This indicates that the unirradiated USE was reported by the licensee from other plants with similar materials to the beltline material.
'~RRC 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.
E iv. to Surv.
Weld This indicates that the unirradiated USE was from the surveillance weld having different weld wire heat number.
~RI<<PI I
This indicates that the unirradiated USE was derived by using the reported value from other plants with the same weld wire heat number.
~Bla Ig indicates that there is insufficient data to determine the unirradiated USE.
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