ML17263A626
| ML17263A626 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/12/1994 |
| From: | Andrea Johnson Office of Nuclear Reactor Regulation |
| To: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| References | |
| GL-92-01, GL-92-1, TAC-M83733, NUDOCS 9404200361 | |
| Download: ML17263A626 (12) | |
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Docket No. 50-244 UNITED STATES NUCLEAR REGULATORY COMMISSION" WASHINGTON, O.C. 20555-0001 April 12, 1994 Dr. Robert C. Mecredy Vice President, Nuclear Production Rochester Gas
& Electric Corporation 89 East Avenue Rochester, New York 14649
Dear Dr. Hecredy:
SUBJECT:
GENERIC LETTER (GL) 92-01, REVISION 1, "REACTOR STRUCTURAL INTEGRITY," ROCHESTER GAS AND ELECTRIC CORPORATION R.E.
GINNA NUCLEAR POWER PLANT (TAC NO. H83733)
By letters dated July 2,
- 1992, and November 29 and April 21,
- 1993, Rochester Gas and Electric Corporation (RG&E) provided its response to GL 92-01, Revision 1.
The NRC staff has completed its review of your responses.
Based on its review, the staff has determined that RG&E 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 Reactor Vessel Integrity Database (RVID).
The RVID contains the following tables:
A pressurized thermal shock (PTS) table for PWRs, a
pressure-temperature limit table for BWRs, and an upper-shelf energy (USE) table for PWRs and BWR's.
Enclosure 1 provides the PTS table(s),
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 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.l.
To demonstrate that the topical reports are applicable to Ginna, you must compare the limiting material properties of the Ginna reactor vessel to the rIn
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April 12, 1994 values reported in the topical reports.
This review will be a plant-specific licensing action.
We further request that you verify that the information you have provided for your facility 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 a 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'egulations.
This action is covered by the Office of Management and Budget Cle'arance Number 3150-0011, which expires June 30, 1994.
'Sincerely, Original signed by:
Allen R.,Johnson, Project Manager Project Directorate,I-3
'ivision of Reactor Projects - I/II
.Office of Nuc]ear Reactor Regulation 1
i
Enclosures:
1.
Pressurized Thermal Shock "
Table 2.
Upper-Shelf Energy Table 3.
Nomenclature Key cc w/enclosures:
See next page DISTRIBUTION:
(Docket File NRC
& Local PDRs PDI-3 Reading SVarga JCalvo WButler SLittle AJohnson DHcDonald JStrosnider EPoteat KBohrer BElliott OGC ACRS (10)
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//0/94 OATE OFFICIAL RECORD COPY DOCUHEMT NAME: A:iM83733.HPA PDI-3:PM 5 AJohnson:dt
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April 12, 1994 values reported in the topical reports.
This review will be a plant-specific licensing action.
Me further request that you verify that the information you have provided for your facility 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 a 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,
Enclosures:
1.
Pressurized Thermal Shock Table 2.
Upper-Shelf Energy Table 3.
Nomenclature Key cc w/enclosures:
See next page All n R. Johnson, Pr ject Manager Proj t Directorate
-3 Divisio Projects I/II Office of Nuclear Reactor Regulation
Dr. Robert C. Hecredy R.E. Ginna Nuclear Power Plant CC:
Thomas A. Moslak, Senior Resident Inspector R.E.
Ginna Plant U.S. Nuclear Regulatory Commission 1503 Lake Road
- Ontario, New York 14519 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Hs.
Donna Ross Division of Policy Analysis 8 Planning New York State Energy Office Agency Building 2 Empire State Plaza
- Albany, New York 12223 Charlie Donaldson, Esq.
Assistant Attorney General New York Department of Law 120 Broadway New York, New York 10271 Nicholas S.
Reynolds Winston 5 Strawn 1400 L St.
N.W.
Washington, DC 20005-3502 Hs. Thelma Wideman
- Director, Wayne County Emergency Management Office Wayne County Emergency Operations Center 7370 Route 31
- Lyons, New York 14489 Ms. Mary Louise Heisenzahl Administrator, Monroe County Office of Emergency Preparedness 111 West Fal-1
- Road, Room ll Rochester, New York 14620
Enclosure i
Summary File for Pressurized Thermal Shock Plant Name 6 irma EOL:
9/18/2009 Beltline Ident.
Nozzle Shell Forging Int. Shell Forging Heat No.
Ident.
123P118VA1 125S255VA]
ID Neut.
Fluence at EOL/EFPY 3.69E18 3.35E19 30'F 20oF Hethod of Determin.
IRT Plant Specific Plant S elfic Chemistry Factor 223.6 44 Nethod of Determin.
CF Table Table
~ 35 0.07 0.68 0.68 Lower Shell Forging Int. to Lower Shell Circ. Held SA-847 Lower Shell to Dutchman Circ. Meld SA-848 125P666VA1 61782 61782 3.35E19 3.35E19 1.6E16 40'F
-5'F
-5'F Plant Specific Generic Generic 26.23 133.94 133.94 Calculated 0.05 Calculated 0%25 Calculated 0.25 0.68 0.54 0.54 Nozz le/ Int Shell Circ. Held SA-1101 References for Ginna 71249 3.72E18 10oF Plant Specific
'80 Table 0.26 0.60
- Fluence, IRT~, and chemical composition data are from July 2, 1992,.letter from R. C. Hecredy (RGKE) to A.
RE Johnson (USNRC), subject:
Reactor Vessel Structural Integrity, 10CFR50.54(f),
Response
to Generic Letter 92-01, Revision 1, R.
G. Ginna Nuclear Power Plant Chemistry Factor for welds fabricated using weld were heat nuaber 61782 were determined from surveillance data from Davis Besse and Ginna, which had welds fabricated using heat nurher 61782.
The data is reported in BAII 1803, Rev. 1.
The copper content for forging 123P118VA1 was a default value in the PTS rule because the licensee did not report a
value.
Enc1osure 2
Summary File for Upper Shelf Energy Plant Name geltline Ident.
Heat Ho.
Naterial Type 1/4T USE at EOL/EFPY 1/4T Neutron Fluence at EOL/EFPY Unirrad.
USE Nethod of Determin.
Unirrad.
USE Ginna EOL:
9/18/2009 Nozzle Shell Forging Int. Shell Forging Lower Shell Forging 123P118VA1 A 508-2 125S255VA1 A 508-2 125P666VA1 A 508-2 70 2.50E18 2.27E19 2.27E19 112 91 119 65K 65X 65K Int. to Lower Shell Circ. Meld SA-847 Lower Shell to Dutchman Circ. Meld SA-848 Nozzle/Int Shell Circ. 'Meld SA-1101 References for Ginna 61782 61782 71249 Linda 80, SAM Linda 80, SAM Linda 80, SA'M ENA*
ENA 2.27E19 1 ~ 08E16 2.52E18 ENA ENA'NA*
Generic Generic Generic Fluence and chemical carposition data are from July 2, 1992, letter from R. C. Necredy (RG&E) to A.
RE Johnson (USHRC), subject:
Reactor Vessel Structural Integrity, 10CFR50.54(f),
Response
to Generic Letter 92-01, Revision 1, R; G. Ginne Nuclear Power Plant UUSE data for forging 125S255VA1 and weld SA-847 are from July 2, 1992, letter from R. C.
Necredy (RG&E) to A. R. Johnson (USNRC), subject:
Reactor Vessel Structural Integrity, 10CFR50.54(f),
Response
to Generic Letter 92-01, Revision 1, R. G. Ginna Nuclear Power Plant UUSE data for forgings 123P118VA1 and 125P666VA1 Mere reported in MCAP-8421, "Analyses of Capsule R."
Licensee must confirm applicability of Topical Reports BAW-'2178P and BAW-2192P
Enclosure 3
PRESSURIZEO THERMAL SHOCK TABLES AND USE TABLES FOR ALL PWR PLANTS NOMENCLATURE Pressurized Thermal Shock Table Column Column Column Column Column Column Io 2:
3
~
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'i This indicates that the IRT was determined from tests on material removed from the same heat of the beltline material.
Column 7:
Column 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.)'o 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.
I Upper Shelf Energy Table Column Column Column Column Column Column
]
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3
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4
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5
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6:
Plant name and date of expir ation 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.
Haterial type; plate types include A 5338-1, A 302B, A 302B Hod.,
and forging A 508-2; weld types include SAW welds using Linde 80,
- 0091, 124,
- 1092, ARCOS-B5 flux, Rotterdam welds using Graw Lo, SHIT 89, LW 320, and SAF 89 flux, and SHAW 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.)
~EH This indicates that the USE issue may be covered by the approved equivalent margins analysis in a topical report.
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 submi,ttals).
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Column 7:
Unirradiated USE.
EHA This indicates that the USE issue may be covered by the approved equivalent margins analysis in a topical report.
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.
65K This indicates that the unirradiated USE was 65K 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.
~RC 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
'his 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 uiv. 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.