ML16342C616
| ML16342C616 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 07/01/1994 |
| From: | Padovan L Office of Nuclear Reactor Regulation |
| To: | Rueger G PACIFIC GAS & ELECTRIC CO. |
| References | |
| GL-92-01, GL-92-1, TAC-M83456, TAC-M83457, NUDOCS 9407120061 | |
| Download: ML16342C616 (26) | |
Text
~C goR REc(
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.0 Docket Nos.
50-275 and 50-323 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 1, 1994 Mr. Gregory M. Rueger Nuclear Power Generation, B14A Pacific Gas and Electric Company 77 Beale Street, Room 1451 P.O.
Box 770000 San Francisco, California 94177
Dear Mr. Rueger:
SUBJECT:
GENERIC LETTER 92-01, REVISION 1, "REACTOR VESSEL STRUCTURAL INTEGRITY," DIABLO CANYON NUCLEAR POWER PLANT, UNIT NO.
1 (TAC NO. M83456)
AND UNIT NO.
2 (TAC NO. M83457)
By letters dated June 30 and December 4,
- 1992, and October 8, 1993, Pacific Gas and Electric Company (PGEE) responded to Generic Letter (GL) 92-01, Revision 1, "Reactor Vessel Structural Integrity, 10 CFR 50.54(f)."
The NRC staff has completed its review of your responses.
We have determined that PGSE provided the information requested in GL 92-01.
The GL is part of our program to evaluate pressurized-water reactor (PWR) and boiling-water reactor (BWR) reactor vessel integrity.
We are using the information provided in response to GL 92-01, including previously docketed information, to confirm that licensees satisfy the requirements and commitments necessary to ensure reactor vessel integrity for their facilities.
Utilities provided large amounts of information in response to GL 92-01, Revision 1.
We entered this information into a computerized data base called the Reactor Vessel Integrity Database (RVID).
The RVID contains the following tables:
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a PWR pressurized thermal shock (PTS) table
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a BWR pressure-temperature limit table
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a PWR/BWR upper-shelf energy (USE) table Enclosure 1 provides the PTS tables, and Enclosure 2 provides the USE tables for your facilities.
Enclosure 3 is a key for nomenclature used in the tables.
The tables include the data necessary to perform USE and RT evaluations.
These data were taken from your GL 92-01 responses and previously docketed information.
The tables reference the source of the data.
We request that you verify we have accurately entered the information you have provided for your facilities in the summary data file.
No response is necessary unless you identify an inconsistency.
If we do not receive any
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Hr. Gregory H. Rueger Ouly 1, 1994 comments from you within 30 days of the date of this letter, we will consider your actions related to GL 92-01, Revision 1, to be complete.
We will then use the information in the 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, Revision 1.
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 No. 3150-0011, which expires September 30, 1994.
Sincerely, Original Signed By L. Hark Padovan, Acting Project Manager Project Directorate IV-2
'ivisi'on of Reactor Projects III/IV 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:
Do'cket -Fi1 e-'D4-2 Reading EAdensam HPadovan OGC
- ABeach, RIV NRC S. Local PDRs JRoe TQuay
" DFoster-Curseen ACRS (10)
- KPerkins, RIV/WCFO LA:DRPW
.PH:PD D:PD4-2 DFoster-Curseen HPadovan:
k TQuay 94
/94
/94 OFFICIAL RECORD COPY DOCUMENT NAME:
DC83456.LTR
r
Hr. Gregory H. Rueger duly 1, 1994 comments from you within 30 days of the date of this letter, we will consider you) actions related to GL 92-01, Revision 1, to be complete.
We will then use the information in the 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, Revision 1.
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 No. 3150-0011, which expires September 30, 1994.
Sincerely, Original Signed By L. Hark Padovan, Acting Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Enclosures:
1.
Pressurized Thermal Shock Tables
?.
".p; o~-Sl elf Energy Tables 3.
NoI.enclature Key cc w/enclosures:
See next page DISTRIBUTION:
Docket File PD4-2 Reading EAdensam HPadovan OGC
- KPerkins, RIV/WCFO LA:DRPW PH PD D:PD4-2 TQuay
/
/94 DFoster-Curseen HPadovan:
k 1l
<9e
/
/94 OFFICIAL RECORD COPY DOCUMENT NAME:
DC83456.LTR
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Mr. Gregory M. Rueger Pacific Gas and Electric Company Diablo Canyon CC NRC Resident Inspector Diablo Canyon Nuclear Power Plant c/o U.S. Nuclear Regulatory Commission P. 0.
Box 369 Avila Beach, California 93424 Dr. Richard Ferguson, Energy Chair Sierra Club California 6715 Rocky Canyon
- Creston, California 93432 Ms.
Nancy Culver San Luis Obispo Mothers for Peace P. 0.
Box 164 Pismo Beach, California 93448 Ms. Jacquelyn C. Wheeler P. 0.
Box 164 Pismo Beach, California 93448 Managing Editor The Count Tele ram Tribune 1321 Johnson Avenue P. 0.
Box 112 San Luis Obispo, California 93406 Chairman San Luis Obispo County Board of Supervisors Room 370 County Government Center San Luis Obispo, California 93408 Mr. Truman Burns Mr. Robert Kinosian California Public Utilities Commission 505 Van Ness, Room 4102 San Francisco, California 94102 Mr. Steve Hsu Radiologic Health Branch State Department of Health Services Post Office Box 942732 Sacramento, California 94234 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Harris Tower 5 Pavillion 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064 Mr. Peter H. Kaufman Deputy Attorney General State of California 110 West A Street, Suite 700 San Diego, California 92101 Christopher J.
- Warner, Esq.
Pacific Gas
& Electric Company Post Office Box 7442 San Francisco, California 94120 Mr. John Townsend Vice President and Plant Manager Diablo Canyon Power Plant P. 0.
Box 56 Avila Beach, California 93424 Diablo Canyon Independent Safety Committee ATTN:
Robert R. Wellington, Esq.
Legal Counsel 857 Cass Street, Suite D
Monterey, California 93940
~ ~t 1
Hr. Gregory H. Rueger comments from you within 30 days of the date of this letter, we will consider your actions related to GL 92-01, Revision 1, to be complete.
We will then use the information in the 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, Revision 1.
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 8udget Clearance No. 3150-0011, which expires September 30, 1994.
Enclosures:
1.
Pressurized Thermal Shock Tables 2.
Upper-Shelf Energy Tables 3.
Nomenclature Key cc w/enclosures:
See next page L. Hark Padovan, Acting Project Hanager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
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ENCLOSURE 1
Summary File for Pressurized Thermal Shock Plant Name DiabLo Canyon 1
EOL:
4/23/2008 Ee'I'.I Ine Ident.
Int. sheLL 84106. 1 Int. shell 84106.2 Int. shell 84106.3 Lowei.
shell 84107.1 Lo~er shell 84107-2 Lower sheLL 84107.3 Int. sheLL aXlel welds 2'42A, 8
Int. shell axial welds 2.442C Heat No.
Ident.
C2884.1 C2854 2 C2793.1 C3121.1 C3131-2 C3131.1 27204 ID Neut.
Fluence at EOL 1.32E19 1.32E19 1.32E19 1.32E19 1.32E19 1.32E19 S.92E18 4.79E18 10'F
.3'F 30'F 15'F 20'F
-22'F
.56'F
.56'F Nethod of Determin.
IRT Plant S
If ic PLant S
if ic Plant S
tfic Plant Speci fic Plant Specific Plant Speci fIc Generic Generic Chemi stry Factor 96.85 65 89.8 82.2 S2.2 Nethod of Determtn.
CF Table Table Table Tabl ~
Table Tabl ~
Table Table 0.14 0.13 0+10 0.13 0.12 0.12 0.20 0.20 0.53 0.50 0.46 0.56 0.56 0.56 1.00 1.00 Lower shell axial welds 3.442A 8
Lour sheLL ax Ial welds 3-442C 27204 7.42E18 1.32E19
.56'F
.56'F Generic Generic Table Table 0.20 0.20 1.00 1.00 lafmrenees Int. to lower shell circ. weld 9.442 21935 1.32E19
.56'F Generic 167.2 Table 0.18 0.6S IRT, and chemical coaposttton data are from Jlala 30, 1992, Letterfrom G. it. Rueger (POTE) to USNRC Docunent Control Desk, subject:
Response
to Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrtty Fluence and WSE for weld 9-442 are free December 4, 1992, Letter froi G. it. Rueger (POLE) to USNRC Docunent Control Desk, subject:
Response
to Generic Letter 92.01, Revision 1, "Reactor Vessel Structural Integri tyw--Supplemental Informat ion
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~I ~
Summary File for Pressurized Thermal Shock PLant Heme Diablo Canyon 2
EOL:
12/9/2010 References Beltl ine Ident.
tnt. sheLL
$5454-1 Int. shell
$5454.2 Int. shelL B5454.3 Lower shell
$5455'1 Lower shell
$5455-2 Lo~er shell
$5455.3 Int. shell axial welds 2-20 IA Int. shell axiaL welds 2-201$,
C Int. to lower shell circ. weld 9.201 Lower shell axial welds 3-201A, C
Lower shell axial welds 3-2018 Heat Ho.
Ident.
C5161.1 C5168.2 C5161.2 C5175.1 C5175 2
C5176.1 12008 and 21935 12008 and 21935 10120 33A277 33A277 ID Naut.
Fluence at EOL 14.6E18 14.6E18 14.6E18 14.6E18 14.6E18 14.6E18 9.34E18 10.1E18 14.6E18 10.1E18 9.34E18 52'F 67'F 33'F
.15'F O'
15'F
.50'F
.50'F
.56'F
-56'F
.56'F Kethod of Oetermin.
IRT Plant S
ific Plant S
if ic Plant S
ific Plant Specif ic Plant Speci fic Plant Speci fic Plant Specific Plant Speci fic Generic Generic Generic Chemistry Factor 110.17 99.55 110.5 98.2 98.2 65.2 223.02 223.02 26.85 129.15 129.15 Kethod of Oetermin.
CF Calculated Table Tabl ~
Table Table Tabl ~
Ca Iouiated Calculated Table Table Table 0.15 0.14 0.15 0.14 0.14 0.10 0.22 0.22 0.04 0.26 0.26
.59 0.62 0.56 0.56 0.62 0.85 0.85 0.03 0.19 0.19 IRT and chemicaL ccepoaition data are from JIRIe 30, 1992, Letter from G. K. Rueger (POTE) to LISNRC Oocunent ControL Desk, subject:
Response
to Generic Letter 92.01, Revision 1, Reactor Vessel Structural Integrity Fluenoe and WSES far Weld 9-20'I and 3-201 are fria Oeoeeher 4, 1992, Letter fry G. K. Rueger (PGSE) tO uSRRC OOCISnenI Control Desk, subject:
Response
to Generic Letter 92.01, Revision 1, "Reactor Vessel Structural Integrity""
Supplenental Information
ENCLOSURE 2
Summary File for Upper Shelf Energy PLant Name SeLtline ident.
Heat No.
Nateri ~ L Type 1/4T USE at EOL 1/4T Ne4tron FL4ence at EOL Unirrad.
USE Nethod of Deterain.
Unirrad.
USE Diablo Canyon 1
EOL:
4/23/200 8
lnt. shell 84106.1 lnt. shell
$4106.2 lnt. shell
$4106 3 Lower shell 84107 1
Lower shell 84107.2 Lower shell
$4107.3 tnt. shell axial we lds 2.442A, 8
lnt. sheLL ax'iaL
<<elds 2.442C C2884.1 C2854-2 C2793.1 C3121-1 C3131.2 C3131.1 27204 A 533$.1 90 A 533$.1 90 A 533$.1 63 A 533$ -1 87 A 533$.1 82 A 533$.1 93 Linda
- 1092, SAV Linda 1092 ~ SAll 7.93E18 7.93E18 7.93E18 7.93E18 7,93E18 7.93E18 5.35E18 2.87E18 116 114 110 116 Direct Direct 65K Direct Oir ect Direct SlÃYo Veld Survo Veld Lower shell axiaL welds 3.442A 8
Lower shell axiaL welds 3.442C int to Lower sheLL circ. weld 9.442 21935 Linda
- 1092, SAV Linda
- 1092, SAV Linda
- 1092, SAV 67 52 4.46E18 7.93E18 7.93E18 S4ryo Veld S4pvo Veld NRC Generic Generic value for welds fabricated by Combustion Engineering using Linde
- 1092, 0091, 124 and Arcos B-5 fluxes (Ref. Letter from S.
Bloom of USNRC to T.L. Patterson of Omaha Public Power District, dated December 3, 1993).
0
Summary File for Upper Shelf Energy Plant Name
~f<~~renc R
Selt line ident.
Neat Ko.
Haterial Type 1/4T USE at EOL 1/4T Neutron F luence at EOL Unirrad.
USE Nethod of Oetermin.
Unirrad.
USE WSE snd chemical coeposition data are fry Joe 30, 1992, letter fros G. N. Rueger (PGLE) to USNRC Docunent Control Desk, subJect:
Response
to Generic Letter 92-01, Revision 1 ~
Reactor Vessel Structural integrity Fluence snd WSE for ueld 9.442 are from December 4, 1992, letter fros G. H. Rueger (PGSE) to USNRC Doc@nant Control Desk, subject:
Response
to Generic Letter 92-01, Revision 1, "Reactor Vessel Structural integrit+.-Supplesental information
Summary File for Upper Shelf Energy Plant Name Diablo Canyon 2 EDL:
12/9/201 0
~Rrrrrrrc Belt line Ident.
Int. sheLL 85454-1 Int. shell 85454.2 Int. sheLL 85454-3 Lover shell 85455.1 Louer shell 85455.2 Louer shell 85455.3 Int. sheLL axial zelda 2.201A Int. shell axial zelda 2-2018, C
Int. to lover shel L
circ. Meld 9.201 Lo~er shell axial uelda 3.201A C
Loaer shell axial uelda 3.2018 Heat No.
C5161-1 C516S.2 C5161.2 C5175-1 C5175.2 C5176.1 12008 and 21935 12008 and 21935 10120 33A277 33A277 Naterial Type A 5338.1 A 533$.1 A 5338-1 A 5338.1 A 533$
1 A 5338.1 Linde
- 1092, SAM Linda
- 1092, SAM Linda
- 0091, SA'M Linda
- 124, SAM Linda
- 124, SAM 1/4T USE at EDL 69 76 81 102 56 57 1/4T Neutron Fluence at EDL S.75818 8.75818 So75818 8.75E18 S.75E18 S.75E18 5.61E18 6.08E18 8.75EI8 6.08E18 5.61E18 Unirrad.
USE 91 112 100 114 114 liethod of Oeterain.
Unlrrad.
USE Direct Direct Direct Direct Direct Direct Survo Meld Surv.
Meld 10'F Data 10 F Data 10'F Oats
- Fluence, UUSE, and cheaicaL caaposition data are froa Jme 30, 1992, Letter frca G. H.
Rueger (POLE) to USNRC Document Control Desk, subject:
Response
to Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity Fluence and UUSEa for veld 9-201 and 3-201 are froe Deceaker 4, 1992, Letter froe G. li.
Rueger (PGBE) to USNRC Document Control Desk, subject:
Response
to Generic Letter 92-01, Revision 1, "Reactor Vessel Structural Integritp--Suppleaental Information
ENCLOSURE 3 PRESSURIZED THERMAL SHOCK TABLES AND USE TABLES FOR ALL PWR PLANTS NOMENCLATURE Pressurized Thermal Shock Table Column Column Column Column Column Column 1
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2:
3
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4o 5
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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).
Plant-S ecific This indicates that the IRT was determined from tests on material removed from the same heat of the beltline material.
Column Column 7
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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.
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y lt Column 9
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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 Column Column Column Column Column 2:
3
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4
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5:
6
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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 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.
EOL upper-shelf energy (USE) at T/4; calculated by usin'g 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 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 attenuatio'n methodology from the ID value reported in the latest submittal (GL 92-01, PTS, or P/T limits submittals).
P
' ~ t Column 7:
Unirradiated USE.
EMA 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.
65/
This indicates that the unirradiated USE was 65N of the USE from a longitudinal specimen.
Generic This indicates that the unirradi ated USE was reported by the licensee from other plants with similar materials to the beltline material.
This indicates that the unirradiated USE was derived by the staff from other plants with similar materials to the beltline material.
IO 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 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.
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