ML17331B363
| ML17331B363 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 04/22/1994 |
| From: | John Hickman Office of Nuclear Reactor Regulation |
| To: | Fitzpatrick E INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| References | |
| GL-92-01, GL-92-1, TAC-M83453, TAC-M83454, NUDOCS 9404270394 | |
| Download: ML17331B363 (14) | |
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t UNITED STATES NUCLEAR REGULATORY CONIMISSION WASHINGTON. D.C. 20555-0001 April 22, 1994 Docket Nos.
50-315 and 50-316 Hr.
E.
E. Fitzpatrick, Vice President Indiana Michigan Power Company c/o American Electric Power Service Corporation 1 Riverside Plaza
- Columbus, Ohio 43215
Dear Hr. Fitzpatrick:
SUBJECT:
GENERIC LETTER (GL) 92-01, REVISION 1, "REACTOR VESSEL STRUCTURAL INTEGRITY," INDIANA MICHIGAN POWER COMPANY, DONALD C.
COOK NUCLEAR PLANT, UNITS 1
AND 2 (TAC NOS.
M83453 AND M83454)
By letters dated July 13,
- 1992, November 29,
- 1993, and January 24,
for D.
C.
Cook Units 1
and 2.
The NRC staff has completed its review of your responses.
Based on its review, the staff has determined that you have 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 Mater 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.
Enclosure 1 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 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.
P We request that you verify that the information you have provided for your facilities has been accurately entered in the summary data files.
No response is necessary unless an inconsistency is identified. If no comments are'eceived 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 your reactor pressure vessel.
9400270394 940422 PDR 'ADOCK: 05000315 P
PDR pm[a RLP It',KMlIESS~
I 1
[
Mr.
E.
E. Fitzpatrick April 22, 1994 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, Original signed by
Enclosures:
1.
Pressurized Thermal Shock Tables 2.
Upper-Shelf Energy Tables 3.
Nomenclature Key cc w/enclosures:
See next page DISTRIBUTION:
John B. Hickman, Project Manager Project Directorate III-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket Fil,e
'NRC 8 Local PDRs JRoe JZwolinski LBMarsh JHickman CJamerson OGC ACRS (10)
W. Kropp cc:
Plant Service list D. McDonald, 14B2 E. Hackett, 7D4 t
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LA:PD31 PD' L
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JHi man:
ll'4 a 94 CJamerson NAME 04 Z~/94 DATE 04/ZZ 94 OFFICIAL RECORD COPY FILENAME:G: iWPDOCSiDCCOOKiC083453.MPA
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Hr.
E.
E. Fitzpatrick Indiana Hichigan Power Company CC:
Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Attorney General Department of Attorney General 525 West Ottawa Street Lansing, Hichigan 48913 Township Supervisor Lake Township Hall Post Office Box 818 Bridgman, Hichigan 49106 Al Blind, Plant Hanager Donald C.
Cook Nuclear Plant Post Office Box 458 Bridgman, Hichigan 49106 U.S. Nuclear Regulatory Commission Resident Inspector Office 7700 Red Arrow Highway Stevensville, Hichigan 49127 Gerald Charnoff, Esquire
- Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.
W.
Washington, DC 20037 Hayor, City of Bridgman Post Office Box 366 Bridgman, Hichigan 49106 Special Assistant to the Governor Room 1
State Capitol Lansing, Hichigan 48909 Nuclear Facilities and Environmental Honitoring Section Office Division of Radiological Health Department of Public Health 3423 N.
Logan Stre'et P. 0.
Box 30195
- Lansing, Hichigan 48909 Donald C.
Cook Nuclear Plant Hr. S.
Brewer American Electric Power Service Corporation 1 Riverside Plaza
- Columbus, Ohio 43215 Dccanbcr 1993
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Sutmaary File for Pressurized Thermal Shock Plant Name Beltline Heat Ko.
ident.
ident.
LD Naut.
Fluence at KOL/EFPY LRT~
Hethod of Deteraiin.
LRT Choe i a try Factor Hethod of XCu Date rain.
CF II01cpNlcc5 o c eggs l ALL date except aa noted belou uere froa the July 13, 1992 Letter fram E.E. Fitspatrick to T.E. Hurley, mgonald C.
Cool Nuclear plant Units 1 and 2... Generic Letter 92.0'1, Rwiaion 1, Reactor vesseL structural integrity."
Lnforsiation regafding chlsicaL coetyoaftiona initi~ L RT>>, and aethodology of deteraination for the, LRJSK for the zzelds a<
froa the Noveaker 29, 1993 Letter free K.E. Fitzpatrick to T.E. Hurley, mgonald C. Cook Nuclear plant Units 1 and 2...
Response
to Request for Additional Lnforsaticn for Generic Letter 92.01, Revision 'i."
No valw for X Cu of ueld d.44'as provided, therefore the default valw of 0.35 uas used.
The UUSE values for uelda 1.442, 2.442 and 3-442 represen>
an NRC staff calculated average of O.C. Cook and Sister plant data (Hcguire, Unit 1 and Diablo Canyon, Unit 2) for uelc uire heats containing heat noa.
13253 and 1200S.
Sullmary File for Pressurized Thermal Shock PLant Name Beltllne ident.
Neat No.'dent.
1D Naut.
Fluence at EOL/EFPY TRT~
Hethod of Choalstry Hethod of XCu Detersin.
Factor Deteratn.
Cook 2 Int. Shell C5556-2 1.71E19 58'F 10.1 Plant S
1flc 108,9 Table 0.15 Oo58 EOLc 12I23I 2017 Louer Shell 9.1 C5540-2 1.71E18
-20'F 1nt. ShelL C5521.2 1.71E19 38'F 10.2 Plant Speci f3c Plant Speclflc 74.6 Table 0.1'I 102.61 Calculated 0.14 0.58 0.64 Lcwer SheLL 9.2 C5592 1
1.71E18
.20'F Plant Speci flc Table 0.14 0.60 lLRfer~~
Louer ShelL Axl~l Molds S3986 C 1 rcess.
S3986 Meld lnt. Shell, S3986 Axial Molds 9.02E18 35'F 6.63E18 35'F 1.71E19
-35'F Plant Speci f1 c Plant S peel fLc Plant S
Lfic 64.4 Calculated 0.05 Calculated 0.05 Calculated 0.05 0.97 0.97 0.97 1n1tial RT and cheeicaL cosposftlon for the plates and are fr~ the July 13, 1992 Letter free E.E. Fltzpatrlck to T.E.
Hurley, "Donald C. Cook Nuclear Plant Units 1 and 2... Generic Letter 92-01, Rwfsfon 1, Reactor Vessel Structural integrity."
Updated fluences, chealcaL coepos{tfon, Lnlt1aL RT, and calcuLated cheafstry factors are fry the April 12, 1993 Letter frocs E.E. Fltzpatrlck to T.E; Hurley, Konald C. Cook Nuclear PLant Unit 2...Updated Reference Temperature and Pressurfzed Thermal Shock Analyses."
Enclosure 2
Suffmfary File for Upper Shelf Energy Plant Name 0 ~ C. Cook 1
EOL:
10/25/
2014 SeLtline ident.
Kozzle ShelL S4405.1 Nozzle Shell S4405.2 Heat No.
Katerial Type A 5338.1 A 5338 1
1/4T USE
~t EOL 1/4T Kwtron FLUence at EOL 8.47E18 8.4?E1d Unirred.
USE 87 Nethod of
- Determfn, Unirrad.
USE 65X 65K Nozzle Shell 84405-3 fnt. Shell 84406.1 lnt. Shall 84406.2 Lnt. SheLL 84406.3 Lcwer SheLL 84407.1 Lour Shell 84407.2 L~r Shell 84407.3 KozzLe Shell Axfal Maids 1.442 Kozzle/int.
Shell Circ.
Meld 8.442 fnt. to Lour Shell Circ. Meld, 9.442 int. Shell Axial Maids 2 442A/C Lever Shell Axial Molds 3.442A/C C1260 13253
~nd 12005 (T) 20291 1P3571 13253 and 12008 LT) 13253 slid 12008
<T)
A 5338 1
A 5338.1 A 5338.1 A 5338 1
A 5338-'I A 5338.1 A 5338-1 Linda 1092'AM Linda
- 1092, SAM Linda
- 1092, SAM Lfnde
- 1092, SAM Linda
- 1092, SAM 74 76 101 dA?K1d 8.47E18 d.4?K1d 8.47E15 8.4?E1d SA?K1d dA7Eld 8.47K15 8.47K1d 8.47E18 8 A?K1d 8.47K1d 110 65K Direct Direct Direct Direct Direct Direct NRC Gener fc Sfster Plant Sister Plant KRC Generic NRC Generic
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SuLLNLary File for Upper Shelf Energy Plant Nese SeltLine Neat No.
Haterial 1/4T USE 1/47 Unirrad.
Kethod of ident.
Type at EOL Neutron USE Oeterain.
FLuence at Unirrad.
EOL USE
~Rf RnceE D.C. Cook 1 All data except as noted belou uere fraa the duly 13, 1992 Letter froe E.E. Fitzpatrick to T.E.
Hurley, 40onald C. Cook Nuclear Plant Units 1 and 2... Generic Letter 92.01, Revision \\,
Reactor Vessel Structural integrity."
information regardinG UUSE and aethodology of deteraination for the UUSE for the uelds are fry the Novcsher 29, 1993 letter free E.E. Fitzpatrick to T.E. Hurley, %onald C. Cook Nuclear Plant Unite 1 and 2... Response to Request for Additional information for Generic Letter 92-01, Revision 1.
No value for X Cu of ueld d-442 uas provided, therefore the default value of 0.35 uas usede The WSE values for uelds 1 442, 2-442 and 3-442 represent the NRC staff calculated average of O.C.
Cook and Sister Plant date (HcGuire, Unit 1 and Oiablo Canyon, Unit 2) for veld uire heats containing heat nos.
13253 and 72008.
The WSE for uelds 9-442 ia free MCAP-12619, "Analysis of the Heine Yankee Reactor Vessel Second Mall Capsule Located at 253'," Harch 1991.
The WSE for zelda d.442 is free MCAP-1078b, the Report for Surveillance Capsule U free HcGuire Unit 1.
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Su@sary File for Upper Shelf Energy Plant Neee Beltlinc Ident.
Neat No.
Haterial Type 1/4T USE at EOL 1/47 Neutron Fluence at EOL Unirrad.
USE Hethod of Deterein.
On)rrad.
USE D. C. Cook 2
EDL:
12/23/
2017 Int. Shell C5556.2 10.1 Int. Shell C5521.2 10-2 LoMer Shell C5540.2 9.1 L~r Shell C5592.1 9.2 Int. Shell S3986 Axial Velds Lcwer Shell S3986 Axial Voids Circ. Veld S3986 A 533$ -1 A 533$
1 A 5338 1
A 533$.1 Linda 124 SAV Linda 124 SAV Linda 124, SAV 67.5 58.4 78.4 1.02E19 1.02E19 1.02E19 1.02E19 5.38E18 3.95E1d 1.02E19 110 Direct Direct 1
Direct Direct Direct Direct Direct WSE data far the plates and uelda are fraa the July 15, 1992 letter frca E.E. Fittpatrick to T.E. Hurley, "Donald C. Cook Nuclear Plans Unite 1 and 2... Generic Letter 92.01, Revision 1, Reactar Vessel Structural Integrity."y Information regarding the elthodology of deterainatian for the WSE af plates C5556.2, CS540.2, and C5592-1 is fry the Novesjber 29, 1993 letter free E.E. Fitzpatrick to T.E. Hurley, Ronald C. Cook Nuclear Plants Units 1 and 2... Response to Request for Additional Information for Generic Letter 92-01, Revision 1.
Updated fluences are froa the April 12, 1993 letter froe E.E. Fitzpatrick to T.E. Hurley, eDonald Cook Nuclear Plant Unit 2...Updated Reference Temperature and Pressurited Thermal Shock Analyses e
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Enclosure 3
PRESSURIZED THERMAL SHOCK AND USE TABLES FOR ALL PWR PLANTS NOMENCLATURE KEY Pressurized Thermal Shock Table Column Column Column 1
2 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 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:
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 temper ature evaluation.
Method of determining chemistry factor.
Tabl e 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 descr'ibed in RG 1.99, Revision 2.
NOMENCLATURE KEYcontinued Column 9:
1 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:
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'ingle wire was used in the SAW process.
(T) indicates tandem wire was used in the SAW process.
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Material type; plate types include A 533B-1, A 3028, A 302B Mod.,
and forging A 508-2; weld types include SAW welds using Linde 80,
- 0091, l24,
- 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 l.99, Revision 2.)
EMA This indicates that the USE issue may be covered by the approved equivalent margins analysis in the BKW Owners Group Topical Reports:
BAW-2178P and BAW-2192-P.
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).
NOMENCLATURE l(EYcontinued Column 7:
Unirradiated USE.
EMA This indicates that the USE issue may be covered by the approved equivalent margins analysis in the B&M Owners Group Topical Reports:
BAM-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 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.
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 unirr adiated 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.