ML20029E609
| ML20029E609 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 05/11/1994 |
| From: | Wang A Office of Nuclear Reactor Regulation |
| To: | Opeka J CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| GL-92-01, GL-92-1, TAC-M83467, NUDOCS 9405190163 | |
| Download: ML20029E609 (9) | |
Text
,.
%c)=ek.44 a a'cg%
eq
(
2 UNITED STATES
{
NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555-0001 g
,g j
May 11, 1994 Docket No. 50-213 Mr. John F. Opeka Executive Vice President, Nuclear Connecticut Yankee Atomic Power Company Northeast Nuclear Energy Company Post Office Box 270 Hartford, Connecticut 06141-0270
Dear Mr. Opeka:
SUBJECT:
GENERIC LETTER (GL) 92-01, REVISION 1, "RE4CTOR VESSEL STRUCTURAL INTEGRITY," CONNECTICUT YANKEE ATOMIC POWER COMPANY, HADDAM NECK PLANT, (TAC NO. M83467)
^
By letters dated September 17, July 19, and December 7, 1993, Connecticut Yankee Atomic Power Company (CYAPC0/the licensee) 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~CYAPC0 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, l
Revision 1.
These data have been entered into a computerized data base 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 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 the information you have provided for your facility has been accurately entered in the summary data file.
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 your reactor pressure vessel.
9405190163 940511 PDR ADOCK 05000213 P
PDR y y g p y y% Q 3,
~ gy2
.M, g
)
Mr. John F. Opeka May 11, 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:
Alan b. Wang, Project Manager Project Directorate I-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosures:
1.
Pressurized Thermal Shock Table 2.
Upper-Shelf Energy Table 3.
Nomenclature Key cc w/ enclosures:
See next page DISTR.1HUTION:
Docket File NRC & Local PDRs 1
PDI-4 Plant SVarga JCalvo SNorris AWang OGC ACRS (10)
LDoerflein, RGI DMcDonald BElliot o"I tt LA:PDI-4 PM:PDI-4 PM:PDI-1 ( [ '0:PDI-4, NE St di AWang:bpMud DMcDonald 1Stolzh#fMr CI[d/94 j /g /94
/ /
fJf{/94 y /{g/94 DATE
/
OFflCIAL RECORD COPY Document Name:
G:\\ WANG \\M83467.GL
Mr. John F. Opeka May 11, 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 tne 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, hm G-Alan B. Wang, Project Manager Project Directorate I-4 Division-of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosures:
1.
Pressurized Thermal Shock Table 2.
Upper-Shelf Energy Table 3.
Nomenclature Key cc w/ enclosures:
See next page 5
e a
v y
ar-,
eewp--
+
_. ~.
4 0-Mr. John F. Opeka Haddam Neck Plant Northeast Nuclear Energy Company cc:
Gerald Garfield, Esquire R. M. Kacich, Director Day, Berry and Howard Nuclear Planning, Licensing & Budgeting Counselors at Law Northeast Utilitiet Service Company City Place Post Office Box 270 Hartford, Connecticut 06103-3499 Hartford, Connecticut 06141-0270 J. M. Solymossy, Director S. E. Scace, Vice President Nuclear Quality and Assessment Services Nuclear Operations Services Northeast Utilities Service Company Northeast Utilities Service Company Post Office Box 270 Post Office Box 270 Hartford, Connecticut 06141-0270 Hartford, Connecticut 06141-0270 Kevin T. A. McCarthy, Director Regional Administrator Monitoring and Radiation Division Region I Department of Environmental Protection U.S. Nuclear Regulatory Commission 79 Elm Street 475 Allendale Road Hartford, Connecticut 06106-5127 King of Prussia, Pennsylvania 19406 Allan Johanson, Assistant Director Board of Selectmen Office of Policy and Management Town Office Building Policy Development and Planning Division Haddam, Connecticut 06438 80 Washington Street Hartford, Connecticut 06106 Resident Inspector Haddam Neck Plant J. P. Stetz, Vice President c/o U.S. Nuclear Regulatory Commission Haddam Neck Plant 361 Injun Hollow Road Connecticut Yankee Atomic Power Company East Hampton, Connecticut 06424-3099 362 Injun Hollow Road East Hampton, Connecticut 06424-3099 Nicholas S. Reynolds Winston & Strawn J. J. LaPlatney 1400 L Street, NW Haddam Neck Unit Director Washington, DC 20005-3502 Connecticut Yankee Atomic Power Company 362 Injun Hollow Road East Hampton, Connecticut 06424-3099 Donald B. Miller, Jr.
Senior Vice President Millstone Station Northeast Nuclear Energy Company Post Office Box 128 Waterford, Connecticut 06385
25 1
Sumary File for Pressurized Thermal Shock i
Plant Beltline Heat No.
10 Neut.
I R T, Method of Chemistry Method of
%Cu
%NI Name ident.
Ident.
Fluence at Determin.
Factor Determin.
EOL/EFPY IR T.
CF Hsddere Noggle A5887 4.2E19 20'F MTER 5 2 58 Table 0.10 0.20 Neck shell W9807-1 EOL:
Nozzle A5897 4.2E19 34'F MTE8 5 2 62 Table 0.11 0.20 6/29/2007 shell W9807 6 Nozzle 80716 4.2E19 10*F MTER 5-2 62 Table 0.11 0.20 shell W9807-8 Int. Shell A5892 9.65E19
-8'F MTER 5 2 50.25 Cateulated 0.10 0.20 W9807-2 Int. Shell A5877 9.65E19 10'F MTER 5-2 68.G9 Calculated 0.12 0.20 W9807 4 i
Int. Shell A5911 9.65E19 10*F MTEB 5 2 51.f.98 Calculated 0.12 0.20 W807 7 Lower 80650 4.03E18 20*F MTER 5-2 67 Table 0.12 0.20 shell 1
W9807 3 i
Lower A5891 4.03E18 10*F MTES 5-2 80 Table 0.15 0.20 Shell W9807-5 Lower P1444 4.03E18 8'F MTES 5 2 75 Table 0.14 0.20 shell W9807-9 Nozzle 9565/
3.63E19
-56*F Generic 104.5 Table 0.22 0.1 Shell 860548 Axlet Wetas Upr 860548 4.2E19
-56'r Generic 104.5 Table 0.22 0.1 Cire. Wold Int. Axlel 850548 5.35E19
-56'F Generic 104.5 Table 0.22 0.1 Welds Low. Cire.
1248 4.03E18
-56'F Generic 112.0 Table 0.22 0.2 Weld Low Axlet 860544 3.49E19
-56*F Ceneric 104.5 Table 0.22 0.1 Welds l
Reference _for Heddam Neck Fluence, chemical compoeltlan, and IRT. date are frae July 6,1992, letter from J. F. Opeke (NNECo) to UsheC Cocument Control Oesk, stbjects Hadden Neck Plant; Millstone Power Station, Units 1, 2, and 3: Reactor vessel structural Integrity,10CFR50.54(f), (Generic Letter 92-01, Revielen 1)
)
Weld surveillance date have not been used I, chemistry factor calculation because one of the two date is not credible.
The chemical conpoettion for weld 1248 la from eleter plants (NMP 1 snd Oyster Creek).
1
1 4
f 28 Sumary File for Upper Shelf Energy Plant Name Beltllrw Heat No.
Natorial 1/4T 8Js2 1/4T Unirred.
Method of Ident.
Type at Neutrcri USE Determin.
EOL/EFPY Fluence at Unfrred.
EOL/EFPY USE Noddme Nottle A5887 A 3022 81 2.22E19 105 Ofrect Neck she1l W9807-1 EOL2 Notate A5897 A 3028 50 2.22E19 65 65%
6/29/2007 shott W9807-6 Nozzle 80716 A 3028 68 2.22E19 90 Direct shelt W9807 8 Int. Shelt A5892 A 3023 61 5.10E19 85 65%
W9807 2 Int. Shett A5877 A 3028 64 5.10E19 92 65%
W9807 4 Int. Shell A5911 A 3025 57 5.10E19 82 65%
W9807-7 Lowert 80650 A 3028 59 2.13J19 78 65%
shott W9807-3 Lower A5891 A 3028 53 2.13E19 74 65%
sheLL W9807-5 Lower P1444 A 3023 56 2.13E19 77 65%
shett W9807-9 Nozzle 9565/
AACOS B5 62 1.92E19 106 sury.
Shett 860545 SAW Wald Axfal Welds Upper 860548 AACOs 85 60 2.22E19 106 sury.
Cire. Veld SAW Weld Int. Axlat 86054J ARCOs 85 53 4.41E19' 106' sury.
1248 ARCOs 85 58 2.13E19 102 Sister Wald SAW Ptant Lou Axlat 860545 AAC03 55 62 1.84E19 106 sury.
Welds SAW Weld Reference Fluence, cheefeat ccupoaltion, and UUSE data are frcus July 6,1992, letter free J. F. Opeka (NNECo) to USNRC Document Control Desk, stbject: Maddme Neck Plant; Mittstone Power station, Units 1, 2, and 3: Reactor Vessel structural Integrity, 10CFt50.54(f), (Generic Letter 92 01, Revision 1)
Nomenclature and Tablet PRESSURIZED THERMAL SH0CK TABLES AND USE TABLES FOR ALL PWR PLANTS 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-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.
Column 4:
End-of-life (E0L) 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).
Column 5: Unirradiated reference temperature.
Column 6: Method of determining unirradiated reference temperature (IRT).
Plant-Specific This indicates that the IRT was determined from tests on i
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.
Generiq 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.
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.
i
> 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 single wire was used in the SAW process.
(T) indicates tandem wire was used in the SAW process.
Column 4:
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.
Column 5:
E0L 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.)
DB This indicata that the USE issue may be covered by the approved equivalent margins analysis in a topical report.
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 10 value reported in the latest submittal (GL 92-01, PTS, or P/T limits submittals).
Column 7:
Unirradiated USE.
DM 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.
s 4
- 1 Direct For plates, this indicates that the unirradiated USE was from a transverse specimen.
For welds, this indicates that the i
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 i
licensee from other plants with similar materials to the beltline material.
NRC aeneric This indicates that the unirradiated USE was derived by the staff from other plants with similar materials to the beltline 4
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.
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.
l
.J