ML20101H830
| ML20101H830 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 06/25/1992 |
| From: | Mccoy C GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20101H833 | List: |
| References | |
| TAC-M83522, TAC-M83523, NUDOCS 9206300215 | |
| Download: ML20101H830 (9) | |
Text
-
l
.r.
l, '~ hi*~4? '
fY
,a l l*l 2 llv
- g f Wp j LO
- ,s
- e.. gia Power c: r.. we6
.Geor
.o.
ew.
June 25, 1992 u
ELV-03648 001530
^
Docket Nos.-50-424 50-425
.U.
S.' Nuclea;, Regulatory Commission
' ATTN: Document Control Desk
- Washingtun,-D..C. -20555-Gentlemen:
V0GTLE ELECTRIC GENERATING PLANT-
-REACTOR VESSEL STRUCTURAL INTEGRITY GENERIC LETTER 92-01 REVISION 1
.-In response to Generic Letter 92-01. Revision I concerning reactor vessel structural-integrity, Georgia Power Company (GPC) is submitting the-enclosed
-information.
Mr. C.:K. McCoy' states'that-he is a vice. president of Georgia Power Company and
~islauthorized to execute this oath on behalf of Georgia Power Company and that,
?
to the best of his knowledge and-belief, the facts set.forth in this letter and enclosure are true.
GEORGIA POWER COMPANY By:
C.:K. McCoy--
Sworn to and subscribed before me this M ay.of' h
, 1992.
J YkOAA4Y.
Notary 9ublic
,,sm_ #
b IO p) w-p A<
wr gassoums 9m p4 s s
,p
Ik
)
- Georgia Power d U. S. Nuclear Regulatory Commission ELV 03648 Page 2
Enclosures:
Response to Generic Letter 92-01 Table 5.3.3-2 Table 5.3.3-3 WCAP-110ll WCAP ll381 CKM/PAH/gmb c(w):.Georaia Power Comoa u Mr. W. B. Shipman Mr. M. Sheibani NORMS U. S. Nuclear Reculatory Commission Mr. S. D. Ebneter, Regional Administrator Mr. D. S. Hood, Licensing Project Manager, NRR Mr. B. R. Bonser, Senior Resident Inspector,.Vogtle e
F l
l l
ENCLOSURE V0GTLE ELECTRIC GENERATING PLANT RESPONSE TO GENERIC LETTER 92-01 REVISION 1 REACTOR VESSEL STRUCTURAL INTEGRITY Generic Letter 92-01 1.
Certain addressees are requested to provide the following information regarding Apperdix H to CFR Part 50:
Addressees who do not have a surveillance program meeting ASTM E 185-73,-79, or -82 and who do not nave an integrated surveillance program approved by the NRC (see Enclosure 2), are requested to describe actions taken or to be taken to ensure compliance with Appendix H to 10 CFR Part 50. Addressees who plan to revise the surveillance program to meet Appendix H to 10 CFR Part 50 are requested to indicate when the revised program will be submitted to the NRC staff for review.
If the surveillance program is not to be revised to meet Appendix H to 10 CFR Part 50, addressa s are requested to indicate when they plan to request an exemption from Appendix H to 10 CFR Part 50 under 10 CFR 50.60(b).
GPC Response The capsule surveillance prograrr for the Vogtle Electric Generating Plant (VEGP) meets ASTM 185-82.
The capsule surveillance program is in compliance with the requirements of Appendix H to 10 CFR Part 50; therefore, a revised program or an exemption is not required.
Generic letter 92-01 2.
Certain addressees are requested to provide the following information regarding Appendix G to 10 CFR Part 50:
- a. Addressees of plants for which the Charpy upper shelf energy is predicted to be less than 50 foot-pounds at the end of their licenses using the guidance in Paragraphs C.I.2 or C.2.2 in Regulatory Guide 1.99, Revision 2, are requested to provide to the NRC the Charpy upper shelf energy predicted for December 16, 1991, and for the end of their current license for the limiting beltline weld and the plate or forging and are requested to describe the actions taken persuant to Paragraphs IV.A.1 or V.C of Appendix G to 10 CFR Part 50.
GPC Response C.1culations for the Vogtle Electric Generating Plant using the guidance of Regulatory Guide 1.99, Revision 2 were performed by Westinghouse Electric Corporation for both reactor vessels. These calculations indicated that the Charpy upper shelf energy is expected to be above 50 ft-lbs, even after 48 effective full power years (EFPY) of operation.
Tables 5.3.3-2 and 5.3.3-3, which are enclosed, were prepared for the next Final Safety Analysis Report (FSAP.) revision and show the results of these calculations for each vessel. 1 l
JF ENCLOSURE (CONTINUED) l V0GTLE ELECTRIC GENERATING PLANT RESPONSE TO GENERIC LETTER 92-01 REVISION 1 REACTOR VESSEL E.TRUCTURAL INTEGRITY Generic letter 92-01 I
b.
Addressees whose reactor vessels were constructed to an ASME Code earl' - than the Summer 1972 Addenda of the 1971 Edition are requested
-to describe the consideration given to the following material properties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph lIII.A of 10 CFR Part 50, Appendix G:
GPC Response:
The reactor vessels at VEGP were constructed to the ASME Summer 1972 addenda of the 1971 edition of the code.
-Generic letter 92-01 (1) the results from all Charpy and drop weight tests for all unirradiated beltline materials, the unirradiated reference temperature for each beltline material, and the method of cetermining the unirradiated reference temperature from the Charpy and-drop weight test; GPC Response The. results1from the Charpy tests. for_ the unirradiated beltline materials for both reactor vessels are found in the FSAR in tables 5.3.2-2, 5.3.2-3,'5.3.2-4, and 5.3.2-5.
The unirradiated reference.
temperature for:the materials is also shown on-these tables. -It should be:noted that the RTNDT for plate B8805-3 (in FSAR table 5.3.2-2 sheet 1 of. 2) should be +300F-.instead of -300F.
This value will be changed in-
-the next FSAR update.
The method of determining the unirradiated reference temperature from the Charpy and drop weight tests is located L in _ WCAP-110ll..- Georgia Power Company Alvin W. Vogtle Unit _No.1 Reactor Vessel Radiation Surveillance Program," and WCAP-ll381, " Georgia Power
- Company Alvin W. Vogtle' Unit No. 2 Reactor Vessel. Radiation Surveillance n
-Program"Lin; sections 3.l'and 3.3.-
-The drop weight test data is-currently maintained by Westinghouse.
The results of the drop weight tests (TNDT values) are shown in. tables-
~
5.3.2-2, 5.3.2-3, 5.3.2-4, and 15.3.2-5 of the FSAR for both reactor vessels..
Ni m
n
=-
w V
I T
ENCLOSURE (CONTINUED)
V0GTLE ELECTRIC GENERATING PLANT RESPONSE TC GENERIC LETTER 92 01 REVISION 1 REACTOR VESSEL STRUCTURAL INTEGRITY Generic letter 92-01 (2) the heat treatment received by all beltline and surveillance materials; SPC Resornse The heat treatment for VE3P Unit 1 is shown in table A-5 of WCAP-110ll; tne heat treatment for VEGP Unit 2 is shown in table A-6 of WCAP-ll381.
Genaric letter 92-01 (3) the heat number for each beltline plate or forging and the heat number of wire and flux lot number u3ed to fabricate each beltline weld; GPC Rup_gnsa The heat number of each beltline plate is on the Combustion Engineering material certification reports maintained by Westir.ghouse.
There are no beltline forging materials.
The heat number of the wire and flux lot used to fabricate each beltline weld is found for VEGP Ik.it 1 in table A-3 of WCAP-110ll and is found for VCGP Unit 2 in tables A-3 and A-4 of WCAP-11381.
Generic letter 92-01 (4) the heat number for each surveillance plate or forging and the heat i
number of wire and flux lot number used to fabricate the surveillance weld; GPC Responsp The heat number of each orveillance plate is maintained by Westinghouse.
The heat number of wire and flux lot number used to f abricate the surveillance weld is found in table A-3 of WCAP-110ll for VEGP Unit I and is found in table A-4 of WCAP-ll381 for VEGP Unit 2.
Ger.eric letter 92-01 (5) the chemical composition, in particular the weight in percent of copper, nickel, phosphorous, and sulfur for each beltline and surveillan e material; and Y
Q
\\
k
)
4 ENCLOSURE (CONTINUED)
V0GTLE ELECTRIC GENERATING PLANT RESPONSE TO GENERIC LETTER 92-01 REVISION 1 REACTOR VESSEL STRUCTURAL INTEGRITY GPC RupanjLe The chemical compositions for VEGP Unit I are found in tables A-1. A-2, and A-3 of WCAP-110ll and for VEGP Unit 2 are found in tables A-1, A-z, A-3, and A-4 of WCAP-ll381.
In WCAP-110ll, table A-1 compares the results of me Comoustion Engineering and Westinghouse chemical analysis of plate B8805-3, and table A-3 compares the results of the chemical analysis of the weld metal used in the core region seam welds.
In WCAP-11381, table A 2 compares the results of the Combustion Engineering and Westinghouse chemical analysis on plate BS523-1, and table A-4 compares the results of the chemical analysis on the weld metal used in the intermediate to lower shell closing girth seam weld.
Gent-rric letter 92-01 (6) the heat number of the wire used for determining the weld metal chemical compositiun if different from items (3) above; GPC Response Not appliable.
Generic letter 92-01 3.
Addressees are requested to provide the following information regarding commitments made to respond to GL 88-11:
a.
How the embrittlement effects of operating at an irradiation temperature (cold leg or recirculation suction temperature) below 5250F were considered.
Iti particular licensees are requested to describe consideration given to determining the effect of lower irradiation temperature oli the reference temperature and on the Charpy upper shelf energy.
GP.G Resnonse The Technical Specifications for the Vogtle Electric Generating Plant require that critical operation occTes at a temperature of 5510F or higher.
Some physics tests are allowed when the reactor coolant system lowest operating loop temperature (Tevg) is greater than or equal to 541'JF.
Critical operation does not occur at temperatures below 5250F, _
Y f
. ENCLOSURE (CONTINUED)
V0 GILE ELECTRIC GENERATING PLANT RESPONSE TO GENERIC LETTER 92-01 REVISION 1 EEACTOR VESSEL STRUCTURAL INTEG41T1 Generic letter 92-01
- b. How their surveillance results on the predicted amount of embrittlement were considered.
GPC Resoonse The mean values of copper and nickel were used for the generation of the chemistry factor in the calculation of the change in RlNDT utilizing Regulatory Guide 1.99, Revision 2.
For VEGP, the surveillance results indicate the changes ic Charny upper shelf energy and the 30 ft-lb transition temperature shift values are less than those predicted -;titring Regulatory Guide 1.99, Revision 2.
Generic letter 92-01 c.
If a measured increase in reference temperature exccads the mean-plus-two standard deviations predicted by Regulatory Guide 1.99, Rev,sion 2, or if a measured decrease in Charpy upper shelf energy exceeds the value predicted using the guidance in Paragraph C.1.2 in Regulatory Guide 1.99, Revision 2, the licensee is requested to report the information and describe the effect of the surveillance results on the adjusted reference temperature and Charpy upper shelf energy for each beltline material as predicted for December 16, 1991, and for the end of its current license.
GPC Resoonse The measured increase in reference temperature does not exceed the mean-plus-two standard deviations predicted by Regulatory Guide 1.99, Revision 2.
The measured decrease in Charpy upper shelf energy does not exceed the value predicted using the guidance in Paragraph C.l.2 in Regulatory Guide 1.99, Revision 2.
Table 5-6 in WCAP-12256, " Analysis of Capsule U from the Georgia Power Company Vogtle Unit 1 Reactor Vessel Radiation Surveillance Program," and table 5-6 in WCAP-13007, " Analysis of Capsule U f,'om the Georgia Power Company Vogtle Electric Generating Plant Unit 2 Reactor Vessel Radiation Surveillance Program," compare the Charpy upper shelf energy valucs and the 30 ft-lb transition temperature shift values to those predicted utilizing Regulatory Guide 1.99, Revision 2. #1530 i
w TABLE 5.3.3-2 UNIT 1 REACTOR VESSEL VALUES FOR ANALYSIS OF POTENTIAL PRESSURIZED THERMAL SHOCK EVENTS (a)
. ___.. sgulaio-~r[Guid,,1.dR2 Irutial
. RTPTS { degrees F) _
Rev. 2 RTND T (deg. F)
Initial Predicted USElft-.lbs)_
Cu Ni RTNDT Dec.16, 32 48 Dec.16, 32 48 USE Dec.16, 32 48
. _... _ Material _ _ _ __ _wt- %.. wt _*/.. (degn _1991_. EFPY_, EFPY. _1991_.. EF ?Y_.EFPY [ft-lbs) _ 1991._.EFPY_ EFPY intermed. Shell Plate,88805-1 0.08 0.59 0
67 100 105 67 1 30 105 90 78 70 68 Intermed. Shell Plate, B8805-2 0.08 0.59 20 87 120 (b) 125 87 20 125 100 87 78 70' Intermed. Shell Plato, B8805-3 0.06 0.60 30 88 112 11C 88
,12 116 107 93 83 81 Lower Shell Plate,88606-1 0.05 0.59 20 74 94 97 74 94 97 116 101 90 88 Lower Shell Plate,88606-2 0.05 0.58 20 74 94 97 74 94 97 113 98 88 86 l
Lower Shell Plate. B8606-3 0.06 0.64 10 68 92 96 68 92 96 118 103 92 90 l
Core Region Longitudinal &
0.04 0.10
-80
-2 20 23
-2 20 23 134 116 105 102 Girth Seams. (c}_
. _a NOTES;
- a. RTPTS and RTNDT values are based on the peak fluence at the vesselinner radius of 2.78 E18, 3.16 E19 and 4.75 E19 for Dec. 16,1991,32 and 48 EFPY, respectrvely. The fluence values for 32 and 48 EFPY were developed assuming thai uprating from 3411 to 3565 MWt would take place during calendar year 1992. &nd that calculated design basis neutron flux levels incident on the reactor vessel were applicable over the 32 EFPY design hfetime as well as for 48 EFPY.
USE was predcted using the 1/4T fluence values based on the peak fluence at the vessel inner radius. The vessel wall thickness is 8.625 inches at the belthne region. Ccpper and nickel values for all materials are based on the results of Combustion Engineering chemical analyses. Surveillance capsule matenal was not used in calculating RTPTS. RTNDT or USE because there has been only one capsule removed from the reactor vessel, hence there es insuffi.ient data at this time.
- b. Limiting vessel material.
- c. All of the core region welds were fabricated from were heat 83653. Two Combustion Engineenng weld quahfications (CE quahfcation codes E3.11 and G1.43) were done for welds containing wire heat 83653.
W w_ _
e yq
._m_
TABLE 5.3.3-3 UNIT 2 REACTOR VESSELVALUES FOR ANALYSIS OF POTENTIAL PRESSURIZED 1 HERMAL SHCCK EVENTS (a) g9 g Rev. 2 RT DT_ (deg. F)_
instial Predicted _USE [ft-bs}_
initial
_ _RTPTS 1 degrees F)_
N Cu Ni RTNDT Dec.16, 32 48 Dec.16 32 48 USE Dec.16, 32 48
{de,q. F) 1991 EFPY EFPY 1991 EFPY EFPY (ft-Ibs) _1991 _
EFPY. EFPY_,
t_~ %
..wt -%
Matenal _ _ _..
q w
intermediate Shell Plate, R4-1 0.06 0.64 10 64 92 96 64 92 96 95 85 74 72 Intermedate Shell Plate, R4-2 0.05 0.62 10 61 04 87 61 84 87 104 93 81 79 Intermediate Shell Plate, R4-3 0.05 0.59 30 81 104 107 81 104 107 84 75 66 64 Lower Shell Plate, B8825-1 0.05 0.59 40 91 114 117 91 114 117 83 74 65 63 Lower Shell Plate, R8-1 0.06 0.62 40 94 122 126 94 122 126 87 77 68 66 Lower Shell Plate,88628-1 0 05 0.59 50 101 124 (b) 127 101 124 127 85 75 66 65 Core Region Longitudinal 0.07 0.13
-10 71 107 111 71 107 111 152 132 112 109 Welds (c)
Intermediate to Lower Shell 0.06 0.12
- 30 49 82 36 49 82 86 90 78 67 65 Girth Weld (c). _ _ _ _ _,__
_____._____1_._
NOTE.S;
- a. RTPTS and RTNDT values are based on the peak fluence at the vesselinner radius of 1.72 E18, 3.17 E19 and 4.76 E19 for Dec. 16,1991,32 and 48 EFPY, respectively. The !!uence values for 32 and 48 EFPY were developed assuming that uprating from 3411 to 3565 MWt would take place during calendar year 1992, and that calculated design basis neutron flux levels incident on the reactor vessel were applicable over the 32 EFPY design lifetime as well as for 48 EFPY.
USE was predicted using the 1/4T fluence values based on the peak fluence at the vessel inner radius. The vessel wall thickness is 8.625 inches at the belti;ne region. Copper and nickel values for all materials are based on the results of Combustion Engineenng chemscal analyses. Surveillance capsule material was not used in calculating RTPTS RTNDT or USE because there has been only one capsule removed from the reactor vessel, hence there is l
inrufficient data at this time.
i l
- b. Limrting vessel material.
- c. All of the core region welds were fabricated from wire heat 87005. Two Combustion Engineenng weld quahfications (CE quahication codes E3.23 and G1.60) were done for welds containing wire heat 87005.
___