ELV-03648, Forwards Response to Rev 1 to Generic Ltr 92-01 Re Reactor Vessel Structural Integrity.Heat Number of Each Beltline Plate Located on C-E Matl Certification Repts Maintained by Westinghouse

Forwards Response to Rev 1 to Generic Ltr 92-01 Re Reactor Vessel Structural Integrity.Heat Number of Each Beltline Plate Located on C-E Matl Certification Repts Maintained by Westinghouse

ML20101H827
Person / Time
Site:	Vogtle
Issue date:	06/25/1992
From:	Mccoy C GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ELV-03648, ELV-3648, GL-92-01, GL-92-1, TAC-M83522, TAC-M83523, NUDOCS 9206300215
Download: ML20101H827 (9)

Text

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June 25, 1992 ELV-03648 001530 Docket Nos. 50-424 50 425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:

V0GTLE ELECTRIC GENERATING PLANT i 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 is authorized to execute this oath on behalf of Georgia Power Com)any and that, to the best of his knowledge and belief, the facts set forth in tais letter and enclosure are true.

GEORGIA POWER COMPANY By: te ,

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, C. K. McCoy Sworntoandsubscribedbeforemethisc28dayof lut'.L , 1992.

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Geoigia Power b 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 Company Mr. W. B. Shipman Mr. M. Sheibani NORMS U. S. Nuclear Reaulatory CLmmission Mr. S. D. Ebneter, Regional Administrator Mr. D. S. Hood, Licensing Project Manager, NRR Mr. 8. R. Bonser, Senior .lesident inspector, Vogtle

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 Appendix 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 have an integrated surveillance program a) proved ty the NRC (see Enclosure 2), are requested to describe actions tar.en 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, addressees are requested to indicate when they plan to request an exemption from Appendix H to 10CFRPart50under10CFR50.60(b).

GPC Response The capsule surveillance program for the Vogtle Electric Generating Plant (VEGP) meets ASTH 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.

Ger,cric 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.1.2 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 pursuant to Paragraphs IV.A.1 or V.C of Appendix G to 10 CFR Part 50.

GEC_fesoonse Calculations 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 (FSAR) revision and show the results of these calculations for each vessel.

ENCLOSURE (CONTINUED)

V0GTLE ELECTRIC GENERATING PLANT RESPONSE TO GENERIC LETTER 92-01 REVISION 1 REACTOR VEESEL STRUCTURAL INTEGRITY Generic letter 92-01

b. Addressees whose reactor vessels were constructed to an ASME Code earlier 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 Ill.A of 10 CFR Part 50, Appendix G:

GPC Responsti The reactor vessels at VEGP were constructed to the ASME Summer 1972 addenda of the 1971 edition of the code.

Ggneric 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 determining the unirradiated reference temperature from the Charpy and drop weight test; GPC Response The results from 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 in WCAP 110ll, " Georgia Power Company Alvin W. Vogtle Unit No.1 Reactor Vessel Radiation Surveillance Program," and WCAP-11381, " Georgia Power Company Alvin W. Vogtle Unit No. 2 Reactor Vessel Radiation Surveillance Program" in sections 3.1 and 3.3.

The drop weight test data is currently maintained by Westinghouse.

The results of the drop weight tests (TNDT valut.) are shown in tables 5.3.2-2, 5.3.2-3, 5.3.2-4, and 5.3.2-5 of the FSAR for both reactor vessels.

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ENCLOSURE (CONTlWVED)

V0G1LE ELEC1RIC GENERATING PLANT RESPONSE TO 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; GPC Response The heat treatment for VEGP Unit 1 is shown in table A 5 of WCAP-110ll; the heat treatment for VEGP Unit 2 is shown in table A-6 of WCAP-ll381.

Generic letter 92-01 (3) the heat number for each beltline plate or forging and the heat number of wire and flux lot number used to fabricate each beltline weld; QPC Response The heat number of each beltline plate is on the Comoustion Engineering material certification reports maintained by Westinghouse. 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 Unit 1 in table A 3 of WCAP-110ll and is found for VEGP Unit 2 in tables A-3 and A-4 of WCAP-ll3Bl.

Generic letter 92-01 (4) the heat number for each surveillance plate'or forging and the heat number of wire and flux lot number used to fabricate the surveillance weld; GPC Response The heat number of each surveillance plate is maintained by Westingheuse. The heat nutber of wtre and flux lot number used to j fabricate the surveillance weld is found in table A-3 of WCAP-11011 for VEGP Unit I and is found in table A-4 of WCAP-ll381'for VEGP Unit 2.

i Generic letter 92-01 i

(5) the chemical composition, in particular the weight' in percent of copper. nickel, phosphorous, and sulfur for each beltline and -

surveillance material; and 3

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ENCLOSURE (CONTINUED)

V0 GILE ELECTRIC GENERATING PLANT RESPONSE 10 GENERIC LETTER-92 01 REVISION 1 E KTOR VESSEL STRUCTURAL INTEGRITY GPC Response The chemical compositions for VEGP Unit 1 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 2, A-3, and A 4 of WCAP ll381. In WCAP-110ll, table A-1 compares the results of the Combustion Engineering and Westinghouse chemical analysis of plate B8805-3, ano table A-3 compares the results of the chemical analysis of the weld metal used in the core region seam welds, in WCAP-ll381, table A 2 compares the results of the Coinbustion Engineering and Westinghouse chemical analysis on plate B86281, and table A 4 compares the results of the chemical l analysis on the weld metal used in the intermediate to lower shell closing girth seam weld.

Generric letter 91M (6) the heat number of the wire used for determining the weld metal chemical composition if different from items (3) abovet GPC Response Not appitable.

Generic letter 92-01

3. Addressees are requeste.1 to provide the following inforation 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. In particular licensees are requested to describe consideration given to determining the effect of lower irradiation temperature on the reference temperature and on the Charpy upper shelf energy.

GPC Response The Technical Specifications for the Vogtle Electric Generating Plant require that critical operation occurs at a temperature of 5510F or higher.

Some physics tests are allowed when the reactor coolant system lowest

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operating loop temperature (Tavg) is greater than or equal to 541of.

Critical operation does not occur at temperatures below 5250F.

ENCLOSURE (CONTINUED)

V0GTLE ELEC1RIC GENERATING PLANT RESPONSE 10 GENERIC LETTER 92-01 REVISION 1 REACTOR VESSEL STRUCTURAL INTEGRITY Generic Letter 92-01

b. Hou their surveillance results on the predicted amount of embrittlement were considered.

GPC Response The mean values of copper and nickel were used for the generation of the chemistry factor in the calculation of the change in RTHDT utilizing Regulatory Guide 1.99, Revision 2. For VEGP, the surveillance results indicate the changes in Charpy upper shelf energy and the 30 ft-lb transition temperature shift values are less than those predicted utilizing Regulatory Guide 1.99, Revision 2.

Generic letter 92-01

c. If a measured increase in reference temperature exceeds the mean plus two stkndard deviations predicted by Regulatory Guide 1.99, Revision 2, or if a measured decrease in Charpy upper shelf energy exceeds the value predicted using the guidance in Paragraph C.I.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 Respon+e 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 valun 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 from the Georgia Power Company Vogtle Electric Generating Plant Unit 2 Reactor Vessel Radiation Surveillance Program," compare the Charpy upper shelf energy values and the 30 ft-lb transition temperature shift values to those predicted utilizing Regulatory Guide 1.99, Revision 2.

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! TABLE 5.3.3-2 UNIT 1 REACTOR VESSELVALUES FOR ANALYSIS OF POTENTIAL PRESSURIZED THERMAL SHOCK EVENTS (a)

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10 CUR 50.61 PiedEted l Regustory Guide I.99, Initial . . _R_TPTS (degrees.F)J Rev. 2 RTNDT_idegg trutial . h' Rigulateiy Guidei.99, R Predicted _USE 111.-lbs)j Cu Ni RTNDT Dec.16 32 48 - Dec.16, 32 l 48 USE l Dec.16,1 32 l 48 l Material wt-% wt-% {deg._Fj_

1 1991 _EFPY EFPY__1991_ EFPYlEFPY _111-lbs)1 1991 1EFPJY EFPYj intermed. Shell Plate, B8805-1 0.08,1 0.59 0 67 100 105 67 ; 100 105 90 78 l 70 t 68 i' t Intermed. She!! Plate B8805-2 0.08 0.59 20 87 120 (b) 125 87 '

120 125 100 87 ~ 78 [ 76 Intermed. Shell Plate, B8805-3 0.06 0.60 30 88 112 116 83 112 , 116 107 93 83 '

81 Lower Shell Plate. B8'X)6-1 0.05 0.59 20 74 94 97 74 94 97 116 101 90 88 Lower Shell Plate.B8606-2 0.05  : 0.58 20 74 94 97 74 94 97 113 98 88 86  ;

Lower Shell Plate B8606-3 0.06 0.64 10 68 92 96 68 92 96 118 103 92 90 I

Core Region Longitudinal & 0 04 0.10 - 80 -2 20 23 -2 20 23 , 134 116 i 105 102 j LGirth Seams (c) __.

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NOTE _S;

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a. RTPTS and RTNDT values are based on the peak Suence at the vesselinner rades of 2.78 E18, 3.16 E19 and 4.75 E19 for Dec.16,1991,32 and 48 EFPY, respectrvely. The !!uence values br 32 and 43 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 incrdent on the reactor vessel were applicable over the 32 EFPY design infetime as well as for 48 EFPY.

USE was predcted usir g the 1/4T fluence values based on the peak f!uence at the vessel inner radius. The vessel wall thickness is 8.625 inches at the belthne region. Copper and nickel values for a!! materials a e based on the results of Combustion Engineenng chemical analyses. Surveillance capsule material was not used in calculating RTFTS, RTNDT or USE because there has been only one capsule removed from the reactor vessel, hence there is insuffice.t data at this time.

b. Limiting vessel material.

c. AS of the core region welds were fabricated from wire heat 83653. Two Combuston Engmeenng weld quahfications (CE quahfication codes E3.11 and G1.43) were done for welds containing wire heat 83653.

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. TABLE 5.3.3-3 i j UNIT 2 REACTOtl VESSEL VALUES FOR ANALYSIS OF POTENTIAL '

PRESSURIZED THERMAL SHOCK EVENT S (a)

F ' 10CFR50.61 Predcted Regulatory Guide 1.99. i Regulatory Gude 1.99, R2{ l i Indial _ MTS1 degrees F)_ prev._2 RTNDT_[deg F)J Initial Predcted USE [ftribs)y Cu Ni RTNDT. Dec.16, 32 [ 48 Dec.16, 32 48 l USE Dec.16, 32 ! 48 j '

_ Material . wt4wtriJdeg. F) 1991 EFPYjEFPY 1991 2 EFPY) EFPy(ft- Ibs(_1991_ EFPY__{EFPY; intermediate Shell Plate, R4-1 0.06 0.64 to 64 92 96 64 92 ' 96 95 85 74 72 .

Intermediate Shell Plate R4-2 0.05 0.62 10 61 84 87 61 84 87 l 104

- 93 81 79 -l Intermediate Shell Plate. R4-3 0.05 0.59 30 81 104 107 81 104 107 84 75 66 64 i Lower Shell Plate,88825-1 l 0.G5 O.59 40 91 114 117 91 114 117 83 74 65 63 Lower Shell Plate R8 - 0.06 _ 0.62. 40 94 122 126 94 122 126 87 77 68 66

• Lower Shell Plate, B8628-1 0.05- 0.59 50 101 124(b)! 127 101 124 127 85 75 66 65 I Core Region Longitudinal 0.07 0.13 -10 71 107 111 71 107  ! 111 5.52 132 112 , 1C9 -

Welds (c) j.  !  : i intermediate to Lower Shell O.06; 0.121 -30 49 82 86 49 82 86 90 78 67 ! 65 ! i

_ Girth Weld {c) ,

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NOTES: l l

a. RTPTS and RTNDT values are based on the peak fluence at the vessel inner radius of 1.72 E18  !

3.17 E19 and 4.76 E19 for Dec.16,1991,32 and 48 EFPY, respectuely The fluence values for '

32 and 48 EFPY were developed assuming that uprating from 3411 to 3565 MWt would talie  !

place during calendar year 1992, and that calculated design basis =e atron flux levels incident on the reactor vessel were applicable over the 32 EFPY design lifenme = weil as for 48 EFPY. I USE was predicted using the 1/4T fluence values based on the peak fluence at the vessel anner radius. The vessel wall thickness is 8.625 inches at the betti ne region. Copper and nickel values for all materials are based on the results of Corrhustion Engineenng chemcal analyses. Surveillance capsule material was not used in calculating RTPTS. RTNDT or USE because there has been only one capsule removed from the react 0r vessel hence there is insufficient data at this time.

b. Limiting vessel material

c. All of the core region welds were fabricated from wire heat 87005. Two Combustion Engineenng weld qualifications (CE grlification codes E323 and G1.60) vare done for welds containing wire heat 87005. [

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