ML20197F903

Forwards Updated Response to Rev 1,Suppl 1 to GL 92-01, Reactor Vessel Structural Integrity. Evaluation of Impact of Info Contained in CE NPSD-1039,Rev 2 on Existing Plant Reactor Vessel Integrity Analysis Encl

ML20197F903
Person / Time
Site:	Vogtle
Issue date:	12/22/1997
From:	Mccoy C SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-92-01, GL-92-1, LCV-0648C, LCV-648C, TAC-M83522, TAC-M83523, NUDOCS 9712300319
Download: ML20197F903 (9)

Text

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'C.K.McCoy Southern Nuclear

.. Vee President Operating Company. Inc, Vogtle Project : 40 inverness Center Parkway P.0 Box 1295 Birmirgham. Alabama 35201

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• Tel205932.7122 fav 2059920403 SOUTHERN -

COMPANY Energy to Sern hurWonF December 22, 1997 Dockd Nos. '50 424 LCV-0648C 50-425 TAC Nos. M83522 M83523 U. S. Nucient Regulatory Commission ATTN: Document Control Desk Washingtoc D. C. 20555-0001 Vogtle Electric Generating Plant Updated Response to Generic letter 92 01, Revision 1, Supplement 1 Reactor Vessel Structural Intenrity Ladies and Gentlenwn:

The NRC issued Generic Ixtter 92-01, Revision 1, Supplement 1 (GL92 01, R1, S I), Reactor Vessel Structural Integrity, on May 19,1995. In the geretic letter supplement, the NRC identified a concert that licensees may not have all of the relevant data pertinent to the evaluation of the structural intcBrity of their reactor prescue vessels. The generic Ltter ~ .

supplement requested licensees to respond within 90 days describing those actions taken or planned to locate all data relevant to the determination of RPV integrity, or an explanation of why the existing data is considered complete as previously submitted.

Additionally, GL92-01, R1, S1 requested licensees to provide the following information within 6 months of the date of the generic letter supplenwnt: .

- an assessment of any change in Lat-estimate chemistry based on consideration of all relevant data;

- a determination regardmg the need to use the ratio procedure described in Position 2.1 of Regulatory Guide 1.99, Revision 2; and L

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- a written nport providias any newly acquired data and the resuhs af any ascessary revisicas to be evaluation d RPV inleyisy is nooardense with tbs . . c' -f o

10 CFR S0.60,10 CFR S0.61, .t;:% 0 and H to 10 CFR $0, av, any potential impact on the L'IDP or P.T limits or a ecer ih that all information previoudy submined remains valid.

Georgia Power Campany (GPC) isster LCV4648, dated August 9,1995, provided tbs 90 day tasponse to GL9241, R1,51, and OPC istler LCV46488, dated Newmber 15,1995, provided the 6 mostb response to GL9241, R1,51, far Vogtle Electric Genwating Plant (VEOP). la letter LCV 06488, OPC advised the NRC of its participation in abe Combustion Engineering Owners Group Ranctor Vessel Working Group (CEOG-RVWG) weld obsmistry variabibly task and consnitted to evaluate the efect of any new data ham the CEOG-RVWG. CEOO report CE NPSD 1039, Revision 2, was issued in June 1997 and a copy has been provided to the NRC by CEOO letter CEOO 97 264, dated July 14,1997. Attachment 1 provides an evaluation of the impact of the information contained in CE NPSD 1039 Revisk,2, on the existing VEOP reactor vessel integrity analysis.

la summery, VBOP 1 and 2 are plate limited with regards to tbs reactor vessel integrity analysis. 'Ilu. revised besNetimate chemistry values have a negligible impact on the current VEOP 1 and 2 reactor vessel integrity analysis and do not reeuk in a new or d;fferent material becoming the limiting bekline material. 'therefore, the cuneet reactor veed analysis for VEOF 1 and 2 remains valid.

Should you have any questions, plasse advise.

SOUTHERN NUCLEAR OPERATING COMPANY By: ,

C. K. oCoy AND SUBSC BSD BEFORE ME

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1 Attaciunsat:  !

1. Updated Response to Generic Lemer 9241,  !

Revision I, Suppkenent 1 oc: Gearsia Poner Campany ,

J. B. Bonaley, Jr.

M. Sheibani .

NORMS  ;

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U. S . N= ' - - m * ^ - v C-- - - '- ' =

L A.Reyes,F-:f-: -' Administrator D. H. Jaflin, Senior Project Manager, NRR NRC Senior Resident Inspector, Vogtle LCV-0648C -

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J ATTACHMENT 1 A

b Vogtle Electric Generating Plant Updated Response to Generic Letter 92-01, Revision 1, Supplement 1 Requested Information 1

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. . _ _ ._ __.. . . . . _ , _ . - ~ , . , . . , - _ . ~ . . . . - . . _ . , , -

. A1TACHb6NT1 Requested lefonmatina (1) Describe those actions takes or plassed to locate au data relevant to the .

determination of reactor vessel lategrity, or as emplanaties of why the I ealstles data is considered evapnete as previously submitted.  !

i Georgia Power Company (GPC) and later, Southem Nuclear Operating Company (SNC), has aggressively sought to obtain the original fabrication '

records for the Vogtle Electric Generating Plant (VEGP) reactor vessels through participation in the Combustion Engineering Reactor Vessel Group '

(CE RVG). In response to Generic Letter 92-01, Revision 1, Supplement 1, '

GPC/SNC participated in the Combustion Engineering Owners Group - Reactor Vessel Working Group (CEOG-RVWG) task to determine the best estimate  :

copper and nickel content of beltline welds contained in reactor vessels fabricated by Combustion Engineering (CE). The resulting best-estimate copper and nickel values determined by the CEOG RVWG and the supporting data are i '

contained in topical report CE NPSD 1039, Revision 2, Best Estimate Copper and Nickel Values in CE Fabricated Vessel Wolds. By CEOG letter CEOG-97 264, dated July 14,1997, a copy of CE NPSD 1039, Revision 2, was provided to the NRC.

(2) as asseesment of any change la best-estimate chemistry based os consideration of au relevant data Tables 1 and 2 below evaluate the impact of the CEOG-RVWG best-estimate ,

copper and nickel values on the current VEGP reactor vessel integrity analysis.

As shown in Table 1, the best-estimate copper value determinal by the CEOG ,

for the VEGP 1 beltline welds is slightly higher than the best-estimate copper value currently used in the VEGP 1 reactor vessel integrity analysis and the

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nickel value is the same in both cases Use of the CEOG best-estimate copper ,

and nickel values result in a slight increase in the chemistry factor (CF) .

determined using Regulatory Guide 1.99, Revision 2 (RG 1.99), Table 1, and 10 CFR $0,61, Table 1, for the VEGP 1 beltline welds firom 33.27 to 34.5'F.

Two surveillance capsules have been withdrawn from VEGP 1 and consistent with Regulatory Position 2.1 of RG 1.99 and 10 CFR 50.61(cX2XiiXA), a CF of 6.5'F was calculated for the VEGP 1 circumfereatial weld based on credible surveillance data.

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ATTACHMENT 1 i,

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l Based on the existing rence t vesselintegrity analysim for VT J . bes&.s Shell Plate B8805 2 is the amiting beltline ma erial with rr/; J ro a.9."d

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refwence temperature (ART) and pressurized thermal shed #Nb 4 % 1 i in Table 1, uw of the best estimate copper and nickel valueo degNdoy dm CEOG do not result in any VEGP 1 beltline weld becoming the amitmg reactor i vessel matwial with regard to ART or PTS. The slight increase in the best-  !

estimate copper value for the VEGP 1 beit!ine welds from 0.039 to 0.042 results  !

in a slight reduction in the EOL USE; however, this cha*3e is not discernible on Figure 2 of RG 1.99 and does not resuh in a EOL USE less than 50 A Ibs in  :

accordance with the requirements of 10 CFR 50, Appendix 0. *Iherefore, the -

results of the existing VEGP 1 reactor vessel integrity analysis and pressure-temperature (P T) limits remain valid.

All VEGP 2 beltline welds were fabricated using the same weld wire heat but I

i two different Aux types were used. As a result, GPC/SNC has historically reported difierent best estimate copper and nickel values for the longitudinal and circumferential beltline welds in the reactor vessel integrity analysis.

According to CE, the Sux types used in the VEOP 2 beltline welds (Aux types '

0091 and 1092) are neutral Aux types that do not effect the copper and nickel cordent of the resulting welds. Therefore, the same best-estimate copper and nickel value is applicable to the VEGP 2 longitudinal and circumferential beltline welds and is used to evaluate the effect of the CEOG-RVWG best estimate values on the cunent reactor vessel integrity analysis.

As shown in Table 2, the best estimate copper value determined by the CEOG-RVWG for the VEGP 2 beltline welds is slightly lower and the nickel value is l slightly higher than those values currently used in the VEOP reactor vessel integrity analysis. Use of the CEOG RVWG best-estimate copper and nickel  :

value results in a slight decrease in the CF for the VEGP 2 longitudinal beltline l welds and a dight increase in the CF for the VEGP 2 circumferential beltline weld as shown in Table 2.

Based on the current reactor vessel integrity analysis for VEGP 2, Lower Shell Plate B8628 1 is the limiting beltline material with regard to ART or PTS. As shown in Table 2, the best estimate copper and nickel values determined by the CEOG-RVWG for the VEGP 2 beltline welds do not result in a new or different material becoming the limiting VEOP 2 reactor vessel material with regard to ART or PTS. The decrease in the best estimate copper content determined by the CEOG-RVWG for the VEGP 2 longitudinal and circumferential welds results in a higher EOL USE tien projected by the existing reactor vessel .

integrity analysis and provides additional margin above the 50 A lbs minimum A2

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i required by 10 CFR 50, AWit G. Therefore, the results of the existing 4 VEGP i reactor vessel integrity analysis and pressure-temperature (P-T) limits j remain valid. j (3) a determination regardlag the need to use the ratio procedere described in l' Position 2.1 of Regulatory Guide 1.99, Revision 2; and

) The surveillance program welds for Vogtle I and 2 were fhbricated using the name heat of weld wire used to fhbricate the reactor vessel beltline welds for the respective units. No copper coated weld wires were used in either the surveillance program welds or the reactor vessel beltline welds for Vogtle 1 and

2. Therefore, the best-estimate copper and nickel content for the Vogtle 1 and 2 surveillance program welds are considered representative of their respective  ;

beltline welds. Therefore, it is not necessary to adjust the surveillance weld  ;

results using the ratio procedure described in Position 2.1 of Regulatory Guide  !

l.99, Revision 2.

, t (4) a written report providleg any newly acquired data and; (a) the results of f any necessary revisions to the evaluaties of RPV lategrity is accordance with the requirements of 10 CFR 50.60,10 CFR 50.61, Appendices G and .

H le 10 CFR 50, and any potential impact on the LTOP or F-T limits or  ;

j (b) a certification that au leformatien previously submitted remains valid. -

As stated above, a copy of CE NPSD-1039, Revision 2, has been provided to the NRC by the CEOG-RVWG. Due to the low copper content of the VEOP 1 and 2 reactor vessel beltline welds, the reactor vessel integrity analysis is limited by the beltline plate properties as opposed to those associated with the '

. beltline welds. As shown in Tables 1 and 2 above, the CEOG-RVWG best-

! estimate values do not result in any of the VEGP 1 and 2 beltline welds becoming the limiting beltline material, nerefore, the current reactor vessel integrity analysis for VEGP 1 and 2 remains valid.

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B8805-2 Insermediane Shell Platem C0613-2 0.083 0.610 53.1 100.7 84.1 1I83 0.083 0.610 53.1 100.7 84.1 IIL3 101-124A latermedtase Shell Axial Weld A 83653 0.039 0.102 33.2 -31.8 -48 3 11.2 0.042 0.102 34 3 -30.0 -47.2 12.6 101-124B Intermedisse ShcIl Axial Weld B 83653 0.039 0.102 33.2 -30.0 -47.0 12.2 0.042 0.102 34.5 -28.0 -45.8 13.6 101-124C Intermedisse SheD Axial Weld C 83653 0.039 0.102 33.2 -30.0 47.0 12.2 0.042 0.102 34 3 ~^s -45.5 13.6 101-142A ImrerShell AxialWeld A 83653 0.039 0.102 33.2 -30.0 -47.0 12.2 0.042 0.102 343 -28.0 -45.8 13.6 101-142B Imrer Shell Axial Weld B 83653 0.039 0.102 33.2 -31.8 -48.5 11.2 0.042 0.102 34.5 -30.0 -47.2 12.6 101-142C Lower Shell Axial Weld C 83653 0.039 0.102 33.2 -30.0 -47.0 12.2 0.042 0.102 34.5 -28.0 -45.8 13 6  !

101-171 Intermediate to Lower Cire. Weld 83653 0.039 0.102 6.5A -68.6 -72.2 -16.1 0.042 0.102 34.5 -19.4 -38.4 17.8 g

Table 1 - Effect of CEOG-RVWG Best-Estimase Chemistry for Beltline Welds on Existag VEGP 1 Reactor Vessei lasegrity Analysism  ;

Notes. m Intermedisse Shell Plase B8805-2 is the limiting behline matenal for VEGP I and is not affected by the GOG-RVWG best.ee- detenanation.

A CF in accord:nce with R.G.1.99. Revision 2. Regulatory Posation 2.I. based on credible surveillance desa. '

M The current VEGP 1 PTS analysas and this compenson are based on the pre-1995 PTS rule. Use of the post-1995 PTS rule resuks in a reduction in the margin term and a -.% -c4/reduction in the projected RTpts values.

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tut %) (=t%) (v) 'as aprvn > gaaserv) tueSe (m%) (T) i ET ET ET ET B0628-1 ImmerShet Fluee88 C3500 2 0.05 0.59 31 110 93 124 0.25 0.59 31 110 93 124 101-124A p SheH AsialWeld A 37005 0.07 0.13 47 31 55 107 0.054 0.151 44.6  %.S 51.4 104 101-1248 p SheII AxialWeldB 37005 0.07 0.13 47 31 55 107 0.054 0.151 44.6 XE 51.4 104 101-124C p 5 hell Axial Weid C 37005 0.07 0.13 47 81 55 107 0.054 0.151 44.6 ME 51.4 104 i

, 101-142A Immer SheE AxialWeld A E7005 0.07 0.13 47 31 55 107 0.054 0.151 44.6 M.8 51.4 104 101-142B ImmerSheE AxialWeld B 37005 0.07 0.13 47 31 55 107 0.054 0.151 44.6  %.S 51.4 104 101-142C IowerSheE AxialWeld C 37005 0.07 0.13 47I 81 55 107 0.054 0.151 44.6  %.3 51.4 104 .

101-171- p toImmer Cisc. Weld 31005 0.06 0.12 43 54 29 32 0.054 0.151 44.6 56.8 31.4 34 Table 2 - FEact of CEDG4VWG BestSw % for Beitimme Weids on Exmang VEGP 2 Ramenor Vessel lmespity AnalyamP l 1

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88 The VEOP 2 rescear vessel isenytty ammlyans - _ C, assummes se peak vessel Summoe for aR amiel weids.

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