RC-95-0266, Forwards Response to Rev 1,Suppl 1 to GL 92-01, Reactor Vessel Structural Integrity

Forwards Response to Rev 1,Suppl 1 to GL 92-01, Reactor Vessel Structural Integrity

ML20094G981
Person / Time
Site:	Summer
Issue date:	11/07/1995
From:	Gabe Taylor SOUTH CAROLINA ELECTRIC & GAS CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-92-01, GL-92-1, RC-95-0266, RC-95-266, NUDOCS 9511130280
Download: ML20094G981 (25)

Text

_ _

8 S uth Circlina El ctric & Gis Comp 1ny Guy J.T or

. Jenkinsville. SC 29065 N > clear Operations (803) 345-4344 SCE8:G escencme November 7,1995 RC-95-0266 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 Gentlemen:

Subject:

VIRGIL C. SUMME R NUCLEAR STATION (VCSNS)

DOCKET NO. 50/395 OPERATING LICENSE NO. NPF-12 RESPONSE TO GENERIC LETTER 92-01 REVISION 1, SUPPLEMENT 1 Pursuant to 10CFR50.54(f), the Nuclear Regulatory Commission (NRC) issued Generic Letter 92-01, Revision 1, Supplent 1, " Reactor Vessel Structural Integrity", requiring that each plant respond to parts (2), (3) and (4) by November 19,1995 providing results of actions taken to locate all data relevant to the determination of Reactor Pressure Vessel (RPV) Integrity.

South Carolina Electric & Gas Company (SCE&G) acting for itself and as agerit for South Carolina Public Service Authority, hereby submits the attached in response to Generic Letter 92-01, Revision 1, Supplement 1. This submittal addresses the action required by parts (2),(3) and (4).

I declare that these statements and matters set forth herein are true and correct to the best of my know: '.e,information, and belief.

Should you have any gestions, please call Mr. Jim Turkett at (803) 345-4047, at your convenience.

Very truly yours, l Gr yl JWT/GJT/nkk Attachment c: J. L. Skolds J. B. Knotts, J r.

O. W. Dixon General Managers R. R. Mahan (w/o attachment) NRC Resident inspector R. J. White Central File System S. D. Ebneter RTS (LTR 920001-1)

5. Dembek File (815.14) g f N(b0 , 0 9511130280 951107 PDR ADOCK 05000395 P PDR

1

, ' documbnt Control Desk LTR 950007 RC-95-0266 Page 2 of 2 l

l STATE OF SOUTHCAROLINA  :

TO WIT :

COUNTY OF FAIRFIELD  :

Ihereby certify that on the 7N day of Alu.b 19 96before me,the subscriber, personally appeared Gary J. Taylor, being duly sworn, states that he is Vice President, Nuclear Operations of the South Carolina Electric & Gas Company, a corporation of the State of South Carolina, that he provides the foregoing response for the purposes therein set forth, that the statements made are true and correct to the best of his knowledge,information, and belief, and that he was authorized to provide the response on behalf ofsaid Corporation. l WITNESS my Hand and Notarial Seal /Notary PubliMor M %

South Carolina MyCommi:sfon ExpiresJufy13,2005 My Commission Expires Date c

7 NUCLEAR EXCELLENCE- A SUMMER TRADITIONI

Atttch'm:nt I to Document Control D:sk Letter  :

' LTR 920001-1  :'

. RC-95-0266

Page 1 of 2,3 4  !

! RESPONSE TO GENERIC LETTER 92-01, l REVISION 1, SUPPLEMENT 1  :

PARTS (2),(3) AND (4) .

i  :

CONTENTS l

REFERENCES .

i i

i

!; RESPONSE TO PART (2) 3  ;

I

- RESPONSE TO PART (3) 3 l 1  !

a i

RESPONSE TO PART(4) 4  :

i RESPONSE - 10CFR 50.60 COMPLIANCE 5  !

> \

RESPONSE - 10CFR 50.61 COMPLIANCE 6 l

TABLES 1-16 8 i

l 4

l i

l 4

4 4

h l

Attrchtnant I to Document Control Desk Letter .

?' - LTR 920001-1 RC-95-0266

< Page 2 of 2,3  ;

References:

i i

j (1) SCE&G letter from J. L Skolds to Document Control Desk, USNRC; dated June i i 30,1992; subject: Virgil C. Summer Nuclear Station, Docket No. 50/395, Operating license Reactor Vessel No. NPF-12, StructuralIntegrity LTR Resp (onse to Generic Letter 92-01, Revision 1!

920001).

, i

(2) Davidson, J.A., Yanichko, S.E., " South Carolina Electric And Gas Company Virgil  :

4 C. Summer Nuclear Plant Unit No.1 Reactor Vessel Radiation Surveillance  !

Program," WCAP-9234, January,1978

! (3) Bogc s, R.S., Fero, A.H., Kaiser, W.T., " Analysis of Capsule U from the South l

Caro ina Electric and Gas Company Virgil C. Summer Unit 1 Reactor Vessel j Radiation Surveillance Program, WCAP-10814, June 1985.

(4) Colburn, D.J., Lamantia, LA., Albertin, L., " Analysis of Capsule V from the  !

South Carolina Electric and Gas Company Virgil C. Summer Unit 1 Reactor Vessel Radiation Surveillance Program, WCAP-11726, January 1988.

l' (5) Chicots, J.M., Lloyd, T.M., Albertin, L, " Analysis of Capsule X from the South Carolina Electric and Gas Company Virgil C. Summer Unit 1 Reactor Vessel Radiation Surveillance Program, WCAP-12867, March 1991.

4

(6) SCE&G letter from O. 5. Bradham to Document Control Desk, USNRC, dated

January 12,1989, subject: Virgil C. Summer Nuclear Station Docket 50/395, 4

Operating License No. NPF 12, Response to Generic Leterr 88-11.

$ (7) Chicots, J. M., Ramirez, M.A., " Evaluation of Pressurizer Thermal Shock for V. C.

Summer," WCAP-13209, March 1992.

2 1

i i

J j

j

' Attschmtnt I to Documant Control Dtsk Letter LTR 920001-1 '

. RC-95-0266 Page 3 of,23 j Generic Letter Reauired information (2):

An assessment of any change in best-estimate chemistry based on consideration of ,

all relevantdata. l

Response

Reference (1) submitted the best-estimate chemistry for the V. C. Summer Nuclear  ;

Station (VCSNS) Reactor Pressure Vessel (RPV) beltline material. This data has been reviewed and verified to be the best-estimate. This data is repeated below with the sources of this information:

DESCRIPTION HEAT NO. %CU %NI SOURCE Intermediate Shell Plate A9154-1 .10 .51 Lukens Steel Company Test Certificate, Dated 3/14/72 Intermediate Shell Plate A9153-2 .09 .45 Lukens Steel Company Test Certificate, Dated 3/14/72 Lower Shell Plate A9923-1 .08 .41 Lukens Steel Company Test Certificate, Dated 1/11/72 Lower Shell Plate A9923-2 .08 .41 Lukens Steel Com; any Test Certificate, Dated 1/11/72 Beltline Weld 4P4784 .05 .91 Average of data from:

Chicago Bridge & Iron Company Certificate of Analysis, Test

1. PT321 A- Tandem Wire, Revised 2/9/73

%CU =.05 %NI=.91 Chicago Bridge & tron Company Certificate of Analysis, Test

1. PT3218- Single Wire, Revised 2/9/73

%CU =.06 %NI=.87 Chemical Analysis of Irradiated Charpy Specimen CW-14 Reported in Ref. (3)

%CU =.04 %NI=.95

l l ' Attachm:nt I to Docum::nt Control Disk Letter LTR 920001-1

, RC-95-0266 Page4 of 23 Generic letter Required information (3):

A determination of the need for use of the ratio procedure in accordance with the established Position 2.1 of the Regulatory Guide 1.99, Revision 2, for those licenseu that use surveillance data to provide a basis for the RPV integrity evaluation.

Response

For the VCSNS Reactor Vessel Radiation Surveillance Program, Chicago Bridge and iron Company furnished sections of SA533 Grade B Class 1 plate used in the core region of the VCSNS Reactor Pressure Vessel, specifically, from the 7.75 inch intermediate shell plate A9154-1. Also su aplied was a weldment made from sections of plate A9154-1 and adjoining lower she I plate C9923-2, using RACO 1 NMM weld wire, heat number 4P4784,linde 124 flux, lot number 3930. This weld wire and flux combination is identical to that used in forming the longitudinal weld seams in the intermediate and lower shell courses and the girth weld joining these two shell courses. Reference (3), Table 4-1, provide a chemical analysis of the RPV surveillance material. This information provides clear evidence that the cop per and nickel content of the surveillance material does not differ from that of the vessel material.

Therefore, the measured values of ARTNDT do not need to be adjusted by multiplying them by the ratio of the chemistry factor for the vessel weld to that for the surveillance weld.

The surveillance data has been fitted to obtain the relationship of ARTNDTto fluence. This fit is calculated in table 1. The values of ARTNDT used in Table 1 have been revised from that submitted in Reference (1). These revised values are included in Tables 2 and 8 and are underlined. An explanation for these revisions is included below in the response to the Required information (4) of the Generic Letter.

Generic Letter Required information (4):

A written report providing any newly acquired data as specified above.

Response

SCE&G has conducted a review of all the data pertinent to the chemical composition and mechanical properties of the materialin the beltline region of the VCSNS Reactor Vessel. As a result of this review, several significant changes were made to the data reported in reference (1). There are two primary reasons for these changes as explained by the following:

- A more exact source of data was obtained for the preirradiated Charpy V-Notch impact Test results for the Reactor Pressure Vessel core region beltline weld metal.

Previously, the data from the Charpy V-Notch Impact Test results for surveillance capsules were plotted on a graph and then a curve was fitted to the data points by hand. In order to standardize the interpretation of the Charpy V-Notch Impact Test results and to avoid individualinterpretation of data, all the results have been replotted using the computer program CVGRAPH, Version 4.0. This program has been developed by the Westinghouse Owner's Group. The CVGRAPH 4.0 Program is a data tool for the fitting of Charpy V- Notch test results.

l

'Atticilment I to Document Control Desk Letter LTR 920001-1 RC-95-0266 Page 5 of 23 The program stores the data and graphically displays the test results. The curve fit routine uses a hyperbolic tangent function. The curve calculator provides results in temperature at 30 and 50 ft-lb impact energy.

Tables 2 through 8 are tables that were previously submitted in reference (1). Tables 2,7, and 8 required revisions due to the reasons noted above. Revised data is indicated by an underline.

Tables 9 and 10 include revisions made to tables of a printout from the database RVID. These revisions are also due to the reasons noted above. Revised data is indicated by a cloud.

Also during'this review, existing industry wide databases were reviewed to identify any " Sister vessels. It was concluded that the intermediate and lower shell axial welds of the Shearon Harris Unit 1 Reactor Vessel were fabricated using the same weld heat material,4P4784, as that used for welds in the VCSNS vessel. Therefore, the Shearon Harris Reactor Vessel is a " Sister" vessel of the VCSNS vessel. In this regard, discussions between these two units have been instigated and are continuing. The intent of these discussions is for a continuing transfer of data on reactor vessel mechanical properties. The Shearon Harris vessel does not contain any surveillance capsules from the weld material, heat 4P4784. Therefore, no surveillance data will be available for use in the VCSNS surveillance program.

However, data from the VCSNS surveillance program will be made available to Shearon Harris.

Additional information reauired by Generic Letter (4)

(1) the results of any necessary revisions to the evaluation of RPV integrity in accordance with the requirements of 10 CFR 50.60,10 CFR 50.61, Appendices G and H to 10 CFR Part 50, and any potential impact on the LTOP or P-T limits in the technical specifications.

Response - 10 CFR 50.60 compliance:

This submittal does not make any changes to the Charpy Upper Shelf Energies (USE) that would affect the evaluation reported in reference (1) for the compliance with Paragra follows:phs IV.A.1 of Appendix G to 10 CFR Part 50. This evaluation is repeated as "A maximum 10 ft-lb, or 7.6%, decrease in USE was determined for plate A9154-1 (longitudinal orientation). This data is plotted on the Regulatory Guide 1.99, 2 Revision 2 curve using the guidance set forth in paragraph C.2.2 of Regulatory Guide 1.99, Revision 2. Based on actual data, the projected decrease in USE at the arojected end of life (32 EFPY) fluence of 3.87X1019, reference 5,is expected to be

,ess than 10%. Using the lowest initial USE value of 75 ft-lb for the intermediate shell plate A9154-1, the resulting end of life USE value is arojected to be 67.5 ft-lb.

This is above the 50 ft-Ib limit and, therefore,is acceptab e."

The P-T limits in the Technical Specifications are to assure that plant operations comply with the requirements of 10 CFR 50.60 and Appendix G. The P-T limits are based on an initial unirradiated value of RTNDT which increases as the material is exposed to fast neutron radiation. The initial value of RTNDT, adjusted for radiation

l

'Att:chment I to Document Control D sk Letter LTR 920001-1

, RC-95-0266 Page 6 of 23

l exposure,is termed ART. To find the most limiting ART at any time period in the l reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT. The current P-T curves in the VCSNS Technical Specifications are based on a value of ART, adjusted for radiation exposure, derived from data obtained from three surveillance ca asules. The derivation of these curves is summarized in reference (5). The value of the limiting ART, and the corresponding P T curves,is based on the following:

EFPY = 14 Controlling Material = Lower Shell Initial RTNDT = 10 F ART at 1/4 T = 96*F I ART at 3/4 T = 83 F Revisions to data included in this submittal will revise the data used to derive the values of ART for some of the beltline materials. Tables 11 through 15 provide a revised set of calculations for the derivation of ART (14 EFPY) for all beltline 4

materials. From thue Tables,it is concluded that the controlling material remains the lower shell plate. The revised values of ART are as follows (note that they remain unchanged):

i Controlling Material (u Jsed) = Lower Shell Initial RTNDT = 10*F l ART at 1/4 T(revised) = 96*F ART at 3/4 T(revised) = 83*F The ASME approach (Part 50 Appendix G sanctioned) for calculating the allowable limit curves for various heatup and cooldown curves specifies that the total stress intensi factor, K i, for the combined thermal and pressure stresses at any time during eatup or cooldown cannot be greater than the reference stress intensity factor, Kig, for the metal temperature at that time.

Ki function (P,T)

Ki < KIR Ki n is obtained from the reference fracture toughness curve given by the following equation:

Ki a = 26.78 + 1.223 exp [0.0145 (T- ART + 160)]

The values of the revised ART at 1/4 T and at 3/4 T a e equal to the values used to develop the current P-T curves. Therefore, the current P-T curves remain valid.

In regards to Low Temperature Overpressure Protection (LTOP), the RHR relief valve setpoints have been verified for compliance with Appendix G based upon the pressure and temperature limits of the current P-T curves in the Technical

/

'Att: chm:nt I to Document Control Desk Letter j LTR 920001-1 l

. RC-95-0266 Page 7 of 2.3 Specification. Therefore, the current setpoints remain valid for the revised data of i this submittal.

It is concluded that the current evaluations contained in reference (5) for compliance with 10 CFR 50.60 (and Appendix G) and for the development of the current P-T curves in the Technical Specifications remain valid for the revised data of this submittal. Therefore, RPV integrity remains in compliance with the requirements of 10 CFR 50.60 and no revisions are necessary for the current P-T curves of the i Technical 5pecifications or the RHR relief valve setpoints for LTOP. l Response - 10 CFR 50.61 compliance:

Using the prescribed PTS rule methodology outlined in 10 CFR 50.61, RTPTS values were generated for all beltline region materials of the VCSNS Reactor Vessel as a i function of end-of-life (32 EFPY) fluence values. Table 16 provide a summary of the RTPTS calculated values for all the beltline region materials. The information on this table updates and revises that of references (6) and (7). The fluence data were generated based on the most recent surveillance capsule program results of I I

reference (5). The chemistry factors for the intermediate shell plate, A9154-1, and the welds were calculated using surveillance capsule data and Regulatory Guide 1.99, Revision 2, methodology as shown in Table 1.  !

Per 10 CFR 50.61, the NRC screening value for RTpTs using the current projected fluence values for the end-of-life (32 EFPY) is 270*F for plates and axial welds and 300 F for circumferential welds. As indicated in Table 16, the maximum projected RTPTs for all alates and axial welds is 113 F for the lower shell plate C9923-1. This is wellbelow11e screening criteria of 270 F. The maximum pro lected RTPTSf or the circumferential welds is 22 F. This is also significantly below t le 300 F screening criteria. Therefore, the PTS Rule evaluation based on the revised data of this submittal concludes a compliance with the requirements of 10 CFR 50.61.

. 'Attechment I to Document Control Desk Letter LTR 920001-1

. RC-95-0266 Page 8 of 2,3 ,

TABLE 1  ;

CALCULATION OF CHEMISTRY FACTORS USING  ;

V. C. SUMMER SURVEILLANCE CAPSULE DATA I

Component Capsule Fluence F_F DRTNDT FF*DRTNDT (FF)2 Int.Shell, A9154-1 U 0.639 0.874 36 36.464 0.765 (Long.) V 1.470 1.107 53 58.671 1.225 l X 2.460 1.242 38 47.196 1.543 Int.Shell, A9154-1 U 0.639 0.874 15 13.110 0.765 (Trans.) V 1.470 1.107 33 36.531 1.225 X 2.460 1.242 26 32.292 1.543 j l

219.264 7.065 Chemistry Factor = 219.264 / 7.065 = 31.035 Component Capsule Fluence Ff DRTNDT FF*DRTNDT (FF)2 Weld Metal U 0.639 0.874 23 20.102 0.765 V 1.470 1.107 47 52.029 1.225 X 2.460 1.242 23 28.566 1.543 100.697 3.533 Chemistry Factor = 100.697 / 3.533 = 28.502 4

- .~ - - - . - _ , _

Attr.chment I to Docum:nt Control Desk Letter '

LTR 920001-1 RC-95-0266 .

Page 9 of 23 .

TABLE 2 .-

EFFECT OF 550*F IRRADIATION ON NOTCH TOUGHNESS PROPERTIES OF THE -

V. C. SUMMER UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE MATERIAL Fluence Average 30 Ft-Lb Average 50 Ft-Lb pt on7t USE'A Material Capsule 1019 N/CM2 Temp. (*F) AT30 Temp. (*F)

^A Decrease Reference Full Shear (Ft-Lb)

Int. Plate A9154-1 unirr. 0 -22 --

7_ 132 --

2 (Longitudinal) U 0.639 14 36 52 131 .8 3 V 1.470 31 53 71 122 7.6 4 X 2.460 16 38 51 125 5.3 5 Int. Plate A9154-1 unirr. 0 28 --

71 75 -

2 (Transverse) U 0.639 43 15 94 75 0 3 V 1.470 61 33 137 76 -

'4 X 2.460 54 26 112 73 2.7 5 Weld Metal unirr. 0 -53 --

-13 91 --

2 U 0.639 -30 23 10 87 4.4 3 V 1.470 -6 47 30 85 6.6 4 X 2.460 -30 23 5 85 6.6 5 HAZ Metal unirr. 0 -93 --

-65 130 -

2 U 0.639 -58 3_5 -28 121 6.9 3 l V 1.470 -43 50 -13 111 14.6 4 X 2.460 -37 56 -5 111 14.6 5

'dttschmtnt I to Docum:nt Control Desk Letter LTR 920001-1

. RC-95-0266 j i

Page.10 of 23' 1 7  !

!, l l

_ TABLE 3 t )

l l '

DESCRIPTION: Reactor Vessel Beltline Plate (Intermediate Shell)

HEAT NO.: A9154-1

SPECIFICATION NO.

SA533 Grade B Clav,1

SUPPLIER

Lukens Steel Company l HEAT TREATMENT: Austenitized at 1625*F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> '

Water Quenched i

! Tempered at 1280*F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />  !

j- Air Cooled l Stress Relieved at 1050*F for 43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br /> '

L Furnace Cooled / Air Cooled COMPOSITION: .10 %Cu, .51 %Ni, .009 % P, .015 %S i

, DROP WElGHT TNDT: -20*F '

RTNDT: 30 F (measured)

, PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE V. C. SUMMER UNIT NO.1 REACTOR PRESSURE VESSELINTERMEDIATE SHELL PLATE A9154-1 i (LONGITUDINAL ORIENTATION) i i" TEST TEMPERATURE IMPACT ENERGY

(*F) (FT-LB) l + 212 121, 144, 144 80, 107, 112

+ 70

+ 40 70, 88, 70 i + 10 54, 50, 48 i -20 17, 10, 15

!- -100 5, 3, 4 i

1 (TRANSVERSE ORIENTATION)

TEST TEMPERATURE IMPACT ENERGY l l

( F) (FT-LB)

~

+ 212 82.5, 76.5, 83 i + 120 56, 64, 74  ;

2 + 90 51.5, 60.5, 58 l l + 70 56, 36, 40

+ 40 36, 41, 38

-20 12.5, 20, 19 l

', l t

4

- ' ~ -

. y . , _ . _ _

i Attcchin::nt I to Document Control Desk Letter .

LTR 920001-1 l

. RC-95-0266  ;

Page,11 of 2,3 TABLE 4  ;

i i DESCRIPTION: Reactor Vessel Beltline Plate (Intermediate Shell) i i HEAT NO.: A9153-2 l SPECIFICATION NO.: SA533 Grade B Class 1 4 SUPPLIER: Lukens Steel Compan

HEAT TREATMENT

Austenitized at 1625*yF for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> l Water Quenched

Tempered at 1280 F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

Air Cooled i i Stress Relieved at 1050 F for 43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br />

! Furnace Cooled / Air Cooled l COMPOSITION: .09 %Cu, .45 % Ni, .006 %P, .016 %S DROPWEIGHTTNDT: -20 F

] RTNDT: -20 (measured)

PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE V. C. SUMMER UNIT NO.1 REACTOR PRESSURE VESSELINTERMEDIATE SHELL PLATE A9153-2 i (LONGITUDINAL ORIENTATION) I TESTTEMPERATURE IMPACT ENERGY

(*F) (FT-LB)

+ 212 142, 146, 136

+ 70 108, 107, 105

+ 40 82, 83, 82

+ 10 68, 56, 74

-20 54, 58, 54

-100 9, 3, 8 (TRANSVERSE ORIENTATION)

TEST TEMPERATURE IMPACT ENERGY

(*F) (FT-LB)

+ 212 101, 108, 111.5

+ 120 93, 90, 94

+ 70 65, 71, 68

+ 40 59, 51, 50

+ 20 43, 42.5, 47

-20 14, 26, 18 l

dttrchrnent I to Docum:nt Control D:sk Lctter LTR 920001-1

. RC-95-0266 l Page.12 of 23 l TABLE 5 DESCRIPTION: Reactor Vessel Beltline Plate (Lower Shell)

HEAT NO.: C9923-2 SPECIFICATION NO.: SA533 Grade B Class 1 ,

SUPPLIER: Lukens Steel Company HEAT TREATMENT: Austenitized at 1600 F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Water Quenched Tempered at 1260*F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />  !

Air Cooled Stress Relieved at 1075"F for 41 hours4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br /> Furnace Cooled / Air Cooled COMPOSITION: .08 %Cu, .41 %Ni, .005 % P, .015 %S DROPWEIGHTTNDT: -10 F RTNDT: 10 (measured)

PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE V. C. SUMMER UNIT NO.1 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE C9923-2 (LONGITUDINAL ORIENTATION)

TEST TEMPERATURE IMPACT ENERGY (F) (FT-LB)

+ 212 164, 164, 155

+ 40 86, 76, 85

+ 10 56, 56, 55

-20 42, 50, 35

-50 8, 6, 12

-100 4, 5, 5 (TRANSVERSE ORIENTATION)

TEST TEMPERATURE IMPACT ENERGY l

( F) (FT-LB) l

+ 212 93, 94, 88 i

+ 120 82, 79, 84

+ 70 55, 50, 51

+ 50 40, 46, 53

+ 40 56, 42.5, 37.5

-10 23, 31, 33.5 I l

Att chm:nt I to Documtnt Control Desk Lettar LTR 920001-1

. RC-95-0266 ,

Page.13 of 23 P

TABLE 6 DESCRIPTION: Reactor Vessel Beltline Plate (Lower Shell)

HEAT NO.: C9923-1 SPECIFICATION NO.: SA533 Grade B Class 1 1

SUPPLIER: Lukens Steel Company HEAT TREATMENT: Austenitized at 1600 F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

! Water Quenched Tempered at 1260*F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> i Air Cooled

, Stress Relieved at 1075 F for 41 hours4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br /> t Furnace Cooled / Air Cooled -

COMPOSITION: .08 %Cu, .41 %Ni, .005 %P, .014 %5 DROP WEIGHT TNDT: -30 F RTNDT: 10 (measured) 1

, j PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE V. C. SUMMER UNIT NO.1

REACTOR PRESSURE VESSEL INTERMEDIATE SHELL PLATE C9923-1 (LONGITUDINAL ORIENTATION)

TEST TEMPERATURE IMPACT ENERGY

( F) (FT-LB)

+ 212 148, 147, 150

+ 40 92, 100, 89 ,

+ 10 92, 70, 82  !

-50 52, 60, 44 l

-75 35, 17, 12

-100 7, 10, 9 (TRANSVERSE ORIENTATION)

TEST TEMPERATURE IMPACT ENERGY

("F) (FT-LB)

~

+ 212 104, 114, 100

-i

+ 120 74.5, 81, 80

+ 70 68, 51, 51

+ 50 47, 44, 43.5

+ 30 37, 42.5,35

-30 11, 9, 12

dttich' :nt m I to Document Control Desk Letter LTR 920001-1

. RC-95-0266 Page.14 of 23 TABLE 7 DESCRIPTION: Reactor Vessel Core Region Beltline Weld Metal WIRE HEAT NO.: 4P4784 WIRE TYPE: RACO 1NMM FLUX TYPE: Linde 124 FLUX LOT: 3930 ,

FABRICATOR: Chicago Bridge and Iron Co. l HEATTREATMENT: 1150*F for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> I WIRE COMPOSITION: .05 %Cu, .91 %Ni, .013 %P, .012 %5 DROPWElGHTTNDT: -50*F RTNDT: -44 (measured) ,

PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE V. C. SUMMER UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION BELTLINE WELD METAL -

TESTTEMPERATURE IMPACT ENERGY

( F) (FT-LB)

+ 212 79, 87, 87

+ 40 64, 70, 66 1 TT5 53, 46, so  :

7 33, 22, 38 I

-T60 8, 7, 8 <

T70 12, 7, 9 l

_ _ _ _ ___1

~

Attachment I to Document Control Desk Lett;r -

LTR 920001-1 RC-95-0266 .

Page 15 of 23 .

TABLE 8 COMPARISON OF V. C. SUMMER UNIT 1 SURVElLLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFT 5 ' .

AND UPPER SHELF ENERGY DECREASES WITH REGULATORY GUIDE 1.99 REVISION 2 PREDICTIONS 30 ft-!b Transition Temp. Shift Upper Shelf Eneray Decrease R. G.1.99 Rev. 2 R. G.1.99 Rev. 2 Fluence (Predicted) Measured (Predicted) Measured Material Capsule 1019 n/cm2 (*F) (*F) (%) (%)

Int. Plate A9154-1 U 0.639 57 36 17 1 (Longitudinal) V 1.47 72 53 21 8 X 2.46 81 38 23 5 int. Plate A9154-1 0 0.639 57 15 17 0 (Transverse) V 1.47 72 31 21 0 X 2.46 81 26 23 3 Weld Metal U 0.639 59 2_1 18 4 V 1.47 75 4_7 22 7 X 2.46 85 23 25 7 HAZ Metal U 0.639 -

35 --

7

~

V 1.47 -

50 --

15 X 2.46 -

56 --

15

Att: chm:nt I to Docum:nt Control Desk Letter -

LTR 920001-1 RC-95-0266 .

Page 16 of 23 TABLE 9 REACTOR VESSELINTEGRITY DATABASE Summary File for Upper Shelf Energy Plant Name 1/4T Method  % Drop Method Beltline Heat No. Material USE @ EOL Neut. Flu Unirr Determ USE @ EOL Determ ident. Ident. Type @ 1/4T @ EOL USE Unirr USE @ 1/4T  % Drop Cu Summer (Continued) Docket No.: 50-395 LOWER SHELL C9923-1 A 5338 81 2.431 106 DIRECT 23.4 % Position 1 0.08 of RG 1.99, Rev.2 LOWER SHELL C9923-2 A 5338 70 2.431 92 DIRECT 23.4 % Position 1 0.08 of RG 1.99, Rev.2 WELDS 4P4784 LINDE 2.431 DIRECT 7.4 % Surveillance 0.05 124 data INTERMEDIATE A9153-2 A 5338 82 2.431 107 DIRECT 23.4 % Position 1 0.09 SHELL of RG 1.99, Rev.2 INTERMEDIATE A9154-1 A 5338 74 2.431 81 DIRECT 8.5% Surveillance 0.10 SHELL data References for Summer

> > > > >GL 92-01 References < < < < <

Fluence, chemical composition, and IRTndt data are from June 30,1992, letter from J. L Skolds (SCE&G) to USNRC Document Control Desk, subject: Response to Generic Letter 92-01, Revision 1. Reactor Vessel 5tructural Integrity UUSE data are from Table A-1 of WCAP-10814

Attachment I to Document Control Desk Letter -

LTR 920001-1 RC-95-0266 ,

Page 17 of 23 TABLE 10

~

REACTOR VESSEL INTEGRITY DATABASE Summary File for PTS Plant Name ID Neut. Method Fluence Method of Method of Beltline Heat No. RTpts @ Fluence Determ ARTndt Factor Chemistry Determin. Determin.

Ident. Ident. EOL @ EOL IRTndt IRTndt at EOL @ EOL Factor CF Margin Margin Cu% Ni%

Summer EOL: 08/06/22 Docket No.: 50-395 INTERMEDIATE A9153-2 92 3.87000 -20 PLANT 5PEC 78.2 1.349 58.00 Table 34.00 l TABLE 0.090 0.450 SHELL INTERMEDIATE A9154-1 3.87000 30 PLANT 5PEC 1.349 j Calculated 17.00 TABLE 0.100 0.510 SHELL LOWER SHELL C9923-1 113 3.87000 10 PLANT SPEC 68.8 1.349 51.00 Table 34.00 TABLE 0.080 0.410 LOWER SHELL C9923-2 113 3.87000 10 PLANT 5PEC 68.8 1.349 51.00 Table 34.00 TABLE 0.080 0.410 WELDS 4P4784 h 3.87000 -44 PLANT 5PEC 1.349 l 2 Calculated 28.00 TABLE 0.050 0.910 References for Summer

> > > > >G L 92-01 References < < < < <

Fluence, chemical composition, and IRTndt data are from June 30,1992, letter from J. L Skolds (SCE&G) to USNRC Document Control Desk, subject: Response to Generic Letter 92-01, Revision 1, Reactor Vessel 5tructural integrity UUSE data are from Table A-1 of WCAP-10814

' Attachment I to Docum:nt Control Dask Letter LTR 920001-1 RC-95-0266 Page 18 of 23 TABLE 11 CALCULATION OF ART FOR V. C. SUMMER REACTOR VESSEL MATERIAL INTERMEDIATE SHELL PLATE A9154-1 Parameter 14 EFPY 1/4T 3/4T Chemistry Factor, CF (*F) 31 31 Fluence, f (1019 n/cm2)(a) 1.08 0.43 Fluence Factor, ff 1.023 0.765 ARTNDT= CF x ff ( F) 31.7 23.7 initial RTNDT,i( F) 30 30 Margin, M ( F) 17 17 Adjusted Reference Temperature, 79 71 ART = Initial RTNDT + ARTNDT + Margin Total ART 14EFPY(b) 96 83 (a) Fluence, f, is based upon fsurf (1019 n/cm2, E > 1 Mev) = 1.73 at 14EFPY (projections determined from Reference (5), Table 6-13). The V. C. Summer reactor vessel wall thickness is 7.75 inches at the beltline region.

(b) The current V. C. Summer heatup and cooldown pressure / temperature limits are based upon these total ART values for 14EFPY,

Attrc'hmtnt I to Document Control Dcsk Letter i LTR 920001-1

. RC-95-0266 i Page 19 of,23 TABLE 12 CALCULATION OF ART FOR V. C. SUMMER REACTOR VESSEL MATERIAL  !

INTERMEDIATE SHELL PLATE A9153 2 ,

Parameter 14 EFPY 1/4T 3/4T Chemistry Factor, CF ( F) 58 58 Fluence, f (1019 n/cm2)(a) 1.08 0.43 Fluence Factor, ff 1.023 0.765 i

ARTNDT= CF x ff ( F) 59.3 44.4 Initial RT NDT,1( F) -20 -20 Margin, M (*F) 34 34 Adjusted Reference Temperature, 73.3 58.4 ART = lnitial RTNDT + ARTN DT + Margin Total ART 14EFPY(b) 96 83 (a) Fluence, f, is based upon fsurf (1019 n/cm2, E > 1 Mev) = 1.73 at 14EFPY (projections determined from Reference (5), Table 6-13). The V. C. Summer reactor vessel wall thickness is 7.75 inches at the beltline region.

(b) The current V. C. Summer heatup and cooldown pressure / temperature limits are based upon these total ART values for 14EFPY.

t l

4

'AttachmInt I to Documznt Control Desk Letter

. LTR 920001-1 RC-95-0266 Page 20 of 23 TABLE 13 CALCULATION OF ART FOR V. C. SUMMER REACTOR VESSEL MATERIAL LOWER SHELL PLATE C99231 Parameter 14 EFPY , , , , ,

1/4T 3/4T Chemistry Factor, CF ("F) 51 51 Fluence, f (1019 n/cm2)(a) 1.08 0.43 Fluence Factor, ff 1.023 0.765 ARTNDT= CF x ff ( F) 52.2 39.0 Initial RTNDT,1 ( F) 10 10 Margin, M (*F) 34 34 Adjusted Reference Temperature, 96 83 ART = Initial RTNDT + ARTNDT + Margin Total ART 14EFPY(b) 96 83 (a) Fluence, f,is based upon fsurf (1019 n/cm2, E > 1 Mev) = 1.73 at 14EFPY (projections determined from Reference (5), Table 6-13). The V. C. Summer reactor vessel wall thickness is 7.75 inches at the beltline region.

(b) The current V. C. Summer heatup and cooldown pressure / temperature limits are based upon these total ART values for 14EFPY.

Attrchm:nt I to Docum:nt Control D:sk Lett:r LTR 920001-1

. RC-95 0266 Page 21 of 23 TABLE 14 CALCULATION OF ART FOR V. C. SUMMER REACTOR VESSEL MATERIAL LOWER SHELL PLATE C9923-2 Parameter 14 EFPY 1/4T 3/4T Chemistry Factor, CF (*F) 51 51 i Fluence, f (1019 n/cm2)(a) 1.08 0.43 j

Fluence Factor, ff 1.023 0.765

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • sn*****

i 1

i I i ARTNDT= CF x ff ( F) 52.2 39.0 l Initial RTNDT,1 ( F) 10 10 Margin, M ( F) 34 34 t

i

~

Adjusted Reference Temperature, 96 83 ART = Initial RTNDT + ARTNDT + Margin

Total ART 14EFPY(b) 96 83 1

4 l l

1 l

Fluence, f, is based upon fsurf (1019 n/cm2, E > 1 Mev) = 1.73 at 14EFPY (a) 1 (projections determined from Reference (5), Table 6-13). The V. C. Summer reactor vessel wall thickness is 7.75 inches at the beltline region.

(b) The current V. C. Summer heatup and cooldown pressure / temperature limits are based upon these total ART 14EFPY Values for 14EFPY.

i i

I

1

^

Attac'hmtnt I to Docum:nt Control Desk Letter LTR 920001-1

. RC-95-0266 -

Page 22 of 23 TABLE 15 CALCULATION OF ART FOR V. C. SUMMER REACTOR VESSEL MATERIAL <

WELD (INTERMEDIATE TO LOWER SHELL) l l

l Parameter 14 EFPY  !

1/4T 3/4T Chemistry Factor, CF (*F) 28.5 28.5 Fluence, f (1019 n/cm2)(a) 1.08 0.43 Fluence Factor, ff 1.023 0.765 l

l ARTNDT= CF x ff ("F) 29.2 21.8 Initial RT NDT,1( F) -44 -44 Margin, M ( F) 28 28 Adjusted Reference Temperature, 13 6 ART = lnitial RTNDT + ARTNDT + Margin Total ART 14EFPY(b) 96 83

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • .************3.*****

(a) Fluence, f,is based upon fsurf (1019 n/cm2, E > 1 Mev) = 1.73 at 14EFPY (projections determined from Reference (5), Table 6-13). The V. C. Summer reactor vessel wall thickness is 7.75 inches at the beltline region.

(b) The current V. C. Summer heatup and cooldown pressure / temperature limits are based upon these total ART values for 14EFPY.

'Attcchmtnt I to Documsnt Control Desk Letter

. LTR 920001-1

, RC-95-0266 Pag,e 23 of,23 TABLE 16 V. C. Summer Reactor Vessel Beltline RTPTs Values At End Of Life (32EFPY)

Margin _ RTPTS Material ARTNDT( F) y initial RTNDT 4 (CF x FF*) ("F) ( F) ( F)

Intermediate Shell (31.0) 1.35 30 17 (89)

Plate, A9154-1 58 1.35 -20 34 92 Intermediate Shell Plate, A9153-2 Lower Shell Plate, 51 1.35 10 34 113 C9923-1 Lower Shell Plate, 51 1.35 10 34 113 C9923-2 1.35 -44 28 (22)

CircumferentialWeld (28.5)

Seam LongitudinalWelds (28.5) 1.08 -44 28 (15)

() Indicates numbers were calculated using surveillance capsule data.

• Fluence factor based upon peak inner surface neutron fluence of 3.87 x 1019 n/cm2 [ Reference (5)], except for longitudinal welds. For longitudinal welds, the fluence factor is based on a neutron fluence of 1.33 x 1019 n/cm2 (Reference (5)] at the inner surface of the weld.

\