ML20203C710

Reactor Pressure Vessel & Surveillance Program Matls Licensing Info for Surry Units 1 & 2

ML20203C710
Person / Time
Site:	Surry
Issue date:	03/31/1986
From:	Lowe A BABCOCK & WILCOX CO.
To:
Shared Package
ML18149A089	List: ML20203C710
References
BAW-1909, NUDOCS 8604210225
Download: ML20203C710 (42)

Text

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I I mum March 1986 I

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atAcron ratssUnz vrsszt AND suRvtIttANcz raocRAM g

MATERIALS LICENSING INFORMATION FOR SURRY UNITS 1 AND 2 I

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g I Babcock &Wilcox 8604210225 860415 a McDermott company I gDR ADO'JK 05000280 PDR

L r' BAW-1909 March 1986 r

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k REACTOR PRESSURE VESSEL AND SURVEILLANCE PROGRAM 7

MATERIALS LICENSING INFORMATION FOR SURRY UNITS 1 AND 2 i

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by A. L. Lowe, Jr., P.E.

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B&W Control No. 77-116380300 B&W Contract No. 583-7375, Task 042 i

Prepared for f Virginia Electric and Power Company Richmond, Virginia by Babcock & Wilcox Nuclear Power Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 a McDermott company

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I CONTENTS I

Page I 1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . 1-1 REACTOR VESSEL PATA BASES 2-1 I

2. . . . . . . . . . . . . . .

2.1. Surry Unit 1 . . . . . . . . . . . . . . . . . . 2-1 2.2. Surry Unit 2 . . . . . . . . . . . . . . . . . . 2-1 2-2 I 2.3.

2.4.

Surveillance Data Bases . . . . . . . . . . . .

Initial Value of Reference Temperature . . . . . 2-2

3. EVALUATION OF REACTOR VESSEL TOUGHNESS . . . . . . . . 3-1

4. REACTOR VESSEL SURVEILLANCE PROGRAMS . . . . . . . . . 4-1 4.1. Surry Unit 1 . . . . . . . . . . . . . . . . . . 4-1 4.2. Surry Unit 2 . . . . . . . . . . . . . . . . . . 4-2 4.3. Spare Capsules . . . . . . . . . . . . . . . . . 4-2

5. INTEGRATED REACTOR VESSEL SURVEILLANCE PROGRAM . . . . 5-1

6.

SUMMARY

. . . . . . . . . . . . . . . . . . . . . . . 6-1

7. REFERENCES . . . . . . . . . . . . . . . . . . . . . . 7-1 List of Tables Table 2-1. Identification of Reactor Vessel Beltline Region Weld Metals - Surry Unit-1 . . . . . . . . . 2-5 l

2-2. Chemical Compositien of Reactor Vessel Beltline Region Weld Metals - Surry Unit-1 . . . . . . . . . 2-6 2-3. Mechanical Properties of Reactor Vessel Beltline Region Weld Metal - Surry Unit-1 . . . . . . . . . . 2-7

.g dj 2-4. Identification of Reactor Vessel Beltline Region Base Materials - Surry Unit-1 . . . . . . . . 2-8 2-5. Chemical Composition of Reactor Vessel Beltline Region Base Materials - Surry Unit-1 . . . . . . . . 2-9 Mechanical Properties of Reactor Vessel Beltline 8 2-6.

Region Base Materials - Surry Unit-1 . . . . . . . . 2-10 2-7. Properties of Surveillance Program Plate and I 2-8.

Weld Material - Surry Unit-1 . . . . . . . . . . . .

Identification of Reactor Vessel Beltline Region Weld Metals - Surry Unit-2 . . . . . . . . .

2-11 2-12 I

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I List of Tables (Cont'd)

Table Page 2-9. Chemical Composition of Reactor Vessel Beltline Region Weld Metals - Surry Unit-2 . . . . . . . . . 2-13 2-10. Mechanical Properties of Reactor Vessel Beltline Region Weld Metal - Surry Unit-2 . . . . . . . . . . 2-14 2-11. Identification of Reactor Vessel Beltline Region Base Materials - Surry Unit-2 . . . . . . . . 2-15 2-12. Chemical Composition of Reactor Vessel Beltline Region Base Materials - Surry Unit-2 . . . . . . . . 2-16 2-13. Mechanical Properties of Reactor Vessel Beltline 3 Region Base Materials - Surry Unit-2 . . . . . . . . 2-17 5 2-14. Properties of Surveillance Program Plate and Weld Material - Surry Unit-2 . . . . . . . . . . . . 2-18 g 3-1. Evaluation of Reactor Pressure Vessel g Fracture Toughness - Surry Unit-1 . . . . . . . . . 3-3 3-2. Evaluation of Reactor Pressure Vessel Fracture Toughness - Surry Unit-2 . . . . . . . . . 3-4 4-1. Revised Surveillance Capsule Withdrawal Schedule - Surry Unit-1 . . . . . . . . . . . . . . 4-3 4-2. Revised Surveillance Capsule Withdrawal Schedule - Surry Unit-2 . . . . . . . . . . . . . . 4-4 List of Ficures Figure 2-1. Location and Identification of Materials Used in the Fabrication of the Belt-Line Region of Surry Unit 1 Reactor Pressure Vessel . . . . . . . . . . . 2-19 .

2-2. Location and Identification of Materials Used in the Fabrication of the Belt-Line Region of Surry Unit 2 Reactor Pressure Vessel . . . . . . . . . . . 2-20 I

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I-Babcock & Wilcox a McDermott company

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1. INTRODUCTION I This report provides a review and update of the materials data and information for the reactor pressure vessels of Surry Units 1 and 2 to ensure that they are in compliance with the require-I ments of 10CFR50, Appendix G.1 In addition, the reactor pressure vessel surveillance programs were reviewed for cobpli-ance with 10CFR50, Appendix H.2 The reactor pressure vessel surveillance capsule withdrawal schedule was modified to meet the intent of ASTM E185-823 as referenced by 10CFR50, Appendix H.

As a result of this review and update the reactor vessels materials data bases for Surry Units 1 and 2 were found to be in I compliance with 10CFR50, Appendix G. The surveillance program materials properties data bases are in compliance with 10CFR50, Appendix H and will provide the material data necessary to I ensure continued licensibility of the reactor vessels.

A new reactor vessel surveillance capsule withdrawal schedule was developed to meet the requirements of ASTM E185-82 as referenced by 10CFR50, Appendix H. This new schedule will provide needed irradiation materials data in a timely manner.

The reason for this review and update is that the reactor pressure vessels were fabricated and the corresponding surveil-lance programs were developed prior to the implementation of 10CFR50, Appendixes G and H. These regulations recognized that the older plants could not meet all the requirements and established guidelines to meet the intent, if not the letter of the regulations. In addition, these regulations have been revised as experience, new data and analysis capability related to reactor vessel integrity have developed. A periodic review and update is necessary to ensure continued compliance with the regulations.

1-1 Babcock & Wilcox a McDermott company

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2. REACTOR VESSEL DATA BASES The establishment of the mechanical and toughness properties I of reactor pressure vessels in accordance with applicable regulations and standards is an essential aspect of the licens-ing process. As these rules are improved it is necessary to ensure that the data used for licensing of the reactor vessels are representative of the best information and materials properties available for each specific reactor vessel. The data are also essential in establishing the normal pressure-tempera-ture operating limitations as required by 10CFR50, Appendix G.

2.1. Surry Unit - 1 The materials and chemical composition data for the Surry Unit 1 reactor vessel are presented in Tables 2-1 through 2-6. These data represent an ac'cumulation of information from various I sources (References 4, 10 and 11) which include the improved chemical composition data for the Linde 80 submerged arc weld metals as reported in BAW-1799.5 In addition, the initial reference temperature data represents the best available data as defined in Section 2.4.

The location and identification of the plates and welds within the belt-line region of the Surry Unit 1 reacter pressure vessel are shown in sigure 2-1.

2.2. Surry Unit 2 I The data for Surry Unit 2 reactor vessel are presented in Tables 2-8 through 2-13. These data represent an accumulation I of information from various sources (References 4, 12 and 13 which include the improved chemical composition data for the Linde 80 submerged arc weld metals as reported in BAW-1799.5 In addition, the initial reference temperature data represents the best available data as defined in Section 2.4.

2-1 I Babcock & Wilcox a McDermott company

I The location and identification of the plates and welds within the belt-line region of the Surry Unit 2 reactor pressure vessel are shown in Figure 2-2.

2.3. Surveillance Data Bases Each reactor vessel has a surveillance program to monitor the neutron radiation damage of the materials in the beltline region. These data for each reactor vessel at Surry were tabulated separately from the main data base as a convenience for easy reference. The data are presented in Tables 2-7 and 2-14.

2.4. Initial Value of Reference Temperature The initial value of reference temperature is not always available for the materials used to fabricate older reactor vessels because it was not an established requirement of the ASME Code. Even for reactor vessels completed after the establishment of the requirement, the value was often unat-tainable because no suitable material was available. The necessary drop weight test data were usually obtained for both plate and forging materials and this provided a reliable data base to establish the initial reference temperature for these materials.

The initial reference temperature of weld metals was not obtained until after it was required by the ASME Code. At that time an effort was made to re-evaluate the weld qualifica-tion to obtain initial reference temperatures. Subsequently, a statistical evaluation was made of the Linde 80 weld metals fabricated after the establishment of the ASME Code require-ments, concurrent with a re-evaluation of old weld metals, to provide a basis for a mean value and standard deviation. A similar approach was used to establish values for plate and forging materials for which the needed actual test data were not available. An acceptable method for establishing the initial reference temperature is presented in the Imc Standard Review Plan, Section 5.3.2.6 It is recognized that the values recom-mended in the Standard Review Plan are very conservative.

2-2 Babcock &Wilcox a McDermott company

I Previous evaluations of the Surry reactor vessels had estab-lished reference temperatures for the base materials which were representative of actual data. However, the weld metals had no data for the Linde 80 submerged-arc weldments made by B&W and I only the initial value of Rotterdam welds established per the NRC Standard Review Plan, Section 5.3.2.

A more recent development is the need to establish the standard deviation of the reactor vessel materials to be used in deter-mining the reference temperature shift as a result of irradia-I tion. Standard deviations for measured initial reference temperature have been established from data obtained since the ASME Code requirement for establishing the reference temperature of all reactor vessel materials.

Listed below are the initial reference temperatures and standard deviations used if actual measured values are not available:

Material Initial RTNDT Description RTNDT,F Std. Dev. Reference SA508,C1. 2 3 130F BAW-10046P7 SA533,Gr.B1 1 26F BAW-10046P Linde 80 Welds -6 19F BAW-18038 RDM Welds 0 120F SECY 82-4659 Another recent development is the need to establish unirradi-ated, or initial, Charpy upper-shelf energy (USE) values for the high-copper Linde 80 submerged-arc weldments made by B&W for which no data are available. A statistical evaluation was made of the upper-shelf energy data for the Linde 83 weld metals fabricated after the establishment of the ASME Code require-ments; this included a re-evaluation of the limited data from the old weld metals, to provide a basis for a mean value and standard deviation.

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3 2-3 I Babcock & Wilcox J McDermott company

I Listed below are the initial Charpy upper-shelf energy value and standard deviations to be used if actual measured values are not available:

Material Initial USE Std.

Description USE, ft-lb. Dev.. ft-lbs. Reference Linde 80 Welds 70 16 BAW-18038 I

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2-4 Babcock & Wilcox J McDermott company

_______________________ __________ _ - . . . . ~

M M M M M M M M M M WWM M~ H M M M M}

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Table 2-1. Identification of Reactor Vessel Beltline Region Weld Metals - Surry Unit-1 Weld Weldirg Weld Wire Flux Identification Weld location Process Type Heat No. Type Int No. Reference l J726* Nozzle Shell to Interm. Sub. Arc SMIT 40 25017 SAF 89 1197 DocPat No.

Shell Circle Seam 50-280(10)

SA1494 Interm. Shell Inngi- Sub. Arc Mn-Mo-Ni 8T1554 Linde 80 8579 tudinal Seams Ia & IA  ;

1 SA1585 Interm, to Iower Shell Sub. Arc Mn-Mo-Ni 72445 Linde 80 8597 Circle Seam (I.D. 40%)

Y SA1560 Interm. to Iower Shell Sub. Arc Mn-Mo-Ni 72445 Lirde 80 8632

• Circle Seam (O.D. 60%)

SA1494 Iower Shell Sub. Arc Mn-Mo-Ni 8T1554 Linde 80 8579 Iorgitudinal Seam L1 SA1526# Iower Shell Sub. Arc Mn-Mo-Ni 299IA4 Lirrie 80 8596 Iongitudinal Seam L2

., I * - Weld made by De Rotterrlawrhe Droogdok Mu (All other welds made by B&W)

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1. - Denotes material included in Reactor Vessel Surveillance Propurs fk

!3r iD iE

v. =

4$M

Table 2-2. 01emical Omnosition of P_cactor Vessel Beltline Recion Weld Metals - Surry Unit-l Weld Olemical Canoosition. Weicdit Perant Identification C Mn P S Si Cr Ni Mo Cu . Reference J726 0.093 1.67 - -

0.27 -

[0.10) 0.44 0.33 Docket No.

50-280(10)

SA1494 0.09 1.52 0.015 0.012 0.44 0.08 0.63 0.37 0.18 BAW-1799 (5)

SA1585 0.08 1.45 0.016 0.016 0.51 0.09 0.59 0.38 0.21 SA1560 0.08 1.45 0.016 0.016 0.51 0.09 0.59 0.38 0.21 0.35 "

SA1526# 0.09 1.53 0.013 0.017 0.53 0.09 0.68 0.42 to E

1. - Denotes material included in Reactor Vessel Surveillance Program

[ ] - Estimate based on review of similar material W

= es Ek

?R g=

UC 8I t=

A 0

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W W W M M M M m M M M M M M M m m ga e

& M D ~~~ M M~M M~~~M ' M D f ~M M Table 2-3. Rhmical Pruerties of Pea _ctor Vessel Deltline P_aligl_}MLd Metal - Surry Unit 1

'Ittrihness Pmperties Tensile Pmnerties Hold 10F Ehergy USE __Sttr h Elong PA Ibst Weld lleat Treatmern.** Reference Identification g T Et-Ibs RPg ,F Ft-lbs yield Ult.  %  %

J726 -

54,77,51 0* -

67.40 82.76 31.0 69.6 S-11300 F - 30 IR-fC Docket No.

50-280(10) 0 "

SA1494 -

54,25,44 -6* 70* -

81.00 - -

S-ll25 i 25 F-4819 SC.

0 "

81.00 S-1125125 F-80ER-FU SA1585 -

50,54,51 -6* 70* - - -

0 "

SA1560 -

46,43,45 -6* 70* 65.00 81.00 30.5 -

S-1125 i 25 F-481R-EU SA15269 -

33,33,33 -6* 70* - 88.00 - -

S-11250 i 25F-481R-FU to 1 * - Estimatal per Secticn 2.4

• • - S = Stress Relievo@RtIna Cooled (EC)

1. - Denotes muterial incitded in Reactor Vessel Surveillance PruJram W

= ts

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!=

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=

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Table 2-4. Identification of Reactor Vessel Beltline portion Base Materials - Surry Unit-1 Material Identification Heat No. Type Ocuponent Sumlier Heat Tmatment* Reference 122V109 SA508 CL.2 Nozzle Shell Forgirg Bethlehem A-15500 F-llHRAQ Do&nt No.

T-1200 0F-22HR/AC 50-280(10) 0 S-ll25 F-40HR/EC Inkens 0 "

C4326-l# SA533,Gr.B1 Inter. Shell Plate A-1675t25 F-9HR/WQ T-12104-9HR/AC 0

S-ll25125 F-60HR/FU C4326-2 SA533,Gr.B1 Inter. Shell Plate Inkens A-1675i254-9HR/WQ T-1210 0F-9HR/AC w S-11251254-60HR/FU 0 "

C4415-1# SA533,Gr.Bl IoWou. Shell Plate Inkens A-1675i25 F-9HR/WQ T-1200-1225 0F-9HR/AC 0

S-1125125 F-60HR/EC C4415-2 SA533,Gr.B1 Iamr Shell Plate Inkens 0 "

A-1675f25 F-9HR/WQ T-1200 0F-9HR/AC S-1125i254-60HR/FU

! # - Denotes material included in Reactor Vessel Surveillance Program t D'

• - A = Austentized/ Water Querxted (WQ), Brine Quenched (BQ) f7

,5 m k. T = Temperal/ Air Cooled (AC), Brine Quenched (BQ)

D S = Stress Relieved / Furnace Cooled (FC)

,$ -I.

• O M

l M M M M M M W W M M M M W W M M M M M

m MTM W mR_ TVWT%f 1 F L J L_J l F7 F L_JV1 D ;

Table 2-5. memical Otmosition of Reactor Vessel Beltline RotTion Base Phterials - Surry Unit-1 Material memical Otm osition. Weictht Percent Identification C Mn P S _Hi_ _lli_ Cr Ho 00 V Ct1 Referery;g 122V109 0.22 0.70 0.010 0.011 0.24 0.74 0.36 0.60 0.010 0.01 [0.09] Docket No.

50-280(10)

C4326-lf 0.23 1.35 0.008 0.015 0.23 0.55 0.069 0.55 0.014 0.001 0.11 C4326-2 0.23 1.35 0.008 0.015 0.23 0.55 0.069 0.55 0.014 0.001 0.11 C4415-1# 0.22 1.33 0.014 0.014 0.23 0.50 0.078 0.55 0.015 0.001 0.11 C4415-2 0.22 1.33 0.014 0.014 0.23 0.50 0.078 0.55 0.015 0.001 0.11 Y

e

1. - Denotes material incltried in Reactor Vessel Surveillance Prapam

[ ] - Estimate based on review of similar material R

5R

?R

!=

=>

$E iif x o M

Table 2-6. Mechanical Properties of Reactor Vessel Beltline Recion Base Materials - Surry Unit 1 Touchness Properties Tensile Properties. RT Material USE, Strenath. Ksi Elong RA Identification TNDT, TNDT,F Ft-Lbs Yield Ult.  %  % Reference 122V109 40 40* 82.5* 76.5 96.5 23.25 65.45 Docket No.

50-280(10)

C4326-lf 10 10* 87.5* 67.5 88.1 26.95 67.70 C4326-2 0 0* 94* 67.6 88.1 26.95 67.70 w C4415-18 20 20* 82* 72.0 94.6 25.00 65.90 C4415-2 0 0* 86.5* 69.5 91.8 25.00 64.10

• - Estimated Per NRC Standard Review Plan Section 5.3.2.

1. - Denotes material included in Reactor Vessel Surveillance Program W

=b N$

ER,.

g

D

$ _E.

M M

M M M m M m M M m mm m m m m m m

M M M M M M M M M M m mmm M m M m m J Mle 2-7. Ptroerties of Surveillarrn ProTram Plate _ ard Weld Material - Surry Unit-1 i l

Tomhness Properties Tensile Properties, RP Material USE, StremL t. Fsi ElorYJ RA Post Weld Identification Yield Ult.  %  % Ileat Treatment ** Reference Tg ,F RPg ,F Ft-Iba 10 115 68.00 90.47 25.75 70.85 A-1650-1700F-91E/WQ WCAP-7723(11)

C4326-1 10 T-1210F-91B/AC S-ll25F-15-1/21R-K C4415-1 20 20 103 71.80 93.77 24.45 69.80 A-1650-1700F-91!IMQ T-1200F-91R/AC S-1125F-15-1/21R-N SA1526 - -6* 70 69.67 83.20 26.50 66.70 S-ll25F-15-1/21R-K f 1

}hterial Chemical Omoosition. Helaht Percent Identification C _Ltr1_ P S _Si_ _Cr_ _El_ Pb. O) V Cu Reference u

I C4326-1 0.23 1.35 0.008 0.015 0.23 0.069 0.55 0.55 0.014 0.001 0.11 WCAP-7723(ll)

[

" I C4415-1 0.22 1.33 0.014 0.014 0.23 0.078 0.50 0.55 0.015 0.001 0.11 SA1526 0.10 1.49 0.011 0.010 0.37 0.08 0.68 0.46 0.001 0.001 0.25

• - Estimated per Section 2.4

• • - A = Austentized/ Water Quenchal (HQ)

T = Tenpercd/ Air Qx> led (AC)

,g S = Stress Relieved /R1rnace Cboled (K) 5R

?R

!r D

85 2=

28M

~ . _ _ _ _ _ _

Table 2-8. Identification of Peactor Vemol Beltline Reaion Weld Metals - Suny Unit-2 Weld Welding Weld Wire Flux Identifica~ica c Weld Incation Process TvDe Heat No. TvDe Iot No. Reference L737 Nozzle Shell to Interm. Sub. Arc S4Mo 4275 SAF 89 02275 DocPat No.

Shell Circle Seam 50-281(12)

SA1585 Interm. Shell Lorgi- Sub. Arc Mn-Mo-Ni 72445 Linie 80 8579 "

tudinal Seams L3 & IA R 3008*# Interm. to Iower Shell Sub. Arc S3Mo 0227 Grau In IH320 Circle Seam WF 4 Iower Shell Iagitudinal Sub. Arc Mn-Mo-Ni 8T1762 Ilnde 80 8597 "

N Seam Il (I.D. 63%)

e Seam I2 (100%)

w WF 8 lower Shell Ingitudinal Sub. Arc Mn h ii 8T1762 Lirde 80 8632 "

Seam L1 (O.D. 37%)

• - Weld made by De Rotterdamsche Drcogdok Mu (All other welds made by B&W)

1. - Denotes material included in Reactor Vessel Surveillance Prupara W

n. t

?R

!=

5D

,i E. -

f%

'o M

M M M M M M M M m M M W m m m m m m

)

Table 2-9. Gemical Otrzmition of Reactor Vessel Beltline Pmion Weld Metals - Surry Unit-2 Weld memical Camosition. Weicht Percent Identification C _Hn_ P S Si Cr Ni Mo Ou Reference L737 0.084 1.74 - -

0.35 -

[0.10] 0.38 [0.35] Docket No.

50-281(12)

SA1585 0.08 1.45 0.016 0.016 0.51 0.09 0.59 0.38 0.21 BAW-1799(5)

R3008# 0.09 1.51 0.017 0.016 0.46 0.10 0.56 0.41 0.19 Docket No.

l 50-281(12) hT 4 0.07 1.45 0.015 0.013 0.43 0.12 0.55 0.41 0.29 BAW-1799(5) 0.29 "

u h7 8 0.06 1.45 0.010 0.009 0.53 0.12 0.55 0.41 b

1. - Denotes material included in Reactor Vessel Surveillance Prugsm

( ) - Estimate based on review of similar material D

N

?R

!=

=>

!E=

EOM

Table 2-10. Mechanical Properties of Reactor Vessel Deltline Rrgion Weld Mqtal - Surry Unit _2 Totxtness Prmerties Tensilg_Etroerties. RT mterial 10F Enertjy USE Stremth. Esi Elory] RA Post Weld Yield Ult.  %  % Ileat Treatment ** Reference Identificatim Tg ,F ft-lbs RTg ,P ft-Ibs 67.6 0 Dociwt No L737 - 75,62,65.5 0* - 66.69 82.62 29.0 S-ll30 F - 241HVEC 50-281(12)

SA1585 - 50,54,51 -6* 70* -

81.00 - -

S-1125 i 25 0F-8GiR/EC R30086 -

65.5,50.5,46 0* - 79.00 89.90 26.0 -

S-ll30 0F - 25ffR/fC "

0 "

WF 4 - 40,31,34 -6* 70* 65.06 82.25 25.0 64.9 S-ll25 1 25 F-8QIR/EC 0 "

WF 8 - 45,38,30 -6* 70* 71.00 85.50 25.0 -

S-ll25 1 25 F-4811R/EC to I

[ * - Estimated per Section 2.4

• • - S = Strtsa Relieved /fbrmco Cboled (EC)

1. - Derntes material incitx5ed in Reactor Vessel Surveillance PruJram W

k EF 60 2R 8

5D iE

=

48M M M M M M M M M M M M m m m m m

l l

l l

Table 2-11. Identification of Reactor Vessel Beltlim Rection Base Mewrials - Surry Unit-2 l l

Material Identification Heat No. Type Ocznponent Supplier Heat Treatment

• Peference 123V303 SA508,CL.2 Nozzle Shell Forging Bethlehem A-1550%-12HR/WQ Docket No.

T-1200%-22HR/AC 50-281(12)

S-ll254-40HR/K C4208-2 SA533,Gr.B1 Inter. Shell Plate Inkens A-1600-1650%-9HR/BQ l T-1200-12254-9HR/BQ 0

S-1125 F-60HR/ M C4339-l# SA533,Gr.B1 Inter. Shell Plate Inkens A-1600-16504-9HR/D2 T-1200-1225%-9HR/BQ 0

w S-ll25 F-60HR/ N i

Inkens 0 "

5 C4331-1 SA533,Gr.Bl IcWer Shell Plate A-1600-1650 F-9HR/D2 T-1200-1225 0F-9HR/BQ 0

S-ll25 F-60HR/ M 0 "

C4339-2 SA533,Gr.Bl Iower Shell Plate Inkeru A-1600-1650 F-9HR/BQ 0

T-1200-1225 F-9HR/BQ 0

S-1125 F-60HR/FC k # - Denotes material included in Reactor Vessel Surveillance Program sR yg * - A = Austentized/ Water Quenched (WQ), Brine Quenched (BQ)

!y T = 'Ibmpered/ Air Cboled (AC), Prine Quenched (BQ) yC3 S = Stress Relieved /R1rnace Cooled (K)

$I 2=

A8M

)

Table 2-12. Gemical Ocznnosition of Reactor Vessel Beltline Rection Base Materials - Surry Unit-a Material Gemical Camposition. Weicht Percent Identification C Mn P S Si Ni Cr Ho Oo V Cu Reference 123V303 0.20 0.63 0.010 0.010 0.24 0.73 0.36 0.58 0.009 0.02 [0.09] Docket No.

50-281(12) 0.15 "

C4208-2 0.21 1.28 0.008 0.013 0.24 0.55 -

0.55 0.020 -

0.11 "

C4339-1# 0.23 1.30 0.012 0.014 0.25 0.54 -

0.54 0.010 -

0.12 "

C4331-2 0.23 1.42 0.009 0.015 0.22 0.60 -

0.55 0.015 -

C4339-2 0.23 1.30 0.012 0.014 0.25 0.54 -

0.54 0.010 -

0.11 w

I 5

1. - Denotes material included in Reactor Vessel Surveillapoe Fregam

[ ] - Estimate based on review of similar material a

6 e 5

?R

!=

5D

,3 -I.

• <)

M M M M W W W W W M M M M M M M m m M

m M M M M M M M M M M M M M M M M M M Table 2-13. Mechanical Properties of Reactor Vessel Beltline Rerrion Base Materials - Surry Unit 2 W a Properties Tensile Properties. RP Material USE,* Stremth. Ksi Elong RA Identification Tg ,F

• RTyg ,F Ft-Ibs Yield Ult.  %  % Refererre 123V303 30 30 103 66.37 87.12 25.00 70.20 Docket No.

50-281(12) 67.35 "

C4208-2 -30 -30 93.5 64.00 85.87 26.55 30 67.00 64.70 "

C4339-l# -10 79.0 90.25 26.95 y C4331-2 -10 10 84.0 68.75 92.25 25.80 65.25 "

C4339-2 -20 10 82.5 68.50 92.12 26.60 66.10 "

• - Estimated frun data in the major workinJ direction per IEC Standard Review Plan Section 5.3.2.

1. - Denotes material inclt&d in Reactor Vessel Surveillance Program IB as 5

?R

!=

5D

$I=

a8M

Table 2-14. Properties of Surveillance Program Plate and Weld Material - Surry Unit-2 a

'Itutness Pmoerties _ Tensile Prtoerties RT Material 12, Stremth FaL ElcavJ RA Pcat Wald Identificaticn Tg ,F Hrg ,F Ft-Lbs Yield Ult.  %  % Heat Treatment

• Referince C4339-1 -10 11. ~s 104 68.15 91.3 26.35 69.55 A-1600-1650F-915VBQ WGP-4085(13)

T-1200-1225F-915VBQ S-1140F-15 1/415VFC

! R3008 0 0 91 70.85 86.50 26.40 67.80 S-1140F-15 1/416VIC "

l Wald es=ical Otrocaitim. Weicht narcent Y M ification c _)tl. P l _S1. A Ni _It2. _QL. Rafarunas C4339-1 0.23 1.30 0.012 0.014 0.25 0.075 0.54 0.54 0.011 NCAP-8085(13)

R3008 0.09 1.51 0.017 0.016 0.46 0.10 0.56 0.41 0.019 "

su I

y * - A = AustentitetVBrino Quendwi (BQ)

T = 'huperwyBrina Quenched (BQ)

S = h Relieved /nzrnmoe cooled (EU)

IB t

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I I Figure 2-1. Location and Identification of Materials Used in the Fabrication of the Belt-Line Region of Surry Unit-1 Reactor Pressure Vessel (Ref. 14)

I CIRC'M[RfMI AL ${ AM$ V{RTICA( $[AMS, p 210*

E Y WO6 L4 9.0"* C4326-1 h

7 45 CDRE Ce=t g _ ,,,.-

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I 2-19 a McDermott company

I Figure 2-2. Location and Identification of Materials Used in the Fabrication of the Belt-Line Region of Surry Unit-2 Reactor Pressure

! Vessel (Ref. 14) l I

l l

C1RetW [n[WTIAL $[AMS VERTICAL $[4MS 270*

e C WO6 (4 C4339-1 9.0 45 Coat CORI 130 C' l

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2-20 Babcock & Wilcox J McDermott compary

3. EVALUATION OF REACTOR VESSEL TOUGHNESS One aspect of reactor pressure vessel licensibility is the l toughness of the materials used in its fabrication. These properties are used to calculate the pressure-temperature l operating limits in accordance with the requirements of 10CFR50, Appendix G. The objective of these limits is to prevent nonductile failure of the reactor vessel during any normal l

operating condition, including anticipated operational occur-rences and system hydrostatic tests.

I The closura head region, the reactor vessel outlet nozzle, l

and the beltline region have been identified as the only g regions of the reactor vessel that regulate the pressure-W temperature limits. Since the closure head region is signifi-cantly stressed at relatively low temperatures (due to mechani-cal loads resulting from bolt preload), this region largely

! controls the pressure-temperature limits of the first several service periods. The reactor vessel outlet nozzle also affects

( the pressure-temperature limit curves of the first several service periods. This is due to the high local stresses at the j inside corner of the nozzle, which can be two to three times the l

membrane stresses of the shell. After the first severa) years of neutron radiation exposure, the RTNDT of the beltline region materials will be high enough that the beltline region of the I reactor vessel will start to control the pressure-temperature limits of the reactor coolant pressure boundary (RCPB) . For the service period for which the limit curves are established, the maximum allowable pressure as a function of fluid temperature is obtained through the point-by-point comparison of the limits imposed by the closure head region, the outlet nozzle, and the beltline region. The maximum allowable pressure is taken to be the lowest of the three calculated pressures.

, 3-1 Babcock & Wilcox a M(Dermott Comparty b

I The unirradiated toughness properties of each reactor pressure vessel were determined for the belt-line region materials in accordance with 10CFR50, Appendix G. For the other belt-line g region materials for which the measured properties are not 5 available, the unirradiated impact properties and residual elements, as originally established for the belt-line region materials, were determined from acceptable data bases using recognized estimating techniques. The adjusted reference temperatures are calculated by adding the predicted radiation-induced shift of the RTNDT to the unirradiated RTNDT including margin. The predicted shift, RTNDI, is calculated using the respective neutron fluence, plus the copper and nickel c on--

tents. The neutron' fluence values are calculated based on functions derived from adjoint functions using transport calculations. The design curves of Regulatory Guide 1.99, Rev.

2,15 were used to predict the radiation-induced shift of RTNDT*

The results of Lne evaluation are presented in Tables 3-1 and 3-2 which show that both reactor pressure vessels have RTNDT values which will permit normal operation to the expira-tion of current licenses.

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M M M M M M M M M M M M M M M M M M M Table _hl._LvalmtimKFeactor_hrmtw3esuel TractwvJts #m4 _$vrrY3nitrl a

RrNor t w riuerum at narrerg 13cwww Expiratirm(c) 1.lonree Expiration Beltline Quical_Qvemitiq[*I Initial Inside Surface T/4 Ioraticn Irmide Surface T/4 Incaticn filterl41_IOCutifiGRL10D Heat No. Type Ra}ics) Incatica Otner w/o Nickel w/ Sm *I "''" d"I ** *c=' ** *'='

SA508,C1.2 Nozzle Shell Fort;ir:3 0.09 0.74 40 34 5.52E18 3.06E18 122 116 122V109 SA533,Gr.B1 Intatin. Shell Plata 0.11 0.55 10 34 1.54E19 1.97E19 142 130 C4326-1 SA533,Gr.B1 0.11 0.55 0 34 3.54E19 1.97E19 132 120 C4326-2 Intaru. Shell Plata SA533,Gr.81 1sk a Shell Plate 0.11 0.50 20 34 3.54E19 1.97E19 151 139 C4415-1 SA533,Gr.B1 0.11 0.50 0 34 3.54E19 1.97E19 131 119 C4415-2 Imer Shell Plata J726 Stie. Art Nm. to Int. Cir. Wald 0.33 0.10 0 69 5.52E18 3.06E18 196 180 Int. to Icu. Cir. Wald 0.21 0.59 -6 68 3.54E19 1.97E19 279 252 SA1585 Stan. Arc Int. lau;it. Wald 0.18 0,63 -6 68 5.71I18 3.17E18 196 180 SA1494 Situ. Arc truer Irrqit. Wald 0.18 0.63 -6 68 5.71E18 3.17E18 196 180 SA1494 Suta. Arc 68 5.71E18 3.17E18 251 225 SA1526 Sim. Arc icuer lauJit. Wald 0.35 0.68 -6 taZE: Wald setal SA1560 is not listed ter=== it is time outside 60% of the inte&,dicts to 1cuar shall citumferential weld.

W I*I'4r Tables 2-1 jb y 2-6 14r nCAP-11015( b IU Per 5hagulatory aside 1.99, Dev. 2 IIN m -

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I I l I 4. REACTOR VESSEL SURVEILLANCE PROGRAMS l

The design of a reactor vessel materials surveillance program  !

is based on the need to monitor the toughness properties of the controlling radiation sensitive material from which the reactor vessel was fabricated. Of equal importance is the benchmarking, or verification, of the-fluence which the reactor vessel experiences.

I .

The extent to which a surveillance program meets these objec-I tives depends on when the reactor vessel was fabricated. This is due to the evolution of surveillance program requirements as more knowledge has been obtained from existing programs. Some of the requirements can be upgraded to meet the current 10CFR50, Appendix H while other reactor vessels will be required to make do with the installed programs. Each of the Surry surveillance programs will be described separately.

4.1. Surry Unit 1 The surveillance program was designed prior to the date when I 10CFR50, Appendix H, established surveillance program require-The fact that the controlling materials were contained ments.

in the program is because the designers were knowledgeable as to the developing requirements. Even though the weld metal is not controlling by current standards, this fact is due to the azimuthal fluence distribution and not by the chemical composi-tion of the weld metal which, if fluence values were equal, would be controlling. The surveillance weld metal was fabri-cated using the same weld wire as the controlling weldment; I therefore, it will monitor the relative neutron radiation sensitivity of the controlling weldment.

The major deficienty of the program is the withdrawal schedule which is not designed to provide timely data as required by the current 10CFR50, Appendix H. A new capsule withdrawal 4-1 Babcock & Wilcox a utoc<most company

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schedule was developed around the current requirements as defined in ASTM E185-82 and the desire to move capsules from low lead factor sites to high lead factor sites only during sched-uled ten year reactor vessel inspections.

Two methods are permitted to determine when a capsule is to be removed for evaluation; i.e., EFPY exposure or cumulative fluence. The cumulative fluence was used to establish the new schedule which will match the materials data obtained at each capsule evaluation to the critical times in the reactor vessel design life. Tha EFPY schedule would produce results with too low cumulative fluence to provide useful irradiation materials data. The new proposed withdrawal schedule for Surry Unit 1 is shown in Table 4-1.

4.2. Surry Unit 2 The surveillance program was designed prior to the date when 10CFR50, Appendix H established surveillance program require-ments. This program is similar to the program for Surry Unit 1.

The controlling materials are contained in the program. The program is updated by establishing a new withdrawal schedule based on accumulative fluence in accordanca with ASTM E185-82.

The new proposed withdrawal schedule for Surry Unit 2 is shown in Table 4-2.

4.3. Soare Capside_ss Extra Surveillance Capsules not required to meet the current requirements of ASTM E185-82 will remain in the reactor vessel.

These capsules can be used in the future to provide data for the verification of reactor vessel fluence calculations or to provide materials data for support of plant life extension. To ensure that the extra capsules will provide useful data in the future they will be moved to maximum lead factor positions during the appropriate inservice inspection as the positions become available, in order to maximize the total accumulated fluence.

I 4-2 Babcock & Wilcox a McDermott company

M N- M M M M, M M M H H M M M M q mM Table 4-1. Revised Surveillancx? Capsule Withdrawal Schedule - Surry Unit-1 l

1 February 1986 R_ - nded Capsule Withdrawal Sdwhile IMr )

10CFR50. Armendix H. and E185-82 Revised Withdrawal Sctwhile*

Capsule Vessel Npml e myile Estimated capsule I14I hted Wv4==I14) 2 Estimated Sequence EFPI Fluence, ycm2 I.D. EFPi fluence, ycm2 RV fluence, rg/cm Withdrawal,Yr First 1.5 SE18 or T** 1.1 2.89E18 1.70E18 -

RP g 50F Seccni 3 1.00E19 W** 3.4 4.31E18 5.49E18 -

V 1.94E19 1.21E19 1986

'Ihird 6 1.97E19 8.0 (DOL T/4) 21.2 3.63E19 2.95E19 2004 Fourth 15 3.54E19 X***

(EDL I.W.)

>3.54E19 Z*** 25.6 4.57E19 3.54E19 2008 Fifth EOL NCTIE: Remainder of capsules moved to mayinn lead factor positiCU durity inservice ingw+1 cms as postions t+>-+ available in order to navinize total fluence.

W

• - Reviewed after eadt fuel cycle and revised as W after each capsule withdrawal and evaluation.

ga Estimated withdrawals based on 18-month fuel cycles and 0.80 plant apacity.

t2 ** - Capsules withdrawn and evaluated.

• • • - Transferred fran 1.1 lead factor location to 1.8 lead factor location during the 20 year inservice

!R jD inspection ( M r 1992/12.53 EFPf).

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• O M

Table 4-2. ReviM Surveillance Orwulle Withdrawal Schedule - Surry Unit-2 1 February 1986 pay _ ..c-rded Capsule Withdrawal Sdudule per 10CFR50. Icoendix H. aM E185-82 Pavised Withdrawal Sdiedule*

Capsule Vessel Capsule Capsule Estimated Capsule I14I hted Wvi==I14) Estimated Sequence EFPi Fluence, rVcm 2 I.D. EFPi fluence,rVcm 2 RV fluence, IVcm 2 Withdrawal,Yr First 1.5 SE18 or X** 1.1 3.01E18 1.84E18 -

RP g 50F SecoM 3 9E18 Wi 8.5 1.30E19 1.21E19 1986

'lhird 6 1.7E19 V 8.5 2.02E19 1.21E19 1986 (BOL T/4) 1 Fourth 15 3.0E19 Y*** 19.7 3.19E19 2.94E19 2001 (BOL I.W.)

Fifth EDL >3.0E19 U*** 25.8 4.38E19 3.04E19 2008 HOIE: Resnalnder of capsules moved to =vi== lead factor positical during inservice inspecticals as positions hamna available in order to = vimize total fluence.

qs * - Reviewed after esd1 fuel cycle and revised as rvwini after each capsule withdrawal and evaluation.

• g Estimated withdrawals based on 18-month fuel cycles and 0.80 plant capacity.

• • - Capsules withdrawn and evaluated.

ga 2R *** - Transferred frun 1.1 lead factor locatim to 1.8 lead factor locatim during the 20 year inservice

$2

~D inspection (April /1993/12.96 EFPi).

1. - Pa>- orded withdrawal of Capsule "V" and leavirg Capsule "W" in reactor since Capsule "V" will

,$ l lead reactor vessel mvi== wall fluence and Capsule "W" will only be equivalent to reactor vessel

!; g =vi== vall fluen . Capsule "W" to be moved to a maxinum lead factor position as it W

• O available.

M E N O O' E O M E W W W: W W M M M

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5. INTEGRATED REACTOR VESSEL SURVEILLANCE PROGRAM The idea of an integrated reactor vessel surveillance program develops whenever two or more nuclear plants share the same site, or when owned by the same utility and share a common design. It is readily apparent that a savings can be recog-nized from reduced capsule evaluations and, of special impor-tance, reduced worker radiation exposure. The requirements set I forth in 10CFR50, Appendix H, as applicable to integrated surveillance programs are as follows:

"C. An integrated surveillance program may be considered for a set of reactors that have similar I design and operating features. The representative materials chosen for surveillance from each reactor in the set may be irradiated in one or more of the reactors, but there must be an adequate I dosimetry program for each reactor. No reduction in the requirements for number of materials to be irradiated, specimen types, or number of specimens I per reactor is permitted, but the amount of testing may be reduced if the initial results agree with predictions. Integrated surveillance program must I be approved by the Director, Office of Nuclear Reactor Regulation, on a case-by-case basis.

Criteria for approval include the following consid-erations:

1. The design and operating features of the reactor in the set must be sufficiently similar I to permit accurate comparisons of the predicted amount of radiation damage as a function of total power output.

2. There must be adequate arrangement for data sharing between plants.

3. There must be a contingency plan to assure that the surveillance program for each reactor will not be jeopardized by operation at reduced I power level or by an extended outage of another reactor from which data are expected.

5-1 Babcock & Wilcox I a McDermott company

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4. There must be a substantial advantage to be gained, such as reduced power outages or reduced personnel exposure to radiation, as a result of not requiring surveillance capsules in all reactors in the set."

The four requirements as set forth above can be met by the Virginia Power plants. The plants are similar in design and power ratings, especially as to current knowledge as to flux g rate effects on neutron radiation damage. Since the plants have 5 the same owner no problem would exist as to data sharing. Each plant could continue to have a plant specific surveillance program as a backup in case the other plants experienced a protracted shutdown. Finally, there would be a gain from the reduced personnel exposed to radiation.

The fault of this approach is the wording in the first para-graph, " ... must be an adequate dosimetry program for each reactor". In recent years, the monitoring of the neutron fluence the reactor vessel receives has developed to be of equal importance with monitoring material damage. In fact, because of the large uncertainties that can be assigned to fluence analysis, if not properly verified (i.e., dosimeters in removed surveillance capsules) , when combined with materials property changes can produce restricted pressure-temperature operations which could more than offset savings to be realized from a surveillance capsule evaluation. On the other hand, if a capsule is removed to benchmark a fluence determination a major portion of the cost is associated with the dosimeter and fluence evaluation. Theretore, it would be more practical to perform the complete capsule evaluation.

In addition, the new revision of Regulatory Guide 1.99, gives credit for obtaining surveillance data for the controlling materials. When two or more credible data points become available from a reactor they may be used to determine the adjusted reference temperature and decrease in Charpy upper shelf energy. With the exception of Surry Unit 1, all the plants have the controlling materials in their surveillance programs. Thus, the data should provide the best evaluation of 5-2 Babcock & Wilcox a McDermott company

I material damage to minimize the effect on operating limita-tions. In the case of Surry Unit 1, the weld metal in the surveillance program is similar to the controlling weld metal I and may be used to provide a high degree of confidence that prediction techniques are not unduly restricting the operating limits.

In summary, while an integrated reactor vessel surveillance program for Surry Units 1 and 2 may be acceptable from a regulatory viewpoint, it would not be practical, since capsules would have to be withdrawn from each unit in order to provide a fluence benchmark. However, this option may become practical in the future and should be re-evaluated after additional capsules have been removed and evaluated.

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6.

SUMMARY

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As a result of this review and update the reactor vessels materials data bases for Surry Units 1 and 2 were found to be in compliance with 10CFR50, Appendix G. The surveillance program materials properties data bases and capsule withdrawal schedules are in compliance with 10CFR50, Appendices G and H and will l

provide the material data necessary to ensure continued compli- l l ance with these appendices.

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7. REFERENCES

1. U.S. Code of Federal Reculations. Title 10. Enerav. Part M, " Domestic Licensing of Production and Utilization Facilities, Appendix G, Facture Toughness Requirements."

2. U.S. Code of Federal Reculations. Title 10. Enerav. Part M, " Domestic Licensing of Production and Utilization Facilities, Appendix H, Reactor Vessel Material Surveil-lance Program Requirements."

3. ASTM Standard E185-82, " Practice for Conducting Surveil-lance Tests for Light Water-Cooled Nuclear Power Reactor Vessels," ASTM Standards 03.01, August 1985.

4. Surry Power Station, Units 1 and 2, Updated Final Safety Analysis Report, Virginia Electric and Power Company, July 16, 1982, as amended.

5. K. E. Moore and A. S. Heller, "B&W 177-FA Reactor Vessel Beltline Weld Chemistry Study," BAW-1799, Babcock &

Wilcox, Lynchburg, Virginia,' July 1983.

6. United States Nuclear Regulatory Commission, Standard Review Plan Branch Technical Position 5-2, Revision 1 NUREG-0800, July 1981.

7. H. S. Palme, et al., " Methods of Compliance With Fracture Toughness and Operational Requirements of 10CFR50, Appendix G," BAW-10046P, Babcock & Wilcox, Lynchburg, Virginia, March 1976.

8. A. S. Heller and A. L. Lowe, J r. , " Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged-Arc Welds," BAW-1803, Babcock & Wilcox, I Lynchburg, Virginia, January 1984.

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9. U. S. Nuclear Regulatory Commission, " Pressurized Thermal Shock (PTS)," SECY-82-465, Nuclear Regulatory Commission, Washington, D. C., November 23, 1982.

10. Letter from C. M. Stallings, Virginia Electric and Power Company to E. G. Case, Office of Nuclear Reactor Regula-tion,

Subject:

Reactor Vessel Material Surveillance Program. Docket No. 50-280, January 23, 1978, Public Document Accession No. 780260154.

11. S. E. Yanichko, " Virginia Electric and Power Co. Surry Unit No. 1 Reactor Vessel Radiation Surveillance Program,"

WCAP-7723, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, July 1971.

12. Letter from C. M. Stallings, Virginia Electric and Power Company to E. G. Case, Office of Nuclear Reactor Regula-tion,

Subject:

Reactor Vessel Material Surveillance Program. Docket No. 50-281, January 23, 1978, Public Document Accession No. 780260154.

13. S. E. Yanichko and D. J. Lege, " Virginia Electric and Power Co. Surry Unit No. 2 Reactor Vessel Radiation g Surveillance Program," WCAP-8085, Westinghouse Electric 5 Corporation, Pittsburgh, Pennsylvania, June 1973.

14. E. L. Furchi et al., "Surry Units 1 and 2 Reactor Vessel Fluence and RT gg Evaluations," WCAP-11015, Westinghouse l Electric Corporation, Pittsburgh, Pennsylvania, December 1985.

15. "Effect of Residual Elements on Predicted Radiation Damage

' to Reactor Vessels," U. S. NRC Regulatory Guide 1.99, Revision 2, Draft dated August 14, 1985.

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