Text

Reactor Pressure Vessel & Surveillance Program Matls Licensing Info for North Anna Units 1 & 2
	ML20203C727
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Site:	North Anna
Issue date:	03/31/1986
From:	Lowe A; BABCOCK & WILCOX CO.
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BAW-1911 March 1986 I

REACTOR PRESSURE VESSEL AND SURVEILLANCE PROGRAM MATERIALS LICENSING INFORMATION FOR I

NORTH ANNA UNITS 1 AND 2 I

I I

I Babcock &Wilcox 8604210230 860415 DR ADOCK 0500 8

a McDermott company L

BAW-1911 March 1986 I

I I

I REACTOR PRESSURE VESSEL AND SURVEILLANCE PROGRAM MATERIALS LICENSING INFORMATION FOR NORTH ANNA UNITS 1 AND 2 I

b.'

A.

L. Lowe, Jr.,

P.E.

I I

I B&W Control No. 77-1163804-00 B&W Contract No. 583-7375, Task 042 I

l Prepared for Virginia Electric and Power Company I

Richmond, Virginia by Babcock & Wilcox Nuclear Power Division P. O.

Box 10935 I

Lynchburg, Virginia 24506-0935 I

g

.e.x a McDermott company

I I

I CONTENTS I

Page 1.

INTRODUCTION.

1-1 2.

REACTOR VESSEL DATA BASES 2-1 2.1.

North Anna Unit 1 2-1 2.2.

North Anna Unit 2 2-1 8

2.3.

Surveillance Data Bases 2-2 2.4.

Initial Value of Reference Temperature.

2-2 3.

EVALUATION OF REACTOR VESSEL TOUGHNESS 3-1 I

4.

REACTOR VESSEL SURVEILLANCE PROGRAMS 4-1 I

4.1.

North Anna Unit 1 4-1 4.2.

North Anna Unit 2 4-2 4.3.

Spare Capsules 4-2

-5.

INTEGRATED REACTOR VESSEL SURVEILLANCE PROGRAM.

5-1 e.

SuMMsm e-1 y

7.

REFERENCEE 7-1 List of Tables 1

Table 2-1.

Identification of Reactor Vessel Beltline Region Weld Metals - North Anna Unit-l 2-4 I

2-2.

Chemical Composition of Reactor Vessel Beltline Region Weld Metals - North Anna Unit-1 2-5 2-3.

Mechanical Properties of Reactor Vessel Beltline I

Region Weld Metal - North Anna Unit-1 2-6 2-4.

Identification of Reactor Vessel Beltline Region Base Materials - North Anna Unit-1 2-7 2-5.

Chemical Composition of Reactor Vessel Beltline I

Region Base Materials - North Anna Unit-1 2-8 2-6.

Mechanical Properties of Reactor Vessel Beltline Region Base Materials - North Anna Unit-1 2-9 I

I Babcock & Wilcox

,j, a McDermott company

I List of Tables (Cont'd)

Table Page 2-7.

Properties of Surveillance Program Plate and Weld Material - North Anna Unit-1 2-10 2-8.

Identification of Reactor Vessel Beltline Region Weld Metals - North Anna Unit-2 2-11 2-9.

Chemical Composition of Reactor Vessel Beltline Region Weld Metals - North Anna Unit-2 2-12 2-10. Mechanical Properties of Reactor Vessel Beltline E

Region Weld Metal - North Anna Unit-2 2-13 5

2-11. Identification of Reactor Vessel Beltline Region Base Materials - North Anna Unit-2 2-14 g

2-12. Chemical Composition of Reactor Vessel Beltline 5

Region Base Materials - North Anna Unit-2 2-15 2-13. Mechanical Properties of Reactor Vessel Beltline Region Base Materials - North Anna Unit-2 2-16 2-14. Properties of Surveillance Program Plate and Weld Material - North Anna Unit-2 2-17 3-1.

Evaluation of Reactor Pressure Vessel E

Toughness - North Anna Unit-1 3-3 3

3-2.

Evaluation of Reactor Pressure Vessel Toughness - North Anna Unit-2 3-4 g

4-1.

Revised Surveillance Capsule Withdrawal g

Schedule - North Anna Unit-1 4-3 4-2.

Revised Surveillance Capsule Withdrawal Schedule - North Anna Unit-2 4-4 List of Fiqures Figure 2-1.

Location and Identification of Materials Used in the-Fabrication of the Belt-Line Region of North Anna Unit 1 Reactor Pressure Vessel 2-18 l

2-2.

Location and Identification of Materials Used in g

the Fabrication of the Belt-Line Region of North 3

Anna Unit 2 Reactor Pressure Vessel 2-19 I

1 l

I I

I l

Babcock & Wilcox g

_ jj _

a McDermott company g

l t

I I

I 1.

INTRODUCTION This report provides a review and update of the materials data I

and information for the reactor pressure vessels of North Anna Units 1 and 2 to ensure that they are in compliance with the requirements of 10CFR50, Appendix G.1 In addition, the reactor pressure vessel' surveillance programs were reviewed for compli-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 s.

As a result of this review and update the reactor vessels materials data bases for North Anna Units 1 and 2 were found to be in 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 I

to ensure continued licensibility of the reactor vessels.

A new reactor vessel surveillance capsule withdrawal schedule I

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 nest the intent, if not the letter of the regulations. In addition, these regulations have been revised as experience, new data and analysis capability related I

to reactor vessel integrity have developed.

A periodic review and update is necessary to ensure continued compliance with the regulations.

1-1 I

Babcock &Wilcox a McDermott company

I I

I 2.

REACTOR VESSEL DATA BASES The establishment of the mechanical and toughness properties 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 I

properties available for each specific reactor vessel.

The data are also essential in establishing the normal pressure-tempera-I ture operating limitations as required by 10CFR50, Appendix G.

2.1, North Anna Unit - 1 The materials and chemical composition data for the North Anna Unit 1 reactor vessel are presented in Tables 2-1 through 2-6.

I These data represent an accumulation of information from various sources (References 4, 8 and 9) and are complete except for the I

limited amount of data for the weld metals.

The current essential data for the weld metals are available.

In addition, the initial reference temperature data represents the best available data as defined in Section 2.4.

The location and identification of the forgings and welds within the belt-line region of the North Anna Unit 1 reactor pressure vessel are shown in Figure 2-1.

2.2.

North Anna Unit 2 The data for North Anna Unit 2 reactor vessel are presented in Tables 2-8 through 2-13.

These data represent an accumula-tion of information from various sources (References 4, 10 and

11) and are complete except for the limited amount of data for the weld metals.

The current essential data for the weld metals are available.

In addition, the initial reference temperature I

2-1 Babcock & Wilcox I

a McDermott company

Il i

data represents the best available data as defined in Section 2.4.

The location and identification of the forgings and welds within the belt-line region of the North Anna 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 North Anna 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 weld metals fabricated after the establishment of the ASME Code requirement, plus reevalua-tion 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 NRC Standard Review Plan, Section 5.3.2.5 It is recognized 2-2 Babcock & Wilcox a McDermott company

I that the values recommended in the Standard Review Plan are very conservative.

Previous evaluations of the North Anna reactor vessels had established reference temperatures for the base materials which were representative of actual data.

However, the weld metals had no data for the submerged-arc weldments made by Rotterdam.

The initial value of Rotterdam welds were 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-10046P6 7

RDM Welds 0

20F SECY 82-465 I

I I

I 2-3 Babcock & Wilcox

. ucoermore company

Table 2-1.

Identification of Reactor Vessel Beltline Region Weld Metals - North Anna Unit-1 Wold Welding Weld Wire Flux Identification Weld Incation Process Type Heat No.

Type Iot No.

Reference WOSA Nozzle Shell to Interm.

Sub. Arc SMIT 40 25295 SMIT 89 1170 Docket No.

Shell Seam (O.D. 94%)

50-338(8)

WO5B Nozzle Shell to Interm.

Sub. Arc S4 Ho 4278 SMIT 89 1211 Shell Seam (I.D. 6%)

WO4#

Interm. Shell to Iower Sub. Arc SMIT 40 25531 SMIT 89 1211 Shell Seam Y

- Denotes material included in Reactor Vessel Surveillance F1.ugam

.I sr ER

!=

ED I$

in

'O M

M M

M m

m m

m

m m

m M

M M

m m

M m

Table 2-2.

Gemical Otmoosition of Reactor Vessel Beltline Reaion Weld Metals - North Anna Unit-1 Material Gemical Otmoosition. Weiaht Percent Jdentification C

Mn P

S Si Cr

. Ni Mo V

Ct1 Reference 0.30 Docket no.

[0.10) 0.37 0.36 W0sA 0.10 1.50 [0.020) 50-338(8)

[0.10] 0.37 0.11 0.33 WOSB O.09 1.49 [0.020]

WO4#

0.06 1.29 0.020 0.012 0.35 0.25 0.11 0.48 0.001 0.086

- Denotes material included in Reactor Vessel Surveillance F1 emu

[ ] - Estimate based on review of similar material 5

6 x?

?R a=

5D

,8 $.

O M

Table 2-3.

Hocknical Properties of Reactor Vessel Deltline Rmion Weld Metal - North Anna Unit 1 Tomhness Pruerties Tensile Properties. RP Wald 10F Ermy,

USE, Stremth, Kai Elorg FA Ibst Weld Identification Tg,F Ft-Ibs RPg,F Ft-Ibs Yield Ult.

Heat Treabnent Referen Docket No.

0**

WOSA 0*

50-338(8)

WO5B 0*

0**

WO48

-13 19 102 u

i

- Estimated per NRC Standard Review Plan Sectica 5.3.2

- - Estimated per Section 2.4

- Denotes material incitded in Reactor Vessel Surveillarce Program 5R

?R

!=

5D S$

2-l 4 oM l

l

mm M

M M

Table 2-4.

Identification of Reactor Vessel Beltline Reaion Base Materials - North Anna Unit-1 Material Identification Heat No.

Type Omxment Chde No.

Supplier Heat Treatment Reference 990286/

SA508 CL.2 Nozzle Shell Forgirq 05 Rhenistahl Docket No.

295213 Huttenwerke 50-338(8) 990311/

SA508 CL.2 Inter. Shell Forging 04 298244 990400/

SA508 CL.2 Iower Shell Forging 03 n

n 292332#

- Denotes material incitried in Reactor Vessel Surveillance Frupam

=!

5R

?R

!=

~D ix

O M

Table 2-5.

Chemical Crmoosition of Reactor Vessel Beltline Recion Base Materials - North Anna Unit-1 Material Chemical (trxxsition. Weicht Percent Identification C

Mn P

S Si Ni Cr Mo 00 V

Cu Reference 0.03 0.16 Docket No.

05 0.20 0.71 0.013 0.012 0.21 0.74 0.39 0.64 50-338(8) 04 0.21 0.75 0.010 0.019 0.21 0.82 0.33 0.64 0.05 0.12 0.02 0.15 03#

0.19 0.68 0.009 0.014 0.22 0.80 0.30 0.63

- Denotes material included in Reactor Vessel Surveillance Fr@ tam T.

E

a. k N

8R

!=

5D 3E iii?

O M

M M

M mm m

m

mW M

M M

M m

mm mm mme M

Table 2-6. Mechanical Properties of Reactor Vessel Beltline Reaion Dase Materials - North Anna Unit 1 Touahness Properties Tensile Properties, RT Material

USE, Strenath, Ksi Elong RA Identification TNDT,F RTNDT,F Ft-Lbs Yield Ult.

Reference Docket No.

05 2

6*

50-338(8) 05

-31 17 92 04#

-13 38 85 u

E

- Estimated Per NRC Standard Review Plan Section 5.3.2.

- Denotes material included in Reactor Vessel Surveillance Program W

=b N$

?R g=

5D

$II

O M

Table 2-7.

Properties of Surveillance Program Plate and Weld Material - North Anna Unit-1 Touahness Prtxerties Tensile Prm erties. RP Material

USE, SLanui. Esi Elong RA Ibst Wald Identification Tg,F RPg,F Ft-Ibs Yield Ult.

Heat Treatment

Reference Forging-03

-12

+15 85 70.68 92.55 18.80 58.85 A-1616-1725-2-1/2HR/W2 WCAP-8771(D T-1202-1292-7-1/2HR/M S-ll30i25-14-3/4HR/FC Weld Metal-WO4

-12

-12 95 64.18 79.35 19.20 71.00 S-1130125-10-3/4HR/K mterial ov=ical n=mitico. Weiaht Percent I_d_entification C

_31_

P S

_S1_

_Cr_

_l{i_

Mo Oo V

Cu Reference 0.16 WCAP-8771(D Forgirg-03 0.20 0.68 0.019 0.011 0.26 0.30 0.79 0.61 Y

Weld Metal-WO4 0.06 1.29 0.020 0.012 0.35 0.025 0.11 0.49 0.086 o

l

- A = Austentized/ Water Quenched (WQ)

T = Stress RelievecyFurnace Cooled (K)

S = Stress Relieved /numce Cooled (K) l e, as

?R 8=

5D 85 2=

a8 M

M M

m M

M M

m ma m

m ma e

m

M M

N M

M M

M Table 2-8.

Identification of Reactor Vessel Beltline Region Weld Metals - North Anna Unit-2 Weld Weldirq Weld Wire Flux Identification Weld Incation Process Type Heat No.

Tvoe Iot No.

Reference WOSA Nozzle Shell to Interm.

Sub. Arc S4 Mo 4278 SMIT 89 1211 Docket No.

Shell Seam (O.D. 94%)

50-339(10)

WOSB Nozzle Shell to Interm.

Sub. Arc S4 Mo 801 SMIT 89 1211 Shell Seam (I.D. 6%)

WO4#

Interm. Shell to Iower Sub. Arc S3 Mo 716126 IH320 26 Shell Seam to I

U

- Denotes material included in Reactor Vessel Surveillance Frogam W

m, as N

?R 8w 5D E$

tr

O M

Table 2-9.

Chemical rwm osition of Reactor Vessel Beltline Reaion Weld Metals - North Anna Unit-2 Weld Chmical rwirosition. Weicht Percent Identification C

Mn P

S Si Cr Ni Mo V

Ct1 Reference 0.11 Docket No.,

(0.10] 0.37 0.33 WOSA 0.09 1.49 g.y3;10; 0.18 WO5B 0.086 1.58 0.012 0.012 0.43

[0.10] 0.51 WO4#

0.08 1.82 0.017 0.011 0.25 0.042 0.084 0.49 0.002 0.088

- Denotes material included in Reactor Vessel Surveillanc 2 Pr@ tam

[ ] - Estimate based on review of similar material

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Table 2-12.

Chemical Wition of Peactor Vessel Beltline Reaion Base Materials - North Anna Unit-2 Material chemical Ctmoosition. Weicht Fermnt Identification C

Mn P

S Si Ni Cr Mo 00 V

Q1 Reference 0.01 0.08 Docket No.

05 0.20 0.68 0.010 0.013 0.25 0.77 0.34 0.60 50-339(l0) 0.02 0.09 04#

0.195 0.78 0.010 0.016 0.24 0.83 0.35 0.62 0.01 0.13 03 0.16 0.66 0.013 0.017 0.15 0.83 0.34 0.59

- Denotes material included in Reactor Vessel Surveillance Fi@me u

a

= es N$

!R

!=

5D 3%

li if

O M

Table 2-13.

Mechanical Properties of Reactor Vessel Beltline Region Base Materials -North Anna Unit 2 Touchness Properties Tensile Properties RT Material

USE, Strenoth. Ksi Elong RA f

Identification TNM, F RTNM, F Ft-Lbs Yield Ult.

Reference Docket No.

05

+5 9*

50-339 00) n 04f

-49 75 74 n

03

-13 56 80 l

- Estimated Per NRC Standard Review Plan Section 5.3.2.

f - Denotes material included in Reactor Vessel Surveillance Program w

i o

a 1

am

?aR a=

,, o

,$ I.

M r

M M

M W

W W

M

Table 2-14.

Properties of Surveillance Program Plate and Weld Material - North Anna Unit -2 Tomhness Prroerties Tensile Prooerties. RP Material

USE, Stremth. Ksi Elong RA Post Weld Ft-Ibs Yield Ult.

Heat Treatment

Reference Identification y g,F OM Forging-04

-48

+57 75 84.90 102.00 19.00 48.20 A-1688-1697-21/21B/WQ WCAP-8772 T-1220-1229-GIR/EC S-ll30125F-14 3/411R/K Weld Metal-WO4

-66

-52 115 76.08 85.90 24.20 69.25 S-ll30t25F-13 1/2!B/m Material Chemical Crmoositim. Weiaht Percent Identificatim c

_}!ti_

P S

_S1_

_Gr_

_Hi_

_}jo__

_Qg_

V Cu Reference 0.11 WCAP-8772(11)

Forging-04 0.19 0.76 0.018 0.011 0.25 0.35 0.86 0.60 y

Weld Metal-WO4 0.08 1.82 0.017 0.011 0.25 0.042 0.084 0.49 0.088 t;

- A = Austentize@' Water Quendied (HQ)

T = Stress RelievecVFurna Cboled (K)

S = Stress Relieved /R1rnace Cooled (K) m

=m b$

?R g=

l 5D F

F a

M

I' Figure 2-1.

Location and Identification of Materials Used in the Fabrication of the Belt-Line Region of North Anna Unit-1 Reactor Pressure Vessel (Ref. 12)

H I

f Weld Seam WO5 pp

[gn I

o Fomim 04 CORI

=

m, m

144.0" g

I s

.5 c1

,o-3 17..

ag Weld Seam WO4 ne-5 4

Forcino 03 c.

E w

4 h

f 3

49.3"

,o.

I I

2-18 Babcock & Wilcox a McDermott company

I Figure 2-2.

Location and Identification of Materials I

Used in the Fabrication of the Belt-Line Region of North Anna Unit-2 Reactor Pressure Vessel (Ref. 12)

I

/

I W.ie S..m W0s g

7,.

iL CORI Forcino 04 e.

2 1so-y I

144.0"

,u 3

gL r$*

ud Weld Seam WO4 t

E r

Forcinc 03 a

tsc-i" 49.3"

,o.

I I

I 2-19 Babcock & Wilcox I

a McDermott company

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

I I

I 3.

EVALUATION OF REACTOR VESSEL TOUGHNESS One aspect of reactor pressure vessel licensibility is the toughness of the materials used in its fabrication.

These properties are used to calculate the pressure-temperature 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 operating condition, including anticipated operational occur-rences and system hydrcstatic tests.

The closure head region, the reactor vessel outlet nozzle, and the beltline region have been identified as the only regions of I

the reactor vessel that regulate the pressure-temperature limits.

Since the closure head region is significantly stressed at relatively low temperatures (due to mechanical loads result-ing 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 inside corner of the nozzle, which can be two to three times the I

membrane stresses of the shell.

After the first several years of neutron radiation exposure, the RT of the beltline region g

I materials will be high enough that the beltline region of the reactor vessel will start to control the pressure-temperature limits of the reactor coolant pressure boundary (RCP).

For the service period for which the limit curves are 1stablished, the maximum allowable pressure as a function of fluici 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 I

a McDermott Company

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 region materials for which the measured properties are not available, the unirradiated impact properties and residual originally established for the belt-line region elements, as materials, were determined from acceptable data bases using Bi recognized estimating techniques.

The adjusted reference 5

temperatures are calculated by adding the predicted radiation-induced shift of the RT e un radiated RT neluding NDT NDT margin.

The predicted shift, RT is calculated using the

NDT, respective neutron fluence, plus the copper and nickel con-tents.

The neutron fluence values are calculated based on function derived from adjoint functions using transport calcula-2,13 tions.

The design curves of regulat'ory Guide 1.99, Rev.

were used to predict the radiation-induced shift of RTNM*

The results of the 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 expiration of current licenses.

I I

3-2 Babcock & Wilcox a McDermott company

M M

Table 3-1.

Evnhmtim of Reactor h1 Fracture Ttxs1hness - North Ama thit-1 Fluence at

% at OnTent

)

Licerine Expiration (c) flosvie Expiratim Msterial Identificatial Daltline Owatical Ozoositim 'I Initial Inside surface T/4 locatim Inside 92rface T/4 tocation I

Heat No.

Type Regim locatim Cy par w/o Nickel w/o (a)

Margin,F(c) yan ya,2

cze ya,2 2

2 Forgirq 05 SA508,C1.2 Nozzle Shall Forgirn 0.16 0.74 6

69 2.49E18 1.46E18 151 142 ftzgirg 04 SA508,C1.2 Interm. Shall Ftrgirq 0.12 0.82 17 34 3.11E19 1.83E19 162 149 ftrgirg 03 SA508,C1.2 laer shall rtrgirn 0.15 0.80 38 34 3.11E19 1.83E19 222 212 Wald 05A S en. Arc Itzt. Int. Cir. Wald 0.30 0.10 0

69 1.46E18 144 Wald 05B Stin. Arc Noz. to Int. Cir. Wald 0.11 0.10 0

69 2.49E18 93 Wald 04 Stim. Arc tower Cir. Wald 0.086 0.11 19 56 3.11E19 1.83E19 142 133 I"I Per Tables 2-2 2-7 NPer N 11016(

IIII ICIPer Regulatcry Qaida 1.99, Rev. 2 u

8u r

IR

=b t&

S^

uR g=

5o IE i=

as

ON i

l i

Table 3-2.

Evaluatim of Reactor PrecsuryJgpsel Fracture Totstaggs - North ArryLB11t-2

% at N Fluence at I

License bpiratim(c) 1Acmas retratim ttiterial Idmtificatim Battline dwatcal Qum61tigo *I Initial Irmide Surfaos T/4 locatim Inside Surface T/4 Eccatim I

2 2

Q,F(a)

Margin,F(CI ycm qcm yan ycm Heat No.

'Iype BEgim locatim OtH er w/o Nicial w/o Fergirq 05 SA508,C1.2 Nozzle Shell 0.08 0.77 9

69 2.72E18 1.60E18 111 107 Pbrgirq 04 SA508,C1.2 Interm. Shall 0.09 0.83 75 24 3.40E19 2.00E19 186 177 Pbrgirq 03 SA508,C1.2 Immr Shall 0.13 0.83 56 34 3.40E19 2.00E19 217 201 Wald 05A Stim. Arc 1t23. to L-It. Cir. Wald 0.11 0.10 0

69 1.60E18 102 Wald 058 Stim. Am Nos. to Int. Cir. Wald 0.18 0.10 0

69 2.72E18 125 Wald 04 Stim. Arc Int. to Im. Cir. Wald 0.088 0.10

-48 56 3.40E19 2.00E19 74 57 (a) Par Tables 2-9 2-14

@I (c)hr WGP-11016(

I13I hr Regulatory niMa 1.99, Rev. 2 LJ I

b M

as CF

~n

?R

!=

3D n=

38 M

M M

M

I I

I 4.

REACTOR VESSEL SURVEILLANCE PROGRAMS I

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.

The extent to which a surveillance program meets these objec-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 I

of the requirements can be upgraded to meet the current 10CFR50, Appendix H while other reactor vessels will be required to make I

do with the installed programs.

Each of the North Anna surveil-lance programs will be described separately.

4.1, North Anna Unit 1 The surveillance program was designed in accordance with I

10CFR50, Appendix H, and ASTM E185-73.

The controlling materi-als are contained in the program.

The program is updated I

by establishing a new withdrawal schedule which 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 scheduled 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.

The EFPY schedule would produce results I

4-1 Babcock & Wilcox I

a McDermott company

I with too low cumulative fluence to provide useful irradiation materials data.

The new proposed withdrawal schedule for North Anna Unit 1 is shown in Table 4-1.

4.2.

North Anna Unit 2 The surveillance program was designed in accordance with 10CFR50, Appendix H, and ASTM E185-73.

The controlling materi-als are contained in the program.

The program is updated by establishing a new withdrawal schedule based on accumulative fluence in accordance with ASTM E185-82.

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

4.3.

Soare Caosules 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 I

fluence.

5 I

I I

I Il l

4-2 Babcock & Wilcox a McDermott company

M M

M Table 4-1.

Revised Surveillance capsule Withdrawal Schedule - North Anna Unit-1 1 February 1986 Revmnanded capsule Withdrawal Sctedule Per 100 R50. Appendix H. and E185-82 Revised Withdrawal Schedule

Capsule vessel capsule capsule Estimated Capsule (12) hted Wvi-(12) h ted 2

2 2

Sequence EFPI Fluence, Wcz I.D.

EFPY fluence, rg/cm RV fluence, I)/cn Withdrawal,Yr First 1.5 SE18 or V**

1.1 3.13E18 1.89E18 g50F Seccni 3

9E18 U

5.8 8.45E18 8.61E18 1937

'Ihird 6

1.8E19 X

8.1 1.88E19 1.13E19 1990 (IDL T/4) i" Fourth 15 3.lE19 W***

16.9 3.05E19 2.16E19 2002 (IDL I.W.)

Fifth EOL

>3.lE19 Y****

25 4.21E19 3.11E19 2011 NOIE: Reminder of capsules moved to mvinnn lead factor position during inservice inspections as positions W availbble in order to = vimize total fluence.

W E$

- Reviewed after each fuel cycle and revised as rwvievi after each capsule withdrawal and evaluation.

Estimated withdrawals based cm 18-month fuel cycles and 0.80 plant capacity.

a oR

- - Capsules withdrawn and evaluated.

g{

- - - Transferred fran 0.82 lead factor location to 1.52 lead factor location during the 10 year inservice o

inspecticn (May 1987/5.89 EFFI).

J!

- - - - Transferred fran 0.82 lead factor location to 1.52 lead factor location durity the 20 year inservice inspection (May 1997/13.58 EFFI)

M

Table 4-2.

Revised Surveillance Capsule Withdrawal Schedule - North Anna Unit-2 1 Febnnry 1986 p m...#rded capsule Withdrawal Sdwble Per 10CFR50. Arxendix H. and E185-82 Revised Withdrawal Schedule

Capsule Vml Capsule Capsulo Estimated Capsule (12)

Estimated Navi==(12) gg 2

2 2

Sequence EFPI Fluence, y m I.D.

EFFI fluence, Wcm RV fluence, Wcm Withdrawal,Yr First 1.5 SE18 or V**

1.0 2.70E18 1.73E18 RP g 50F Seccmd 3

1.0E19 U

6.1 1.03E19 8.62E18 1989

'Ihird 6

2.lE19 X

7.1 1.78E19 1.04E19 1990 (IDL T/4)

Fourth 15 3.4E19 W***

17.0 3.60E19 2.43E19 2004 (EDL I.W.)

Fifth EOL

>3.4E19 Y***

23.8 5.25E19 3.40E19 2011 PETIE: Remainder of capsules moved to navi== lead factor position during inservice inspecticris as positims h available in order to mavim1 e total fluence.

Z W

{y

- Reviewed after each fuel cycle and revised as rwvivi after each capsule witMrawal and evaluaticn.

6a Estimatal withdrawals based cm 18-month fuel cycles and 0.80 plant capacity.

jE

- - Capsules withdrawn and evaluated.

o*

- - - Transferred frus 0.82 lead factor location to 1.52 lead factor locaticn durirg the 10 year inservice inspection (August 1990/7.15 EFFI).

r, =

E ou M

M M

M

I I

I 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 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 1

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.

I 5-1 Babcock &Wilcox I

a McDermott company

I 4.

There must be a substantial advantage to be g

gained, such as reduced power outages or 3

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 rate effects on neutron radiation damage.

Since the plants have 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 g

fluence the reactor vessel receives has developed to be of 5

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.

Therefore, it would be more practical to perform the complete capsule evaluation.

In addition, the new revision of Regulatory Guide 1.99, gives g

credit for obtaining surveillance data for the controlling a

materials.

When two or more credible data points become available from a reactor they may be used to determine the adjustuu 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 comparty

I material damage to minimize the effect on operating limita-tions.

In the case of North Anna 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 I

program for North Anna 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.

I I

I 5-3 Babcock & Wilcox I

A McDermott company

I I

I 6.

SUMMARY

As a result of this review and update the reactor vessels materials data bases for North Anna 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 to the extent practical and will provide the material data neces-sary to insure continued compliance with these appendices.

I I

I 6-1 Babcock & Wilcox I

s W<Dermott company

I I

I 7.

REFERENCES 1.

U.S.

Code of Federal Reculations. Title 10. Enerav. Part

5_Q,

" Domestic Licensing of Production and Utilization Facilities, Appendix G, Facture Toughness Requirements."

2.

U.S.

Code of Federal Reculations. Title 10. Enerav. Part

H_Q,

" 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.

North Anna Power Station Units 1 and 2,

Updated Final Safety Analysis Report, Virginia Electric and Power Company, July 20, 1982, as amended.

5.

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

6.

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.

7.

U.

S.

Nuclear Regulatory Commission, " Pressurized Thermal Shock (PTS)," SECY-82-465, Nuclear Regulatory Commission, Washington, D.

C.,

November 23, 1982.

8.

Letter from C. M. Stallings, Virginia Electric and Power Company to Harold B.

Denton, Office of Nuclear Reactor Regulation,

Subject:

Pressure Vessel Fracture Toughness Properties, North Anna Power Station, Unit Nos. 1 and 2,

Docket No. 50-338, December 11, 1978, Public Document I

Accession No. 7812150277.

7-1 Babcock & Wilcox a ucoermoir <ompany

I 9.

J. A. Davidson and J. H. Phillips, " Virginia Electric and Power Company North Anna Unit No. 1 Reactor Vessel Radiation Surveillance Program," WCAP-8771, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, September 1976.

10.

Letter from C. M. Stallings, Virginia Electric and Power Company to Harold B.

Denton, Office of Nuclear Reactor Regulation,

Subject:

Pressure Vessel Fracture Toughness Properties, North Anna Power Station, Unit Hon. 1 and 2,

E Docket No. 50-339, December 11, 1978, Public Document 5

Accession No. 7812150277.

11.

J.

A.

Davidson, et al.,

" Virginia Electric and Power Company North Anna Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-8772, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, November 1976.

12.

E.

L.

Furchi, et al.,

" North Anna Units 1 and 2 Reactor Vessel Fluence and RT Evaluations" WCAP-11016 Revision gg 1, Westinghouse Electric Corporation, Pittsburgh, Pennsyl-vania, December 1985.

13.

"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.

I I

-2 Babcock & Wilcox a McDermott company

. - _ _

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