ML20101A875

Non-proprietary Evaluation of Pressurized Thermal Shock for Vogtle Electric Generating Plant (VEGP) Unit 2

ML20101A875
Person / Time
Site:	Vogtle
Issue date:	03/07/1996
From:	Boyle D, Terek E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20101A873	List: ML20101A870 ML20101A875
References
WCAP-14534, NUDOCS 9603130323
Download: ML20101A875 (24)

Text

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-14534

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l EVALUATION OF PRESSURI7FD THERMAL SHOCK FOR VOGTLE ELECTRIC GENERATING PLANT (VEGP) UNIT 2 J

P. A. Grendys February 1996 Work Performed Under Shop Order GXOP-108 PreyesM by the Westinghouse Electric Corporation l

for the Southern Nuclear Operating Company Approved D. E. Boyle, Manager !

Reactor Equipment & Materials Engineering WESTINGHOUSE ELECTRIC CORPORATION Systems and Major Projects Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 01996 Westinghouse Electric Corporation i

All Rights Reserved 9603130323 960307 DR ADOCK 05000425 PDR

PIEFACE This report has been technically reviewed and verified by:

E. Terek

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TABLE OF CONTENTS Section lillE East l

LIST OF TABLES................................................iii LIST OF FIGURES..............................................iii

1.0 INTRODUCTION

...............................................I l

2.0 PRESSURI7FD THERMAL SHOCK RULE............................ 2 J

l 3.0 METHOD FOR CALCULATION OF RTm....

..................... 4 l

4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES........... 7 l

5.0 NEUTRON FLUENCE VALUES.......................

.......... 12 6.0 DETERMINATION OF RT VALUES FOR ALL BELTLINE REGION m

MATERIALS.................................................. 13 i

7.0 CONCLUSION

S................................................ 18

8.0 REFERENCES

................................................. 19 1

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LIST OF TABLES l

Iattle Titic Eags I

Calculation of Average Cu and Ni Weight Percent Values for Base Materials....... 9 2

Calculation of Average Cu and Ni Weight Percent Values for Weld Materials....

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Vogtle Unit 2 Reactor Vessel Beltline Region Material Properties............... I1 Fluence (10" n/cm, E > 1.0 MeV) on the Pressure Vessel Clad / Base 2

4 Metal Interface for Vogtle Unit 2...

.......... 12 5

Interpolation of Chemistry Factors from Regulatory Guide 1.99, Revision 2, Position 1.1

...................................................14 6

Calculation of Chemistry Factors Using Credible Surveillance Capsule Data Regulatory Guide 1.99, Revision 2, Position 2.1

......................15 7

RTm Calculations for Vogtle Unit 2 Beltline Region Materials at 32 EfPY...... 16 8

RTm Calculations for Vogtle Unit 2 Beltline Region Materials at 54 EFPY...... 17 LIST OF FIGURES Figure Iille Eage 1

Identification and Location of Beltline Region Materials for the Vogtle Unit 2 Reactor Vessel............

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- - = - - - -. - - - - -

SECTION

1.0 INTRODUCTION

t A Pressurized 'Ihermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant

.l pressure in the reactor vessel. A PTS concern arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation.

Such an event may produce the propagation of flaws postulated to exist near the inner wall surface, 1

thereby potentially affecting the integrity of the vessel.

l The purpose of this report is to determine the RT values for the Vogtle Unit 2 reactor vessel using m

the results of the surveillance Capsule Y evaluation. Section 2.0 discusses the PTS Rule and its requirements. Section 3.0 provides the methodology for calculating RTm. Section 4.0 provides the l

l reactor vessel beltline region material properties for the Vogtle Unit 2 reactor vessel. The neutron fluence values used in this analysis are presented in Section 5.0. The results of the RTm calculations are presented in Section 6.0. The conclusion that all PTS screening criteria are satisfied at end-of.

i license (EOL) and references for the PTS evaluation follow in Sections 7.0 and 8.0, respectively.

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SECTION 2.0 PRESSURIZED THERMAL SHOCK RULE The Nuclear Regulatory Commission (NRC) recently amended its regulations for light-water-cooled nuclear power plants to clarify several items related to the fracture toughness requirements for reactor pressure vessels, including pressurized thermal shock requir ments. The revised FTS Rule ",10 CFR l

Part 50.61, was published in the Federal Register on December 19,1995, with an effective date of January 18,1996 This amendment to the FTS Rule makes three changes:

1. The rule incorporates in total, and therefore makes binding by rule, the method for determining the reference temperature, RTm, including treatment of the unirradiated RT, value, the margin term, and the explicit definition of " credible" surveillance data, which is currently described in Regulatory Guide 1.99, Revision 25

2. The section is restructured to improve clarity, with the requirements section giving only the requirements for the value for the reference temperature for end of life fluence, RTm-

3. 'Ihermal annealing is identified as a method for mitigating the effects of neutron irradiation, thereby reducing RTm-The FTS Rule requirements consist of the following:

For each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall have projected values of RTm, accepted by the NRC, for each reactor vessel beltline material for the EOL fluence of the material.

• The assessment of RT must use the calculation procedures given in the FTS Rule, and must m

specify the bases for the projected value of RTm for each vessel beltline material. The report must specify the copper and nickel contents and the fluence values used in the calculation for each beltline material.

This assessment must be updated whenever there is significant change in projected values of RT rrs or upon the request for a change in the expiration date for operation of the facility. Changes to RT values are significant if either the previous value or the current value, or both values, exceed m

the screening criterion prior to the expiration of the operating license, including any renewal term, if applicable for the plant.

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o The RTm screening criterion values for the beltN region are:

270 F for plates, forgings, and axial weld materials, and j

300*F for circumferential weld materials.

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d SECTION 3.0 METHOD FOR CALCULATION OF RTm RT must be calculated for each vessel beltline material using a fluence value, f, which is the EOL m

fluence for the material. Equation 1 must be used to calculate values of RTxor for each weld and a

plate, or forging, in the reactor vessel beltline.

RT

=RT

+M+ ART (I)

Nor xmtu) xer l

RTsurm = reference temperature for a reactor vessel material in the pre-service or unirradiated condition M = Margin to be added to account for uncertainties in the values of RTuore, copper and nickel contents, fluence and calculational procedures. M is evaluated from Equation 2.

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M-2) o* + o

• u A (2) t I

o is the standard deviation for RTuore.

u o = 0*F when RTworm is a measured value u

i o = 17 F when RTworm is a generic value u

f is the standard deviation for ARTuor-0 3 For plates and forgings:

03 = 17 F when surveillance capsule data is not used o, = 8.5 F when credible surveillance capsule data is used For welds:

I 03 = 28 F when surveillance capsule data is not used l

03 = 14 F when credible surveillance capsule data is used o3 not to exceed one-half of ART ur-s i

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ARTum is the mean value of the transition temperature shift, or change in RTum, due to irradiation,

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and must be calculated using Equation 3.

ART,=(CF)*f "'"

(3) 8 CF ('F) is the chemistry factor, which is a function of copper and nickel content. CF is given in Table I for welds and Table 2 for base metal (plates or forgings) of 10 CFR Part 50.61. If i

i surveillance data is deemed credible, it must be used to determine the material-specific value of CF.

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A material-specific vdue of CF is determined in Equation 5.

f is the best estimate neutron fluence, in units of 10" n/cm (E > 1.0 MeV), at the clad-base-metal.

2 interface on the inside surface of the vessel at the location where the material in question receives the highest fluence. De EOL fluence is used in calculating RTm-Equation 4 must be used for detennining RTm using Equation 3 with EOL fluence values for determining ARTm and Equation 2 for determining M.

RT =RT

+M+ ART

(#)

m nontn m

To verify that RTum for each vessel beltline material is a bounding value for the specific reactor vessel, licensees shall consider plant-specific information that could affect the levels of embrittlement.

His information includes but is not limited to the reactor vessel operating temperature and any related surveillance program results. Results from the plant specific surveillance program must be integrated into the RTum estimate if the plant-specific surveillance data has been deemed credible.

When credible surveillance data is available, a material-specific value of CF is determined from Equation 5.

J (o.2s-0.tologry IA*I 3

s s CF=

(5)

E([,(o.ss-o.2ciasr)}

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In Equation 5, "A," is the measured value of ARTuur and "f," is the fluence for each surveillance data l

point. If there is clear evidence that the copper and nickel content of the surveillance weld differs 1

from the vessel weld, i.e., differs from the average for the weld wire heat number associated with he l

vessel weld and the surveillance weld, the measure values of ARTuur must be adjusted for differences 1

l in copper and nickel content by multiplying them by the ratio of the chemistry factor for the vessel I

material to that for the surveillance weld.

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1 SECTION 4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES

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l Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties for the Vogtle Unit 2 vessel was performed. The beltline region of a reactor vessel, per the PTS Rule, is defined as "the region of the reactor vessel (shell material including welds, heat-affected zones and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to 1

radiation damage." Figure 1 identifies and indicates the location of all beltline region materials for the Vogtle Unit 2 reactor vessel.

j Material propeny values were obtained from material test certifications from the original fabrication as well as the additional material chemistry tests performed as part of the surveillance capsule testing DI program. The average copper and nickel values were calculated for each of the beltline region materials using all of the available material chemistry information as shown in Tables 1 and 2. A summary of the pertinent chemical and mechanical properties of the beltline region plates and weld materials of the Vogtle Unit 2 reactor vessel are given in Table 3.

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R4-3 f101-124B M

R4-2 A

300 0

1800 0

5 101-124A 4

9 CORE R4-1 2700 101-124C

( - 101-171 101-142A 0

R8-1 90 B8628-1 0

0 180

.J 0

o, 101-142C 101-142B i

i RGURE1 Identification and Location of Beltline Region Materials for the Vogtle Unit 2 i

Reactor Vessel 1

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l TABLE 1 Calculation of Average Cu and Ni Weight Percent Values for Base Materials l

Inter. Shell Inter. Shell Inter. Shell lower Shell lower Shell Lower Shell Plate R4-1 Plate R4-2 Plate R4-3 Plate R8-1 Plate B8628-1*

Plate B8825-1 Ref.

Cu %

Ni %

Cu%

Ni %

Cu%

Ni %

Cu%

Ni %

Cu %

Ni %

Cu %

Ni %

4 0.07 0.61 0.07 0.59 0.05 0.60 5

0.07 0.64 0.07 0.63 0.07 0.64 3

0.06 0.64 0.05 0.62 0.05 0.59 0.06 0.62 0.05 0.59 0.05 0.59 3

0.05 0.59 1

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0.053 0.598 i

7 0.049 0.549 l

Avg.

0.07 0.63 0.06 0.61 0.05 0.60 0.07 0.63 0.05 0.59 0.06 0.62

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• Surveillance program base metal material.

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TABLE 2 Calculation of Average Cu and Ni Weight Percent Values for Weld Materials Weld Material Weighting

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Reference Cu %

Ni %

Factor 3: WCAP-11381 0.07 0.13 I

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3: WCAP-11381 0.06 0.12 1

3: WCAP-ll381 0.04 0.17 2

Tandem Arc

• 6: WCAP-13007 0.045 0.091 2

Tandem Arc

• 7: Cap. Y Chem. Analysis 0.039 0.127 2

Tandem Arc

• 7: Tandem Arc

• 0.037 0.118 2

7: Tandem Arc

• 0.040 0.137 2

m 8: GL 92-01, Supp.1 0.03 1

R3493 8: Supplier Analysis 0.04 1

8: D18140 0.04 0.15 1

8: D18142 0.04 0.16 1

8: D18138 0.04 0.17 1

I 8: D18139 0.04 -

0.17 1

8: D21867 0.05 0.15 1

8: D18143 0.05 0.19 1

8: D21865 0.05 0.20 1

8: D21866 0.05 0.21 1

8: D18141 0.06 0.27 1

Average 0.04 0.15 NOTE.

• Per the request of the Southern Nuclear Operating Company (SNC), the Cu and Ni weight percent values will be determined per the methods used by SNC in the Vogtle Unit 2 Generic 12tter 92-01, Revision 1, Supplement I submittal'l.

l Surveillance program weld metal specimens are of the tandem are weld type, and therefore, will use a weighting factor of 2 in the calculation of the average Cu and Ni valces for the weld metal.

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m TABLE 3 Vogtle Unit 2 Reactor Vessel Beltline Region Material Propenies Material Description Cu (%) )

Ni (%) )

RTm (*F)

• Intermediate Shell Plate R4-1 0.07 0.63 10 l

l Intermediate Shell Plate R4-2 0.06 0.61 10 Intermediate Shell Plate R4-3 0.05 0.60 30 Lower Shell Plate R8-1 0.07 0.63 40 Lower Shell Plate B8628-1 0.05 0.59 50 Lower Shell Plate B8825-1 0.06 0.62 40 Inter, and Lower Shell Long. Welds 0.04 0.15

-10 Circumferential Weld 0.04 0.15

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NOTES.

(a) Average values of copper and nickel as indicated in Tables 1 and 2 on preceding pages.

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(b) The RT values for the plates and welds are measured values per U.S. NRC Standard Review I lanm, m

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1 SECTION 5.0 NEUTRON FLUENCE VALUES L

'lhe calculated fast neutron fluence (E > 1.0 MeV) values at the inner surface of the Vogtle Unit 2 reactor vessel are shown in Table 4. These values were projected using the results of the Capsule Y radiation analysis'73 TABLE 4 Fluence (10" n/cm, E > 1.0 MeV) on the Pressure Vessel Clad / Base Metal 2

Interface for Vogtle Unit 2 t:rrY O'

15' 30"

30**)

30'")

45' 4.83 0.175 0.254 0.306 0.200 0.166 0.304 16 0.581 0.842 1.02 0.663 0.551 1.01 1

32 1.16 1.69 2.03 1.33 1.10 2.01 48 1.74 2.53 3.05 1.99 1.65 3.02

{

54 1.%

2.84 3.43 2.24 1.86 3.40 l

NOTES, (a) Indicates location in octants with a 12.5* neutron pad span (no surveillance capsules).

(b) Indicates location in octants with a 20.0* neutron pad span (single capsule holder).

l (c) Indicates location in octants with a 22.5* neutron pad span (dual capsule holder).

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SECTION 6.0 DETERMINATION OF RT VALUES FOR ALL BELTLINE REGION MATERIALS m

Using the prescribed PTS Rule methodology, RTm values were generated for all beltline region materials of the Vogtle Unit 2 reactor vessel for fluence values at the EOL (32 and 54 EFPY).

Each plant shall assess the RTm values based on plant-specific surveillance capsule data. Vogtle Unit 2 plant-specific surveillance capsule data for lower shell plate B8628-1 and the weld metal is provided for the following reasons:

1) There have been two capsules removed from the reactor vessel. and the data is deemed credible per 10 CFR Part 50.61. (See Appendix B of WCAP-14532, Reference 7.)

2) The surveillance capsule materials are representative of the actual vessel plates and circumferential and longitudinal weld materials.

Per the Southern Nuclear Operating Company, the beltline region materials of the Vogtle Unit 2 reactor vessel are not contained in any other commercial plant reactor vessel surveillance program.

As presented in Table 5, chemistry factor values for Vogtle Unit 2 based on average copper and nickel weight percent were calculated using Tables 1 and 2 from 10 CFR 50.61 H. Additionally, chemistry t

factor values based on credible surveillance capsule data are calculated in Table 6. Tables 7 and 8 contain the RTm calculations for all beltline region materials for 32 and 54 EFPY, respectively.

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l TABLE 5 Interpolation of Chemistry Factors from Regulatory Guide 1,99, Revision 2, Position 1.1 Material Ni, wt %

Chemistry Factor, *F i

Intermediate Shell Plate R4-1 0.63 44 Given Cu wt% = 0.07 l

l Intermediate Shell Plate R4-2 0.61 37 Given Cu wt % = 0.06

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Intermediate Shell Plate R4-3 0.60 31 Given Cu wt % = 0.05 Lower Shell Plate 0.63 44 E.B::1 i

Given Cu wt % = 0.07 l

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Lower Shell Plate B8628-1 0.59 31 Given Cu wt % = 0.05 Lower Shell Plate B8825-1 0.62 37 4

Given Cu wt % = 0.06 Weld Metal 0.15 38.3 Given Cu wt % = 0 04 M

The weld metal CF was determined using a Cu weight percent value of 0.04% and interpolating between Ni weight percent values of 0% and 0.20% from 10 CFR Part 50.61.

Table 1. Specifically, the CF for Ni = 0% is 24'F and Ni = 0.20% is 43'F respectively.

Therefore, for a Cu weight percent value of 0.04% and a Ni weight percent value of 0.15%,

the weld metal CF value is interpolated to be 38.3'F.

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i TABLE 6 Calculation of Chemistry Factors Using Credible Surveillance Capsule Data Regulatory Guide 1.99, Revision 2, Position 2.1 Material Capsule Capsule f FF ART,a FF* A RT,a FF" Lower Shell Plate B8628-1 U

0.422 0.760 2.12 1.61 0.578 (Longitudinal)

Y 1.13 1.03 5.76 5.%

l.07 lewer Shell Plate B8628-1 U

0.422 0.760 0.00 0.00 0.578 (Transverse)

Y 1.13 1.03 1.93 2.00 1.07 Sum:

9.56 3.30 CF = 1(FF

• RTm) + I(FF ) =

2.9 2

Weld Metal U

0.422 0.760 0.00 0.00 0.578 Y

1.13 1.03 18.59 19.15 1.07 Sum:

19.15 1.65 CF = 1(FF

• RT,a) + I(FF ) =

11.6 2

NOTES.

f = fluence (10" n/cm i All updated fluence values taken from Section 6.0 of the Capsule Y analysis, WCAP-14532m, 2

FF = fluence factor = f a2s.ainso ART,a values obtained frem CVGRAPH Vers' ion 4.0. (See WCAP-14532, Reference 7.) These values differ from those previously reported in WCAP-13007 since those were hand-fit using engineering judgement.

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TABLE 7 RTers Calculations for Vogtle Unit 2 Beltline Region Materials at 32 EFPY Material CF f

FF RT,orn M

ARTm RTm EOL 32 EFPY inter. Shell Plate R4-1 44.0 2.03 1.19 10 34 52.5 i

Inter. Shell Plate R4-2 37.0 2.03 1.19 10 34 44.1 88 i

inter. Shell Plate R4-3 31.0 2.03 1.19 30 34 37.0 101 Lower Shell Plate R8-1 44.0 2.03 1.19 40 34 52.5 126 1.ower Shell Plate B8628-1 31.0 2.03 1.19 50 34 37.0 121 2

using sury, capsule data 2.9 2.03 1.19 50 3.5 3.5 57 Lower Shell Plate B8825-1 37.0 2.03 1.19 40 34 44.1 118 Inter. Shell lang. Weld 101-124A 38.3 1.16 1.04

-10 39.9 39.9 70 1

Inter. Shell Long. Weld 101-124B 38.3 1.33 1.08

-10 41.3 41.3 73 Inter. Shell Long. Weld 101-124C 38.3 1.10 1.03

-10 39.3 39.3 69 Lower Shell Long. Weld 101 142A 38.3 1.16 1.04

-10 39.9 39.9 70 Lower Shell lang. Weld 101-142B 38.3 2.03 1.19

-10 45.7 45.7 81 Lower Shelllong. Weld 101 142C 38.3 2.03 1.19

-10 45.7 45.7 81 Circumferential Weld 101-171 38.3 2.03 1.19

-30 45.7 45.7 61 1

using sury. capsule data 11.6 2.03 1.19

-30 13.8 13.8 2

NOTES.

FF = fluence factor = f "8 * **

• RTeerm values are measured values.

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TABLE 8 RTm Calculations for Vogtle Unit 2 Beltline Region Materials at 54 EFPY Material CF f

FF RTwo M

ARTm RTm l

54 EFPY Inter. Shell Plate R4-1 44.0 3.42 1.32 10 34 58.1 102 Inter. Shell Plate R4-2 37.0 3.42 1.32 10 34 48.9 93 Inter. Shell Plate R4-3 31.0 3.42 1.32 30 34 41.0 105 1

Lower Shell Plate R8-1 44.0 3.42 1.32 40 34 58.1 132 Lower Shell Plate B8628-1 31.0 3.42 1.32 50 34 41.0 125 l

using sury, capsule data 2.9 3.42 1.32 50 3.8 3.8 58 lower Shell Plate B88251 37.0 3.42 1.32 40 34 48.9 123 Inter. Shell long. Weld 101-124A 38.3 1.96 1.18

-10 45.3 45.3 81 Inter. Shell long. Weld 101-124B 38.3 2.25 1.22

-10 46.7 46.7 83 Inter. Shell Long. Weld 101 124C 38.3 1.85 1.17 10 44.8 44.8 80 Lower Shell long. Weld 101-142A 38.3 1.96 1.18

-10 45.3 45.3 81 Lower Shell Long. Weld 101-142B 38.3 3.42 1.32

-10 50.6 50.6 91 Lower Shell long. Weld 101-142C 38.3 3.42 1.32

-10 50.6 50.6 91 Circumferential Weld 101 171 38.3 3.42 1.32

-30 50.6 50.6 71 using surv. capsule data 11.6 3.42 1.32

-30 15.3 15.3 I

l NOTES FF = fluence factor = f nasm o RTau values are measured values.

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i SECTION 7.0

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CONCLUSIONS As shown in Tables 7 and 8, all of the beltline region materials in the Vogtle Unit 2 reactor vessel have EOL RT values well below the screening criteria values of 270 F for plates and longitudinal m

welds and 300 F for circumferential welds at EOL (32 and 54 EFPY).

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l SECTION

8.0 REFERENCES

1.

10 CFR Pan 50.61, " Fracture Toughness Requirements for Protection Against Pressurized

'Ihermal Shock Events", Federal Register, Volume 60, No. 243, dated December 19,1995, effective January 18,1996.

2.

Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials,"

U.S. Nuclear Regulatory Commission, May 1988.

3.

WCAP-ll381, " Georgia Power Compi.ny Alvin W. Vogtle Unit No. 2 Reactor Vessel Radiation Surveillance Program", L. R. Singer, April 1986.

4.

Metallurgical Research and Development Dept., Materials Cenification Reports, Vendor -

Lukens Steel Company, Contract No. 7372, Code Nos. R-4-1, R-4-2, and R-4-3, J. M. Arnold, April 5,1974.

5.

Metallurg; cal Research and Development Dept., Materials Cenification Repons, Vendor -

Lukens Steel Company, Contract No. 7372, Code Nos. B-8628-1,11-8825-1, and R-8-1, C. E.

Bingham, March 12,1975.

6.

WCAP-13007, " Analysis of Capsule U from the Georgia Power Company Vogtle Electric Generating Plant Unit 2 Reactor Vessel Radiation Surveillance Program" E. Terek, et al.,

August 1991.

7 WCAP-14532, " Analysis of Capsule Y from the Georgia Power Company Vogtle Unit 2 Reactor Vessel Radiation Surveillance Program", P. A. Grendys, et al. February 1996.

8.

LCV-0648B, "Vogtle Electric Generating Plant Response to Generic Leuer 92-01, Revision 1, Supplement 1, Reactor Vessel Structural Interrity", C. K. McCoy, dated 11/15/95.

9.

" Fracture Toughness Requirements", Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, Rev.1 July 1981.

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