ML20080S526
| ML20080S526 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/31/1995 |
| From: | Peter P, Rishel R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20080S520 | List: |
| References | |
| WCAP-14197, NUDOCS 9503100432 | |
| Download: ML20080S526 (31) | |
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i WESTINGHOUSE CLASS 3 (Non-Pmprietary)'
, WCAP-14197 l
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f EVALUATION OF PRESSURran THERMAL SHOCK FOR FARLEY UNITS 1 & 2 i
.E I
P. A. Peter
.j i
March 1995 j
I f
Work Performed Under Shop Orders AWEP-108 and ALPP-0733 l
t Prepared by Westinghouse Electric Corporation -
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for the Southern Nuclear Operating Company
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Approved by:
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.t R. D. Rishel, Manager j
Metallurgical & NDE Analysis i
WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355
'l i
O 1995 Westinghouse Electric Corporation All Rights Reserved
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PREFACE P
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' This report has been technically reviewed and verified by:_
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4 E. Terek t
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. TABLE OF CONTENTS Section lillt _.
- Enitt
- LIST OF TABLES iii LIST OF FIGURES.................................................. v L
~
1.0 INTRODUCTION
1 l
2.0 PRESSURIZFD THERMAL SHOCK....................................
2 3.0 METHOD FOR CALCULATION OF RTm...
4 i
i i
4.0 VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES....
6 5.0 NEUTRON FLUENCE VALUES.................
9 1
i i
i 6.0 DETERMINATION OF RTm VALUES FOR ALL BELTLINE REGION l
1
'1
' M ATE RI ALS................................................... 10 7.0 CONCLUS IONS................................................ 21 i
8.0 REFERENCES
.................................................... 23 L
11
a LIST OF TABLES lahJle Title P.agg 1
Neutron Exposure (n/cm, E>1.0 MeV) Projections at Peak Location on the 2
Farley Unit 1 Pressure Vessel Clad / Base Metal Interface......................
9 2
Neutron Exposure (n/cm, E>1.0 MeV) Projections at Peak Location on the 2
Farley Unit 2 Pressure Vessel Clad / Base Metal Interface..
9 3
PTS Calculations for Farley Unit I at 12.43 EFPY 12 4
FTS Calculations for Farley Unit I at 32 EFPY.........
12 5
PTS Calculations for Farley Unit I at 48 EFPY............................ 13 6
l'TS Calculations for Farley Unit 2 at i1.201 EFPY.........
13 7
PTS Calculations for Farley Unit 2 at 32 EFPY.
14
{
t t
8 PTS Calculations for Farley Unit 2 at 48 EFPY...............
14 9
Calculation of Chemistry Factors Using Farley Unit 1 Surveillance Capsule Data..... 15 10 Calculation of Chemistry Factors Using Farley Unit 2 Surveillance Capsule Data..... 16 11 PTS Calculations for Farley Unit I at 12.43 EFPY - Using Credible Surveillance 17 Capsule Data.
12 FTS Calculations for Farley Unit I at 32 EFPY - Using Credible Surveillance 17 Capsule Data.
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in LIST OF TABLES r
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- I sb]E Tint
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- 13 FTS Calculations for Farley Unit I at 48 EFPY - Using Credible Surveillance Capsule Data..................................................... 18 l
b 14:
FTS Calculations for Farley Unit 2 at 11.201 EFPY - Using Credible Surveillance Capsule Data..................................................... 18-
[
.+
'f 15 1*fS Calculations for Farley Unit 2 at 32 EFPY - Using Cmlible Surveillance j
Capsule Data................................................... 19 J
16 FTS Calculations for Farley Unit 2 at 48 EFPY - Using Credible Surveillance Capsule Data...................................................... 19 l
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. LIST OF FIGURES 3
ElglEE Tillt-:
P.aan
+
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Identification and Location of Beltline Region Materials for the Farley Unit 1 i
. Reactor Ves sel.................................................
.7-
. i
'3'
. Identification and Location of Beltline Region Materials for the Farley Unit 2 Reactor Vessel.......................
8' "l
t s
'k; 3
RTm versus Fluence Curves for Farley Unit 1 Limiting Material-Lower Shell i
lf Plates B6919-1 and B6919-2........................................ 21 4
RTm versus Fluence Curves for Farley Unit 2 Limiting Material - Intermediate.
Shell Plate B7212-1..........................
22.
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SECTION
1.0 INTRODUCTION
Pressurized thermal shock (PTS) events are system transients in a pressurized water reactor (PWR) that can cause severe overcooling concurrent with or followed by significant pressure in the reactor vessel.
The thermal stresses caused by rapid cooling of the reactor vessel inside surface combined with the pressure stresses increase the potential for fracture if an initiating flaw is present in low toughness material.tri In 1985 the Nuclear Regulatory Commission (NRC) issued a formal ruling on FTS. It established screening criteria on pressurized water reactor vessel embrittlement as measured by the nil-ductility reference temperature, termed RTm"1 RTm screening values were set for beltline axial welds, forgings or plates, and beltline circumferential weld seams for the end-of-license plant operation. The screening criteria were determined using conservative fracture mechanics analysis techniques. All PWR vessels in the United States have been required to evaluate vessel embrittlement in accordance with the criteria through end-of-license.
The NRC recently amended its regulations for light water nuclear power plants to change the procedure for calculating radiation embrittlement. The revised FTS Rule (10 CFR Part 50.61) was published in the Federal Register on May 15,1991 with an effective date of June 14,1991r21. This amendment makes the procedure for calculating RTns values consistent with the methods given in Regulatory Guide 1.99, Revision 2I31 The purpose of this report is to determine the RTns values for the Farley Units 1 and 2 reactor vessels to address the revised FTS Rule. Section 2.0 discusses the Rule and its requirements. Section 3.0 provides the methodology for calculating RTm. Section 4.0 provides the reactor vessel beltline region material properties for the Farley Units 1 and 2 reactor vessels. The neutron fluence values used in the analyses are presented in Section 5.0. The results of the RTm calculations are presented in Section 6.0. The conclusions and references for the FTS evaluations follow in Sections 7.0 and 8.0, respectively.
I
SECTION 2.0 PRESSURIZED THERMAL SHOCK
'Ihe FTS Rule requires that the FTS submittal be updated whenever there are changes in core loadings, surveillance measurements, or other information that indicates a significant change in projected RTm values. The Rule outlines regulations to address the potential for PTS events on pressurized water reactor vessels in nuclear power plants that are operated with a license from the United States Nuclear Regulatory Commission (USNRC). PTS events have been shown from operating experience to be transients that result in a rapid and severe cooldown in the primary system coincident with a high or increasing primary system pressure. The FTS 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 result in the propagation of flaws postulated to exist near the inner wall surface, thereby potentially affecting the integrity of the vessel.
The Rule establishes the following requirements for all domestic, operating PWRs:
All plants must submit projected values of RTm for reactor vessel beltline materials by giving values for time of submittal, the expiration date of the operating license, and the projected expiration date if a change in the operating license or renewal has been requested. This assessment must be submitted within six months after the effective date of this Rule if the value of RTm for an:r material is projected to exceed the screening criteria. Otherwise,it must be submitted with the next update of the pressure-temperature limits, or the next reactor vessel surveillance capsule report, or within 5 years from the effective date of this Rule change, whichever comes first. These values must be calculated based on the methodology specified in this rule. The submittal must include the following:
I 1) the bases for the projection (including any assumptions regarding core loading patterns), and 2) copper and nickel content and fluence values used in the calculations for each beltline material. (If these values differ from those previously submitted to the NRC, justification must be provided.)
2
7
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I
'Ihe RTm screening criteria for the reactor vessel belthne region are:
l
. 270*F for plates, forgings, axial welds; and i
300*F for circumferential weld materials.
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,'Ihe following equations must be used to calculate the RTm values for each weld, plate l
or forging in the reactor vessel beltline:
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Equation 1:
RTm = I + M + ARTm j
Equation 2:
ARTm = CF
- f * *'"'8 0 All values of RTm must be verified to be bounding values for the specific reactor vessel.
In doing this, each plant should consider plant-specific information that could affect the
_j I
level of embrittlement.
1 i
Plant-specific FFS safety analyse; are required before a plant is within 3 years of reaching the screening criteria, including analyses of alternatives to minimize the PTS concern.
r NRC approval for operation beyond the screening criteria is seguired.
i
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3 I
SECTION 3.0 METHOD FOR CALCULATION OF RTm In 10 CFR 50.61(b)(2), the NRC Staff has selected a conservative and uniform method for determining plant-specific values of RTm at a given time. For the purpose of comparison with the screening criteria, the value of RTrrs for the reactor vessel must be calculated for each weld and plate or forging in the beltline region as follows.
RTm = I + M + ARTm, where ARTm = CF
- FF I=
Initial reference temperature (RTum) in F of the unirradiated material M = Margin to be added to cover uncertainties in the values of initial RT ur. COPPER and nickel N
contents, fluence and calculational procedures, in F.
M = 66*F for welds and 48'F for base metal if generic values of I are used M = 56 F for welds and 34 F for base metal if measured values of I are used FF = fluence factor = f e28 aioios o, where f = Neutron fluence in units of 10" n/cm (E>l.0 MeV) at the clad / base metal interface 2
Paragraph (b)(3) of 10 CFR 50.61 requires licensees to verify that values of RTm calculated using the above method are bounding values for the specific reactor vessel by considering plant-specific information, including vessel operating temperature and surveillance results, that could affect the level of embrittlement. Regulatory Guide 1.99, Revision 2, provides guidance regarding the application of plant-specific information. Specifically, Regulatory Guide 1.99, Revision 2, provides the methodology for adjusting the margin term (M) and calculating a chemistry factor (CF) based on the surveillance capsule results for use in the same equation shown above from 10 CFR 50.61 (all other terms are the same as defined in 10 CFR 50.61 shown above).
4
9y 4;!!
p L Per Regulatory _ Guide 1.99, Revision 2:
~
8 2
M =' margin = 2 0
+ o, op 3
' oi = 0*F when I is a measured value ci= 17'F when I is a generic value -
- l t
For plates and forgings:
l o, = 17'F when surveillance capsule data is not used j
i 4 = 8.5'F when surveillance capsule data is used I
0 For welds:
03 = 28'F when surveillance capsule data is not used j
03 = 14'F when surveillance capsule data is used
]
03 not to exceed 0.5* ARTrn j
t CF - If plant-specific surveillance data has been deemed credible per Regulatory Guide 1.99, Revision 2, it may be considered in the calculation of the chemistry factor per Position 2.1 of Regulatory l
Guide 1,99, Revision 2.
.k Values of RTrrs modified using the procedures of paragraph (b)(3) of 10 CFR 50.61 are subject to the approval of the Director, Office of Nuclear Reactor Regulation.
6 1
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SECTION 4.0.
VERIFICATION OF PLANT-SPECIFIC MATERIAL PROPERTIES Before performing the pressurized thermal shock evaluations, a review of the latest plant-specific material properties for the Farley Units 1 and 2 vessels was performed. 'Ihe beltline region is defined by the PTS Rulem to be "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 t
regions of the reactor vessel that are predicted to experience sufficient neutron irradiation damage to be considered in the selection of the most limiting material with regard to radiation damage." Figures l
1 and 2 identify and indicate the location of all beltline region materials for the Farley Units 1 and 2 reactor vessels.
j i
Material property values were obtained from material test certifications from the original fabrication as i
well as the additional material chemistry tests performed as part of the surveillance capsule testing j
programs *33 The copper and nickel values used in the calculations were obtained from Reference 6.
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CIRCUMFERENTIAL SEAMS VERTICAL SEAMS 86903-3 19-894B M _ 10-894 45 '
8.4"
(
CORE h
l CORE l
1
=2
)
m 144.0" 19-894A B6903 o
.5 C
P 20.1"11-894
(
~ '
- B6919-2 i
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- 45
[o CORE y
~
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3 48.75"
.J B6919-1 20-894A Figure 1 Identification and Location of Beltline Region Materials for the Farley Unit 1 Reactor Vesselm 4
7
1 1
CIRCUMFERENTIAL SEAMS VERTICAL SEAMS B7212-1 19-9238
- q _ 10-923 I
45 8.4" l
I CORE m
h i
r CORE
=
R m
l o
144.D" 3
19-923A i
-o B7203-1
-c
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g 20.1"11-923 4
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i 20-9238 B7210-2 7
'45 i
m CORE U
g
.3 n
48.75" l
f 20-923A B7210-1 Figure 2 Identification and Location of Beltline Region Materials for the Farley Unit 2 Reactor Vesselm i
8 i
i
SECTION 5.0 NEUTRON FLUENCE VALUES The calculated peak fast neutron fluence (E>1.0 MeV) values at the inner surface of the Parley Units 1 and 2 reactor vessels are shown in Tables 1 and 2, respectively. These values were projected using M
the results of the Farley Unit 1 Capsule W radiation surveillance program and the Farley Unit 2 m
Capsule X radiation surveillance program, based on the assumption that the average exposure rates characteristic of the analyzed cycles would continue to be applicable throughout the plant lives. The RTm calculations were performed using peak fluence values, which occur at the 0 and 45 azimuthal angles of the Farley Units 1 and 2 reactor vessels.
TABLE 1 2
Neutron Exposure (n/cm, E>l.0 MeV) Projections at Peak Location on the Farley Unit 1 Pressure Vessel Clad / Base Metal Interface
5.059 x 10
32 3.550 x 10" 1.216 x 10
48 5.193 x 10
l.797 x 10
I 1
TABLE 2 2
Neutron Exposure (n/cm, E>l.0 MeV) Projections at Peak Location on the M
Farley Unit 2 Pressure Vessel Clad / Base Metal Interface C*
45' EFPY Base Metal & Cire. Weld Longitudinal Welds 11.201" i 1.360 x 10
4.625 x 10
32 3.462 x 10
1.289 x 10 '
j 48 5.079 x 10" l.925 x 10
9 l
p SECTION 6.0 l
DETERMINATION OF RTm VALUES FOR ALL BELTLINE REGION MATERIALS i
F
-.Using the methodology of 10 CFR 50.61(b)(2), RTm values were generated for all beltline region materials of the Farley Units 1 and 2 reactor vessels. Values for RTm are provided for Unit I at
.12.'43 EFPYm (surveillance capsule withdrawal),32 EFPY, and 48 EFPY in Tables 3,4, and 5, respectively. Unit 2 values for RTm are provided at 11.20lm(January 1995),32 EFPY, and 48 f
g
-- EFPY in Tables 6,7, and 8, respectively.
i b
' Plant-specific information that could affect the level of embrittlement was considered in the determination of RTm values in accordance with 10 CFR 50.61(b)(3). Based on plant-specific information, values of RTm calculated in accordance with 10 CFR 50.61(b)(2) are bounding values
- for the Parley Unit I and Unit 2 reactor vessels.
Plant-specific information for Farley Unit 1 is based on; test results from four surveillance capsules that have been removed from the Farley Unit I reactor
- vessel, surveillance data that has been deemed credible per Regulatory Guide 1.99, Revision 2, and
[
the materials contained in the suiveillance capsules are representative of the vessel plate B6919-1, i
intermediate shell longitudinal welds, and heat-affected zone (HAZ) between the intermediate
[
l shell plates B6903-2 and B6903-3.
Plant-specific information for Farley Unit 2 is based on:
test results from three surveillance capsules that have been removed from the Farley Unit 2 reactor vessel, surveillance data that has been deemed credible per Regulatory Guide 1.99, Revision 2, and e
the materials contained in the surveillance capsules are representative of the vessel plate B7212-1, j
intermediate shell longitudinal weld 19-923B, and heat-affected zone (HAZ) between the intermediate shell plates B7203-1 and B7212-1.
i Chemistry factors based on surveillance capsule results were calculated in accordance with Position 2.1 of Regulatory Guide 1.99, Revision 2, for Farley Unit I and Unit 2 and are provided in Tables 9 and f-10, respectively. Tables 11 through 13 provide Farley Unit I values of RTm determined using the l
methodology of Regulatory Guide 1.99, Revision 2, to deterrnine the chemistry factor (CF) and margin 10 i
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.l
- (M) terms.. Similarly, Tables 14 through 16 provide the correspondag Unit 2 RTm values determmed in accordance with Regulatory Guide 1.99, Re' vision 2.
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' TABLE 3 '
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-j PTS Calculations for Farley Unit I at 12.43 EFPY Material
-Cu%
Ni%
CF f'
FF.
1-M-
-ARTm RTm
{
12.43 EPPY '
I inser. Shell Plate B6903-2 0.13
- 0.60 91.0 1.539.
1.119 0
34 101.9 =
135.9
-j Inter. Shell Plate B6903 0.12 0.56 82.2
~ 1.539 L 1.119
.10 34 92.0 --
136.0 '
Lower Shell Plane B6919 0.14 0.55 97.8 1.539 -
1.119 15 34-109.5 -
158.5
- )
14wer Shell Plate B6919-2 0.14 0.56 98.2 1.539 1.119 ~
5-34 109.9 148.9 i
~ '
Inser. Shell leng. Welds '
. 894A/B 0.25
' O.21 127.1-0.5059 0.810
-56 66 102.9 112.9 Cire. Weld 11-894 0.225 0.20 114.5 1.539 '
1.119
-56 66 128.2 138.2
{
i Lower Shell Long. Welds 20-894A/B 0.17 0.20 92.0 0.5059 0.810 56 -
66 74.5 84.5 l
l St.LL Initial RT,str values of the base rnetal plates are measured values, while the weld metal values are generic.
-l f
TABLE 4 1
F13 Calculations for Farley Unit I at 32 EFPY q
Material Cu%
Ni%
CF f
. FF 1
M ARTm RTm f
i 32 EFFY inter. Shell Plate B6903-2 0.13 0.60 91.0 3.550 1.330 0
34 121.0 155.0 Inter. Shell Plate B6903-3 0.12 0.56 82.2 3.550 1.330 10 34 109.3
' 153.3 lower Shell Plate B6919-1 0.14 0.55 97.8 3.550 1.330 15 34 130.1 179.1 i
i Lower Shell Plate B6919-2 0.14 0.56 98.2 3.550 1.330 5
34 130.6 169.6 Inter. Shell Long. Welds 19-894A/B 0.25 0.21 127.1 1.216 1.055
-56 66 134.0 144.0 1
Cire. Weld 11894 0.225 0.20 114.5 3.550 1.330
-56 66 152.3 162.3
'{
i Lower Shell Long. Welds 20 894A/B 0.17 0.20 92.0 1.216 1.055
-56 66 97.0 107.0
. san Initial RT,,,r values of the base metal plates are measured values, while the weld metal values are generic.
l i
12
- i A
.~. -
. ~.
N1
=
1
.4 9
i TABLE 5 l
PTS Calculations for Farley Unit 1 at'48 EFPY i'
. Material Cu%
Ni%
CF f
FF 1
M ARTm RTm 48 EFPY inter. Shell Plate B6903-2 0.13 0.60 91.0 5.195 1.410 0
34 128.3 162.3 j
1mer. Shell Plate B690313
' O.12 -
0.56 82.2 5.195 1.410 10 34 115.9 159.9 -
{
lower Shell Plate B6919-1 0.14 0.55 97.8 5.195 1.410 '
15-34 137.9' 186,9 l
Lower Shell Plate B6919-2 0.14 0.56
. 98.2 5.195 1.410 -
5 34 138.4 177.4 Inter. Shell long. Welds i
19-894A/B 0.25 0.21 127.1 1.797 1.161
-56 66 147.6 157.6 l
i Cire. Weld 11-894 0.225 0.20 114.5 5.195 1.410
-56 66 161.4 171.4 lewer Shell 1.ong. Welds i
20-894A!B 0.17 0.20 92.0 1.797 1.161
-56 66.
106.8
. 116.8 l
N i
Initial RT,arr values of the base metal plates are measured values, while the weld rnetal values are generic.
i TABLE 6 PTS Calculations for Farley Unit 2 at 11.201 EFPY j
I Material Cu%
Ni%
CF f
FF I
M ARTm RTm j
11.201 EFPY
..i Inter. Shell Plate B7203-1 0.14 0.60 100.0 1.360 1.085 15 34 108.5 157.5 Inter. Shell Plate B7212-1 0.20 0.60 149.0 1.360 1.085
-10 34 161.7 185.7 i
lower Shell Plate B72101 0.13 0.56 89.8 1.360 1.085 18 34 97.5 149.5 1Awer Shell Plate B7210-2 0.14 0.57 98.7 1.360 1.085 10 34 107.1 151.1 Cire. Weld 11-923 0.13 0.20 76.0 1.360 1.085
-40 56 82.5 98.5 l
lower Shell Long. Welds 0.05 0.20 49.0 0.463 0.785
-70 56 38.5
- 24.5 20-923A/B Inter. Shell long. Welds 19-923A 0.02 0.%
27.0 0.463 0.785
-56 66 21.2 31.2 L!
j Inter. Shell long. Welds 19-923B 0.02 0 93 27.0 0 463 0.785
-60 56 21.2 17.2
,illlL Initial RT,arr values of the beltline region materials are measured values, except for 1mer. Shell Long. Weld 19-923A.
-j l
i 13
TABLE 7 FTS Calculations for Farley Unit 2 at 32 EFPY Material Cu%
Ni%
CF f
F7 1
M ARTm RTm 32 EFPY Inter. Shell Plate B7203-1 0.14 0.60 100.0 3.462 1.324 15 34 132.4 181.4 Inter. Shell Plate B7212-1 0.20 0.60 149.0 3.462 1.324
-10 34 197.3 221.3 lower Shell Plate B7210-1 0.13 0.56 89.8 3.462 1.324 18 34 118.9 170.9 lower Shell Plate B7210-2 0.14 0.57 98.7 3.462 1.324 10 34 130.7 174.7 Cire. Weld 11-923 0.13 0.20 76.0 3.462 1.324
-40 56 100.6 116.6 lower Shell Long. Welds 0.05 0.20 49.0 1.289 1.071
-70 56 52.5 38.5 20-923A/B Inter. Shell long. Welds 19-923A 0.02 0.96 27.0 1.289 1.071
-56 66 28.9 38.9 Inter. Shell tone. Welds 19-923B 0 02 0 93 27.0 1.289 1.071
-60 56 28.9 24.9 NOTE:
Initial RTc values of the beltline region materials are measured values, except for Inter. Shelllong. Weld 19-923A.
TABLE 8 PTS Calculations for Farley Unit 2 at 48 EFPY Material Cu%
Ni%
CF f
FF 1
M ARTm RTm 48 EFPY Inter. Shell Plate B7203-1 0.14 0.60 100.0 5.079 1.405 15 34 140.5 189.5 Inter. Shell Plate B7212-1 0.20 0.60 149.0 5.079 1.405
-10 34 209.4 233.4 1Awer Shell Plate B7210-1 O.13 0.56 89.8 5.079 1.405 18 34 126.2 178.2 Lower Shell Plate B7210-2 0.14 0.57 98.7 5.079 1.405 10 34 138.7 182.7 Cire. Weld 11923 0.13 0.20 76.0 5.079 1.405
-40 56 106.8 122.8 Lower Shell long. Welds 0.05 0.20 49.0 1.925 1.179
-70 56 57.8 43.8 20-923A/B Inter. Shell Long. Welds 19-923A 0.02 0.96 27.0 1.925 1.179
-56 66 31.8 41.8 Inter. Shell tone Welds 19-923B 0 02 0 93 27 0 1.925 1.179
-60 56 31 8 27.8 hM Initial RTm values of the belthne region materials are measured values. except for Inter. Shell long. Weld 19-923A.
)
14 i
6 j
m
(.
.?-
- TABLE.9
. Calculation of Chemistry Factors Using Farley Unit 1 Surveillance Capsule Datal.
[
t -
li Material.
Capsule
- f-FF A RT,,,
' FF* A RT,,,
FF' i
Lower Shell Plate B6919 Y-0.6014.
0.858 85 72.897 0.735 -
l (IAngitudinal)
U l.817 1.164 105
' 122.203 1.355 X-3.062 1.296 135 174.902 1.679 W
4.040 1.358 155
-210.547
' l.845
- }
Lower Shell Plate B6919-1 Y
0.6014 0.858.
55 47.168 0.735 -
}
(Transverse)
U 1.817 1.164 90 104.746 1.355 X
3.062 1.2%
105 136.035 1.679 W
4.040 1.358 -
145 196.963 1.845 l
1065.461 11.227 2
0.6014 0.858 80 68.609 0.735 U
1.817 1.164 80
- 93.107 1.355 X
3.062 1.296 100 129.557 1.679 W
4.040 1.358' 95 129.045 1.845 -
.dWlfsh.
I f = fluence + 10" n/cm'; All values taken fan Capsule W analysis. WCAP-141965 FF - noence factor - t * **
- l i
. Calculated Chemistry Factor for Lower Shell Plate B6919-1 based on surveillance capsule data =
94.9'F j
.. Calculated Chemistry Factor for Weld Metal based on surveillance capsule data = 74.9'F t
.l 15
~
d
< TABLE 10 '
tssi -
Calculation of Chemistry Factors Using Farley Unit 2 Surveillance' Capsule Data Material C-31e -
f FF ART FF* ART FF' Inter. Shell Plate B7212-1 U
0.612 0.862 103 88.834 0.744 (Longitudinal)
- W 1.67 1.141 165-188315 1303 X
3.02 1.292-180 232.613 1.670
' Inter. Shell Plate B7212-1
'U 0.612 0.862
-133 114.708 0.744 (Fransverse)
W 1.67 1.141 165 188315 1303 X
3.02 1.292 190 245.536 1.670 1058321 7.433 CF = E(FF
- RTer) + E(FF') = 142.4 i
Weld Metal 11-923 U
L 0.612 0.862 10 8.625 0.744 W
l.67 1.141 10 11.413 1303 X
3.02 1.292 10 12.923 1.670 32.%1 3.716 CF = E(FF
- RTa) + E(FF') = 8.9 dsara.
f = fluence"" + 10* n/cm' (E > 1.0 MeV)
FF = nuence Getor = f ""N o Calculated Chemistry Factor for Intermediate Shell Plate B7212-1 based on surveillance capsule data = 142.4'F
. Calculated Chemistry Factor for Weld Metal based on surveillance capsule data = 8.9'F 16
l TABLE 11 PTS Calculations for Farley Unit 1 at 12.43 EFPY - Using Credible Surveillance Capsule Data Material Cu%
Ni%
CF f
FF I
M ARTm RTm 12.43 EFPY Inter. Shell Plate B6903-2 0.13 0.60 91.0 1.539 1.119 0
34 101.9 135.9 Inter. Shell Plate B6903-3 0.12 0.56 82.2 1.539 1.119 10 34 92.0 136.0 Lower Shell Plate B6919-1 0.14 0.55 97.8 1.539 1.119 15 34 109.5 158.5 using S/C data 94.9 1.539 1.119 15 17 106.2 138.2 Lower Shell P Ax B6919-2 0.14 0.56 98.2 1.539 1.119 5
34 109.9 148.9 Inter. Shell Long. Welds 19-894A/B 0.25 0.21 127.1 0.5059 0.810
-56 66 102.9 112.9 using S/C data 74.9 0.5059 0.810
-56 44 60.7 48.7 Cire. Weld 11-894 0.225 0.20 114.5 1.539 1.119
-56 66 128.2 138.2 Lower Shell Long. Welds 20-894A/D 0.17 0.20 92.0 0.5059 0.810
-56 66 74.5 84.5 NOTL Initial RT., values of the base metal plates are measured values, while the weld metal values are generic.
Margins per Regulatory Guide 1.99, Revision 2.
TABLE 12 FTS Calculations for Farley Unit I at 32 EFPY - Using Credible Surveillance Capsule Data Material Cu%
Ni%
CF f
FF I
M ARTm RTm 32 EFPY inter. Shell Plate B6903-2 0.13 0.60 91.0 3.550 1.330 0
34 121.0 155.0 Inter. Shell Plate B6903-3 0.12 0.56 82.2 3.550 1.330 10 34 109.3 153.3 Lower Shell Plate B6919-1 0.14 0.55 97.8 3.550 1.330 15 34 130.1 179.1 using S/C data 94.9 3.550 1.330 15 17 126.2 158.2 Lower Shell Plate B6919-2 0.14 0.56 98.2 3.550 1.330 5
34 130.6 169.6 Inter. Shell long. Welds 19-894A/B 0.25 0.21 127.1 1.216 1.055
-56 66 134.0 144.0 using S/C data 74.9 1.216 1.055 56 44 79.0 67.0 I
Oire. Weld 11-894 0.225 0.20 114.5 3.550 1.330 56 66 152.3 162.3 Lower Shell Long. Welds 20-894A/B 0.17 0.20 92.0 1.216 1.055
-56 66 97.0 107.0 laitial RTmy values of the base metal plates are measured values. while the weld metal values are generic.
Margins per Regulatory Guide 1.99. Revision 2.
17 l
1
TABLE 13 PTS Calculations for Farley Unit I at 48 EFPY - Using Credible Surveillance Capsule Data Material Cu%
Ni%
CF f
FF I
M ARTm RTm 48 EFPY Inter. Shell Plate B6903-2 0.13 0.60 91.0 5.195 1.410 0
34 128 3 1623 Inter. Shell Plate B6903-3 0.12 0.56 82.2 5.195 1.410 10 34 115.9 159.9 14wer Shell Plate B6919-1 0.14 0.55 97.8 5.195 1.410 15 34 137.9 186.9 using S/C data 94.9 5.195 1.410 15 17 133.8 165.8 Lower Shell Plate B6919-2 0.14 0.56 98.2 5.195 1.410 5
34 138.4 177.4 Inter. Shell leng. Welds 19-894A/B 0.25 0.21 127.1 1.797 1.161
-56 66 147.6 157.6 using S/C data 74.9 1.797 1.161
-56 44 87.0 75.0 Cire. Weld 11-894 0.225 0.20 114.5 5.195 1.410
-56 66 161.4 171.4 Lower Shell leng. Welds 20-894A/B 0.17 0.20 92.0 1.797 1.161
-56 66 106.8 116.8 J.2IIL Initial RT, values o'.ae base metal plates are measured values, while the weld metal values are generic.
Margins per Reguitary Guide 1.99, Revision 2.
TABLE 14 PTS Calculations for Farley Unit 2 at 11.201 EFPY - Using Credible Surveillance Capsule Data Material Cu%
Ni%
CF f
FF I
M ARTm RT-11.201 EFPY Inter. Shell Plate B7203-1 0.14 0.60 100.0 1360 1.085 15 34 108.5 157.5 j
Inter. Shell Plate B7212-1 0.20 0.60 149.0 1360 1.085
-10 34 161.7 185.7 l
1 using S/C data 142.4 1360 1.085
-10 17 154.6 161.6
]
1 lower Snell Plate B7210-1 0.13 0.56 89.8 1360 1.085 18 34 97.5 149.5 1
1 lower Shell Plate B7210-2 0.14 0.57 98.7 1360 1.085 10 34 107.1 151.1 Cire. Weld 11923 0.13 0.20 76.0 1360 1.085
-40 56 4
82.5 98.5 lower Shell leng. Welds 0.05 0.20 49.0 0.463 0.785
-70 38.5 38.5 7.0 20-923A/B Inter. Shell long. Welds 19-923A 0.02 0.96 27.0 0.463 0.785 56 40.1 21.2 5.3 Inter. Shell leng. Welds 19-923B 0.02 0.93 27.0 0.463 0.785
-60 21.2 21.2
-17.6 usine S/C data 89 0 463 0.785
-60 7.0 7.0
-46 0 NOTE Initial RTer values of the beltline region materials are measured values, except for Inter. Shell Long. Weld 19-923A.
Margins per Regulatory Guide 1.99, Revision 2.
I8 I
TABLE 15 PTS Calculations for Farley Unit 2 at 32 EFPY - Using Credible Surveillance Capsule Data Matedal Cu%
Ni%
CF f
l FF 1
M ARTm RTm 32 EFW Inter. Shell Plate B7203-1 0.14 0.60 100.0 3.462 1.324 15 34 132.4 181.4 Inter. Shell Plate B7212-1 0.20 0.60 149.0 3.462 1.324
-10 34 197.3 221.3 using S/C data 142.4 3.462 1.324
-10 17 188.6 195.6 lower Shell Plate B7210-1 0.13 0.56 89.8 3.462 1.324 18 34 118.9 170.9 lower Shell Plate B7210-2 6.14 0.57 98.7 3.462 1.324 10 34 130.7 174.7 Cire. Weld 11-923 0.13 0.20 76.0 3.462 1.324
-40 56 100.6 116.6 lower Shell Long. Welds 0.05 0.20 49.0 1.289 1.071
-70 52.5 52.5 34.9 20 923A/B Inter. Shell long. Welds 19-923A 0.02 0.96 27.0 1.289 1.071
-56 44.6 28.9 17.5 Inter. Shell long. Welds 19-9238 0.02 0.93 27.0 1.289 1.071
-60 28.9 28.9
-2.2 using S/C data 89 1.289 1.071
-60 9.5 9.5
-40 9 DREL Initial RT, values of the beltline region materials are measured values, except for Inter. Shell Long. Weld 19-923 A.
Margins per Regulatory Guide 1.99, Revision 2.
TABLE 16 PTS Calculations for Farley Unit 2 at 48 EFPY - Using Credible Surveillance Capsule Data l
Material Cu%
Ni%
CF f
FF I
M ARTm RTm 48 EFPY Inter. Shell Plate B7203-1 0.14 0.60 100.0 5.079 1.405 15 34 140.5 189.5 Inter. Shell Plate B7212-1 0.20 0.60 149.0 5.079 1.405
-10 34 209.4 233.4 using S/C data 142.4 5.079 1.405
-10 17 200.1 207.1 lower Shell Plate B7210-1 0.13 0.56 89.8 5.079 1.405 18 34 126.2 178.2 lower Shell Plate B7210-2 0.14 0.57 98.7 5.079 1.405 10 34 138.7 182.7 Cire. Weld 11-923 0.13 0.20 76.0 5.079 1.405
-40 56 106.8 122.8 lower Shell long. Welds 0.05 0.20 49.0 1.925 1.179
-70 56 57.8 43.8 20-923A/B Inter. ShcIl long. Welds 19-923A 0.02 0.96 27.0 1.925 1.179
-56 46.6 31.8 22.4 Inter. Shell Long. Welds 19-923B 0.02 0.93 27.0 1.925 1.179
-60 31.8 31.8 3.7 using S/C data 8.9 1.925 1.179
-60 10.5 10.5
-39.0 M!IR Initial RTa values of the beltline region materials are measured values, except for Inter. Shell Long. Weld 19-923A.
Margins per Regulatory Guide 1.99. Revision 2.
19
mc 1
m 40 1
37 T
4
- Ill of the beltline materials in the Farley Units 1 and 2 reactor vessels have RTm values below the
-j
. FTS screening criteria at 32 and 48 EFPY. Plots of RTm versus fluence (Figures 3 and 4) also -
'{
'; illustrate the available snargm, where RTm = I + M + ARTm-l 1
.i r
l l(
'i l
- ]
i 1-i o
20
p..
SECTION
7.0 CONCLUSION
S As shown in Tables 3 through 8 and Tables 11 through 16, all RTm values remain below the NRC screening values for PTS using fluence values for the present time, end of license (32 EFPY), and 48 EFPY. Plots of the RTm values versus fluence, shown in Figures 3 and 4, illustrate the available I
margin for the most limiting material in the Parley Units 1 and 2 reactor vessel beltline regions.
j J
I 300 Screening Criteria 250 C
w 186.9 1
179.1
=
b 1
g 177 4 158.5 169.6 150 148.9 100 1.539E+19 3.550E+19 5.195E+19 Fluence (n/cm2)
_._ Limiting Mat'l B6919-1, per 10CFR50.61 Limiting Mat'l B6919-2, considering credible S/C results per RG 1.99, R2 r
Figure 3 RTers versus Fluence Curves for Farley Unit 1 Limiting Material - Lower Shell Plates B6919-1 and B6919-2 21
300 2M Screening Criteria 260
{ 240 w
233.4 gg 220 221.3 g
2 E 200 207.1 195.6 180 185.7 160 161.6 140 1.360E+19 3.462E+19 5.079E+19
]
Fluence (n/cm2) j
_._ Limiting Mat'l B7212-1, per 10CFR50.61
_._. Limiting Mat'l B7212-1, considering credible S/C results per RG 1.99, R2 Figure 4 RTm versus Fluence Curves for Farley Unit 2 Limiting Material - Intermediate Shell Plate B7212-1 1
l 22
SECTION
8.0 REFERENCES
[1]
10 CFR Pan 50, " Analysis of Potential Pressurized Thermal Shock Events," July 23,1.985.
[2]
10 CFR Part 50.61, " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," May 15,1991. (FIS Rule)
[3]
Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials,"
U.S. Nuclear Regulatory Commission, May 1988.
[4]
WCAP-8810. " Southern Alabama Power Company Joseph M. Farley Nuclear Plant Unit No.1 Reactor Vessel Radiation Surveillance Program", J. A. Davidson, et al., December 1976.
[5]
WCAP-8956, " Alabama Power Company Joseph M. Farley Nuclear Plant Unit No. 2 Reactor Vessel Radiation Surveillance Program", J. A. Davidson and S. E. Yanichko, August 1977.
[6]
Southern Nuclear Operating Company letter to the U. S. Nuclear Regulatory Commission,
" Joseph M. Farley Nuclear Plant Responses to Open Issues Regarding Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity", J. D. Woodard, dated June 21,1994.
[7]
WCAP-11047, "J. M. Farley Units 1 and 2 Reactor Vessel Fluence and RTm Evaluations", T.
Congedo, et al., January 1986.
j l
r
[8]
WCAP-14196, " Analysis of Capsule W from the Alabama Power Company Farley Unit 1 Reactor Vessel Radiation Surveillance Program", P. A. Peter, et al., February 1995.
[9]
WCAP-12471, "Analysi. 'f Capsule X from the Alabama Power Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program", E. Terek, et al., December 1989.
[10]
Per Southem Nuclear Company (Mason Dove), EFPY value as of 1/20/95, stated 1/20/95.
[11]
WCAP-14044, " Westinghouse Surveillance Capsule Neutron Fluence Reevaluation", E. P.
Lippincott, April 1994.
23 L
__