ML20082M582
| ML20082M582 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 08/31/1991 |
| From: | Albertin L, Shaun Anderson, Terek E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20082M574 | List: |
| References | |
| WCAP-13007, NUDOCS 9109050260 | |
| Download: ML20082M582 (170) | |
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WESTINGHOUSE CLASS 3 WCAP-13007 4
r ANALYSIS OF CAPSULE U FROM THE GEORGIA POWER COMPANY V0GTLE ELECTRIC GENERATING PLANT UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM E. Terek S. L. Anderson L. Albertin August 1991 Work Performed Under Shop Order GURP-106 Prepared by Westinghouse Electric Corporation for the Georgia Power Company 0
g Approved by: h' NN T. A. Meyer, Maniger Structural Reliability and Plant Life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 0 1991 Westinghouse Electric Corp.
'J PREFACE This report has been technically reviewed and verified.
4 Reviewer:
r Sections 1 through 5, 7, 8 and J. M. Chicots [ 78, a Appendix A v
Section 6 E. P. Lippincott /M'hm
__1 k' N
- f Appendix B N. K. Ray t, W
_K e
0 9
TABLE OF CONTENIS Title Eagg Section 1.0
SUMMARY
OF RESULTS 1-1 4
2.0 INTRODUCTION
2-1
- 3-1 4
3.0 BACKGROUND
4.0 DESCRIPTION
OF PROGRAM 4-1 5.0 TESTING OF SPECIMENS FROM CAPSULE U 5-1 5.1 Overview 5-1 5.2 Charpy V-Notch Impact Test Results 5-4 5.3 Tension Test Results 56 5.4 Compact Tension Tests 5-7 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6.1 Introduction 6-1 6.2 Discrete Ordinates Analysis 6-2 6.3 Neutron Dosimetry 6-7 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1
8.0 REFERENCES
8-1
.O APPENDIX A - LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS 0
APPENDIX B - EVALUATION OF PRESSURIZED THERMAL SH0CK FOR
- V0GTLE ELECTRIC GENERATING PLANT UNIT 2 il 1 l
LIST OF TABLES lable Title P_iLqt 4-1 Chemical Composition of the Vogtle Electric Generating 4-3
, Plant Unit 2 F=' actor Vessel Surveillance Material 4-2 Heat Treatment History of the Vogtle Electric Generating 4-4
- Plant Unit 2 Reactor Vessel Core Region Shell Plates and Weld Seams 4-3 Chemical Composition of Two Vogtle Electric Generating 4-5 Plant Unit 2 Charpy Specimens Removed from Surveillance Capsule U 4-4 Chemistry Results from the NBS Certified Reference Standards 4-6 5-1 Charpy V-Notch Impact Data for the Vogtle Electric Generating 5-8 Plant Unit 2 Lower Shell Plate B8628-1 Irradiated at 550*F, Fluence 4.44 x 10 18 n/cm2 (E > 1.0 MeV) 5-2 Charpy V-Notch Impact Data for the Vogtle Electric 5-9 Generating Plant Unit 2 Reactor Vessel Weld Metal and HAZ Metal Irradiated at 550*F, Fluence 4.44 x 10 18 n/cm2 (E > 1.0 MeV) 5-3 Instrumented Charpy mpact Test Results for the Vogtle 5-10 Electric Generating Plant Unit 2 Lower Shell Plate B8628-1 Irradiated at 550*F, Fluence 4.44 x 10 18 n/cm 2
." (E > 1.0 MeV)
.. 5-4 Instrumented Charpy Impact Test Results for the Vogtle 5-11 Electric Generating Plant Unit 2 Weld Metal and Heat-Affected-Zone (HAZ) Metal, Irradiated at 550*F, Fluence 4.44 x 10 18 n/cm2 (E > 1.0 MeV) iii
LIST OF TABLES Table Title P_iLqn 5-5 Effect of 550*T Irradiation to 4.44 x 1018 n/cm 2
5-12 (E > 1.0 MeV) on the Notch Toughness Properties of the Vogtle Electric Generating Plant Unit 2 Reactor Vessel Surveillance Materials 5-6 Comparison of the Vogtle Electric Generating Plant Unit 2 5-13 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 E? i: ion 2 Predictions 5-7 Tensile Properties for the Vogtle Electric Generating 5-14 Plant Unit 2 Reactor Vessel Surveillance Materials Irradiated at 550*F to 4.44 x 1018 n/cm2 (E > 1.0 MeV) 6-1 Calculated Fest Neutron Exposure Parameters at the 6-14 Surveillance Capsule Center 6-2 Calculated Fast Neutron Exposure Rates at the 6-15 Pressure Vessel Clad / Base Metal Interfacc 6-3 Relative Radis ' Distributions of Neutron Flux 6-16 (E > 1.0 MeV) within the Pressure Vessel Wall 6-4 Relative Radial Distributions of Neutron Flux 6-17 (E > 0.1 MeV) within the Pressure Vessel Wall ..
6-5 Relative Radial Distributions of Iron Displacement Rate 6-18 (dpa) within the Pressure Vessel Wall .
6-6 Nuclear Parameters for Neutron Flux Monitors 6-19 iv i
LIST OF TABLES P
Iahle Tit 1e hgg ;
6-7 Monthly Thermal Generation During the First Fuel 6-20 Cycle of the Vogtle Electric Generating Plant i Unit 2 Reactor 6-8 Measured Sensor Activities and Reactions Rates 6-21 6-9 Summary of Neutron Dosimetry Results 6-23 6-10 Comparison of Measured and FERRET Calculated Reaction 6-24 Rates at the Surveillance Capsule Center 6-11 Adjusted Neutron Energy Spectrum at the Surveillance 6-25 Capsule Center
~
6-12 Comparison of Calculated and Measured Exposure Levels 6-26 for Capsule U 6-13 Neutron Exposure Projections at Key Locations on the 6-27 Pressure Vessel Clad / Base Metal Interface 6-14 Neutron Exposure Values for use in the Generation of 6-28 Heatup/Cooldown Curves 6-15 Updated Lead Factors for Vogtle Electric Generating 6-29 Plant Unit 2 Surveillance Capsules V
LIST OF ILLUSTRATIONS Fioure Iltle fjLat 4-1 Arrangement of Surveillance Capsules in the Vogtle 4-7 Electric Generating Plant Unit 2 Reactor Vessel '
'i 4-2 Capsule U Diagram Showing Location of Specimens, Thermal 4-8 Monitors and Dosimeters ' ' '
5-1 i Charpy V-Notch Impact Properties for Vogtle Electric 5-15 {
Generating Plant Unit 2 Reactor Vessel Lower Shell Plate B8628-1 (Longitudinal Orientation) 5-2 Charpy V-Notch Impact Properties for Vogtle Electric 5-16 1
Generating Plant Unit 2 Reactor Vessel Lower Shell Plate B8628-1 (Transverse Orientation) 5-3 Charpy V-Notch Impact Properties for Vogtle Electric 5-17 Generating Plant Unit 2 Reactor Vessel Surveillance i
Weld Metal i
5-4 Charpy V-Notch Impact Properties for Vogtle Electric 5-18 Generating Plant Unit 2 Reactor Vessel Weld Heat-Affected-Zona Metal 5-5 Charpy Impact Specimen Fracture Surfaces for Vogtle 5S Electric Generating Plant Unit 2 Reactor Vessel Lower (
Shell Plate B8628-1 (Longitudinal Orientation) 5-6 Charpy impact Specimen Fracture Surfaces for Vogtle 5-20 Electric Generating Plant Unit 2 Reactor Vessel Lower Shell Plate BS628-1 (Transverse Orientation) vi
.- -- .- . - . - - - - - _ - - - _ _ _ =
LISI 0F ILLUSTRATIONS (Cont)
, Fioure Title Eagg
.- 5-7 Charpy impact Specimen fracture Surfaces for Vogtle 5-21 Electric Generating Plant Unit 2 Reactor Vessel 3
Surveillance Weld Metal
, 5-8 Charpy Impact Specimen Fracture Surfaces for Vogtle 5-22 Electric Generating Plant Unit 2 Reactor Vessel Weld Heat-Affected-2one Metal 5-9 Tensile Properties for Vogtle Electric Generating Plant 5-23 Unit 2 Reactor Vessel Lower Shell Plate B8628-1 (Longitudinal Orientation) 5-10 Tensile Properties for Vogtle Electric Generating Plant 5-24 Unit 2 Reactor Vessel Lower Shell Plate B8628-1 (Transverse Orientation) 5-11 Tensile Properties for Vogtle Electric Generating Plant 5-25 Unit 2 Reactor Vessel Surveillance Weld Metal 5-12 Fractured Tensile Specimens from Vogtle Electric Generating 5-26 J Plant Unit 2 Reactor Vessel Lower Shell Plate B8628-1 -
(longitudinal Orientation)
, 5-13 Fractured Tensile Specimens from Vogtle Electric Generating 5-27 l Plant Unit 2 Reactor Vessel Lower Shell Plate B8628-1 l
- (Transverse Orientation) ,
5-14 . Fractured Tensile Specimens from Vogtle Electric Generating 5-28 Plant Unit 2 Reactor Vessel Surveillance Weld Metal t
i vil
LIST OF ILLUSTRATIONS (Cont) !
f_igur_t Tit 1e "
fjtqe 5-15 Engineering Stress-Strain curves for P? ate B8628-1 '
5 Tensile Specimens BL1 and BL2 (Longitudinal Orientation)- ,
S-16 Engineering Stress-Strain Curves for Plate B8628-1 5-30 Tensile Specimens BL3 (longitudinal Orientation) and BTl '
(Transverse Orientation) i i
5-17 Engineering Stress-Strain Curves for Plate B8628-1 5-31 Tensile Specimens BT2 and BT3 (Transverse Orientation) 5-1B Engineering Stress-Strain Curves for Weld Metal i 5-32 Tensile Specimens BW1 and BW2 i
5-19 Engineering Stress-Strain Curve for Weld Metal 5-33 Tensile Specimen BW3 6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13 i
t e
t e
4 e
viii ,
I SECTION 1.0 I
SUMMARY
OF RESUlls The analysis of the reactor vessel materials contained in surveillance Capsule U, the first capsule to be removed from the Georgia Power Company Vogtle Electric Generating Plant Unit 2 reactor pressure vessel, led to the following conclusions:
o The capsule received an average fast neutron fluence (E > 1.0 MeV) of 4.44 x 10 18 n/cm2 af ter 1.18 EFPY of plant operation.
c Irradiation of the reactor vessel lower shell plate B8628-1 Ciiarpy specimens, oriented with the longitudinal axis of the specimen parallel to the major rolling direction (longitudinal orientation),
to 4.44 x 10 18 n/cm2 (E > 1.0 MeV) resulted in no 30 ft-lb transition temperature increase and in a 50 ft-lb transition temperature increase of 5'F. This results in a 30 f t-lb transition temperature of -10'F and a 50 ft-lb transition temperatura of 50*F for longitudinally oriented specimens, o Irradiation of the reactor vessel lower shell plate B8628-1 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major rolling direction (transverse orientation), to 4.44 x 1018 n/cm2 (E > 1.0 MeV) resulted in no 30 ft-lb transition temperature increase and in a 50 ft-lb transition temperature increase of 5'F. This results in a 30 ft-lb transition o
temperature of 30*F and a 50 ft-lb transition temperature of 80*F for transversely oriented specimens.
o The weld metal Charpy specimens irradiated to 4.44 x 10 18 n/cm 2
.. (E > 1.0 MeV) resulted in no 30 and 50 ft-lb transition temperature increases. This results in a 30 ft-lb transition temperature of
-15*F and a 50 ft-lb transition temperature of 5'F for the weld metal.
1-1
. -- - __ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ i
.. - - - ~. .
t-o Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal s Charpy specimens to 4.44 x 10 18 n/cm2 (E > 1.0 MeV) re'sulted in i no 30 and no 50 ft-lb transition temperature increases. This results
in a 30 ft-lb transition temperature of -80*f and a 50 ft-lb transition temperature of -45'F for the weld HAZ metal, f ':
o' The average upper shelf energy of lower shell plate B8628-1 ;
(longitudinal orientation) resulted in a energy increase of 10 ft-lb '
after irradiation to 4.44 x 1018 n/cm2 (E > 1.0 MeV). This results in an upper shelf energy of 99 ft-lb for longitudinally !
oriented specimens.
o The average upper shelf energy of lower shell plate B8628-1 (transverse orientation) resulted in a energy increase of 9 ft-lb +
after irradiation to 4.44 x 1018 n/cm2 (E > 1.0 fieV) . This results in an upper shelf energy of 79 f t-lb for transversely oriented specimens.
L J
o The average upper shelf energy of the weld metal increased 6 ft-lb after irradiation to 4.44 x 1018 n/cm2 (E > 1.0 MeV). This j results in an upper shelf energy of 98 f t-lb for the weld metal.
o The average upper shelf energy of the weld HAZ metal increased 16 ft-lb after irradiation to 4.44 x 1018 n/cm2 (E > 1.0 MeV). This results in an upper shelf energy of 122 ft-lb for the weld HAZ metal, o The surveillance capsule V test results indicate that the surveillance material 30 ft-lb transition temperature changes and upper shelf energy decreases are less than the Regulatory Guide 1.99 Revision 2 predictions.
- e-1-2 I
(~;..
I l
o The surveillance capsule materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are expected to maintain an upper shelf energy of no less than 50 ft-lb
. throughout the life (32 EFPY) of the vessel as required by 10CFR50, Appendix G.
a o The calculated etid-of-life (32 EFPY) maximum neutron fluence (E > 1.0 MeV) for the Vogtle Electric Generating Plant Unit 2 reactor vessel is as follows:
Vessel inner radius * - 3.04 x 10 19 n/cm 2
19 2 Vessel 1/4 thickness - 1.66 x 10 n/cm 18 2 Vessel 3/4 thickness - 3.59 x 10 n/cm Clad / base metal interface o All RTPTS values remah below the NRC screening values for PTS.
The PTS values for the limiting beltline region lower shell plate B8628-1 for 32 EFPY and 48 EFPY are 124'F and 127'F,
- respectively.
9 e
1-3
- - --- _ _ _ _ _ _ _ _ _ ._ .l
SECTION 2.0 !
INTRODUCTION :
. t This report presents the results of the examination of Capsule U, the first <
e capsule to be removed from the reactor in the continuing surveillance program ,
which monitors the effects of neutron irradiation on the Georgia Power Company Vogtle Electric Generating Plant Unit 2 reactor pressure vessel materials under actual operating conditions.-
The surveillance program for the Georgia Power Company Vogtle Electric Generating Plant Unit 2 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials is presented in PCAP-ll381 entitled " Georgia Power
- Company Vogtle Electric Generating Plant Unit 2 Reactor Vessel Radiation Surveillance Program" by L. R. Singer Ill. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was h based on ASTM E185-82, " Standard Practice for Conducting Surveillance Tests for
. Light-Water Cooled Nuclear Power Reactor Vessels". Westinghouse Power Systems personnel were contracted to aid in the preparation of procedures for removing capsule "U" from the reactor and its shipment to the Westinghouse Science and Technology Center Hot Cell Facility, where, the postirradiation mechanical !
testing of the Charpy V-notch impact and tensile surveillance specimens was !
performed.
This-report summarizes the testing of and the postirradiation data obtained ,
from surveillance capsule "U" removed from the Georgia Power Company Vogtle Electric Generating Plant Unit 2 reactor vessel and discusses the analysis of
- the data. !
l i
l i
2-1 :
l
SECTION
3.0 BACKGROUND
The ability of the large steel pressure vessel containing the reactor core and
, its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the inost critical region of the vessel because it is
~
subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as A533 Grade B Class 1 (base material of the Georgia Power Company Vogtle Electric Generating Plant Unit 2 reactor pressure vessel lower shell plate B8628-1) are well documented in the literature.
Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a dacrease in ductility and toughness under certain conditions of irradiation.
A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in Protection Against Nonductile f ailure,"
Appendix G to Section 111 of the ASME Boiler and Pressure Vessel Code W .
The method uses fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT)-
RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208) or the temperature 60*F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the plate. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G to the ASME Code. The KIR curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the KIR curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowabie operating limits can then be determined using these allowable stress intensity factors.
3-1 1
RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement changes in mechanical propecties of a given reactor pressure vessel steel can be monitored by a reactor surveillance !
program, such as the Vogtle Electric Generating Plant Unit 2 Reactor Vessel '
Radiation Surveillance Programill, in which a surveillance capsule is ,j periodically removed from the operating nuclear reactor and the encapsulated ;
specimens are tested. The increase in the average Charpy V-notch 30 ft-lb l
temperature (ARTNDT) due to irradiation is added to the original RTNDT to adjust the RTNDT for radiation embrittlement. This adjusted RTNDT (RTNDT initial + ARTNDT) is used to index the material to the KIR curve and, in turn, to set operating limits for the nuclear power plant which i take into account the effects of irradiation on the reactor vessel materials.
h
)
e[
i f
3-2 r
SECTI,0N 4,0 ;
DESCRIPT")N OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Vogtle Electric Generating Plant Unit 2 reactor pressure vessel core region l material were inserted in the reactor vessel prior to initial plant start-up. f
. The six capsules were positioned in the reactor vessel between the neutron pads ;
and the vessel wall as shown in Figure 4-1. The vertical center of the f
capsules is opposite the vertical center of the core. ;
Capsule U was removed after 1,18 Effective full Power Years (EFPY) of plant ,
operation. This capsule contained Charpy V-notch, tensile, and 1/2 T compact {'
tension (CT) specimens (Figure 4-2) from the lower shell plate B8628-1 and submerged arc weld metal identical to the weldment used in the girth seam I between the intermediate and lower shell plates and Charpy V-notch specimens i from weld HAZ material.
Test material obtained from the lower shell plate (after the thermal heat l treatment and forming of the plate) was taken at least one plate thickness from -
the quenched ends of the plate. All test specimens were machined from the 1/4 "nickness (a] location of the plate after performing a simulated postweld, :
stress-relieving treatment on the test material and also from weld and !
heat-affected-zone (HAZ) metal of a stress-relieved weldment joining lower shell plate B8628-1 and adjacent lower shell plate B8825-1. All HAZ test l specimens were taken from weld and weld HAZ of lower shell plate B8628-1. ;
Base metal Charpy V-notch impact and tension' specimens were oriented with the longitudinal axis of the specinen parallel to the major rolling direction of I the plate (longitudinal orientation) and also normal to the major rolling j direction (transverse orientation). Charpy V-notch and tensile specimens from [
the weld metal were oriented such that the long dimension of the specimen was '
normal to the welding direction. j
- a. The compact test specimens were obtained from the 3/4 T-thickness location of the plate. !
4-1 t'
Capsule U, al<,0, contained 1/2T CT test specimens from the lower shell' plate B8628-1 and were machined in both the longitudinal and transverse orientations. The 1/2T CT Test specimens from the weld metal were machined '
with the notch oriented in the directior, of welding, All CT specimens were fatigue precracked according to ASTM E399.
The chemice.1 composition and heat treatment of the surveillance material is presented in Tables 4-1 threugh 4-4. The chemical analysis reported in Table 4-1 was obtained from unirradiated material used in the surveillance programill, In addition, a chemical analysis using Inductively Coupled Plasma Spectrometry (ICPS) was performed on irradiated Charpy specimens BL-1 (plate meterial) and BW-2 (weld metal). The chemistry results from the NBS certified reference standards are reported in Table 4-4.
Capsule U contained dosimeter wires of pure copper, iron, nickel, and aluminum-0.15 weight percent cobalt wire (cadmium-shielded and unshielded). In addition, cadmium shielded dosimeters of neptunium (Np237) and uranium (U238) were placed in the capsule to measure the integrated flux at specific
~
neutron energy levels.
Thermal monitors made from the two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule. These thermal monitors were are used to define the maximum temperature attained by the test specimens during irradiation. The composition of the two alloys and their melting points are as follows:
2.5% Ag, 97.5% Pb Melting Point: 579'F (304*C) ,
1.5% Ag, 1.0% Sn, 97.5% Pb Melting Point: 590*F (310*C)
The arrangement of the various mechanical specimens, dosimeters and thermal ,
monitors contained in capsule U are shown in Figure 4-2.
4-2
TABLE 4-1 CHEMICAL COMPOSITION OF THE V0GTLE ELECTRIC GENERATING ,
PLANT UNIT.2 REACTOR VESSEL SURVEILLANCE MATERIAL Chemical Composition (wt%) !
Element Lower Shell Plate B8628-1 Weld Metal (d) ;
. C 0.24 3 0.23 3 0.0753 0.099)
Mn 1.34l 1.30l 1.27l 1.25l P 0.007l 0.007l 0.007l 0.008l S 0.016l 0.014l 0.010l 0.013l Si 0.25l 0.23l 0.50l 0.43l Ni 0.59l 0.59l 0.12l 0.17l 7 Mo 0.59l 0.50l 0.52l 0.47l I I
Cr 0.02l(a) 0.07l(b) 0.07l(a) 0.061l(b)
Cu 0.05l 0.05l 0.06l 0.040l Al 0.029l 0.034l l 0.015l Co 0.004l 0.008l l 0.002l l Pb (c)l <0.07l --
l <0,01l l W <0.01l <0.05l --
l <0.01l .
Ti <0.01l 0.005l l <0.001l i
. Zr <0.001l 0.03l --
l <0.01l V 0.004l <0.005l 0.004l <0.004l !
Sn 0.017l 0.007l l <0.001l As 0.007l 0.008l l 0.003l +
Cb <0.01-l <0.05l --
l <0.002l ;
N 0.008l 0.007l l 0.002l .
B <0.001) 0.008) -J 0.009) i I
- a. Chemical Analysis by Combustion Engineering, Inc.Ill ,
i* b. Chemical Analysis by Westinghouseill
- c. Not detected !
- d. Representative of the closing girth seam. Weld wire Heat No. 87005, Linde 124 Flux Lot No. 1051.
?
4-3 f
TABLE 4-2 i HEAT TREATMENT HISTORY OF THE V0GTLE ELECTRIC GENERATING PLANT UNIT 2 l REACTOR VESSEL CORE REGION SHELL PLATES AND WELD SEAMSill Temperature Time -
Material (*F) (hr) Cooling !
4~ustenitizing: 4 I'l Water-quenched 1600 i 25 Intermediate (871
- C)
Shell Plates Tempered: 4 I*I Air cooled R-4 1 1225 i 25 R-4-2 (663*C) .
R43 Stress Relief: 16.5@l Furnace-coolet.
1150 i 50 (621 *C) 4ustenitizing: 4 I"I Water quenched 1600 t 25 Lower (871 'C)
Shell Plates Tempered: 4 l*I Air cooled B88251 1225 1 25 R-8-1 (663* C) -
B86281 Stress Relief: 12.0PI Furnace-cooled 1150 50 (621 *C) ,
Intermediate Shell Longitudittal Stress Relief: 16.5N1 Furnace-cooled Seam Welds 1150 1 50 -
~ (621 *C) - - - -
Lower Shell Longitudinal 12.0@l Furnace-cooled Seam Welds Local Intermediate to Stress Relief: 5.0 Furnace-cooled Lower Shell Girth 1150 50 Seam Weld (621 *C)
_DC ,
surveillance Program Test Material Surveillance 7 Program ,
,'ese d, , Stress Relief: 6.0 *I I
Furnace-cooled c.osing G.rm suami 1150 t 50 -
(621 *C) a Luens Steel CompanyTComoustion Engineenng, inc. Certibcation Reports b Stress Aeket includes the intermediate to lower Shell Closing G rth Seam Post Weld Heat Treatment.
c The Stress Rehet Heat Treatment received by the Surveillance Test Weidment has been simulated
! 4-4 4
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TABLE 4-3
{
Chemical Compostion of Two Vogtle Electric Generating Plant Unit 2 Charpy Specimens Removed from Surveillance Capsule U i
.- i Concentration in Weigh 1 Percent l l
e Specimen No. BL-1 BW-2 ,
Material PLATE WELD !
Fe KATRIX ELEMENT: Remainder by Difference ,
Mn 1.199 1.111 ;
Cr 0.061 0.061 l
Ni 0.598 0.091 j Mo 0.546 0.529 ;
Co 0.020 0.025 l Cu 0.053 0.045 P 0.0084 0.0124 .
V <0.010 <0.010 .l C 0.181 0.074 j S 0.0098 0.0106 ;
Si 0.201 0.484 [
i f
o Analyses Method of Analysis Metals IrPS, Inductively coupled Plasma Spectrometry i l.* Carbon EC-12, LECO Carbon Analyzer -
Sulfur Combustion / titration l Silicon Dissolution / gravimetric
,- j i
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4-5 I .
3 n -
w
TABLE 4-4 Chemistry Results from the NBS Certified Reference Standards -
Material ID Low Alloy Steel: NBS Certified Reference Standards NBS 361 NBS 362 Certified Measured Certified Measured Metals Concentration in Weicht Percent Fe
- 95.60 (matrix) 95.30 (matrix) I Mn 0.66 0.632 1.04 1.036 .!
Cr 0.694 0.690 0.30 0.314 Ni 2.00 above cal 0.59 0.601 Mo 0.19 0.212 0.068 0.069 ;
Co 0.032 0.040 0.30 0.336 Cu 0.042 0.042 0.50 0.495 P 0.014 0.0184 0.041 0.0384 V 0.011 0.011 0.040 0.039 C 0.383 0.378 0.160 0.162 S 0.014 N.A. 0.036 0.0364 i Si 0.?22 0.203 0.39 N.A. T Material 10 Low Alloy Steel: NBS Certified Reference Standards NQS 363 NBS 364 Certified Measured Certified Measured Metals Concentration in Weiaht Percent Fe 94.4 (matrix) 96.7 (matrix) ,
Mn 1.50 1.500 0.255 0.242 ,
Cr 1.31 1.358 0.063 0.063 Ni 0.30 0.275 0.144 0.101 Mo 0.028 0.028 0.49 0.490 Co 0.048 0.025 0.15 0.157 '
Cu 0.10 0.100 0.249 0.243
- P 0.029 0.0361 0.01 0.0104 V 0.31 0.310 0.105 0.111 0 0.62 N.A. 0.87 N.A.
S 0.0068 N.A. 0.0250 0.0197 >
Si 0.74 0.724 0.065 N.A.
Matrix element calculated as difference for material balance. ( )
N.A. - Not analyzed 4-6
O' REACTOR VESSEL I
p~ ~ -- CORE BARREL j
. . g y NEUTRON PAD CAPSULE U (58 5')
~
L (301.5') Z I 1 f+"$8.5 ' l
- 58.5' N [ :
/' / )
o 61' ,
270' - ----
90' f
-/l
/
(241 ') y
\ l
/
X- ~2 (238.5')
~
REACTOR VESSEL 180' s
~
PLAN VIEW
- VESSEL s
r s ,/~~ W ALL I l L k CAPSULE
^
CORE N
{f!Il!\ll!llE
..) .
CORE MIDPLANE g [ s s i l T , N
' /
[ Q i L D NEUTRON PAD
) ts k '
/ t - CORE BARREL ELEVATION VIEW Figure 4-1. Arrangement of Surveillance Capsules in the Vogtle Electric Generating Plant Unit 2 Reactor Vessel 4-7
t'
"' us wacn_ en cum, .n e. ,,, ,,, ,, _
em e. is wie me
>9 em e.
em w.
E e em em e*, m. >,4 ea u en em we me u u mi .c n,. em em e., mi wo me >= w- en w. ewi LEGEND:BL . l.OWER SHELL PLATE 080281 (LONGITUDIN AL)
BT LOWER SHELL PLATE B80281 (TRANSVERSE)
BW WELD METAL BH HEAT AFFECTED-ZONE MATERIAL O
e
- '* tuir s , ,
-_m__m.-_m-____._____m__________ _ _ _ _ _ _ _ _ , , _
i-1 I
I i
I
.,,, l
, , m. m. . - = . e -me i ===as j u n,. nii me mir m u nj _
me mI u
m -
l mj
=
= n,. . n, .,, m u u m u ,,. m . .,, m t
.. m, au m, . m, u ... ; .. n, *, .1, l t t
1 i
SI l '
APERTURE CARD .
[
i Aloe lvadlable On ?
Apertwre Card }!
l il Figure 4-2 Capsule U Diagram Showing j location of Specimens, {
Thermal Monitors and i Dosimeters v
1 0l090SOZloO b) l
- 4-0 #
.---~~
, s
. , _ . . - , , , , , . _ , , _ , - . > - . , _ _ , , - - - , - , , , , , , , . - -.--.-.,,,,,,,,,,c, v,-.. x~, , _ ---
, . , , - , - - , , - - - ~ , _ . ,
I . SECTION 5.0 TESTINGOFSPECIMENSFROMCAPSULEU i l
5.1 Overview i
- i. l The post-irradiation mechanical testing of the Charpy V-notch and tensile
)
- specimens was performed at the Westinghouse Science and Technology Center hot 4
cell with consultation by Westinghouse Power Systems personnel, lesting was performed in accordance with 10CFR50, Appendices G and HI23, AS1H Specification E185-82[63, and Westinghouse Remote MetallograPhic facility i
! (RMF) Procedure 8402, Revision 1 as modified b; RMF Proces1ures 8102, Revision 1 and 8103, Revision 1.
Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and 2 checked against the master list in WCAP-ll381 Ill. No discrepancies were found.
Examination of the two low-melting point 579'F (304*C) and 590*F (310'C) eutectic alloys indicated no melting of either type of thermal
. monitor. Based on this examination, the maximum temperature to which the test specimens were e nosed was less than 579'F (304'C).
The Charpy impact tests were performed per ASTM Specification E23-88I73 and RMF Procedure 8103, Revision 1 on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy machine is instrumented with an Effects Technology ;
Model 500 instrumentation system. With this system, load-time and energy-time
, signals can be recorded in addition to the standard measurement of Charpy energy (ED ). From the load-time curve (Appendi_x A), the load of general yielding (Pgy), the time to general yleiding (tcy), the maximum load (Pg), and the time to maximum load (tg) can be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (Pp), and the load at which fast fracture terminated is identified as the arrest load (PA)*
5-1
The energy at maximum load (Eg) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is roughly ;
~
equivalent to the energy required to initiate a crack in the specimen. t Therefore, the propagation energy for the crack (E p
) is the difference ,
between the total energy to fracture (E )Dand the energy at maximum load. ;
The yield stress (oy) was calculated from the three-point bend formula having the following expression: .
. i oy - Pgy * (L/[B*(W-a)2*C)) (1) where L - distance between the specimen supports in the impact testing machine; 8 = the width of the specimen measured parallel to the notch: W = height of the specimen, measured perpendicularly to the notch; a = notch depth. The constant C is dependent on the notch flank angle (d), notch root radius (p), and the type of loading (i.e., pure bending or three-point bending).
In three-point bending a Charpy specimen in which p = 45' and p =
0.010", Equation 1 is valid with C = 1.21. Therefore (for L = 4W),
oy - Pgy * (L/[B*(W-a)2*l.21)) = [3.3P GYW )/[B(W-a)2] (2)
For the Charpy specimens, B = 0.394 in., W = 0.394 in., and a = 0.070 in.
Equation 2 then reduces to: j L
oy = 33.3 x Pay (3) l where oy is in units of psi and Pgy is in units of lbs. The flow
! stress was calculated from the average of the yield and maximum loads, also using the three-point bend formula. =.[
Percent shear was determined from post-fracture photographs using the ;
ratio-of-areas methods in compliance with ASTM Specification A370-89[8), [
The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.
5-2 ;
r
_ . _ - . . ___ _. _ _ __ _ ._ . _. - _ . - - _~., __ _ . _ _ _ . _ _ _ _ _
Tension testr w m w formed on a 20,000-pound instron Model 1115, split-console test machine, per ASTM Specification E8-89bl91 and E21-79 (1988)(10), and RMF procedure 8102, Revision 1. All pull rods, grips, and
.. pins were made of Inconel 718 hardened to HRC45. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests !
- were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.
Deflection measurements were made with a linear variable displacement ,
transducer (LVOT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-85(Ill. -
Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air. ;
Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperature.
. Chromel-alumel thermocouples were inserted in shallow holes in the center and i each end of the gage section of a dummy specimen and in each grip. In the test configuration, with a slight load on the specimen, a plot of specimen -
temperature versus upper and lower grip and controller temperatures was developed over the range of room temperature to 550'F (288'C). The <
upper grip was used to control the furnace temperature. During the actual testing the grip temperatures were used to obtain desired specimen !
, temperatures. Experiments indicated that this method is accurate to 12'F. j The yield load, ultimate load, fracture load, total elongation, and uniform [
- elongation were determined directly from the load-extension curve. The yield !
strength, ultimate strength, and fracture strength were calculated using the 7 original cross-sectional area. The final diameter and final gage length were ;
determined from post-fracture photographs. The fracture area used to calculate ;
the fracture stress (true stress at fracture) and percer' reduction in area was [
computed using the final diameter measurement. f 5-3 !
i m -u--+-w ' ' % yr *T - -*Wr-tr'= ? 7 +t-M-17 *y-g'*T'Wt t- uT --'-
7--7+w + N-- S'- = s g-, , ,
5.2 Charmy V-Notch Impact Test Resultj The results of the Charpy V-notch inipact tests performed on the various materials contained in Capsule V, which was irradiated to 4.44 x 10 18 n/cm 2
^
(E > 1.0 MeV), are presented in Tables 5-1 through 5-4 and are compared with unirradiated resultslll as shown in Figures 5-1 thrcugh 5-4. The transition ,
temperature increases and upper shelf energy decreases for the Capsule U materials are summarized in Table 5-5.
Irradiation of the reactor vessel lower shell plate B8628-1 Charpy specimens oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal orientation) to 4.44 x 10 18 ;
n/cm2 (E > 1.0 Mev) at 550*f (figure 5-1) resulted in no 30 ft-lb transition temperature increase and in a 50 ft-lb transition temperature increase of 5'F. This resulted in the 30 ft-lb transition temperature remaining at 10*f and a 50 ft-lb transition temperature of 50*f ,
(longitudinal orientation).
The average upper shelf energy (USE) of the lower shell plate B8628-1 Charpy specimens (longitudinal orientation) resulted in a energy increase of 10 ft-lb ,
after irradiation to 4.44 x 1018 n/cm2 (E > 1.0 MeV) at 550*F. This results in an average USE of 99 ft-lb (Figure 5-1).
Irradiation of the reactor vessel lower shell plate B8628-1 Charpy specimens oriented with the longitudinal axir of the specimen normal to the major rolling direction of the plate (transverse orientation) to 4.44 x 1018 nicm2 (E > ;
1.0 MeV) at 550'f (Figure 5-2) resulted in no 30 ft-lb transion temperature increase and in a 50 ft-lb transition temperature increase of 5'F. This resulted in the 30 ft-lb transition temperature remaining at 30*f and a 50 ft-lb transition temperature of 80'F (transverse orientation).
- 5 5-4
The average upper shelf energy (USE) of the lower shell plate B8628-1 Charpy specimens (transverse orientation) resulted in an energy increase of 9 ft-lb ,
after irradiation to 4.44 x 10 18 n/cm2 (E > 1.0 MeV) at 550'F. (
This resulted in an average USE of 79 ft-lb (Figure 5-t). ,
Irradiation of the reactor vessel core region weld metal Charpy specimens to 4.44 x 1018 n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-3) resulted in no 30 and 50 ft-lb transition temperature increases. The 30 ft-lb transition temperature remained at -15'F and the 50 ft-lb transition temperature remained at 5'F i
The average upper shelf energy (USE) of the rcactor vessel core region weld metal resulted in an energy increase of 6 ft-lb after irradiation to 4.44 x ;
1018 n/cm2 (E > 1.0 MeV) at 550*F. This resulted in an average USE of ;
98 ft-lb (Figure 5-3).
Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal specimens .
to 4.44 x 1018 n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-4) resulted in no 30 and 50 ft-lb transition temperature increases. The 30 ft-lb transition I
temperature remained at -80'F and the 50 ft-lb transition temperature remained at -45'F ,
i The average upper shelf energy (USE) of the reactor vessel weld HAZ metal experienced an energy increase of 16 ft-lb after irradiation to 4.44 x 10 18 n/cm2 (E > 1.0 MeV) at 550'F. This resulted in an average USE of 122 ft-lb (Figure 5-4).
t '
The fracture appearance of each irradiated Charpy specimen from the various materials is shown in figures 5-5 through 5-8 and show an increasingly ductile ;
,o or tougher appearance with increasing test temperature, j I
5-5
A comparison of the 30 ft-lb transition temperature increases and upper shelf energy decreases for the various Vogtle Electric Generating Plant Unit 2 surveillance materials with predicted vilues using the methods of NRC Regulatery Guide 1.99, Revision 2(33 is presented in Table 5-6. This comparison indicates that the transition temperature increases and the USE decreases resulting from irradiation to 4.44 x 10 18 n/cm2 (E > 1.0 MeV) are .
less than the Regulatory Guide predictions.
I
~
The load-time records for the individual instrumented Charpy specimen tests are shown in Appendix A.
5.3 Tension Test Results l
The results of the tension tests performed on the various materials contained in capsule U irradiated to 4.44 x 10 18 n/cm2 (E > 1.0 MeV) are presented in Table 5-7 and are compared with unirradiated resultsill as shown in Figures 5-9 through 5-11.
1he results of the tension tests performed on the lower shell plate B8628-1 ,
(longitudinal orientation) indicated that irradiation to 4.44 x 1018 n/cm 2
(E > 1.0 MeV) at 550*F caused less than a 3 ksi increase in the 0.2 percent offset yield strength and less than a 3 ksi increase in the ultimate tensile strength when compared to unirradiated datall) (Figure 5-9).
The results of the tension tests performed on the lower shell plate B8628-1 (transverse orientation) indicated that irrad.ation to 4.44 x 1018 n/cm2 (E t
> 1.0 MeV) at 550*F caused less than a 3 ksi increase in the 0.2 percent offset yield strength and less than a 2 ksi increase in the ultimate tensile strength when compared to unirradiated datalll (Figure 5-10).
- 5-6
. - _ . 1
The results of the tension tests performed on the reactor vessel core region l8 weld metal indicated that irradiation to 4.44 x 10 n/cm2 (E > 1.0 MeV) at 550*F caused less than a 5 ksi increase in the 0.2 percent offset yield strength and less than a 4 ksi increase in the ultimate tensile strength when compared to unirradiated dataill (Figure 5-11).
The small increases in 0.2% yield strength and tensile strength exhibited by the lower shell platt B8628-1 and the weld metal indicate that this material is not highly sensitive to irradiation to 4.44 x 10 18 n/cm2 (E > 1.0 MeV), as is also indicated by the Charpy impact test results.
The fractured tension specimens for the lower shell plate B8628-1 material are shown in Figures 5-12 and 5-13, while the fractured specimens for the weld metal are shown in Figure 5-14.
The engineering stress-strain curves for the tension tests are shown in Figures 5-15 through 5-19.
5.4 Comoact Tension Tests Per the surveillance capsule testing program with Georgio Power Compariy,1/2 T-compact tension fracture mechanics specimens will not be tested and will be stored at the Westinghouse Science and Technology Center.
5-7
TABLE 5-1 CHARPY V-NOTCH IMPACT DATA FOR THE V0GTLE ELECTRIC GENERATING PLANT UNIT 2 LOWER SHtLL PLATE 88628-1 1RRADI ATED AT 550*f, FLVENCE 4.44 x 10 18 n/cm2 (E > 1.0 MeV)
Temperature Impact Energy Lateral Expansion Shear Sample No. O'F ('0) (ft-lb) 121 (sils) (mm) (%) ,
Longitudinal Orientation BL9 -75 -
5.0 7.0 5.0 0.13 5 BL10 -45 -
14.0 19.0 10.0 0.25 10 BL12 -15 -
12.0 16.5 9.0 0.23 10 BL1 0 -
36.0 49.0 27.0 0.69 25 BLS 15 -
36.0 49.0 27.0 0.69 25 BL2- 25 49.0 66.5 34.0 0.86 40 BLIS 40 48.0 65.0 40.0 0.86 40 BL8 50 31.0 42,0 32.0 0.81 40 BL5 65 51.0 69.0 36.0 0.91 50 BL7 80 54.0 73.0 44.0 1.12 55 BL4 100 74.0 100.5 49.0 1.24 70 BL11 125 73.0 99.0 50.0 1.27 70 BL14 150 102.0 138.5 74.0 1.88 100 -
BL13 200 97.0 131.5 68.0 1.73 100 BL3 250 1 99.0 134.) 65.0 1.65 100 Transverse Orientation BT5 -75 -
6.0 8.0 4.0 0.10 5 BT1 -50 -
12.0 16.5 10.0 0.25 10 BT12 -25 -
17.0 23.0 12.0 0.30 10 BT13 0 -
23.0 31.0 22.0 0.56 15 BT3 25 36.0 49.0 30.0 0.76 30 BT7 35 38.0 51.5 29.0 0.74 35 BT14 50 40.0 54.0 34.0 0.86 40 BT8 65 45.0 61.0 38.0 0.97 45 BT10 80 43.0 58.5 38,0 0.97 60 -
BTS 100 50.0 68.0 42.0 1.07 65 BT2 120 66.0 89.5 55.0 1.40 80 BT9 150 65.0 88.0 52.0 1.32 95 BT11 200 74.0 100.5 61,0 1,55 100 -
BT4 275 1 82.0 111.0 55.0 1.40 100 BT15 375 1 81.0 110.0{ , 54.0 1.37 100 5-8
= ___ _________ __ _ - -. - - _. .-.
. TABLE 5-2 I CHARP'Y V-NOTCH IMPACT DATA FOR THE V0GTLE EL[CTRIC GENERATING PLANT UNIT 2 react 0R VE$SEL WELO ME1AL AND HAZ HETAL 1RRADIATED AT 550'F, FLUENCE 4.44 x 10 18 n/cm2 (E > 1.0 MeV) i Temperature Impact Energy Lateral Expansion Shear Sanple No.(*F) ('C)(f t-lb) (J) (mils) Imm1 (%)
Weld Vetal BW2 -75 -
7.0 9.5 3.0 0.08 5 BW14 -50 -
14.0 19.0 9.0 0.23 10 BW11 -25 -
20.0 27.0 19.0 0.48 15 BW9 -20 -
58.0 78.5 38.0 0.97 55 BW12 -10 -
54.0 73.0 38.0 0.97 55 (
BW8 0 -
55.0 74.5 39.0 0.99 55 !
BW3 10 -
63.0 85.5 40.0 1.02 65 BW15 25 - 57.0 77.5 39.0 0.99 60 BW7 40 64.0 87.0 48.0 1.22 85 BW1 60 82.0 111.0 C2.0 1.57 90 BW5 80 84.0 114.0 61.0 1.55 90 BW4 100 83.0 112.5 61.0 1.55 90 BW10 150 96.0 130.0 73.0 1.85 100 BW13 200 95.0 129.0 71.0 1.80 100 BW6 275 1 104.0 141.0 76.0 1.93 100 RAZ Vetal BH9 -150 -101) 17.0 23.0 7.0 10 BH3 -125 -
22.0 30.0 11.0 {0.18 0.28 15 BBS -110 -
32.0 43.5 17.0 0.43 25 BH15 -100 -
51.0 69.0 25.0 0.64 45 BH1 -80 -
40.0 54.0 19.0 0.48 40 BH6 -75 - 39.0 53.0 23.0 0.58 40 BH2 -50 -
52.0 70.5 35.0 0.89 50
., BH13 -40 -
99.0 134.0 67.0 1.70 90 BH10 -25 -
112.0 152.0 60.0 1.52 95 BH12 -25 -
112.0 152.0 60.0 1.52 90 DR11 0 -
83.0 112.5 49.0 1.24 80 BH4 25 108.0 146.5 82.0 1.57 100 BH7 75 125.0 169.5 70.0 1.78 100 BH14 110 128.0 173.5 69.0 1.75 100 BH8 150 126.0 171.0 60.0 1.52 100 5-9
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TABLE 5-4 INSTRUMENTED CHARPY IMPACT TEST RESUETS FOR THE V0GTLE ELEC1RIC GENERATING PLANT UNIT 2 WELD METAE A I8 HEAT-AFFECTED-ZONE (HAZ) METAE, IRRADIATED AT 550*F, FLUENCE 4.44 X 10 n/cm2 (E > I.0 Mev)
Nor==11 sed Enerries Isami cm Time to Fracture Arreet. Yield Flow Test Charpy Charpy Maximum Prop YMd Time Load Maximum Load Load Streme Stree Ee/A Ep/A Load to Yield (peec) (kipe) (keil (kei)
Sample Temp Energy Ed/A fueec) (kipe) (kipe)
Nueber (*F) (ft-lb) (ft-Ib/ ins) (ktpe) weld Metal 200 3.55 0.01 91 106 27 2.75 60 3.65 131
-75 7.0 56 29 195 4.05 0.01 125 BW2 31 3.80 85 4.15 120 133
-50 14.0 113 82 330 4.40 0.01 BW14 11 3.65 SS 4.40 132 151 BW11 -25 20.0 161 150 5.10 500 4.80 0.90 262 206 4.00 85 117 135 BW9 -20 58.0 467 4.65 670 4.50 0.20 324 111 3.55 85 1.00 IIS 134 BW12 -10 54.0 435 4.60 605 4.35 443 288 155 3.50 95 1.10 123 141 BWs O 55.0 80 4.80 610 4.50 63.0 507 314 193 3.70 4.35 0.85 100 129 BW3 to 3.05 55 4.80 640 57.9 459 302 157 665 4.65 1.50 107 133 BW15 25 3.25 75 4.85 40 64.0 515 327 388 655 3.85 2.25 99 122 BW7 3ct 3.00 75 4.40 120 60 82.0 660 300 605 3.70 1.75 94 BW1 408 2.85 75 4.35 129 BWS 80 84.0 676 269 4.65 665 3.75 2.85 103 308 360 3.10 80 --. 74 107 BW4 100 83.0 668 4.25 730 --.
n 310 463 2.25 50 83 115 UW10 150 96.0 773 4.40 650 --* --.
765 2ES 477 2.55 55 --. 93 123 BW13 200 25.0 50 4.6u 725 -.
(( 837 340 4G7 2.80 BW6 275 104.0 HAZ Metel l 195 4.70 0.01 154 158 54 4.65 ISO 4.90 147 BH9 -150 17.0 137 83 295 5.60 0.01 107 153 25 3.25 60 5.65 155 170 BH3 -135 22.0 177 90 5.60 335 5.60 0.25 258 181 76 4.65 BHS -110 32.0 -- -- -- --
BH15 -100 51.0 411 COMPUTER MALFUNCTION .--85 -- 5.30 275 5.20 0.70 147 162 322 151 172 4.45 4.70 0.60 132 145 BH1 -80 40.0 90 4.80 435 314 217 97 4.00 4.55 1.25 123 138 BH6 -75 39.0 3.75 90 4.60 530
-50 52.0 419 251 186 500 1.40 0.90 139 162 BH2 515 4.20 85 5.60 149 99.0 797 2e2 575 5.15 0.01 128 BH13 -40 605 3.90 90 5.15 136 154
-25 112.0 902 301 500 2.10 1.00 BH12 633 4.10 a6 5.20 120 138
-25 112.0 902 2s9 575 3.35 1.95 BH10 3MO 3.60 80 4.75 96 130 BH11 O 83.0 668 286 4.95 575 --* --.
285 585 2.90 60 115 139 BH4 25 108.0 870 4.95 600 --- -*
292 715 3.50 80 83 116 BU7 75 125.0 1007 75 4.50 605 --. --.
128.0 1031 263 768 2.50 --. --* 103 131 BH14 110 3.10 60 4.80 655 1015 328 687
! BH8 ISO 126.0 I
. Fully ductile fractures no arrest load
.. Caused by data memory drop-out
?
- j. .'
l TABLE 5-5 EFFECT or $50*F 1RRADIATIon To a.sa = 10 38 n/c 2 (E
- 1.0 Mev) l Ott THE NOTCH TOLGiMESS PROPERTIES OF TM V0 GILE ELECTRIC GENERATI% PLANT iMIT 2 RIACTM VE55EL SLRVEILLAaCE MATERIALS i
Average 30 ft-1b II Average 35 mi1 III Ave age 50 ft-1b II8 Average Energy III Transitton Lateral Erpension Transitim Absorption at Temperature (*F) Temperature (*F) Temperature (*F) Full $heer (ft-lb)
Material Unirradiated Irradiated AT untreadiated Irradiated AT untreadiated irradiated AT unirradiated teradiated Atft-ib) '
E Plate 88528-1 10 10 0 SS 40 5 45 50 5 89 99 + 10 (longitudinal) !
4 j Plate B8628-1 30 30 0 40 45 5 75 80 5 70 79 + ? l Y ;
- g (Transverse) :
l i
' Weld metal - 15 - 15 0 - 5 - 5 0 $ $ 0 32 95 + 6 I HAZ Ntal - 83 - 80 0 - 50 - 50 0 - 45 - 45 0 106 12Z + 16 s
(1) " AVERAGE" is def 6ned as the value read from the curve fitted ttwough the data peints of the Charry tests (Figures 5-1 throups 5-4).
i l
l
. s e n
, + *
[
, , e
= w.-e - ,y. n ~ ~.. 3 ,*y v 3, -,++,re.m ,,%.-..*.-+e---.- ,* .--,e- w--,-y w.,--r -w,,,,--= g- ----.,,c--#+,, .v-p- n. --+~---wm,,-----.-,-,---w,_.y.-,,5-,.- -
. . e e TABLE 5-6 COMPARISON OF THE V0GTLE ELECTRIC GENERATING PLANT UNIT 2 SURVEILLANCE MATERIAL 30 FT-LB TRANSITION TEMPERATURE SHIFTS AND UPPER SHELF EhrRGY DECREASES WITH REGULATORY GUIDE 1.99 REVISION 2 PREDICTIONS 30 ft-lb Transition Tee . Shift Uccer Shelf Enercy Decrease i R.G. 1.99 Rev. 2 R.G. 1.99 Rev. 2 Fluence (Predicted)(a) Measured (Predicted) Measured Material Capsule 10 I9 n/cm 2 (*F) (*F) (%) (%)
) Plate B8628-1 U 0.444 24 0 16 0
) (longitudinal) j*." P1 ate B8628-1 0 0.444 24 0 16 0 (Transverse)
I i Weld Metal U 0.444 28 0 16 0 i
l HAZ Metal U G.444 --
0 --
0 1
a) Mean wt. % values of Cu and Ni were used to calculate the chemistry factors for the surveillance material.
3 4
y ,- . , , . ~ - , s _ , . - , ~ ~ _ , - , , , , - m-_._ - - -
1 1
TABLE 5-7 i r
t 1
TENSILE PROPERTIES FOR TliE V0GTLE ELECTRIC GENERATING PLANT UNIT 2 REACTOR VESSEL SURVEILLANCE
, MATERIALS IRRADIATED AT 550*F TO 4.44 X 10I9 n/m2 (E > 1.0 MeV) ,
t
! l 1 r
} i i i I
] ;
i Test O.25 Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction f Semple Temp. Strength Strength Load Strese Strength Elongation Elongation in Ares
' [
Material Number. L*D, (kei) (hel) (klo) (kel) (kei) (%) (%) (W)
Plate Bee 2s-1 IE.1 70 71.s 93.7 3.06 168.1' '82.1 12.0 24.0 83 !
(Longitudinal) BL2- ISO 88.8 SS.O 2.30 139.3 67.0 9.9 22.7 69 BL3 660 63.2 89.6 2.20 141.0 86.2 9.6 21.6 64 [
- Plate B8628-1 BT1 60 72.3 93.7 3.30 168.2 67.2 12.0 23.1 67 ;
(Traneverse) BT2 ISO 88.2 87.8 3.00 142.0 61.1 11.4 22.1 67 i BT3 660 83.8 87.8 3.40- '192.4 69.3 9.6 13.4 e4 Wald BW1 60 73.2 as.e 2.96 191.s 60.1 12.0- 23.1 e9 ,
! u, BW2 ISO e7.2 83.6 2.70 170.6 65.0 12.0 24.6 es !
j4 BW3 650 83.2 83.6 3.00 167.0 61.1 9.9 1s.9 61 [
t
! v h
i r
i r ;
1 b
f F
1 k
b
. . e a !
, .
- s
(
C'C)
--150 -100 -50 0 50 100 150 200 ggg _ _
_k'
. ea -
60
. .0 -
20 -
-r 0
1 bei}
t i I !
100 2.5
}"
60
,=gJ -
u 1.5 40 -
rr -
1.0 0.5 3 20 '
b 0I o temnuu n I I I I I O
a murun cWn ruixt tm a d8vcn' 120 160 100 --
,-p ,
o -
120
- 80 - %' G -- ,
' i D 60 - -
80 O
- r.,
g ci 40 -
40 ,
20 -
- ! l l l I I '
0 O
-200 -100 0 100 200 300 400 500 TEMPERATURE (*f)
Figure 5-1. Charpy V-Notch Impact Properties for Vogtle Electric Generating Plant Unit 2 Reactor Vessel Lower Shell Plate B8528.1 l (Longitudinal arientation) l i 5-15 !
r
! i
i (SC)
-150 -100 -50 0 50 100 150 200 100 -
' k ='$ ==
'a
- 8 80 60 -
, l 1
40 ,
O I b 3 I 1 1 I 4
100 2.5 h80 , ,
2.0 60 -
s c a,e e a 40 - I tr -
1.0 3 20 t - 0,5 I I I I I I O
l O
' O lNBRABMYJ e anWnu ase'n ruru tess a u"vd 120 go _
o a =
- O O f* - .
2 r 60 - hs5 g 3
_ go 5 _ s'r ._, , S 40 -
- r -
o 40 20 -
e o l l f '
! i 0 -- 0 -
-200 -100 0 100 200 300 400 500 TEMPERATURE ('F)
Figure 5-2. Charpy V-Notch Impact Properties for Vogtle Electric Generating Plant Unit' 2 Reactor Vessel Lower Shell Plate B8628-1 (Transverse Orientation) 5-16
('C)
-150 -100 -50 0 50 100 150 200 i i j l
3 1 ) l Js l I .
100 e-o-oo-
~
e'
. B 60 {
$ 60 -
e 40 o
t 20
('
0 1 '\ .Le I i t i i 100 2.5 80 -
6-(3- -
2.0 g 60
- 1.5 m
]40 1.0 3 20
- e go --
0.5 0
I - k'~ I I I I -
I O
o wr>Tu o noto esn'n ntas uit a ts%'
120 160 100 -
IN % e ,
,PN r \/
0 o
h s a -
.g -
80 S
~
o a 40 -
o g 40 20 -
~
0 l T~' l l l l 0
-200 -100 0 100 200 300 400 500 TEMPERATURE (*f)
Figure 5-3. Charpy V-Notch Impact Properties for Vogtle Electric Generating Plant Unit 2 Reactor Vessel Surveillance Weld Metal 5-17
(SC)
-150 -100- -50 0 50 100 150 200 ;
100 b --- ._.}- h ,
t o
8 90 '
60 - o
{
H 40 -
- 4 20 0
k(# -
-t 1 I I i 1
.. j 100 2.5
{60 l off 2.0
--Ie 1.5 60 -
8 e - 1 -
- 1.0 el 40 0 -
0.5 320 I E I I I I I j
0 O l
o amansuiu e Im45p1E3 esen IVENE 6444 a [ vin' l 160 !
- 200 140 , l
- o !
120 q o
- M o .
h100-g e g5 o o -
j 120 g t
- 90 - 0 G i 5 60 - -
80 5 eo o ;
- f 40 S -
40 :
20 e s i l l l l I I 0 0 h
-200 -100 0 100 200 300 400 ,
TEMPERATURE ('f) l d
t t
figure 5-4, Charpy V-Notch Impact Properties for Vogtle Electric Generating f
Plant Unit 2 Reactor Vessel Weld Heat-Affected-Zone Metal !
}
t 5-18 i i
. _ _ . = _ _ _ _ .-_ . _ _ - _ _ _ _ .
I P I
i
- i. f 1 - ,-
Eb' EC i
BL9 BL10 BL12 BL1 BLO !
1
,_ ~ ,m 7. . ,- -
y-
- ., y s - 3^,>'
, a ^ 1; ps ).
- [ ~ k.
N t
i
' {iM
. , %'eg $hl
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. g, ; .;;- yy - ~ ~ ~m.Q1 .
> w.p.. '
a
% .g y.g 3 .., . g .
1 ,4 ...,
o -
. BL2 BLIS DL8 UL5 BL7 i
! . . . _ m e- !
o )~ ..
bt.- K .
i
. - _ p a r;i jc..-
nv..
.:n -?
.s4
,yu:
v.. 4 4 g..g j y~ .- , . . -
7.m .s uu.
. , - ,,, uw-. ng: , e. _
. vy * ).(f. 23
_ Q.; . --
2a a n.-
1 BIA BL11 UL14 BL13 BL3 ;
t I
i figure 5-5. Charpy impact Specimen Fracture Surfaces for Vogtle Electric f
Generating Plant Unit 2 Reactor Vessel Lower Shell Plate B8628-1 j (longitudinal Orientation) l f
l 5-19 >
s
, f I- r
, . . . . ~ . . . - . . . _ , .- _ . . , . _ . - - , , , - v-.--,~- - - * - "- ~~~.~~-*~ +- - - - ---- -- "----~----~--'-~'#
v
- .- ; n,.- .e nm >
t t.
I f !
\
-p. .i, . ., ,;t j
2g h w...,
.. . a .
s, s
t y.
i 10 ,-_ ; .
.s a _.
BT5 BT1 BT12 BT13 BT3 .
- g. . r- ,
3 4
4 ,
s
.[i. ,
.,{
+ '
y .
L t. .
~%g z.. ,
g [
-p.h e j$v . ( -+
...f4, s'.,6, )
4
- W.,
b<
f
,p . 4 J t k, ,
i i- % y, y - ,
num m essard w--- - s_: n ,L - IMbr d ; - .. c BT7 BT14 BT8 DT10 BTS .
I
-staw - "***4*'d , .g 3 weer str" m
4 ,
- t I I. ' -j T I ?
'* i p f ,, -
?-
6 P d 5pali. 4 l l ,} $
g '
j .
9y-[c.
.. ' & fs. ^_ {
s' k*
, r.;.y, gy. -
,' - g l
, nn ,
- 5
,2M&th i66 ~ (gjg.4 ,
-rg . . . , ,
i+4 r, ,;. t, , i j - i .<<, a ssu m _. m.,a ma e wu : BT2 BT9 BT11 BT4 BT15 i i f i Figure 5-6. Charpy impact Specimen fracture Surfaces for Vogtle Electric Generating Plant Unit 2 Reactor Vessel Lower Shell Plate B8628-1 i (Transverse Orientation) i l l 5-20 t l r-m *v- Nrr=w- ---"v T- -p-a g g - , , ,-,. .% r
1
)
ti a-hf l
' t l~.- *. i l g y, s ~-,, .
. . , y ;
, m:. ~ ,
a 9.' .
._ eer
, h,, .
+ (- .i u.,, .; -
% l,
^Y ga =,.4-wm..., a.: Qh -
BW2 BW14 DW11 BWO BW12 ! i
't ' M. 4.'_:,
7, ,,p ,,,,,,, 7- ., s i l s l
%+ #
m i f, /w e e' , 4
. w ir , ;
og~h ,.... s . : wL+t.: . nummenn
, , _ ~
r
- u. u -
*1 p:rt -
.am vJ~ e
~
r l , - f a [ I
,; e' ,- j i p 4d A sa M T:A l BW8 BW3 BW15 BW7 DW1 i i
[
- .w.m ,
+
p di - .
. g' .. J ~
% f,' r'w.g -- ;
- l. 9, j
( 4 i r.
%._ 1
,. E Lt
- 4. g r. , i..
e.m
%,w , g -
~
;~ tw
{
) 'g i Jh lj [( ,
[ ! 1
=u MWWMM*
~
( wy@/SA%=had 1 t
.j; J' ' 534As.ar: ' MMbeA
, l%._, _ ' '
B'A 5 Bw4 BW10 BW13 BW6 ! I i r t t, Figure 5-7. Charpy impact Specimen f racture surf aces for Vogtle Elect ric l Generating Plant Unit 2 Poactor Vessol Surveillance Weld Metal 5-?) ! i f
?
i
-, . - - - - . . - . . . . - -- n.- -- -- -- , - - - , . . - - - . . - . , , . - -
., 1" i , ., c. j r ,
. , 13 t
.t 3 . ,
77 a
"J; , .
. o !y i.
Y
, w t
v; f y_ Y& Q k- %, i s , 31 -. ; t r;.= > <
'3'
, ,9 I L
----.m.
BH9 BH3 BBS BH15 BH1 g.,,,,,,. -,m i y' [ kl p ,-- ,b { s ; w ,*L 1 hhe. , i.i #; pny , ,
' l f ..
y py:n **]lg ?" m wm r. i.L, s0 : acw . , s.u . _ . ;, as.m
'('( ;
e s , r
',,4 J C. '
p\ ,<
~'
Q)- '~"
- c. \
rew * - -
. -n .,~ %, . , e m ,2 BBS BH2 Bill 3 BR10 Bill 2 @
y acm ma -. f-* , , , . -., m .-- m aa b ' L . 'j ' f. 4 \ I. ,p, _
.l
\ v ; 2... -
=
w vo A. ,, , a , 4 '(i , *_ [ y j,9, .h QT ./*u a J',. t4*1ho WWF = ..
" % _,, hyg.
'e -
- . r.* - .19t++ v
) .hE4- kddna'- -
* * . Nf-g f 'lilJ( j l((' i[ ]tr g[
k -g 4
. ,S f i,
, ej '
, _m f-<.enN _:
27W etna b ,e db . BH11 BH4 BH7 BH14 BH8 Figure 5-8. Charpy Impact Specimen fracture Surfaces for Vogtle Electric Generating Plant Unit 2 Reactor Vessel Weld Heat-Affected-lone Metal 5-22
i PC) 0 50 100 150 200 250 300 120 g ; j l l l ja 800 110 700 100 o uttwit Ttxtu stawk .
- 3 90 -
Ob e._-m- #- 600
,5
/
g' -R i 1 M 80 t W -- 500
- G 70 -
g'
'h&%'N_1 A-- . --- &
60 -
'\ - 400 in Ygu !!anTH
)
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SECTION 6.0 RADIATION ANALYSIS AND NEUIRON 00SIMETRY 6.1 Introduction Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons. First, in order to interpret the neutron radiation-induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter
. information is derived solely from analysis, The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition, in recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and
. positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall.
Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, " Analysis and Interpretation of Light Water React 7r 6-1
Surveillance Results." recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) _to provide a- data base for future _ reference. The energy dependent dpa function to be used for this evaluation is specified , in ASTM Standard Practice E693, " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom." ThE application of the dpa , parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to the
- Regulatory Guide 1.99, " Radiation Damage to Reactor Vessel Materials."
This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance Capsule U. Fast neutron exposure parameters in terms of fast neutror, fluence (E > 1.0 MeV), f ast neutron fluence (E > 0.1 Mev), and iron atom displacements (dpa) are established for the capsule irradiation history. The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel itself. Also uncertainties associated with the derived exposure parameters at the surveillance capsule and with the projected exposure of the pressure vessel are provided. . 6.2 Discrete Ordinates Analysis A plan view of the reactor geometry at the core midplane is shown in Figure 4-1. Six irradiation capsules attached to the neutron pads are included in the reactor design to constitute the reactor vessel surveillance program. The capsules are-located at azimuthal angles of 58.5', 61.0*, 121.5*,-238.5*, 241.0', and 301.5* relative to the core cardinal axes as shown in Figure-4-1, A-plan view of a dual surveillance capsule holder attached to the neutron pad is shown in Figure 6-1. The stainless steel specimen containers are 1.182 by 1-inch and approximately 56 inches in height. The containers are positioned - axially such that the specimens are centered on the core midplane, thus spanning the central 5 feet of the 12-foot high reactor core. - 6-2
From a neutron transport standpoint, the surveillance capsule structures are significant. They have a marked effect on both the distribution of neutron flux and the neutron energy spectrum in tne water annulus between the neutron i pad and the reactor vessel, in order to properly determine the neutron l . environment at the test specimen locations, the capsules themselves must be included in the aalytical model. In performing the fast neutron exposure evaluations for the surveillance !
, capsules and reactor vessel, two distinct sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters {d(E > 1.0 Mev), p(E > 0.1 Mev), and dpa) through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsule as well as for the determination of exposure parameter ratios; i.e.,
dpa/p(E > 1.0 MeV), within the pressure vessel geometry, The relative radial gradient information was required to permit the projection of measured . exposure parameters to locations interior to the pressure vessel wall; i.e., the 1/4T, 1/2T, and 3/4T locations. The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux (E > 1.0 MeV) at surveillance capsule positions, and several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core. The importance functions generated from these adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance
. functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for each cycle of irradiation; and established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles. It is important to note that the cycle specific neutron source distributions utilized
' in these analyses included not only spatial variations of fission rates within the reactor core; but, also accounted for the effects of varying neutron yield 6-3 L
per fission and fission spectrum introduced by the build-in of plutonium as the burnup of individual fuel assemblies increased. The absolute cycle specific data fron, the adjoint evaluations together with relative neutron energy spectra and radial distribution information from the , forward calculation provided the means to:
- 1. Evaluate neutron dosimetry obtained from surveillance capsule locations. ,
- 2. Extrapolate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel wall.
- 3. Enable a direct comparison of analytical prediction with measurement.
- 4. Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves.
The forward transport calculation for the reactor model summarized in Figures . 4-1 and 6-1 was carried out in R, O geometry using the DOT two-dimensional discrete ordinates code [12) and the SAILOR cross-section library [133 The . SAILOR library is a 47 group ENDFB-IV based data set produced specifically for light water reactor applications. In these analyses anisotropic scattering was treated with a P3 expansion of the cross-sections and the angular discretization was modeled with an S8 order of angular quadrature. The reference core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 4-loop , plants. Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2a - uncertainty derived from the statistical evaluation of plant to plant a d c.ycl e to cycle variations in peripheral power was used. Since it is unlikely that a - single reactor would have a power distribution at the nominal +2a 6-4
level for 'a large number of fuel cycles, the use of tais reference distribution 1 is expected to yield somewhat conservative results. All adjoint analyses were also carried out using an S8 order of angular quadrature and the P3 cross-section approximation from the SAILOR library. Adjoint source-locations were chosen at several azimuthal locations along the pressure vessel inner radius as well as the geometric center of each surveillance capsule. Again, these calculations were run in R, 0 geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, 4 (E > 1.0 MeV). Having the ' importance functions and appropriate core source distributions, the response of interest could be calculated as: R(r,0)=fr[0[E I(r, 0, E) S (r, 8. E) r dr de dE where: R (r, 0) = d (E > 1.0 MeV) at radius r and azimuthal angle 0 I (r, 0, E) = Adjoint importance function at radius, r, azimuthal angle 0, and neutron source energy E. S (r, 0, E) Neutron source strength at core location r, 0 and energy E. Although the adjoint importance fuactions used in the analysis were based on a response function defined by the threshold neutron flux (E > 1.0 MeV), prior calculations have shown that, while the implementation of low leakage loading
? patterns significantly impact the magnitude and the spatial distribution of the > neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the ratio of dpa/d (E > 1.0 MeV) is insensitive to changing core source distributions. In the application of these adjoint importance functions to the Vogtle Electric Generating Plant Unit 2 f
reactor, therefore, the iron displacement rates (dpa) and the neutron flux (E > 0.1 MeV) were computed on a cycle specific basis by using dpa/d (E > 1.0 MeV) and'd (E > 0.1 MeV)/4 (E > 1.0 MeV) ratios from the forward analysis in conjunction with the cycle specific d (E > 1.0 MeV) solutions f from the individual adjoint evaluations. 6-5 i
,,_j
The reactor core power distribution used. in the plant specific adjoint l calculations was taken from the fuel cycle design report for the first ! operating cycle of :tle Electric Generating Plant Unit 2Il43 , Selected results from the neutron transport analyses are provided in Tables 6-1 , through 6-5. The data listed in these tabM establish the means for absolute comparisons of analysis and measurement for the capsule irradiation period and + provide the means to correlate dosimetry results with the corresponding neutron exposure of the pressure vessel wall. , in Table 6-1, the calculated exposure parameters [p (E > 1.0 MeV), d(f > 0.1 MeV), and dpa) are given at the geometric center of the two surveillance capsule positions for both the design basis and the plant specific core power distributions. The plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis. Th9 design basis data derived from the forward calculation are provided as a point of reference against which plant specific fluence evaluations can be compared. Similar data is given in Table 6-2 for the pressure vessel inner radius. Again, the three pertinent exposure , parameters are listed for both the design basis and the cycle 1 plant specific power distribution. It is important to note that the data for the vessel inner . radius were taken at the clad / base metal interface; and, thus, represent the maximum exposure levels of the vessel wall itself. Radial gradient information for neutron flux (E > 1.0 MeV), neutron flux (E > 0.1 MeV), and iron atom displacement rate is given in Tables 6-3, 6-4, and 6-5, respectively. The data, obtained from the forward neutron transport calculation, are presented on a relative basis for each exposure .
-parameter at several azimuthal locations. Exposure parameter distributions within the wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data given in Tables 6-3 -
through 6-5. 6-6 i
For example, the neutron flux (E > 1.0 MeV) at the 1/4T position on the 45' azimuth is given by: pl/4T(45') = 4(220.27, 45') F (225.75, 45') where: Projected neutron flux at the 1/4T position on dl/4T(45*) the 45' azimuth 4 (220.27,45') = Projected or calculated neutron flux at the vessel inner radius on the 45' azimuth. F (225.75, 45') = Relative radial distribution function from Table 6-3. Similar expressions apply for exposure parameters in terms of ; (E > 0.1 MeV) and dpa/sec, The DOT calculations were carried out for a typical octant of the reactor. However, for the neutron pad arrangement in Vogtle Electric Generating Plant Unit 2, the pad extent for all octants is not the same. For the analysis of the flux to the pressure vessel, an octant was chosen with the neutron pad extending from 32.5 - 45.0 degrees which produces the maximum flux. Other octants have neutron pads spanning larger azimuthal sectors which provlot sre shielding. For the octant with 12.5 degree pad, the maximum flux to the vessel occurs near 25 degrees and the values in the tables for the 25 degree angle are vessel maximum values. Exposure values for 0, 15, and 45 degrees can be used for all octants; values in the tables for 25 and 35 degrees are maximum values and only apply to octants with a 12.5 degree neutron pad. 6.3 Neutron Dosir,etry The passive neutron sensors included in the Vogtle Ele:tric Generating Plant Unit 2 surveillance program are listed in Table 6-6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the capsule and the subsequent determination of the various exposure parameters of interest [4 (E > 1.0 Mev), 4 (E > 0.1 MeV), dpa]. I l 6-7
The relative locations of the neutron sensors within the capsules are shown in Figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within .
- the capsules. The cadmium-shielded-neptunium and uranium fission monitors were accommodated within the dosimeter block located near the center of the capsule. .
The use of panive monitors such as those listed in Table 6-6 does not yield a a direct measure of the energy dependent flux level at the point of interest. Rather, the actlyation or fission process is a measure of the integrated effect , that the time- and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known, In particular, the following variables are of interest: i o The specific activity of each monitor, o The operating history of the reactor, o The energy response of the monitor. o The neutron energy spectrum at the monitor location. . o The physical characteristics of the monitor. 4 The specific activity of each of the neutron monitors was determined using established ASTM procedures (15 through 28] . Following sample preparation and weighing, the activity of each monitor was determined by mears of a
- lithium-drifted germanium, Ge(Li), gamma spectrometer. The irradiation history I
of the Vogtle Electric Generating Plant Unit 2 reactor during cycle I was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report" and the Nucleonics Week Data Sheets for the applicable period. , The irradiation history applicable to Capsule U is given in Table 6-7. Measured and saturated reaction product specific activities as well as measured - . full power reaction rates are listed in Table 6-8. Reaction rate values were derived using the pertinent data from Tables 6-6 and 6-7. - i Values of key fast neutron exposure parameters were derived from the maasured reaction rates using the FERRET least squares adjustment code [29] The . I f 4 6-8
' FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at the the center of the surveillance capsule as input and proceeded to adjust a priori (calculated) group fluxes to produce a best fit (in a least squares sense)_to the reaction rate data. The exposure parameters along with associated uncertainties where then obtained from the adjusted spectra.
In the FERRET evaluations, a log-normal least-squares algorithm weights both
> the a priori values and the measured data in accordance with the assigned uncertainties and correlations. In general, the measured values f are linearly related to the flux 4 by some response matrix A:
(s,a) (s) (a) f =I A 4 l ' 9 ig 9 i ?- where i indexes.the measured vaiues belonging to a single data set s, 9
-designates the energy group and a delineates spectra that may be simultaneously adjusted. For example, l
R -I o & 1 i 9 19 9 i relates a set of measured reaction rates R$ to a single spectrum pg by ! the multigroup cross section ajg . (In this case, FERRET also adjusts the i cross-sections.) The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with the large assigned uncertainties. In the FERRET analysis of the dosimetry data, the continuous quantities (i.e., fluxes and cross-sections) were approximated in 53 groups. The calculated fluxes from the discrete ordinates analysis were expanded into the FERRET group structure using the SAND-II code [30]. This procedure was carried out by first expanding the a priori soectrum into the SAND-II 620 group structuro using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not__ coincide. The 620-point spectrum was then easily collapsed 6-9
. ~ . _ __. . . ._ _ _
to the group scheme esed in FERRET. The cross-sections were also collapsed into the 53 energy-group structure using SAND II with calculated spectra (as expanded to 620 groups) as weightino > functions. The cross sections were taken from the ENDF/B-V dosimetry file. ,
'acertainty estimates and 53 x 53 covariance matrices were constructed for each cross section. Correlations between cross sections were neglected due to data * -and code limitations, but are expected to be unimportant.
For each set of data or a priori values, the inverse of the corresponding ! telative covariance matrix M is used as a statistical weight. In some cases, as for the cross sections, a multigroup covariance matrix is used. More ofter, i a simple parameterized form is used: ! M gg,=Rf+R g R,P g gg, , i where RN specifies an overall fractiona; normalization uncertainty (i.e., j complete correlation) for the corresponding set of values. The fractional - uncertainties Rg specify additional random uncertainties for group g that are . correlated with a correlation matrix: - l Pgg, = (1 - 0) 6gg, + 0 exp [ ] The first term specifies purely random uncertainties while the second term describes short-range correlations over a range a (0 specifies the a strength of the latter term). For the a priori calculated fluxes, a short-range correlation of a - 6 ' groups was used. This choice implies that neighboring groups are strongly correlated when 0 is close to 1. Strong long-range correlations (or anticorrelations) were justified based on information presented by R.E. Maerker[31]. Maerker's results are closely duplicated when a - 6. For the integral reaction rate covariances, simple normalization and random uncertainties were combined as deduced from experimental uncertainties. 6-10
integrated exposure of 4.44 x 10 n/cm (E > 1.0 MeV) with an associated uncertainty of ! 8%. Also reported are capsule exposures in terms of fluence (E > 0.1 MeV) and iron atom displacements (dpa). Summaries of the fit of the j adjusted spectrum are provided in Table 6-10. In general, excellent results were achieved in the fits of the adjusted spectrum to the individual experimental reaction rates. The adjusted spectrum itself is tabulated in Table 6-11 for the FERRET 53 energy group structure.
- A summary of the measured and calculated neutron exposure of Capsule V is presented in Table 6-12. The agreement between calculation and measurement
, falls within 24% for all fast neutron exposure parameters listed. The thermal neutron exposure calculated for the exposure period undepredicted the measured value by 61 percent.
Neutron exposure projections at key locations en the pressure vessel inner radius are given in Table 6-13. Along with the current (1.18 EFPY) exposure derived from the Capsule U measurements, projections are also provided for an exposure period of 16 EFPY and to end of vessel design life (32 EFPY). In the evaluation of the future exposure of the reactor pressure vessel the
. design basis exposure rates from Table 6-2 were employed. Since the Vogtle Electric Generating Plant Unit 2 reactor has operated for only one fuel cycle and equilibrium fuel management has not been fully established, the use of these design basis values is still appropriate. The use of the design basis values should result in conservative predictions of future vessel exposure that can be refined as additional dosimetry becomes available.
In the calculation of exposure gradients for use in the development of heatup and cooldown curves for the Vogtle Electric Generating Plant Unit 2 reactor
, coolant system, exposure projections to 16 EFPY and 32 EFPY were also employed. Data based on both a fluence (E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wall are provided in Table 6-14.
e 6-11 j
~ In order to access RTNDI vs. fluence trend curves, dpa equivalent fast , neutron fluence levels for the 1/4T and 3/41 positions were defined by the relations , p' (1/4T) p (Surface) (dp (Stl e)} , L d' (3/4T) ' - d (Surface) (dp (Su e)) . Using this approach results in the dpa equivalent fluence values listed in Table 6-14. In Table 6-15 updated lead factors are listed for each of the Vogtle Electric Generating Plant Unit 2 surveillance capsules. These data may be used as a guide _in establishing future withdrawal schedules for the remaining capsules. e 4 e 6-12
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\ \ '
4 Figure 6-1. Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13
TABLE 6-1 l CALCULATED FAST NEUTRON EXPOSURE PARAMETERS , AT THE SURVEILLANCE CAPSULE CENTER j d(E > 1.0MeV) d(E > 0.1Mev) Iron Displacement Rate 2 In/cm -secl In/cm2-sec1 Idoa/seci 29.0* 31 5* 29.0* 31.5* 29.0* 31.5* DESIGN BASIS 1.13 X 10 11 1.21 X 1011 5.08 X 10 11 5.44 X 10 ll 2.21 X 10-10 2.37 X 10-10 CYCLE 1 8.47 X 10 10 9.03 X 10 10 3.81 X 10ll 4.06 X 10 11 1.66 X 10-10 1.77 X 10-10 i e . ,
- e , * . * *
, - , . . - , - .-,.v... . , , . . . , . - . ~ . . . , . . - . . . . . .
TABLE 6-2 , CALCULATED FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE 6(E > lJMeV) In/cm2 -secl 5 0.0* E.9
- lid' 35.0' 45.01 DESIGN BASIS 1.78 X 10 10 2.66 X 1010 3.01 X 10 10 2.45 X 1010 2,81 X 10 10 CYCLE 1 1.36 X 10 10 2.00 X 10 10 2.29 X 1010 1.86 X 10 10 2.13 X 1010 2
6(E > 0.lMeV) in/cm _sE J 0.0* 15.0* liJ' 35.0* 45.01 DESIGN BASIS 3.70 X 1010 5.60 X 10 10 8.22 X 1010 6.96 X 1010 7.04 X 10 10
. CYCLE 1 2.83 X 1010 4.21 X 1010 6.25 X 1010 5.28 X 1010 5.34 X 10 10 Iron Atom Disolacement Rate Idoa/sec1 0.0* 15J.'. Zid' 35.0* 45.0*
DESIGN BASIS 2.77 X 10-I1 4.12 X 10-11 5.04 X 10~11 4.15 X 10-11 4.48 X 10-11 CYCLE 1 2.12 X 10-11 3.10 X 10-11 3.83 X 10-11 3.15 X 10-11 3,40 X 10-11 4 e l 6-15 5 l
TABLE 6-3 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 1.0 MeV) WITHIN THE PRESSURE VESSEL WALL Radius _ (cm) 0* 15' 25' 35' 45' 220.27(I) 1.00 1,'00 1.00 1.00 1.00 220.64 0.976 0.979 0.980 0.977- 0.979 221.66 0.888 0.891 0.893 0.891 0.889 222.99 0.768 0.770 0.772 0.770 0.766 224.31 0.653 0.653 0.657 0.655 0.648 225.63 0.551 0.550 0.554 0.552 0.543 l 226.95 0.462 0.460 0.465 0.463 0.452 228.28 0.386 0.384 0.388 0.386 0.375 229.60 0.321 0.319 0.324 0.321 0.311 230.92 0.267 0.263 0.275 0.267 0.257
-232.25 0.221 0.219 0.225 0.221 0.211 233.57 0.183 0.181 0.185 0.183 0.174 234.89 0.151 0.149 0.153 0.151 0.142
. 236.22 0.124 0.122 0.126 0.124 0.116 , 237.54 0.102 0.100 0.104 0.102 0.0945 I 238.86 0.0828 0.0817 0.0846 0.0835 0.0762 240.19 0.0671 0.0660 0.0689 0.0679 0.0608
- 241.51 0.0538 0.0522 0.0550 0.0545 0.0471 242.17(2) 0.0506 0.0488 0.0518 0.0521 0.0438
-NOTES: 1) Base Metal Inner Radius-
- 2) Base Metal Outer Radius
- 6-16
TA8LE 6-4 RELATIVE RADIAL DISTRIBUTIONS OF NEUTRON FLUX (E > 0.1 MeV) WITHIN THE PRESSURE VESSEL WALL Radius _ (cm) 0' 15* 25' 35' 45' t 220.27(l) 1.00- 1.00 1.00 1.00 1.00 220.64 1.00 1.00 1.00 1.00 1.00 221.66 1.00 1.00 1.00 0.999 0.995 222.99 0.974 0.969 0.974 0.959 0.956 224.31 0.927 0.920 0.927 0.907 0.901 225.63 0.874 0.865 0.874 0.850 0.842 226.95 0.818 0.808 0.818 0.792 0.782 228.28 0.761 0.750 0.716 0.734 0.721 229.60 0.705 0.693 0.704 0.677 0.662 230.92 0.649 0.637 0.649 0.621 0.605 232.25 0.594 0.582 0.594 0.567 0.549 233.57 0.540 0.529 0.542 0.515 0.495 234.89 0.487 0.478 0.490 0.465 0.443 236.22 0.436 0.428 0.440 0.416 0.392 237.54 0.386 0.380 0.392 0.369 0.343 238.86 0.337 0.333 0.344 0.324 0.295 240.19 0.289 0.287 0.298 0.279 0.248 241.51 0.244 0.238 0.249 0.233 0.201 242.17(2) 0.233 0.226 0,237 0.223 0.188 NOTES: 1) Base Metal Inner Radius
. 2) Base Metal Outer Radius 6-17 i
TABLE 6-5 RELATIVE RADIAL DISTRIBUTIONS OF 1RON DISPLACEMENT RATE (dpa) WITHIN THE PRESSURE VESSEL WALL Radius (cm) O' 15' 25' _ 35' 45' 220.27(l) 1.00 1.00 1.00 1.00 1.00 220.64 0.984 0.981 0.984 0.983 0.984 221.66 0.912 0.909 0.917 0.921 0.915 , 222.99 0.815 0.812 0.826 0.833 0.821 224.31 0.722 0.719 0.737 0.747 0.730 225.63 0.638 0.634 0.656 0.668 0.647 226.95 0.563 0.559 0.584 0.597 0.572 228.28 0.497 0.493 0.519 0.533 0.506 229.60 0.439 0.435 0.462 0.475 0.447
^
230.92 0.387 0.383 0.410 0.423 0.394 232.25 0.341 0.338 0.364 0.376 0.347 233.57 0.300 0.297 0.322 0.334 0.305 234.89 0.263 0.261 0.285 0.295 0.266 236.22 0.230 0.228 0.250 0.260 0.231 237.54 0.199 0.198 0.218 0.227 0.199 238.86 0.171 0.170 0.189 0.196 0.169 240.19 0.145 0.144 0.161 0.167 0.140 241,51 0.121 0.119 0.135 0.139 0.113 242.17(2) 0.116 0.113 0.128 0.134 0.106 NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius .
?
6-18
= -
i , TABLE 6-6 NUCLEAR PARAMETERS FOR NEUTRON FIUX MONITORS Target Fission Reaction Weight Response Product Yield Monitor of Interest Fraction Ranoe Hal f-Life (%) Material Cu63(n,a)Co60 0.6917 E > 4.7 MeV 5.272 yrs Copper Fe54(n,p)Mn54 0.0582 E > 1.0 MeV 312.2 days Iron NiS8(n p)CoS8 0.6830 E > 1.0 MeV 70.90 days Nickel 1.0 E > 0.4 MeV 30.12 yrs 5.99 Uranium-238* U?38(n,f)Csl37 l 1.0 E > 0.08 MeV 30.12 yrs 6.50 Neptunium-237* Np237(n,f)Csl37 CoS9(n,d)Co60 0.0015 0.4ev>E> 0.015 MeV 5.272 yrs Cobalt-Aluminum
- CoS9(n,B)Co60 0.0015 E > 0.015 MeV 5.272 yrs Cobalt-Aluminum l
- Denotes that monitor is cadmium shielded.
TABLE 6-7 MONTilLY THERMAL GENERATION DURING THE FIRST FUEL CYCLE 0F THE V0GTLE ELECTRIC GENERATING PLANT UNIT 2 REACTOR , THERMAL GENERATION t!QEj (MW-bri_ 4/89 475504 5/89 617966 6/89 2450888 7/89 2452023 8/89 2526703 9/89 2439109 10/89 2034639 11/89 2350213 12/89 2335572 , 1/90 2503482 2/90 2289775 . 3/90 2332504 4/90 2367059 5/90 2394293 6/90 2119961 7/90 1730098 8/90 1413553 9/90 524543 i 6-20
_ - ~ . . . _ . _ . . _ . TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES Measured Saturated Reaction
.- Moaitor and Activity Activity Rate Axial location Idis/sec-omi Idi12sec-om) (RPS/ NUCLEUS) i Cu-63 (n,a) Co-60 Top- 5.51 x 10 4 4.01 x 10 5 Middle 4.96 x 10 4 3.6i x 10 5 Bottom 4.86 x 10 4 3.53 x 10 5 l Average 5.11 x 10 4 3.71 x 10 5 5.67 x 10-17 Fe-54(n p) Mn-54 . Top 1.88 x 10 6 4.05 x 10 6 Middle 1.65 x 10 6 3.55 x 10 6 - Bottom 1.67 x 10 6 3.59 x 10 6 4
Average 1.73 x 10 6 3.73 x 10 6 5.94 x 10-15 Ni-58'(n,p) Co-58 Top 1.86 x 10 7 6.04 x 10 7 7 L Middle 1.70 x 10 7 5.52 10 ' Bottom 1.65 x 10 7 5.36 x 107 Average 1.74 x 10 7 5.64 x 10 7 8.05 x 10-15
.o.
U-238 (n,f) Cs-137 (Cd) Middle 1.59 x 10 5 5.99 x 10 6 3.95 x 10-14 6-21
1 TABLE 6-8 HEASURED SENSOR ACTIVITIES AND REACTION RATES - cont'd
, Heasured Saturated Reaction
) Monitor and Activity Activity Rate , b.1111 Location idhLag-qm_.1 (dis /see-aml (RPS/NUCLEUM t e Np-237(n,f) Cs-137 (Cd) Middle 1,56 x 106 5.86 x 10 7 3.55 x 10'l3 C0-59 (n,0) Co-60 Top 1.13 '0 7 8.21 x 10 7 Middle 1.26 x 107 9.16 x 10 7 Bottom 1.15 x 107 8.36 x 107 Average 1.18 x 10 7 8.58 x 10 7 5.60 x 10-12 , Co-59 (n,3) Co-60 (Cd) . Top 5.99 x 106 4.35 x 10 7 Middle 6.21 x 10 6 4.51 x 107 Bottom 6.19 x 106 4.50 x 107 Average 6.13 x 106 4.46 x 10 7 2.91 x 10-12 e 6-22
I TABLE 6-9
SUMMARY
OF NEUTRON DOSIMETRY RESULTS TIME AVERAG{D EXPOSURE RATES ) l
- 2 d (E > 1.0 MeV) (n/cm -sec) 1.19 x 10ll i 8%
2 5.19 x 10Il d (E > 0.1 MeV) (n/cm -sec) 1 15% dpa/sec 2.26 x 10-10 y 33g 4 (E < 0.414 eV) (n/cm -sec) 2 1.11 x 10ll i 21% INTEGRATE 0_LAPsulE EXPOSURE 2 4.44 x 1018 3 gx
+ (E > 1.0 MeV) {n/cm )
2 1.94 x 1019
, 4 (E > 0.1 MeV) (n/cm ) i 15%
, dpa 8.43 x 10-3 3 3 3y, 4 (E < 0.414 eV) {n/cm )
2 4.14 x 10 18 1 21% - NOTE: Total Irradiation Time - 1.18 EFPY e e 6-23
1 TABLE 6.10 COMPARISON Of MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER i Adjusted
- Rgaction tiguf,gd Calculation [13
~
Cu-63 (n,a) Co-60 5.67x10-17 5.77x10-17 1.02 Fe-54 (n.p) Mn-54 5.94x10-15 5.92x10'l6 1.00 Ni-58(n.p)Co-58 8.05x10-15 8.09x10-15 1.00 , U-238 (n.f) Cs-137 (Cd) 3.95x10-14 3.60x10'I4 0.91 Np-237 (n.f) Cs-137 (Cd) 3.55x10-13 3.67x10-13 1.03 , Co-59 (n,a) 00-60 (Cd) 5.60x10-12 5.56x10-12 0.99 I Co-59 (n,8) Co-60 2.91x10-12 2.92x10-12 1.00 9
-+
F 6-24 i
TABLE 6-11 l ADJUSTED NEUTRON ENERGY SPECTRUM Al lHE SURVEILLANCE CAPSULE CENTER Energy Energy Adjusgedflux Adjusgedflux Group (Mev) (n/cm -sec) Group (Mev) (n/cm -sec) I 1.73x10 I 8.10x10 6 28 9.12x10-3 2.23x10 10 2 1.49x10 I 1.82x10 7 29 5.53x10-3 2.89x1010 3 1.35x10I 7.01x10 7 30 3.36x10-3 9.00x10 9 4 1.16x10 I 1.57x10 8 31 2.84x10-3 8.59x10 9 5 1.00x10 I 3.47x10 8 32 2.40x10-3 8.28x10 9 6 8.61x10 0 5.98x10 8 33 2.04x10-3 2.33x10 10 7 7.41x10 0 1.39x10 9 34 1.23x10-3 2.14x10 10 8 6.07x10 0 2.03x10 9 35 7.49x10-4 1.99x10 10 9 4.97x10 0 4.37x10 9 36 4.54x10-4 1.89x1010 10 3.68x10 0 5.94x10 9 37 2.75x10-4 2.03x1010 11 ?.87x10 0 1.28x1010 38 1.67x10-4 2.17x10 10 12 2.23x10 0 1.81x1010 39 1.0lx10-4 2.19x10 10 13 1.74x10 0 2.59x10 10 40 6.14x10-0 2.18x10 10 14 1.35x10 0 2.89x10 10 41 3,73x10-5 2.13x10 10 15 1.lix10 0 5.30x10 10 42 2.26x10-5 2.07x1010 16 8.21x10-I 6.04x10 10 43 1.37x10-5 2.02x1010 17 6.39x10-l 6.23x1010 44 8.32x10-6 1.92x1010 18 4.98x10-1 4.48x1310 45 5.04x10-6 1.77x10 10 19 3.88x10-1 6.24x1010 46 3.06x10-6 1.66x1010 20 3.02x10-I 6.35x1010 47 1.86x10-6 1.53x10 10 3 21 1.83x10-I 6.22x10 10 48 1.13x10-6 1.13x1010 10 22 1.llx10-I 4.92x10 49 6.83x10-7 1.45x1010 23 6.74x10-2 3.39x10 10 50 4.14x10-7 1.93x10 10 24 4.09x10-2 1.91x10 10 51 2.51x10-7 1.92x1010 25 2.55x10-2 2.49x10 10 52 1.52x10-7 1.83x10 10 26 1.99x10-2 1.22x10 10 53 9.24x10-8 5.45x10 10 27 1.50x10-2 1.55x10 10 NOTE: Tabulated energy levels represent the upper energy of each group. 6-25
6 TABLE 6-12 : COMPARISON OF CALCULATED AND MEASURED EXPOSURE LEVELS FOR CAPSULE U Calculated Measured []B , 2 f(E > 1.0 MeV) (n/cm ) 3.37 x 1018 4.44 x 1018 0.76 2 ' f(E > 0.1 MeV) {n/cm ) 1.52 x 1019 1.94 x 1019 0.78 dpa 6.60 x 10~3 8.43 x 10~3 0.78 f(E < 0.414 eV) (n/cm2 ) 1.61 x 1018 4.14 x 10 18 0.39 t S e 4 6-26
+ ,
TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON TiiE PRESSURE VESSEL CLAD / BASE METAL INTERFACE 1.18 EFPY 25* 35* 45* 1.05 X 10 I8 0* 15* 6.68 X 10 17 9.86 X 10 17 1.13 X 10 18 9.17 X 10 I7
+ (E > 1.0 Mev)
[n/cm2] 2.00 X 10 18 2.98 X 10 18 2.53 X 10 18 2.54 X 10 I8
+ (E > 0.1 MeV) 1.34 X 10 18
[n/cm2] 1.48 X 10-3 1.83 X 10-3 1.51 X 10-3 1.62 X 10-3 Iron Atom Displacements 1.01 X 10-3 [dpa] 1 16.0 EFPY 15* 25* 35* 45' 0* 18 1.34 X 10 I9 1.52 X 10 I9 1.24 X 10 19 1.42 X 10 19 4 (E > 1.0 Mev) 8.99 X 10 [n/cm2] 3.51 X 10 19 3.55 X 10 I9 1 1.86 X 10 19 2.82 X 10 19 4.14 X 10 19
? f (E > 0.1 Mev) 3 [n/cm2]
2.07 X 10-2 2.54 X 10-2 2.09 X 10-2 2.26 X 10-2 Iron Atom Displacements 1.40 X 10-2 [dpa) 32.0 EFPY 15* 25* 35* 45* 0* 1.80 X 10 I9 2.69 X 10 19 3.04 X 10 19 2.47 X 10 I9 2.84 X 10 19 f (E > 1.0 Mev) [n/cm2] 19 5.65 X 10 19 8.29 X 10 I9 7.02 X 10 19 7.10 X 10 I9
+ (E > 0.1 MeV) 3.73 X 10
[n/cm2] 4.16 X 10-2 5.08 X 10-2 4.19 X 10-2 4.52 X 10-2 Iron Atom Displacements 2.80 X 10-2 [dpa)
TABLE 6-14 NEUTRON EXPOSURE VALUES FOR USE IN THE GENERATION OF HEATUP/COOLDOWN CURVES 16 EFPY NEUTRON FLUENCE (E > 1.0 MeV) SLOPE dea SLOPE 2 (n/cm ) (equivalent n/cm2) Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T 0* 8.99 x 10 18 4.88 x 10 18 1.04 x 10 18 8.99 x 10 18 5.67 x 10 I8 1.97 x 10 18 15* 1.34 x 10 l9 7.28 x 10 18 1.53 x 10 18 1.34 x 10 I9 8.42 x 10 18 2.91 x 10 18 25 (a) 1.52 X 10 l9 8.30 X 10 18 ' 79 X 10 18 1.52 X 10 I9 9.87 X 10 18 3.62 X 10 18 l 35' 1.24 x 10 19 6.73 x 10 18 1.44 x 10 18 1.24 x 10 I9 8.19 x 10 18 3.07 x 10 I8 45* 1.42 x 10 I9 7.59 x 10 18 1.53 x 10 18 1.42 x 10 l9 9.08 x 10 I8 3.11 x 10 I6 I 32 EFPY [ NEUTRON FLUENCE (E > 1.0 MeY) SLOPE doa SLOPE 2 (n/cm ) (equivalent n/cm2 ) l Surface 1/4 T 3/4 T Surface 1/4 T 3/4 T O' l.80 x 10 19 9.76 x 10 18 2.09 x 10 18 1.80 x IC I9 1.13 x 10 I9 3.94 x 10 I8 15* 2.69 x 10 19 1.46 x 10 19 3.06 - 10 18 2.69 x 10 l9 1.68 x 10 19 5.83 x 10 l8 l 3.04 X 10 I9 1.66 X 10 I9 3.59 X 10 18 3.04 X 10 I9 1.97 X 10 I9 7.24 X 10 I8 l 25 (a) 35* 2.47 x 10 19 1.35 x 10 l9 2.87 x 10 I8 2.47 x 10 19 1.64 x 10 l9 6.14 x 10 I8 45* 2.84 x 10 19 1.52 x 10 19 3.07 x 10 18 2.84 x 10 l9 1.82 x 10 I9 6.22 x 10 I8 i (a) Maximum point on the pressure vessel
TABLE 6-15 UPDATED LEAD FACTORS FOR V0GTEE ELECTRIC GENERATING PLANT UNIT 2 SVRVEILLANCE CAPSULES Caosule lead Factor V 3.94(a) Y 3.75(b) V 3.75(b) W 4.02(b) X 4.02(b) Z 4.02(b) (a) Plant specific evaluation based on end of cycle 1 calculated fluence. (b) Projection based on design basis flux. U ) l 4; I e 4 6-29
. SECTION 7.0 SURVEILLANCE CAFSULE REMOVAL SCHEDULE The following removal schedule meets ASTM E185-82 and is reconsnended for future capsules to be removed from the Vogtle Electric Generating Plant Unit 2 reactor > vessel:
S , Capsule Estimated Location Lead Fluence 2 Capsule (deg.) Factor Removal Time (b) (n/cm ) U 58.S 3.94 1.18 (Removed)(a) 4.44 x--1018-(Actual) i- Y 241.0 3.75 5,0 1.78 x 10I9 (c)
-V 61.0 3.73 9.0 3.21 x 1019 (d)
' X 238.5 4.02 15.0 5.74 X 1019 W 121.5 4.02 $tandby --- 301.5 4 . 11 2 Standby --- (a) Plant Specific Evaluation (b) Effective full Power Years (EFPY) from plant startup. 3 (c) Approximate fluence at 1/4 thickness of reactor. vessel wall at end of life (32 EFPY). (d) Approximate fluence at reactor vessel inner wall at end of life (32 EFPY).. 7-1 1
- - - - - _ _ - - _ - - _ _ - - . , _ , , - - , , . .-,....,-,..~m,.1-- - ,<-v- ,--~.emr .- w - - r e m nen,,,
-v-- en--- -~,,-..--.e -,-w _,w. , _ . . . , _
o2_4L.A 4 A-A 4.4.-.AaJ A-- A__.444% 4-'-' '4--- Lha- a---+-4--Ahw MM M- & 4 be hh 4+4' i m m 44h.h'J JA,e-4. A- m.JL-. 4e n- 4-4#448,4.ha'.4=db 4J A h e --s-
, AaK.J8 m+ E-G-.A_a-4 --p a M.-M-eAh,s-Y u
I l f t P F f I I I 1 i 9 4 n e > r I t t
)
W y i I D 6 h R I b I
?
s 6 1 h t
" .iI f
s & J
)
+
N O t t t e
- I i
i i t.
-w-, wwwy- ww yve mv'yy v'
SECTION
8.0 REFERENCES
- 1. L.R. Singer, et. al., " Georgia Power Company Alvin W. Vogtle Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-ll331, April 1986.
- 2. Code of federal Regulations, 10CFR50, Appendix G, " Fracture Toughness Requirements", and Appendix H, " Reactor Vessel Material Surveillance
, Program Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.
- 3. Regulatory Guide 1.99, Revision 2,
- Radiation Embrittlement of Reactor Vessel Materials", U.S. ' uclear Regulatory Commission, May,1988.
- 4. Section 111 of the ASME Boiler and Pressure Vessel Code, Appendix G,
" Protection Against Nonductile failure "
- 5. ASTM E208, " Standard Test Method for Conducting Drop-Weight lest to Determine Nil-Ductility Transition Temperature of Ferritic Steels."
- 6. ASTM E185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (lf)."
- 7. ASTM F23-88, " Standard Test Methods for Notched Bar impact Testing of Metallic Materials."
- 8. ASTM A370-69, " Standard Test Methods and Definitions for Mechanical 6 Testing of Steel Products "
- 9. ASTM EB-89b, " Standard Test Methods of Tension Testing of Metallic Materials."
- 10. ASTM E21-79 (1988), " Standard Practice for Elevated Temperature Tension Tests of Metallic Materials."
8-1
- 11. ASTM E83-85, " Standard Practice for Verification and Classification of Extensometers."
- 12. R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler, " Nuclear Rocket Shielding Methods, Modification, Updating and input Data .
Preparation. Vol. 5--Two-Dimensional Discrete Ordinates Transport lechnique", WANL-PR(LL)-034, Vol. 5, August 1970. *
- 13. "0RNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded, 47 ,
4 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors".
- 14. K.W. Bonadio, et. al., "The Nuclear Design and Core Physics j Characteristics of the Alvin W. Vogtle Unit 2 Nuclear Power Plant - Cycle !
1", WCAP-12099, December 1988. (Proprietary l I
- 15. ASTM Designation E482-82, " Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA. .
1984. l
- 16. ASTM Designation E560-77, " Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM Standards, Section i 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 17. ASTM Designation E693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)", in ASTM Standards, Section 12, American Society- for Testing and Materials, a Philadelphia, PA, 1984.
- 18. ASTM Designation E706-81a, " Standard Master Matrix for Light-Water Reactor -
Pressure Vessel Surveillance Standard", in ASTM Standards, Section l', American Society for Testing and Materials, Philadelphia, PA,1984.
- l 8-2
- 19. AS1H Designation E853-84, " Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984
- 20. ASTM Designation E261-77, " Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984, r
- 21. ASTM Designation E262-77, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques", in ASTM Standards, Section 12 American Society for Testing and Materials, Philadelphia, PA,1984.
- 22. ASTM Designation E263-82, " Standard Method for Determining fast-heutron Flux Density by Radioactivation of Iron", in ASTM Standards, Section ;2, American Society for Testing and Materials, Philadelphia, PA,1984.
- 23. ASTM Designation E264-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Nickel", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984,
- 24. ASTM Designation E481-78, " Standard Method for Measuring Neutron-Flux Density by Radioactivation of Cobalt and Silver", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 25. ASTM Designation E523-82, " Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Copper", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.
- 26. ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standa)Js, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.
8-3
I
- 27. -ASTM Designation E705-79, " Standard Method for Measuring fast-Neutron Flux l Density by Radioactivation of Neptunium-237", in ASTM Standards, Section !
12, American Society for Testing and Materials, Philadelphia, PA, 1984. , f
- 28. ASTM Design.stion E1005-84, " Standard Method for Application and Analysis .
[ of Radiomotric Monitors for Reactor Vessel Surveillance", in ASTM [ Standards, Section 12, American Society for Testing and Materials. , Philadelphia, PA, 1984. l l
- 29. F. A. Schmittroth, FERRET Data Analysis Core, HEDL-TME 79-40. Hanford l Engineering Development Laboratory, Richland, WA, September 1979. :
- 30. W. N. McElroy, S. Berg and T. Crocket, A Comouter-Automated Iterative f Method of Neutron Flux Scectra Determined by Foll Activation, .
AFWL-TR 7-41, Vol. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, } July 1967. ! i EPal-NP-2188, " Development and Demonstration of an Advanced Methodology 31. for LWR Dosimetry Applications", R. E. Maerker, et al., 1981. - ! e t i r d [ l . t i ;
'I 8-4 i g :
1
s
. I APPENDIX A Load-Time Records for Charpy Specimen Tests 4
) A-0
._ i
- W g
& = Wp r i
r gP a MAXIMUM LOAD L Pg = F R;.CTURE LOAD Pgy - GENER AL g YIELD LOAD l l 0
<t i O
s o i I I
> PA = ARREST LOAD i
g i / I I g i I ; l
'l I I I I
I l 1 l l t i
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l E+00 ' O.0000E+00 3.0000E+02 'i.6000E+03 2.4000E+03 Time, microseconds i i Figure A-2. Charpy impact test load-time record fcr Specimen BL9. i
, - , , , - - , , - - , . . . . - - , . . . . - , , , , _ . . ~ . . . , , . ..-- ,. .,--.. --,-m . ,+ ,-,,.-,,--.--.-,,,,,_,_.,..,-,,---.,.,-,,..,--~,.,,...,_,.,...x,. ., ,-,.---,,m., ,,._,--,y... -.
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4 r i l' i Figure A-3. Charpy i:npact test load-time record for Specimen BL10. 5 s 4 1 . 4 , b S g I D . 6
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Figure A-4. Charpy impact test load-time recocd for Specimen BL12.
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'i f p re A-7. Charpy impact test load-time record for Specimen BL2.
I
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I COMPUTER MALFUNCTION - NO LOAD TIME RECORD y (Data memory drop-out) l e 1 i Figure A-8. Charpy impact test load-time record for Specimen BLIS.
l I I i t t
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Figure A-11. Charpy impact test load-time record for Specimen EL7. i
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-.................. ........................v..
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o.000,0
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a i E+03 s O a . 1.5000 .c E+03 -> L 0.0000 . r- e. v. -A n.; s.. e..,, we.ee.a., a_._
. ,. c__ z_,.,____z;._ ei E+00 0.0000E+00 3.0000E+02 1.6000E+03 2.4000E+03 Time, microseconds Figure A-17. Charpy impact test load-time record for Specimen BT5.
s . , ,
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9.0000E+02 1.6000E+03 2.4000E+03 E+00 ' O.0000E400 Time, microseconds l Figure A-18. Charpy impact test load-time record for Specimen BT1.
l l l t l l E4.07 s a x 4.5000 F.+0.3 o
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6.0000................................................................................................................................ L. + 0.,a
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a u - . b l t
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~3,5 5+03 0,0000E+00 B,0000E+0$" "'~~~7's,,. ,}$5U0dk~+5 E+b0 Time, microseconds Figure A-22. Charpy impact test load-time record for Specimen BT7.
- - - _ - - - _ - _ _ _ _ _ _ _ _ _ = - _ _ _ _ - _ _ _ -
i i o.u0u,,............................................................................................................................... v t+va 4.5000 E+03 .. r, .s .,-
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fi fil)f)fi . k+bb ' O.0000E+00 8.0000E+di ~~ i ld56E56i ~~ T I0IE 705-~"'. Time, microseconds Figure A-23. Charpy impact test load-time record for Specimen BT14.
a . .. o.vuev r- ...v ~ - b._ . !.,_I ._
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- -:cvvvv fifiiiii .
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. h
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. r . .
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= - _- - . _ - - - - _ _ _ _ _ - _. _ _
I Y _ . .............................. .... ....... ............. ...........:..;.c...; . :2. .......u.. ....... ... , t= . 0000,1 E+03.l 4.5000 ',.,.r m F+0.3_ +: - : ? -' . y .- l'
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. a . .
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1; B.0000E+6E""~'""I.~5UUUN'+'UE"" .I000E53 k+in n ni 4hh '0s0000E+00 Time, microseconds Figure A.-30. Charpy impact test load-time record for Specimen BT4.
E+03 I I 4.5000 E+03 ' e,v'. a . I N r s./ "s,
.o .
-sa m j .*
a, , - T i0
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E+00 ' O.0000E+00 B.0000EI65~ Ild55E[+[UE~~~~'53666Idi~~~' Time, microseconds Figure A-31. Charpy impacat test load-time record for Specimen BT15. __________a _
- me 9 . V '.i n nt. V 2_ T 1,, 4I .h ,g b
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1
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. Figure A-33. Charpy imptct test load-timo record for Specimen BW14.
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E:.0000E+02 1.6000E+03 2.4000E+03 E+00 0.0000E+00 Tice, microseconds Figure A-34. Charpy impact test load-time record for Specimen BW11. e
i 1 1 C+n3 m vm
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. s.
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l t 1 i l l ._, l b.00,,..............................................................................................................................I
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, . = *
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w
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Time, microseconds i 1 Figure A-47. Charpy impact test load-time record for Specimen BH9. l i e
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~
y (Data memory drop-out) E Figure A-50. Charry impact test load-time record for Specimen BB15.
b,.0000 E+03
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t o.vuuu!i Itv0 A 4.5000
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1 i l t % e
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l
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APPENDIX B G EVALVATION OF PRESSVRIZED TiiERMAL SHOCK FOR V0G1LE ELECTRIC GENERATING PLANT UNIT 2 < k 4 4 4 B-0 ,
INil,000CT10N A limiting condition on reactor vessel integrity known as pressurized thermal shock (PTS) may occur during a severe system transient such as a ' loss-of-coolant-accident (LOCA) or a steam line break. Such transients may ' challenge the integrity of a reactor vessel under the following conditions: 6
- severe overcooling of the inside surface of the vessel wall followed by high repressurization significant degradation of vessel material toughness caused by radiation embrittlement
- the presence of a critical-size defect in the vessel wall Fracture mechanics analysis can be used to evaluate reactor vessel integrity under severe transient conditions.
In 1985 the Nuclear Regulatory Commission (NRC) issued a formal ruling on , pressurized thermal shock, it established screening criterion on pressurized water reactor (PWR) vessel embrittlement as measured by the nil-ductility , reference temperature, termed RTp . The NRC guidelines for calculating the RTPTS are defined in the PTS Rulei l. RTPTS screening values were set for beltline axial welds and plates and for beltline circumferential weld seams for end-of-life 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-life. The Nuclear Regulatory Commission has amended its regulations for light water nuclear power plants to change the $ procedure for calculating the radiation embrittlement. This revised PTS Rule was put'11shed in the federal Register, May 15, 1991 with an effective date of - June 14, 1991. This amendment makes the procedure for calculating RTPTS values consistent with the one given in Regulatory Guide 1.99, Revision 2I23 - The purpose of this report is to determine the reference temperature for B-1 , l
1 pressurized thermal shock (RTPTS) values for the Vogtle Electric Generating Plant Unit 2 reactor vessel to address the Pressurized Thermal Shock (PTS) Rule, l o PRESSURIZED THERHAL SHOCK 4 The PTS Rule requires that the PTS submittal be updated whenever there are changes in core loadings, surveillance measurements or other information that
- indicates a significant change in projected values.
The Rule outlines regulations to address the potential for PTS events on pressurtred water reactor (PWR) vessels in nuclear power plants that are operated with a license from the United States fluclear Regulatory Commission (USNRC). PTS events have been shown from operating experience to be transients l that result in a rapid and severe cooldown in the primary system coincident l with a high or increasing primary system pressure. The 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, thereby potentially affecting the integrity of the vessel.
4 The Rule establishes the following requirements for all domestic, operating PWR5: All plants must submit projected values of RTPTS 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 by six months after the efective date of this Rule if g the value of RTpys for any material is projected to exceed the screening criterie. Otherwise, it should be submitted with the next update of the pressure-temperature limits, or the next 3 reactor vessel surveillance report, or 5 years from the effective date of this Rule, whichever comes first. These i 1 - ! values must be calculated based on the methodoingy specified in this rule. The submittal must include the following:
- 1) the bases for the projection (including any assumptions regarding core loading patterns)
- 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 14RC, justification must be provided.)
! B-2
o The RTPTS (measure of fracture resistance) Screening Criterions for the reactor vessel beltline region is 270'F for plates, forgings axial welds 300*F for circumferential w, eld materials The following equations should be used to calculate the RTp valuesforeachweld,plateorforginginthereactorvesseIS beltline.
- Equation 1: RTPTS = 1 + H + ARTPTS Equation 2: ARTPTS - (CF)f(0.28-0.10 log f)
- All values of RTpTS rnust be verified ib be bounding values for the specific reactor vessel. In doing this each plant should
- consider plant-;pecific information that could affect the level l
of embrittlement.
- Plant-specific PTS safety analyses are required before a plant is within 3 years of reaching the Screening Criterinn, including analyses of alternatives to ninimize the PTS concern. l l
- NRC approval for operation beyond the Screening Criterion is !
required. j HETHOD FOR CALCULATION OF RTPTS In the PTS Rule, the NRC Staff has selected a conservative and uniform method for determining plant-specific values of RTPTS at a given time. , For the purpose of comparison with the Screening Critierion, the value of RTpj3 for the reactor vessel must be calculated for each weld and plate or forging in the beltline region as given below. RTPTS = 1 + H + ARTPTS, where ARTPTS - (CF)f(0.28-0.10 log f) I- Initial reference temperature (RTNDT) of the unirradiated material k H= Hargin to be added to cover uncertainties in the values of a initial RTNDT, copper and nickel contents, fluence and calculational procedures. H = 66*F for welds and 48'F for base metal if generic values of I are used. . H = 56*F for welds and 34*F for base metal if measured values of I are used. B-3
f= Neutron fluence, n/cm2 (E > IMeV at the clad / base metal interface), divided by 10 19 CF = Chemistry factor from tables (2) for welds and for base metals (plates and forgings) VERiflCATION OF PLANT-SPECIFIC HATERIAL PROPERTIES s Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties was performed. The beltline region is de ;ned by the PTS Rulelll to be "the region of the reactor vessel (shell material including weids, 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 irradiation damage to be considered in the selection of the most limiting material with regard to radiation damage " figure 1 identifies and indicates the location of all beltline region materials for the Vogtle Electric Genertting plant Unit 2 reactor vessel. Material property values were derived from vessel fabrication test certificate results. Fast neutron irradiation-induced changes in the tension, fracture and impact properties of reactor vesssi materials are largely dependent on chemical composition, particularly in the copper concentration, The variability in irradiation-induced property changes, which exists in general, is compounded by the variability of copper concentration with the weldments, A summary of the pertinent chemical and mechanical properties of the beltlip g region plate and weld materials of the Vogtle Electric Generating Plant Unit 2 Unit I reactor vessel are given in Table 1. All of the initial RTNDT values (1-RTNDT) are also presented in Table II33, B-4 l
P R4 101-124B '! y 9 R4-2 l d J00 , ; 0 0 __ 1300 f g . 101-124A 4 9
~ i R4-1 2700 101-124C q - 101-171 101 142A 0
RB-1 90 B8628-1 h i 0 ! 180 0 a[N t o 101-142C.) 101-142B 0 270 88825-1 I Figure 1. Identification and location of Beltline Region Material for for the Vogtle Electric Generating Plant Unit 2 Reactor Vessel B-5 l
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TABLE 1 i V0GTLE ELECTRIC GENERATING PLANT UNIT 2 REACTOR VESSEL BELTLINE REGION MATERIAL PROPERTIES G CU NI l-RTNDT
- Material Description (%) (%) ('f)
, Intermediate Shell, R4-1 0.06 0.64 10 Intermediate Shell R4-2 0.05 0.62 10 Intermediate Shell, R4-3 0.05 0.59 30 v Lower Shell, B8825-1 0.05 0.59 40 Lower Shell, R8-1 0.06 0.62 40 Lower Shell, B8628-1 0.05 0.59 50 Longitudinal Welds 0.07 0.13 -10 l Circumferential Weld 0.06 0.12 -30 NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E>l MeV) at the inner surface of the Vogtle Electric Generating Plant Unit 2 reactor vessel is shown in figure 2.
These values were projected using the results of the Capsule U radiation surveillance program (See Section 6 of this report) and are presented in Table 2. TABLE 2 NEUTkON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE y '!0GTLE ELECTRIC GENERATING PLANT UNIT 2 PRESSURE VESSEL CLAD / BASE HETAL INTERFACE FOR 32 EFPY l w l
- O' 15' 25' 35' 45' fluence x 1019 n/cm 2 1.80 2.69 3.04 2.47 2.84 (E > 1 HeV)
B-6
DETERMINATION OF RTPTS VALUES FOR ALL BELTLINE REGION MATERIALS Using the prescribed PTS Rule methodology, RTPTS values were generated for all beltline region materials of the Vogtle Electric Generating Plant Unit 2 ,- reactor vessel as a function of end-of-life (32 EFPY) and 48 EFPY fluence values. The fluence data were generated based on the most recent ' surveillance capsule program 3 results. Table 3 provides a summary of the RTPTS values for all beltline reg:on < materials for the end-of-life (32 EFPY) and 48 EFPY using the PTS Rule. The PTS Rule requires that each p1&nt assess tia RTp73 values based on plant specific surveillance capsule data under certain conditions. These conditions are: Plant specific surveillance data has been deemed credible 1 defined in Regulatory Guide 1.99, Revision 2, and RTPTS values change significantly. (Changes to RTPTS values are considered significant if the value determined with RTPTS equations (1) and (2), or that using capsule data, or both, exceed the screening criteria prior to the expiration of the operating license, including any renewed term, if applicable, for the plant.) For Vogtle Electric Generating Plent Unit 2, the use of plant specific surveillance capsule data does not arise because of the following retsons:
- 1) Capsule V it the first capsule removed from the reactor vessel, hence the data is not credibles per Reference 2.
f
- 2) Based on the material chemistry and projected fluence the RTPTS values are not expected to exceed the RTPTS screening criteria up to the end of license life.
B-7 l
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- 14 - $ 12 -
b 10 - h 8-6- 4-2- 0-C i , , , , , 0 5 10 15 20 25 30 35 EFPY Figure 2. Fluence vs. Effective Full Power Years for Vogtle Electric Generating Plant Unit 2 TABLE 3 RTPTS VALUES FOR V0GTLE ELECTRIC GENERATING PLANT UNIT 2 MATERIAL 32 EFPY 48 EFPY Intermediate Shell Plate, R4-1 92 95 . Intermediate Shell Plate, R4-2 84 87 Intennediate Shell Plate, R4-3 104 107 Lower Shell Plate, B8825-1 114 117 Lower Shell Plate, R8-1 122 125 Lower Shell Plate, B8628-1 124 127 Longitudinal Welds 107 111 Circumferential Weld 82 86 B-8
CONCLUSIONS As shown in Table 3, all the RTPTS values remain below the NRC screening values for PTS using the projected fluence values for both the end-of-life (32 EFPY) and 48 EFPY. Using the PTS Rule, the highest RTPTS values at 32 , EFPY and 48 EFPY are at the lower shell plate, (B8628-1). These values are l 124'F and 127'F, respectively. Plots of the RTPTS values versus '
)
the fluence are shown in Figures 3 and 4. These plots indicate that none of
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REFERENCES [1] 10CFR Part 50, " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events", June 14, 1991. [2] Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1989. [3] MT/ SMART-213(88), "Vogtle Units 1 & 2 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation", N. K. Ray, November, 1988. I B-12
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