ML20081H285
| ML20081H285 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 03/31/1995 |
| From: | Fero A, Peter P, Zawalick S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20081H259 | List: |
| References | |
| WCAP-14228, NUDOCS 9503240157 | |
| Download: ML20081H285 (133) | |
Text
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.a Westinghouse Class 3 (Non-Proprietary) l WCAP-14228 l
ANALYSIS OF CAPSULE X FROM THE COMMONWEALTH EDISON COMPANY i
BRAIDWOOD UNIT 2 REACTOR VESSEL i
RADIATION SURVFTI T.ANCE PROGRAM r
P. A. Peter
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S. S. Zawalick A.H. Fero J. F. Williams i
i March 1995 f
Work Performed Under Shop Order BWSP-106A i
Pirpared by Westinghouse Electric Corporation j
for the Commonwealth Edison Company
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Approved by:
. LV. _ Ap l
R. D. Rishel, Manager j
Metallurgical & NDE Analysis WESTINGHOUSE ELECTRIC CORPORATION j
Nuclear Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 l
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01995 Westinghouse Electric Corp.
1 All Rights Reserved I
9503240157 950321 ADOCK0500g6 PDR P
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PREFACE i
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t his report has been technically reviewed and verifed.
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. I Reviewer:
Sections 1 through 5,7,8 and Appendix A E. Terek Section 6 S. L. Anderson
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TABLE OF CONTENTS Section Title Eass
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LIST OF TABLES iii t
I LIST OF ILLUSTRATIONS vii-l f
1.0
SUMMARY
OF RESULTS 1-1 I
2.0 INTRODUCTION
2-1 i
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3.0 BACKGROUND
3-1
4.0 DESCRIPTION
OF PROGRAM 4-1 f
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5.0 TESTING OF SPECIMENS FROM CAPSULE X 5-1 t
5.1 Overview 5-1 5.2 Charpy V-Notch Impact Test Results 5-3
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5.3 Tension Test Results 5-5 5.4 Fracture Toughness Tests 5-6 i
6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 f
6.1 Introduction 6-1 t
6.2 Discrete Ordinates Analysis 6-2 i
6.3 Neutron Dosimetry 6-6
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6.4 Projections of Pressure Vessel Exposure 6-10 i
7.0 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE 7-1
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8.0 REFERENCES
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APPENDIX A - LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS A-0 i
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1 LIST OF TABLES Iahk
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. Heat Treatment of the Braidwood Unit 2 Reactor Vessel Surveillance Materials -
4-3 i
f 2' Original Chemical Composition of the Braidwood Unit 2 Reactor Vessel 4-4 Surveillance Materials 4-3 Chemical Composition of Braidwood Unit 2 Charpy Specimens Removed from.
4-5 q
i Capsule X E
f 4-4 Chemistry Results from the Low Alloy Steel NIST Cenified Reference Standards 4-8 I
4-5 Calculation of Average Cu and Ni Weight Percent to be Used in Surveillance 4-9 j
Capsule Calculations l
5-1 Charpy V-notch Impact Data for the Braidwood Udt 2 Lower Shell Forging 5-7 (Tangential Orientation) Irradiated to a Fluence of 1.126 x 10 n/cm2 (E > 1.0 MeV) i 5-2 Charpy V-notch Impact Data for the Braidwood Unit 2 Lower Shell Forging 5-8 (Axial Orientation) Irradiated to a Fluence of 1.126 x 10 n/cm (E > 1.0 MeV):
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5-3 Charpy V-notch Impact Data for the Braidwood Unit 2 Surveillance Weld Metal 5-9 l
Irradiated to a Fluence of 1.126 x 10 n/cm (E > 1.0 MeV) f 2
5-4 Charpy V-notch Impact Data for the Braidwood Unit 2 Heat-Affected-Zone (HAZ) 5 i Metal Irradiated to a Fluence of 1.126 x 10 n/cm (E > 1.0 MeV) 2 5-5 Instrumented Charpy Impact Test Results for the Braidwood Unit 2 Lower Shell 5-11 Forging (Tangential Orientation) Irradiated to a Fluence of 1.126 x 10" n/cm -
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(E > 1.0 MeV) l e
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LIST OF TABLES (CONTINUED) na rn em 5-6 Instrumented Charpy Impact Test Results for the Braidwood Unit 2 Lower Shell 5-12 Forging (Axial Orientation) Irradiated to a Fluence of 1.126 x 10 n/cm2 (E > 1.0 MeV) l 5-7 Instrumented Charpy Impact Test Results for the Braidwood Unit 2 Surveillance 5-13 Weld Metal Irradiated to a Fluence of 1.126 x 10 n/cm (E > 1.0 MeV) 2 5-8 Instmmented Charpy Impt.ct Test Results for the Braidwood Unit 2 Heat-Affectei-5-14 2
Zone (HAZ) Metal Irradiated to a Fluence of 1.126 x 10 n/cm (E > 1.0 MeV) 5-9 Effect of Irradiation to 1.126 x 10 n/cm (E > 1.0 MeV) on the Notch
.5-15 2
Toughness Properties of the Braidwood Unit 2 Reactor Vessel Surveillance Materials 5-10 Comparison of the Braidwood Unit 2 Surveillance Material 30 ft-lb Transition 5-16 i
Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 5-11 Tensile Properties for the Braidwood Unit 2 Reactor Vessel Surveillance 5-17 Materials Irradiated to 1.126 x 10 n/cm (E > 1.0 MeV) 2 I
l 6-1 Calculated Fast Neutron Expost e Rates at the Surveillance Capsule Center 6-14 l
6-2 Calculated Azimuthal Variation of Fast Neutron Exposure Rates at the 6-15 Pressure Vessel Clad / base Metal Interface i
6-3 Relative Radial Distribution of & (E > 1.0 MeV) Within the Pressure Vessel 6-16 Wall iv
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I LIST OF TABLES (CONTINUED)
Iakk Illk Eagn 6-4 Relative Radial Distribution of & (E > 1.0 MeV) Within the Pressure Vessel 6-17 Wall i
65 Relative Radial Distribution of dpa/see Within the Pressure Vessel Wall 6-18 j
6-6 Nuclear Parameters Used in the Evaluation of Neutron Sensors 6-19 i
6-7 Monthly Thermal Generation During the First Four Fuel Cycles of the 6-20 Braidwood Unit 2 Reactor 6-8 Measured Sensor Activities. Saturated Activities and Reaction Rates 6-21
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Surveillance Capsule U 6-9 Measured Sensor Activities, Saturated Activities and Reaction Rates 6-22 Surveillance Capsule X i
6-10 Summary of Neutron Dosimetry Results Surveillance Capsules U and X 6-23 l
6-11 Comparison of Measured and FERRET Calculated Reaction Rates at the 6-24 i
t Surveillance Capsule Center i
6-12 Adjusted Neutron Energy Spectrum at the Center of Surveillance Capsule U 6-25 6-13 Adjusted Neutron Energy Spectrum at the Center of Surveillance Capsule X 6-26 l
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6-14 Comparison of Calculated and Measured Neutron Exposure Levels for Braidwood 6-27 Unit 2 Surveillance Capsules U and X 6-15 Neutron Exposure Projections at Key Locations on the Pressure Vessel 6-30 Clad / base Metal Interface v
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LIST OF TABLES (CONTINUED) l
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.6-16 Neutron Exposure Values for the Braidwood Unit 2 Reactor Vessel 6-29 i
6-17 Updated Lead Factors for Braidwood Unit 2 Surveillance Capsules 6-30 7-1 Surveillance Capsule Withdrawal Schedule 7-1 f
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LIST OF ILLUSTRATIONS EiEHm Illis East 4-1 Arrangement of Surveillance Capsules in the Braidwood Unit 2 Reactor Vessel 4-11 4-2 Capsule X Diagram Showing Location of Specimens, Thermal Monitors, 4-12 and Dosimeters 5-1 Charpy V-Notch Impact Properties for Braidwood Unit 2 Reactor Vessel Lower 5-18 Shell Forging (Tangential Orientation) 5-2 Charpy V-Notch Impact Properties for Braidwood Unit 2 Reactor Vessel Lower 5-19 Shell Forging (Axial Orientation) 5-3 Charpy V-Notch Impact Properties for Braidwood Unit 2 Reactor Vessel 5-20 Surveillance Weld Metal 5-4 Charpy V-Notch Impact Properties for Braidwood Unit 2 Reactor Vessel Weld 5-21 Heat-Affected-Zone (HAZ) Metal 5-5 Charpy Impact Specimen Fracture Surfaces for Braidwood Unit 2 Reactor Vessel 5-22 Lower Shell Forging (Tangential Orientation) 5-6 Charpy Impact Specimen Fracture Surfaces for Braidwood Unit 2 Reactor Vessel 5-23 Lower Shell Forging (Axial Orientation) 5-7 Charpy Impact Specimen Fracture Surfaces for Braidwood Unit 2 Reactor Vessel 5-24 Surveillance Weld Metal 5-8 Charpy Impact Specimen Fracture Surfaces for Braidwood Unit 2 Resctor Vessel 5-25 Weld Heat-Affected-Zone (HAZ) Metal vii
LIST OF H.LUSTRATIONS (CONTINUED) 7 F.isiin:
T.111e Eass 5-9 Tensile Properties for Braidwood Unit 2 Reactor Vessel Lower Shell Forging 5-26 (Tangential Orientation) 5-10 Tensile Properties for Braidwood Unit 2 Reactor Vessel Lower Shell Forging.
5-27 (Axial Orientation) 3 t
5 Tensile Properties for Braidwood Unit 2 Reactor Vessel Surveillance Weld Metal ~
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f 5-12 Fractured Tensile Specimens from Braidwood Unit 2 Reactor Vessel Lower Shell 5-29 i
Forging (Tangential Orientation) l 5-13 Fractured Tensile Specimens from Braidwood Unit 2 Reactor Vessel lower Shell 5-30 Forging (Axial Orientation)
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5-14 Fractured Tensile Specimens from Braidwood Unit 2 Reactor Vessel Surveillance 5-31 W eld M etal
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5-15 Engineering Stress-Strain Curves for Forging Tensile Specimens FL10 and FL11.
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(Tangential Orientation)
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5-16 Engineering Stress-Strain Curve for Forging Tensile Specimen FL12 5-33 l
(Tangential Orientation) j 5-17 Engineering Stress-Strain Curves for Forging Tensile Specimens FTIO and Fril 5-34~
j (Axial Orientation) 5-18 Engineering Stress-Strain Curve for Forging Tensile Specimen FT12 5-35'
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LIST OF ILLUSTRATIONS (CONTINUED) i Esius Tills Eass l
5-19 Engineering Stress-Strain Curves for Wcld Metal Tensile Specimens FW10 5-36 and FW11 5-20 Engineering Stress-Strain Curve for Weld Metal Tensile Specimen FW12 5-37 t
6-1 Plan View of Dual Reactor Vessel Surveillance Capsule 6-12 6-2 Axial Distributi'an of deutron Fluence (E > 1.0 MeV) Along the 45 Degree 6-13 Azimuth j
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i SECTION 1.0
SUMMARY
OF RESULTS De analysis of the reactor vessel materials contained in surveillance Capsule X, the second capsule to be removed from the Commonwealth Edison Company Braidwood Unit 2 reactor pressure vessel, led to the following conclusions:
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he capsule received an average fast neutron fluence of 1.126 x 10 n/cm (E > 1.0 MeV) after 4.215 Effective Full Power Years (EFPY) of plant operation.
Irradiation of the reactor vessel lower shell forging Charpy specimens, oriented with the j
o longitudinal axis of the specimen parallel to the major rolling direction (tangential orientation), to 1.126 x 10 n/cm (E > 1.0 MeV) resulted in a 30 ft-lb transition i
2 temperature increase of 3 F and a 50 ft-lb transition temperature increase of 0*F. This results in an irradiated 30 ft-lb transition temperature of-7 F and an irradiated 50 ft-lb j
transition temperature of 15 F.
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I radiation of the reactor vessel lower shell forging Charpy specimens, oriented with the j
longitudinal axis of the specimen normal to the major rolling direction (axial orientation), to 1.126 x 10" n/cm (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of f
2 35 F and a 50 ft-lb transition temperature increase of 35'F. This results in an irradiated 30 ft-lb transition temperature of 10 F and an irradiated 50 ft-lb transition temperature of 35*F.
Irradiation of the weld metal Charpy specimens to 1.126 x 10 n/cm (E > 1.0 MeV) 2 o
resulted in a 30 ft-lb transition temperature increase of 20*F and a 50 ft-lb transition i
temperature increase of 5*F. His results in an irradiated 30 ft-lb transition temperature of 0 F and an irradiated 50 ft-lb transition temperature of 45*F for the weld metal.
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Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal Charpy specimens to o
1.126 x 10 n/cm (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of j
2 0*F and a 50 ft-lb transition temperature increase of 0*F. This results in an irradiated 30 ft-lb transition temperature of -135*F and an irradiated 50 ft-lb transition temperature of
-105'F for the weld HAZ metal.
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o Irradiation oflower shell faging ('a====*i=1 orientation) to 1.126 x 102' n/cm (E > 1.0 2
MeV) resulted in an irradiated average upper shelf energy decrease of I ft-lb, resulting in an
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irradiated upper shelf energy of 167 ft-lb.
I Irradiation oflower shell forging (axial orientation) to 1.126 x 10 n/cm (E > 1.0 MeV) -
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i resulted in an irradiated average upper shelf energy decrease of 8 ft-lb, resulting in an i
arradiated upper shelf energy of 145 ft-lb.
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The average upper shelf energy of the weld metal decreased I ft-lb after irradiation to 1.126 o
r 10 n/cm (E > 1.0 MeV). This results in an irradiated upper shelf energy of 70 ft-lb for 2
die weld metal sperimanc f
o The average upper shelf energy of the weld HAZ metal decreased 30 ft-lb after irradiation to 1.126 x 10 n/cm (E > 1.0 MeV). 'Ihis results in an irradiated upper shelf energy of 2
t 125 ft-lb for the weld HAZ metal.
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The surveillance Capsule X test results indscate that all measured values are less than the f
o Regulatory Guide 1.99, Revision 2"1. predictions for the 30 ft-lb transition temperature shift.
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The surveillance Capsule X test results indicate that the measured upper shelf energy o
decreases of all surveillance materials are less than the Regulatory Guide 1.99, Revision 2,'
predictions (Table 5-10). Additionally, all belthne materials exhibit a more than adequate upper shelf energy level for mnnnued safe plant operation and are expected to maintain an upper shelf energy of no less than 50 ft-lb throughout the life of the vessel (32 EPPY) as
-l required by 10CFR50, Appendix Gm,
)l The calculated end-of-life (32 EPPY) maximum neutron fluence (E > 1.0 MeV) for the o
Braidwood Unit 2 reactor vessel is as follows:
Vessel inner radius * - 2.199 x 10 n/cm2 Vessel 1/4 thickness - 1.201 x 10 n/cm2 l
Vessel 3/4 thickness - 2.595 x 10 n/cm2 3j
- Clad / base metal interface I
1-2 j
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SECTION
2.0 INTRODUCTION
This report presents the results of the examination of Capsule X, the second capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Commonwealth Edison Company Braidwood Unit 2 reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Braidwood 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 are presented in WCAP-lll88, entitled " Commonwealth Edison Company Braidwood Station Unit No. 2 Reactor Vessel Radiation Sumillance Program" by L. R. SingeA The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-82,
" Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels". Estinghouse Energy Systems personnel were contracted to aid in the preparation of procedures for removing Capsue X 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 X removed from the Commonwealth Edison Company Braidwood Unit 2 reactor vessel and discusses the analysis of the data.
2-1
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3.0 BACKGROUND
De 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 most critical region of the vessel because it is subjected to significant fast neutron bombardment. He overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA 508 Class 3 (base material of the Braidwood Unit 2 reactor pressure vessel lower shell forging) are well documented in the literature.
Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.
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A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in " Protection Against Nonductile Failure", Appendix G to Section XI of the ASME Boiler W
and Pressure Vessel Code. The method uses fracture mechanics concepts and is based on the reference nil-ductility temperature (RTym).
RTym is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208W) or the temperature 60*F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens with the longitudinal axis oriented normal (axial orientation) to the major working direction of the forging. The RTum of a given material is used to index that material to a reference stress intensity factor curve (K, curve) which appears in Appendix G to the ASME Code. The K curve is a lower bound of dynamic, crack arrest, and static fracture u
l toughness results obtained from several heats of pressure vessel steel. When a given material is l
indexed to the Ka curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.
RTum and, in turn, the operating limits of nuclear power plants can be adjusted to account for the l
cftects of radiation on the reactor vessel material properties. The radiation embrittlement changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program, such as the Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program,
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in which a surveillance capsule is periodically removed from the operating nuclear reactor and the I
encapsulated specimens tested. The increase in the average Charpy V-notch 30 ft-lb temperature 3-1
(ARTum) due to irradiation is added to the original RT,, to adjust the RT,, for radiation embrittlement. 'Ihis adjusted reference temperature (ART) (ART = Initial RT,, + ART,m) is used '
to index the material to the K curve and, in tum, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.
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4.0 DESCRIPTION
OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Braidwood Unit 2 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant start-up. The six capsules were positioned in the reactor vessel between the neutron shielding pads and the vessel wall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.
Capsule X was removed after 4.215 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 forging 50D102-1/50C97-1 and weld metal identical to the upper (49D963-1/49C904-1) to lower shell beltline weld seam of the reactor vessel. Capsule X also contained Charpy V-notch specimens from the weld Heat-Affected-Zone (HAZ) of forging 50D102-1/50C97-1.
Test material obtained from the lower shell forging (after thermal heat treatment and forming of the forging) was taken from at least one forging thickness from the quenched ends of the forging. All test specunens were machined from the 1/4-thickness location of the forging after performing a simulated post-weld stress-relieving treatment on the test material. Test specimens were also removed from weld and heat-affected-zone metal of a stress-relieved weldment joining the upper and lower shell forgings.
All heat-affected-zone specimens were obtained from the weld heat-affected-zone of the lower shell forging.
Charpy V-notch specimens from the lower shell forging were machined in both the tangential orientation (longitudinal axis of the specimen parallel to the major working direction) and axial orientation (longitudinal axis of the specimen perpendicular to the major working direction). The Charpy V-notch specimens from the weld metal were machined such that the longitudinal axis of the specimen was transverse to the weld direction. 'Ibe notch was machined such that the direction of crack propagation is in the weld direction.
Tensile specimens from the lower shell forging were machined with the longitudinal axis of the specimens both parallel and nonnal to the major working direction of the forging. Weld specunens were oriented transverse to the weld direction.
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Compact tension test specimens from the lower shell forging were machined in both tangential and axial orientations. Compact 'ension test specimens from the weld metal were machined normal to the weld direction with the notch oriented in the direction of the weld. All compact tension test specimens were fatigue pre-cracked according to ASTM E399.
The heat treatment and chemical composition of the unirradiated surveillance materials is presented in Tables 4-1 and 4-2, respectively. In addition, a chemical analysis using Inductively Coupled Plasma Spectrometry (ICPS) was performed on sixteen irradiated Charpy specimens,15 weld metal and one lower shell forging metal, and is reported in Table 4-3. The chemistry results from the NIST certified reference standards are reported in Table 4-4. The results were obtained from Westinghouse Electric Corporation Nuclear Services Division CMT Analytical Laboratory under analytical request number 15482. Table 4-5 provides the calculation of the average Cu and Ni weight percent values of the reactor vessel beltline materials, which are used in the Braidwood Unit 2 surveillance Capsule X calculations.
Capsule X contained dosimeter wires of pure iron, copper, nickel, and aluminum 0.15 weight percent cobalt wire (cadmium-shielded and unshielded). In addition, cadmium-shielded dosimeters of Neptunium (Np ") and Uranium (U2") used to measure the integrated flux at specific neutron energy 2
levels were included in the capsule.
i Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule. These thermal monitors 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:
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 test specimens, dosimeters and thermal monitors contained in Capsule X is shown in Figure 4-2.
4-2 i
i TABLE 4-1 i
Heat Treatment of the Braidwood Unit 2 Reactor Vessel Surveillance Materials"1 I
i Material Temperature ('F)
Tune (hours)
Coolant Upper Shell Forging Austenitizing 6.75" Water-quenched 49D9631/49C904-1 1600
- 25 (MK-24 2)
Tempered 12.25'"
Air cooled 1225 1 25 Stnss Relief 11.75*
Furnace-cooled 1150 i 50 t
Lower Shell Forging Austenitizing 6.FS Water-quawhed i
50D102-1/50C97-1 1600 *25 (MK 24-3)
Tempered 12.25'd Air cooled 1225
- 25 Stress Relief 11.75*
Furnace-cooled 1150 *SO Upper to Lower Stress Relief 11.75*
Furnace-cooled Shell Closing Girth 1150 1 50 Weld Seam Gleat # 442011. Flux Linde 80. Imt # 0344)
Surveillare Pmgram Test Material Surveihre Program Post Weld Stress 14.25Sd Furnace. cooled Test Forging Relief 50D102-1/50C97-1 1150 i 50 s
Surveihre Pmgram Post Weld Stress 12.5 "
Furnace cooled Test Weldment Relief 1150 i50 t
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I (a) Data obtahed from Japan Steel Worts, tad. Matenal Test Reports (b) Data frurn Babcock and Wilcox. Co.
(c) The Stress Relief Heat Treatment moeived by the Surveinance Test Forging and Weldenent have tnen shoulated.
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TABLE 4-2 i
Original Chemical Composition of the Braidwood Unit 2 Reactor Vessel Surveillance Materials
- Element Upper Shell Forgmg Lower Shell Forging Weld Metal 49D9631/49C9041 50D1021/50C971 (Heat # 442011. Linde 80 flux.
Lot # 0344)
C-0.20*
0.22
- 0.24
- 0.066"
0.069'"
Mn 133 1.30 1.38 1.44 1.45 l
P 0.007 0.006 0.013 0.015 0.011 S
0.007 0.004 0.009 0.012 0.013 I
Si 0.25 0.28 0.30 0.48 0.53 Ni 0.71 0.75 0.77 0.67 0.64 Mo 0.53 0.49 0.56 0.44 0.46 Cr 0.08 0.08 0.095 0.10 0.082 Cu 0.03 0.06 0.057 0.04 0.040 A1 0.024 0.025 0.024 0.004 0.007 Co 0.012 0.011 0.008 0.011 0.004 Pb 0.0003 max 0.0003 max
<0.001 0.0006
<0.001 W
0.005 max 0.005 max
<0.01 0.010
<0.01 T
0.001 max 0.005 max 0.004 0.007 0.003 Zr 0.005 max 0.005 max
<0.002 0.003
<0.002 V
0.01 max 0.01 max
<0.002 0.005
<0.002 t
Sn 0.009 0.007 0.004 0.005 0.004 f
As 0.006 0.008 0.007 0.004 0.004 Cb 0.005 max 0.005 max
<0.002 0.004
<0.002 N
0.0097 0.0084 0.009 0.013 0.012 2
B Not Reported Not Reported
<0.001 0.007
<0.b01 dE-5 (a) Data repor1ed here is the unirradiated chemistry reschs irported in WCAP Il188,
l (b) Ormical Analysis by Japan Steel Works, Ltd.
(c) Westinghouse Analyses frcen tbc Surveillance Program Test Plate.
(d) Owmical Analysis of Tdler Wirr Qualification Tem" by liabcock and Wilcox. Cornpany. Test No. %T-562.
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TABLE 4-3 Chemical Composition of Braidwood Unit 2 Charpy Specimens Removed from Surveillance Capsule X Weld Metal Bement FW-51 FW-54 FW-57 FW-53 FW-55 i
Fe 96.411
%.425 96.469
%.199
%.354 Co
<0.009
<0.009
<0.009
<0.011
<0.009 Cr 0.094 0.093 0.095 0.104 0.097 Cu 0.033 0.034 0.033 0.038 0.035 Mn 1.634 1.642 1.591 1.731 1.663 Mo 0.481 0.474 0.479 0.520 0.491 Ni 0.724 0.711 0.714 0.780 0.737 P
0.017 0.015 0.017 NA NA Ti
<0.007
<0.007
<0.007 NA NA V
<0.003 0.003 0.006 NA NA Al
<0.015
<0.016
<0.015 NA NA i
As
<0.022
<0.023
<0.021 NA NA l
B 0.006
<0.005 0.006 NA NA Nb
<0.008
<0.006
<0.004 NA NA Ta NA NA NA NA NA Pb NA NA NA NA NA Sn
<0.034
<0.035
<0.033 NA NA W
<0.001
<0.001
<0.001 NA NA Zr 0.001 0.001 0.001 NA NA Carbon 0.064 0.069 0.%9 NA NA Sulfur 0.0080 0.0096 0.0090 NA NA Silicon 0.517 0.492 0.509 NA NA NA represents elements not requested for analysis 4-5 l
TABLE 4-3 (CONTINUED)
Chemical Composition of Braidwood Unit 2 Charpy Specimens i
t Removed from Surveillance Capsule X Weld Metal l
Element FW-56 FW-58 FW-59 FW-60 FW-46 Fe
%.380
%.330
%.336
%.412
%.425 Co
<0.009
<0.009
<0.009
<0.008
<0.009 Cr 0.092 0.097 0.095 0.096 0.095 Cu 0.033 0.032 0.032 0.031 0.032 l
Mn 1.669 1.678 1.682 1.630 1.643 Mo 0.483 0.488 0.492 0.482 0.474 Ni 0.728 0.752 0.743 0.730 0.711 l
P NA NA NA NA NA
'Il NA NA NA NA NA V
NA NA NA NA NA M
NA NA NA NA NA M
NA NA NA NA NA B
NA NA NA NA NA Nb NA NA NA NA NA Ta NA NA NA NA NA Pb NA NA NA NA NA Sn NA NA NA NA NA W
NA NA NA NA NA Zr NA NA NA NA NA Carbon NA NA NA NA NA Silicon NA NA NA NA NA Sulfur NA NA NA NA NA NA represents elements not requested for analysis 1
4-6 i
TABLE 4-3 (CONTINUED) l Chemical Composition of Braidwood Unit 2 Charpy Specimens Removed from Surveillance Capsule X 1
l Base Weld Metal Metal Element FW-47 FW-48 FW-49 FW-50 FW-52 FIA7 Fe
%.338
%.461
%.610
%.406
%.515 95.964 Co
<0.008
<0.008
<0.011
<0.009
<0.009
<0.010 Cr 0.097 0.091 0.089 0.093 0.091 0.099 Cu 0.032 0.031 0.032 0.033 0.033 0.056 Mn 1.701 1.633 1.476 1.674 1.595 1.411 Mo 0.480 0.468 0.457 0.471 0.454 0.539 Ni 0.728 0.703 0.687 0.703 0.695 0.804 P
NA NA NA NA NA 0.010 T1 NA NA NA NA NA
<0.007 V
NA NA NA NA NA
<0.003 Al NA NA NA NA NA 0.029 As NA NA NA NA NA
<0.020 B
NA NA NA NA NA 0.007 Nb NA NA NA NA NA
<0.007 Ta NA NA NA NA NA NA Pb NA NA NA NA NA NA Sn NA NA NA NA NA
<0.031 W
NA NA NA NA NA
<0.001 Zr NA NA NA NA NA 0.001 Carbon NA NA NA NA NA 0.223 Sulfur NA NA NA NA NA 0.0034 Silicon NA NA NA NA NA 0.280 NA represents elements not requested for analysis 4-7
.i TABLE 4-4 rhemicry Results from the IAw Alloy Steel NIST Certified Reference Standards
]
w Concentration in Weight Percent NIST 361 NIST 362 NIST 363 NIST 364 Metal Certified Meas.
Certified Meas.
Certified Meas.
Certified Meas.
Fe 95.600 95.5 %
95.300 95.295 94.400 94.131
%.700 97.028 Co 0.032 0.029 0300 0355 0.048 0.049 0.150 0.151 Cr 0.694 0.698 0300 0342 1310 1.415 0.063 0.063 Cu 0.M2 0.046 0.500 0.609 0.100 0.113 0.249 0.249 Mn 0.660 0.656 1.040 1.188 1.500 1.588 0.255 0.252 Mo 0.190 0.197 0.%8 0.070 0.028 0.025 0.490 0.474 Ni 2.000 2.054 0.590 0.663 0390 0332 0.144 0.136 P
0.014 0.017 0.041 0.041 0.029 0.034 0.010 0.010 Ti 0.020 0.018 0.084 0.028 0.050 0.055 0.240 0.272 V
0.011 0.008 0.040 0.044 0310 0341 0.105 0.109 l
Al 0.021 0.024 0.095 0.100 0.240 0.282 0.008 NA As 0.017
<.0.022 0.092 0.097 0.010
<0.024 0.052 0.039 B
0.000
<0.0M 0.003
<0.005 0.001
<0.005 0.011 0.013 Nb 0.022 0.018 0.290
<0.010 0.049 0.043 0.152 0.030 Ta 0.020 NA 0.200 NA 0.053 NA 0.110 NA Pb 0.000 NA 0.000 NA 0.002 NA 0.023 NA Sn 0.010
<0.034 0.016
<0.039 0.lM 0.107 0.008
<0.031 W
0.017 0.M1 0.200 0.274 0.046 0.044 0.100 0.100 g
I Zr 0.009 0.009 0.190 0.253 0.049 0.M7 0.068 0.068 C
0383 0383 0.160 0.161 0.620 NA 0.067*
0.067
(
S 0.0140 NA 0.036 0.0354 0.0068 NA 0.025*
0.0211 Si 0.222 0.222 0390 0.415 0.74 NA 0.065*
NA NA represents elements not requested for analysis
- NIST 121 4-8
TABLE 4-5 Calculation of Average Cu a.d F W,ight Percent to be Used in Surveillance ( -
A-'i tions
. f'.
r _ 27r tJtl2 Upper Shell Fe.pm, Lower Shell Forging 49D963-1/49C904-1 50D102-1/50C97-1 Weld Metal WF562 "
Reference Cu (wt. %)
Ni(wt.%)
Cu (wt. %)
Ni(wt.%)
Cu (wt. %)
Ni(wt.%)
3 0.03*
0.71*
0.06*
0.75*
0.G4*
0.67*
3 0.057 0.77 0.04 0.64 6
0.049 0.745 0.032 0.704 6
0.034 0.754 6
0.032 0.698 6
0.026 0.623 6
0.028 0.635 6
0.031 0.679 6
0.029 0.644 6
0.032 0.699 6
0.034 0.765 6
0.031 0.673 6
0.034 0.724 6
0.035 0.747 6
0.033 0.711 6
0.031 0.688 6
0.035 0.750 6
0.031 0.685 7
0.03 0.71 8
0.06 0.75 9
0.04 0.67 10 0.056 0.804 0.033 0.724 10 0.034 0.711 10 0.033 0.714 10 0.038 0.780 10 0.035 0.737 4-9
Upper Shell Forging Lower Shell Forging 49D963-l/49C904-1 50D102-1/50C97-1 Weld Metal WF562 "
Reference Cu (wt. %)
Ni (wt. %)
Cu (wt. %)
Ni(wt.%)
Cu (wt. %)
Ni(wt.%)
10 0.033 0.728 i
10 0.032 0.752 10 O.032 0.743
.i 10 0.031 0.730 10 0.032 0.711 10 0.032 0.728 10 0.031 0.703 10 0.032 0.687 10 0.033 0.703 10 0.033 0.695 Average 0.03 0.71 0.06 0.77 0.03 0.71 491FA.
Not used in Averag: calculation since already reported in snatenal certifications; reponed only for completeness.
- Subrnerged are weldrnent fabncated using 5/32-inch weld filler wire. Heat Number 442011 and Linde 80 flux, let Number 0344.
)
identical to that used in the closing girth weld scam between the upper shell forging and the limiting lower shell forging.
i 4-10 l
oo REACTCR VESSEL CORE BARREL NEUTRON PAD (301.5 *)
Z CAPSULE U (58.5 *)
g:< '58.5
- 58.5*
N 61*
I 270*
90*
(241 *)
y
]
I I
(238.5 * ) X W (121.5 *)
REACTOR VESSEL 1
180*
l PLAN VIEW f
VESSEL
\\
[ WALL s
CAPSULE s
ELEVATION VIEW CORE s/
ASSEMBLY l IIlllll !
lI !
CORE f.
j [ {
MIDPLANE d
E N
- *N NEUTRON PAD n
1 CORE BARREL f
I Figure 4-1 Arrangement of Surveillance Capsules in the Braidwood Unit 2 Reactor Vessel 4-11
- e,-- a LEGEND:
FL-LOWER SHELL FORGING 50D102-1/50C07-1 (TANGENTIAL)
FT-LOWER SHELL FORGING S00102-1150C97-1 (AXIAL)
FW-WELD METAL FM-HEAT-AFFECTED-ZONE MATERIAL a
WIGE amzR TENBILES C000mACTS COGAPACTS CHARPTS CHARPYS C8WIPYS COMRACTS CORAPACTS CHA wYS CH UIP a
,w,,,us, E 's ma PWen PHee M
PHs7 N1 MS MI N4 N3 M
M M
PHOS FL16 FL1S FL14 PL13 E
FMIB E
M x
PW10 PWEB PH00 PWEE PHOS M
pygg M h J L J Lj Cu 8l ll!-
Al.15%Co Cu I Il lf
!l gi li I i
8 1i I !I* 8 LLlO Udc Fe Fe 579'F M r"1 m 590*F mnc MONITOR 1 ll 1 i
Al.15%Co (Cd)
MONITOR i 18 10 I ll l 1
g a l 10 lig i I -
ni g Il IB
! h if I I! !Y CENTER REGI TO TOP OF VESSEL 4
gg MM
-W I
t fB DOSIMETER TENSILES CHARPYS CHARPYS CHARPYS CHARPYS CHARPYS COMPACTS COMPACTS TENSILES 448 FL12 FT60 FL80 FT57 FL57 FT54 FL54 FT51 FL51 FT48 FL48 FT12
~
447 558 R11 FT59 FL59 FT56 FL56 55 FL53 FT50 FL50 FT47 FL47 FT16 FT15 FT14 FT13 FT11 MS FL10 FT58 FL54 FT55 FL55 FT52 FL52 FT49 FL49 FT46 FL46 FT10 J L Al.15%Co
' Cu -
-q 8
- g i Al 15%Co I
I' 3 I
I 18 I r,s e
J udu Al.15%Co (Cd) f
-h g]
Al.15%Co (Cd)
I I !! ll I l
N1 II I t:
Ni 1
!I,l il I 3N OF VESSEL TO BOTTOM OF VESSEL Figure 4-2 Capsule X Diagram Showing Location of Specimens, Thennan Monitors, and Dosimeters I
5'O32 O/S
~
4 I
. SECTION 5.0 1
TESTING OF SPECIMENS FROM CAPSULE X j
1
'l 5.1 Overview I
t De post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at j
i the Westinghouse Science and Technology Center hot cell with consultation by Westinghouse Energy
)
Systems personnel. Testing was performed in accordance with 10CFR50, Appendix HI"3, ASTM Specification E185-82n23, and Westinghouse Remote Metallographic Facility (RMF) Procedure 8402, j
Revision 2, as modified by RMF Procedures 8102, Revision I and 8103, Revision 1.
l Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully l
removed, inspected for identification number, and checked against the master list in WCAP-11188I'l.
No discrepancies were found.
Exammation of the two low-melting point 579*F (304*C) and 590*F (310*C) eutectic alloys indicated j
that no melting had occurred in either type of thermal monitor. Based on this examination, the l
temperatures to which the test specimens were exposed did not exceed 579*F (310*C).
{
I he Charpy impact tests were performed per ASTM Specification E23-93a"'I and RMF Procedure i
8103, Revision 1, on a Tinius-Olsen Model 74,358J machine. De tup (striker) of the Charpy j
machine is instrumented with a GRC 830-1 instrumentation system, feeding informatson into an IBM compatible 486 computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (Eo). From the load-time curve (Appendix l
A), the load of general yielding (P y), the time to general yielding (tcy), the maximum load (P.), and a
. the time to maximum load (t ) can be determined (see the Idealized load-time record, Figure A-1 in j
u Appendix A). Under some test conditions, a sharp drop in load indicative of fast fracture was observed. De load at which fast fracture was initiated is identified as the fast fracture load (P,), and the load at which fast fracture terminated is identified as the arrest load (P).
j The energy ai.-aximum load (E ) was determined by comparing the energy-time record and the l,
load-time recos. The energy at maximum load is approximately equivalent to tie energy required to
.l
]
I I
i 5-1 l
l
i
.. ~
initiate a asck in the specimen Therefore, the propagation energy for the crack (E) is the difference between the total energy to fracture (IQ and the energy at mawimum load. (Ed.
)
f The yield stress (c ) was cabilarad imm the three-point bend formula having the followmg y
i expression:
I L
e,=P"A(F-s)*C (1) where L = distance between the spacimen supports in the impact testag machine; B = the width of -
j the aparimen measured parallel to the notch: W = height of the aparimen meassed perpawlirsitarly
-l t
to the notch; a = notch depth 1he ennerant C is by t-i on the notch flank angle ($), notch root j
radius (p), and the type ofloading (i.e., pme bending or three-poim bending). In three-post bandag l
a Charpy spaniman in which $ = 45' and p = 0.010". M* 1 is valid with C = 1.21. Therefore (for L = 4W),
_ ' w' L
o,.r"A(F-s)*1.21 3(F-s)*
(2) l
'i l
For the Charpy specunens, B = 0.3% in., W = 0.394 in., and a = 0.079 in. Equation 2 then reduces j
1 w.
I
'r*
w (3) l l
where o is in units of psi and Pay is in units of pounds. 'Ihe flow stress was calenlanad from the y
average of the yield and maximum loads, also using the three point bend formula.
Symbol A in columns 4,5, and 6 of Tables 5,6,7, and 8 is the cross-section area under the notch of 1
the Charpy specimene A M 6 1241
,q,,,
(4)
Percent shear was determinad from post-fractme photographs using the ratio of areas methods in compliance with ASTM Specificapon A370-92M. The lateral expansion was measmed using a dial gage rig similar to that shown in the same specification.
5-2 I
eg
~ _ _ _ _. _ - _
c Tensile tests were performed on a 20,000-pound Instron Model 1115, split-console test machine, per ASTM Specification E8-93"53 and E21-92V'3, and RMF Procedure 8102, Revision 1. All pull rods, grips, and pins were made ofInconel 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 j
crosshead speed of 0.05 inches per minute throughout the test. -
]
l i
Extension measurements were made with a linear variable displacement transducer extensometer.
The extensometer knife edges were spring-loaded to the specimen and operated through specimen t
i failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per i
ASTM E83-93"73 j
f 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 l
a thermocouple directly to the specimen, the following pmcedure was used to monitor specimen temperature. Chromel-alumel thernmcouples were inserted in shallow holes in the center and 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 i
upper grip was used to control the fumace temperature. During the actual'testmg the grip i
temperatures were used to obtain desired specimen w.w-dres. Experiments indicated that this f
I method is accurate to 2 2*F.
l The yield load, ultimate load, fmeture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and i
fracture strength were calculated using the original cross-sectional area. The final diameter and l
final gage length were determined from post-fracture photographs. The fracture area used to l
calculate the fracture stress (true stress at fracture) and percent reduction in area was computed
[
using the final diameter measurement.
5.2 Nrny V-Notch Imret Test Reculte The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule X, which was irradiated to 1.126 x 10" n/cm (E > 1.0 MeV), are presented in Tables 5-1 f
2 i
I 5-3 I
[
f I
through 5-8 and are compared with unirradiated resultsm as shown in Figures 5-1 through 5-4. De i
transition temperature increases and upper shelf energy decreases for the Capsule X materials are summarized in Table 5-9.
i hadiation of the reactor vessel lower shell forgmg Charpy specimens oriented with the longitudinal i
axis of the specimen parallel to the major rolling direction of the forging (tangential orientation) to f
1.126 x 10" n/cm (E > 1.0 MeV) (Figure 5-1) resulted in a 30 ft-lb transition temperature incream 2
of 3'F and in a 50 ft-lb transition LWure increase of O'F. This results in an irradiated 30 ft-lb transition temperature of -7'F and an irradiated 50 ft-lb transition temperature of 15'F.
De average Upper Shelf Energy (USE) of the lower shell forging Charpy specimens (tangential orientation) decreased I ft-lb after irradiation to 1.126 x 10 n/cm (E > 1.0 MeV). his results in l
2 an irradiated average USE of 167 ft-lb (Figure 5-1).
Irradiation of the reactor vessel lower shell forging Charpy specimens oriented with the longitudinal axis of the specimen normal to the major rolhng duection of the forging (axial orientation) to 1.126 l
x 10 n/cm (E > 1.0 MeV) (Figure 5-2) resulted in a 30 ft-lb transition L.,
.me increase of j
2 35'F and in a 50 ft-lb transition temperature increase of 35'F. This results in an irradiated 30 ft-Ib transition temperature of 10*F and an irrman'ad 50 ft-lb transition temperature of 35'F.
The average USE of the lower shell forging Charpy specimens (axial orientation) decreased 8 ft-Ib
{
after irradiation to 1.126 x 10 n/cm (E > 1.0 MeV). This results in an irradiated average USE of 2
145 ft lb (Figure 5-2).
Irradiation of the surveillance weld metal Charpy specimens to 1.126 x 10" n/cm' (E > 1.0 MeV)
(Figure 5-3) resulted in a 20'F increase in 30 ft-lb transition temperature and a 50 ft-Ib transition l
temperature increase of 5'F. His results in an irradiated 30 ft-lb transition temperature of O'F and an irradiated 50 ft-lb transition temperature of 45'F.
s The average USE of the reactor vessel core region weld metal decreased I ft-lb after irradiation to 1.126 x 10 n/cm (E > 1.0 MeV). His results in an irradiated average USE of 70 ft-lb (Figure
)
2 5-3).
5-4
. = -
Iw h:
1 Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal specimens to 1.126 x 10
n/cm (E >~ 1.0 MeV) (Figure 5-4) resulted in a 30 ft-Ib transition temperature increase of O'F and a l
2 50 ft-lb transition temperature increase of O'F. His results in an irradiated 30 ft-lb transition i
i temperature of -135'F and an irradiated 50 ft-lb transition temperature of -105'F.
l
.l l
De average USE of the reactor vessel weld HAZ metal decreased 30 ft-Ib after irramation to 1.126 x 10 n/cm' (E > 1.0 MeV). This results in an irradiated average USE of 125 ft-lb (Figure 5-4).
~!
he 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 or tougher appearance with increasing test l
temperature, r
t l
t A comparison of the 30 ft-lb transition temperature increases and upper shelf energy decreases for j
the various Braidwood Unit 2 surveillance materials with predicted values using the methods of NRC Regulatory Guide 1.99, Revision 2, is presented in Table 5-10. His comparison indicates I
i that, for the Capsule X surveillance matenals, the measured 30 ft-lb transition temperature increases i
for all surveillance materials are less than the Regulatory Guide 1.99, Revision 2, predictions and j
the measured upper shelf energy decreases of all surveillance materials are less than the Regulatory l
Guide 1.99, Revision 2, predictions.
I The load-time records for individual instrumented Charpy specimen tests are shown in Appendix A.
I 5.3 Tancion Test Rau*c i
ne results of the tension tests performed on the various materials contained in Capsule X, l
irradiated to 1.126 x 10 n/cm (E > 1.0 MeV), are presented in Table 5-11 and are compared with j
2 unirradiated results 'l as shown in Figures 5-9 through 5-11.
I t
e De results of the tension tests performed on the lower shell forging (tangential orientation) indicate
. that irradiation to 1.126 x 10" n/cm (E > 1.0 MeV) caused less than a 7 ksi increase in the 0.2 2
percent offset yield strength and less than a 8 ksi increase in the ultimate tensile strength when l
[
compared to unitradiated data (Figure 5-9).
{
1 i
5-5 j
- =
. i
m y
The results of the tension tests performed on the lower shell forging (axial orientation) indicate that 2
irradiation to 1.126 x 10" n/cm (E > 1.0 MeV) caused less than a 4 ksi increase in the 0.2 percent offset yield strength and less than a 6 ksi increase in the ultimate tensile strength when compared to ts unirradiated data (pigure 5-10).
De results of the tension tests performed on the surveillance weld metal indicate that irradiation to 1.126 x 10 n/cm (E > 1.0 MeV) caused less than a 6 ksi increase in the 0.2 percent offset yield 2
strength and less than a 9 ksi increase in the ultimate tensile n....g;h when compared to umrradiarad data '1 (Figure 5-11).
l De fractured tension specimens for the lower shell forgmg 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 tensile tests are shown in Figures 5-15 through 5-20.
5.4 Fracture Toughness Tests Per the surveillance capsule testing contract with the Can=anwealth Edison Company, the 1/2-T compact tension fracture mechanics specimens will not be tested and will be stored at the Westinghouse Science and Technology Center Hot Cell.
5-6 t
p-E i'
TABLE 5-1 i-Charpy V-notch Impact Data for the Braidwood Unit 2 Lower Shell Forging (Tangential Orientation) Irradiated to a Fluence of 1.126 x 10" n/cm' (E > 1.0 MeV)
Sample Temperature Impact Energy Lateral Expansion Shear Number T
- C ft.lb.
Joules mils mm FL57
-50
-46 9
12 4
0.10 0
FL58
-25
-32 13 18 9
0.23 5
FL59
-10
-23 40 54 29 0.74 5
FL52 0
-18 66 89 48 1.22 10 FL55 5
-15 58 79 42 1.07 15 FIA9 10
-12 82 111 54 1.37 30 FL54 25
-4 53 72 38 0.97 20 FL51 40 4
100 136 60 1.52 30 FIA8 75 24 66 89 47 1.19 25 F140 80 27 99 134 66 1.68 40 FL50 105 41 152 206 89 2.26 85 FL46 125 52 145 197 86 2.18 80 FL47 150 66 166 225 86 2.18 100 FL56 200 93 166 225 86 2.18 100-FL53 250 121 169 229 88 2.24 100 5-7
l TABLE 5-2 Charpy V-notch Impact Data for the Braidwood Unit 2 Lower Shell Forging (Axial Orientation) Irradiated to a Fluence of 1.126 x 10 n/cm (E > 1.0 MeV) 2 i
f Sample Temperature Impact Energy Lateral Expansion Sheer Number T
- C ft-Ib Joules mils l
nun FT59
-10
-23 11 15 8
0.20 i
FT49 0
-18 20 27 16 0.41 10 FT54' 5
-15 FT46 5
-15 29 39 24 0.61 10 FT60 15
-9 22 30 22 0.56 10 i
FT52 25
-4 54 73 40 1.02 15 FT55 30
-l
$1 69 36 0.91 15 FT56 50 10 49 66 40 1.02 20 FT48 80 27 74 100 54 1.37 25 FT51 110 43 90 122 66 1.68 40 FT47 150 66 114 155 76 1.93 80 FT53 200 93 131 178 85 2.16 90 FT58 250 121 152 206 88 2.24 100 FT57 300 149 149 202 90 2.29 100 FT50 350 177 134 182 87 2.21 100
-- 6pecimen alignment error - Data is not valid.
5-8
TABLE 5-3 Charpy V-notch Impact Data for the Braidwood Unit 2 Surveillance Weld Metal Irradiated to a Fluence of 1.126 x 10" n/cm (E > 1.0 MeV)
{
2 Sample Temperature Impact Energy Lateral Expansion Sheer Number
'F T
ft lb l Joules mils mm FW53 50
-46 10 14 8
0.20 FW52
-25
-32 23 31 19 0.48 10 i
FW47
-15
-26 8
11 10 0.25 10 FW54 0
-18 20 41 30 0.76 20 i
FW49 5
-15 21 28 19 0.48 15 FW56 15
-9 45 61 38 0.97 55 FW50 30
-1 37 50 31 0.79 25 FW59 50 10 55 75 49 1.24 80 FW57 75 24 66 89 52 1.32 90 FW55 100 38 57 77 51 1.30 90 FW48 125 52 62 84 54 1.37 95 FW46 150 66 65 88 61 1.55 100 FW60 200 93 70 95 64 1.63 100 FW58 250 121 75 102 69 1.75 100 FW51 300 149 70 95 67 1.70 100 b
i L
5-9
TABLE 5-4 i
Charpy V-notch Impact Data for the Braidwood Unit 2 Heat-Affected-Zone (HAZ) Metal 2
hradiated to a Fluence of 1.126 x 10" n/cm (E > 1.0 MeV)
Saample Temperature Impact Energy Lateral Fr--
Shear f
Neanbar
'F
'C ftlb Joules adis nun FH57
-150
-101 33 45 22 0.56 5
FH46
-130
-90 56 76 30 0.76 10 FH47
-120
-84 44 60 25 0.64 10 l
FH51
-100
-73 88 119 56 1.42 40 FH60
-100
-73 65 88 30 0.76 30
[
FH48
-75
-59 59 80 23 0.58 20 FH58
-60 51 60 81 38 0.97 25 FH59
-50
-46 IES 183 73 1.85 55
~
FH49
-40
-40 142 193 74 1.88 65 FH55 0
-18 117 159 63 1.60 60 i
FH50 50 10 144 195 78 1.98 80 FH54 100 38 126 171 67 1.70 70 FH53 150 66 104 141 63 1.60 100
{
FH52 200 93 103 140 65 1.65 100 FH56 250 121 169 229 86 2.18 100 5-10
TABLE 5-5 Instrumented Charpy Impact Test Results for the Braidwood Unit 2 Lower Shell Forging (Tangential Orientation)
Irradiated to a Fluence of 1.126 x 10" n/cm (E > 1.0 MeV) 2 9.emeMnd Ea.g-Semple Test Cheryy ft Ih/in8 Yield Dee to Max.
Mne to Fracture Arnst Ylete Flow Number Temp Energy Charpy Mez.
Prop.
leed Yleid imod Mas.
Leed tend Steves Strees
('F)
(ft-th)
EdfA eda Ep/A
@)
(msee)
@)
(msec)
@)
@)
(ksi)
(ksi)
FL57
-50 9
72 38 35 3705 0.16 3705 0.16 3705 51 123 123 FL58
-25 13 105 61 44 3515 0.14 3821 0.2 3821 272 117 122
~
R59
-10 40 322 301 21 3445 0.15 4711 0.63 4711 119 114 135 FL52 0
66 531 407 124 3440 0.15 4791 0.81 4614 85 114 137 FL55 5
58 467 359 108 3336 0.I8 4749 0.76 4695 68 lit 134 IL49 10 82 660 349 311 3632 0.16 4971 0.69 48 %
1594 121 143 FL54 25 53 427 332 95 3332 0.14 4728 0.69 4718 591 111 134 FL51 40 100 805 406 399 3282 0.14 4706 0.82 4244 1186 109 133
~
FL48 75 66 531 3%
135 3151 0.14 4549 0.83 4508 1077 105 128 FL60 80 99 797 3%
401 3140 0.14 4562 0.83 4009 2206 104 128 FL50 105 152 1224 403 821 3156 0.14 4628 0.83 1336 760 105 129 FL46 125 145 1168 385 782 2985 0.14 4488 0.83 2157 1125 99 124 FL47 150 166 1337 414 922 2897 0.19 4363 0.94 N/A N/A 96 121 FL56 200 166 1337 363 974 2809 0.14 4260 0.82 N/A N/A 93 117 FL53 250 169 1361 368 992 2685 0.15 4172 0.86 N/A N/A 89 114 N/A - Fully ductile fracture. No arrest load.
i l
l i
l TABLE 5-6 I
Instmmented Charpy impact Test Results for the Braidwood Unit 2 Lower Shell Forging (Axial Orientation)
Irradiated to a Fluence of 1.126 x 10 n/cm (E > 1.0 MeV) 2 NeesneNeed Energies Seenple Test Cherry ft-Mkint Yleid Theeto Men.
Tinee to Fractere Arrest Yleid Flow Number Teamp Energy Cheryy Men.
Prop.
Lead Yleid Imed Men.
leed Lead Stress
'Stevne
('F)
(R-Ib)
EWA E./A EvA Ob)
(moec)
(Ib)
(smsec)
(Ib)
Ob)
(ksi)
(ks0 FT59
-10 11 89 42 46 3580 0.17 3580 0.17 3580 171 119 119 FT49 0
20 161 125 36 3537 0.19 4166 0.34 4166 212 117 128 FT54' 5
FT46 5
29 234 208 25 3420 0.15 4521 0.48 4521 99 114 132 FT60 15 22 177 137 40 3348 0.15 4158 0.36 4158 385 111 125 Fr52 25 54 435 330 105 3398 0.16 4716 0.69 4706 299 113 135 FT55 30 51 4I1 327 83 3290 0.15 4651 0.69 4614 823 109 132 FT56 50 49 395 323 72 3214 0.14 4520 0.7 4463 643 107 128
/
FT48 80 74 5%
397 199 3224 0.15 4562 0.83 4364 1477 107 129 FT51 110 90 725 386 339 3096 0.14 4463 0.82 4197 1790 103 126 FT47 150 114 918 379 539 3008 0.I4 4373 0.83 3682 2604 100 123 FT53 200 131 1055 364 690 2904 0.14 4268 0.82 3104 2073 119 FT58 250 152 1224 316 908 2774 0.19 4174 0.77 N/A N/A 92 115 FT57 300 149 1200 346 854 2588 0.19 4063 0.86 N/A N/A 86 110 FT50 350 134 1079 313 766 2707 0.2 4026 0.78 N/A N/A 90 112
--- Specimen alignment error - Data is not valid N/A - Fully ductile fracture. No arrest load.
6 r
_-_._--,,,-.-e
-,.4 c
.-..m
. s
.w TABLE 5-7 Instrumented Charpy Impact Test Results for the Braidwood Unit 2 Surveillance Weld Metal Irradiated to a Fluence of I.126 x 10" n/cm (E > 1.0 MeV) 2 Meesenheed Energies a
Sample Test Cheryy R-IMn YleM Theeto Man.
Thee to Frectore Arved YleW Flow Nember Temp Energy Cheepy Max.
Prop.
leed YleM tmd Men.
Reed Imed Stress Serves
(*F)
(n-m)
EarA Em/A F#A Ob)
(msec)
Ob)
(mecc)
Ob)
Ob)
(td)
(kd)
FW53
-50 10 81 57 23 3886 0.22 3886 0.22 3886 126 129 129 FW52
-25 23 185 144 41 3478 0.15 4271 0.36 4271 259 116 129 FW47
-15 8
64 15 49 2198 0.11 2198 0.11 2198 277 73 73 FW54 0
30 242 185 57 3492 0.16 4305 0.44 4305 955 116 129 FW49 5
21 169 124 45 3381 0.14 4097 0.33 4097 341 112 124 FW56 15 45 362 214 148 3301 0.15 4124 0.51 3907 938 110 123 FW50 30 37 298 215 83 3266 0.14 4166 0.51 3989 401 108 123
?
FW59 50 55 443 217 226 3171 0.14 4093 0.52 3933 2565 105 121 U
FW57 75 66 531 266 265 3115 0.14 4098 0.62 324I 2221 103 120 FW55 100 57 459 211 248 3123 0.14 4073 0.51 3085 2112 104 120 FW48 125 62 499 217 282 2907 0.19 4042 0.56 3136 2305 97 115 FW46 150 65 523 205 319 3073 0.15 4007 0.51 N/A N/A 102 118 FW60 200 70 564 218 345 2879 0.14 3847 0.55 N/A N/A 112 FW58 250 75 604 220 384 2902 0.15 3912 0.55 N/A N/A 113 FW51 300 70 564 231 333 2731 0.19 3748 0.63 N/A N/A 91 108 i
l l
l l
N/A - Fully ductile fracture. No arrest load.
TABLE 5-8 Instrumented Charpy Impact Test Results for the Braidwood Unit 2 Heat-Affected-Zone (HAZ) Metal Irradiated to a Ruence of I.126 x 10 n/cm (E > l.0 MeV) 2
- 4ermeNeed Energere Sample Test Cheryy ft-IMn' Yleid Time to Men.
Theese Fracture Artuut Yield new Number Temp Energy Cheryy Men.
Prop.
teed YleW teed Mas.
Imed tend strees Stress
(*F)
(A-Ib)
EdfA eda EpfA
@)
(nroec)
@)
(mecc)
@)
@)
(kd)
(kd)
R157
-150 33 266 246 20 4703 0.21 5409 0.49 5409 130 156 168 R146
-130 56 451 376 75 4303 0.17 5188 0.69 5096 99 143 158 R147
-120 44 354 275 79 4268 0.17 5084 0.54 5050 106 142 155 RISI
-100 88 709 323 386 4172 0.21 5090 0.65 4308 187 139 154 R160
-100 65 523 369 154 4264 0.17 4095 0.69 4999 1096 142 139 RI48
-75 59 475 345 130 3726 0.15 4863 0.68 4747 105 124 143 R158
-60 60 483 356 127 3932 0.17 5024 0.69 4952 754 131 149 R159
-50 135 1087 377 710 4084 0.17 5071 0.72 2937 374 136 152 y
R149
-40 142 1143 375 769 3721 0.19 4915 0.75 2636 327 124 143 E
HISS 0
117 942 422 520 3453 0.15 4763 0.84 3279 1754 115 136 RISO 50 144
!!60 418 742 3291 0.14 4688 0.85 3056 1581 109 133 R154 100 126 1015 401 614 3272 0.14 4627 0.83 2901 1813 109 131 R153 150 104 837 401 437 3154 0.14 4522 0.85 MA N/A 105 127 R152 200 103 829 394 436 3086 0.15 4387 0.86 N/A N/A 103 124 MI56 250 169 1361 354 1007 2994 0.16 4296 0.8 N/A N/A 99 121 N/A - Fully ductile fracture. No arrest load.
S
..s TABLE 5-9
^'
Effect of Irradiation to 1.126 x 10 n/cm (E > 1.0 MeV) on the Notch Toughness Properties 2
of the Braidwood Unit 2 Reactor Vessel Surveillance Materials
- t Averese 3e M4 (a)
Average 35 asR (e)
Average 50 fte*'
Average Energy Absorpelen Transisten Tennyerstere (*F)
Laterol Eeyension Tesaperatore (*F)
Trh Teamperature (*F) et FtsE Sheer (R4)
I 1
Meterial Unirradiated irradlated AT Unirradiated Bereessed AT Unirredleted Irredissed AT Umleredissed Irreented AE Lower Shell Forging 10
-7 3
10 10 0
15 15 0
168 167
-1 i
' Jigentiet) z l
Lower Shell Forging 25 10 35 5
30 35 0
35 35 153 145
-8 vi (Axlel)
Weld Metal
-20 0
20 5
25 20 40 45 5
71 70
-I HAZ Metal
-135
-135 0
-80
-80 0
-105
-105 0
155 125
-30 (a) " Average" is defined as the value read from the curve fit through the data points of the Charpy tests (see Figures 5-1 through 5-4).
j r
i L
I I-L P
.. _..,_- _.,.. _...- _,, -... _ -. _ _.-_ _.. - _._,. u - _.;,.,. _ ~ _ _,.._ ___. _. _ _ _ _ _. _.. _ _ _ _.... _ _ _ _ _
TABLE 5-10 Comparison of the Braidwood Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-Ib Transition Upper SheK Energy Temperature Shift Decrease d
MmM N*
MmW Meterial Capsule 8
(10 n/cm, E>1.0MeV)
('F)
('F)
(%)
(%)
Lower Shell U
03933 27 0
15 0
Forging (Tangential)
X 1.126 38 3
20 i
Lower Shell U
03933 27 5
15 10
?
Forging 5
(Axlet)
X 1.126 38 35 20 5
Weki Metal U
03933 30 0
15 13 X
1.126 42 20 20 i
HAZ Metsi U
03933 0
0 X
l.126 0
19 (a) Based on Regulatory Guide 1.99, Revision 2. methodology.
l l
e
.-.rm---
w e
.---..,,,----.v-,
w--
-, - - -. - -.~., - - - -
<e y-,--,-,--+,,s.-
r---
--.w,--,-.---...-.---r...
. -... -,, ~.,
M
.s o
o TABLE 5-11 Tensile Properties for the Braidwood Unit 2 Reactor Vessel Surveillance Materials Irradiated to 1.126 x 10" n/cm (E > 1.0 MeV) 2 Semple Test Temp e.2% Ylehl Ultimate Fracture Fractere Fracture Uniform Totet.
Reduction Meterial Number
('F)
Strength Strength Lead Stress Strength Elongetton Elongation in Area (kd)
(kni)
(kle)
(kal)
(ksi)
(%)
(%)
(%)
Lower Stell Forging ILIO 35 73.8 95.7 2.93 197 3 59.7 12.0 26.9 70 50D102-1 FLII 125 69.8 91.1 2.80 158.4 57.0 113 26.0 64
/50C97-1 FL12 550 64.7 91.1 2.85 166.8 58.1 9.8 22.4 65 (tangential)
Lower Shell y
Forging FTIO 75 713 92.7 3.00 149.2 61.1 12.0 25.1 59 G
$0D102-1 Fril 175 68.2 87.6 2.70 163.5 55.0 10.5 23 3.
66
/50C97-1 FTl2 550 63.2 90.2 3.00 175.6 61.1 9.8 213 65 (axial)
FW10 25 80.5 84.5 335 152.0 68.2 9.0 20 3 55 Wekt IWil
.100 77.4 90.7 3.18 148.7 64.8 9.8 20.7 56 FW12 550 713 86.6 335 121 3 68.2 7.5 15.9 44
_. -. _ _ _ _ _.. _., _.. _ _ _ _..... _ - -. - -, - - - - -.. - - -. - - - - - - - - - - - - - - - - - - - - - - ~ ~ - - --
~ ~ - - - - - - - ~ ~ ~ ~ ~ ' ' - - - ~
DPERATURE ( C )
-100
-50 0
50 100 150 200 250 l
i i
i i
i l
i 3
2.
i 100 2
T
\\
T n
3 2
2 g
g x
2 g
2, l
2 2
\\
20 0
10 I
I I
I i
i l
100 2.5
"^ -
^
2.0 gg 1.5 ^
g b#
1.0 ~
A0 d
E 5
0 O
E i
i i
i i
i i
g 180 2#
2,a m
a o
g Jg 200 at l#
UNNUGA 130
?
120 8
160 IE s
I#
e IIRADIAD 10 jg 8
2 E3 m
1.126 X 10 n/an 100 -
t m
N.
gg t
m
^0 2
g as 8
3 0
0
-200
-100 0
100 200 300
- 0 500 BPERATURE (* F )
Figure 5-1 Charpy V-Notch Impact Properties for Braidwood Unit 2 Reactor Vessel Lower Shell Forging (Tangential Orientation) 5-18
BPERATURE ( C )
-100
-50 0
50 100 150 200 250 j
i i
i I 3 i
i i
i 2
100 z -q 3
3
^
g g
o 60 g0 2
o g
E 0
120 3.0
)
i i
i 2.5 100 j
80 E
j:
2.0 33 2
7 60 1.5 e b
40 8
1.0 W
E a5 o
5 0
0 M
i i
i i
i i
i g
180 20
)
- N
,/
4s - 220 160 200 r
I#
180 UERADIAD o
- e 120 160 140 e
100 o
- 3 s
120 g EADIAE TD g
80 1.12s x to" n/cm2 100 -
8 5
W
,,o E
W g
40 E
o 0
0
-200
-100 0
100 200 300 400 500 HPERATURE (* F )
Figure 5-2 Charpy V-Notch Impact Pmperties for Braidwood Unit 2 Reactor Vessel Lower Shell Forging (Axial Orientation) l 5-19 l
DPERAME (* C )
-100
-50 0
50 100 150 200 250 I
I I
I
-l i
I 8
100 2
m-
\\
I T
m r
M e
3 3
3 g
ha M
2
\\'
0 0
I I
I I
I I
I 100 2.5 i
E
(
\\
2.0 am 60 15 g
-b#
1.0 m 8
i 5
0' O
M i
i i
i i
i 1
g 15 2g i
2 IM 30 140 180 e
120 160 g
UMUUGAD 140 e
1 120 n i
3 80 4~
41 - 100 -
g g
E as E
suvaAn m 60 8
2 a5 1.13 X 10 n/an g
20 0
0
-200
-100 0
100 200 300 400 500 UPERAME (* F )
Figure 5-3 Charpy V-Notch Impact Properties for Braidwood Unit 2 Reactor Vessel Sun'eillance Weld Metal 5-20
HPERATURE (* C )
-100
-50 0
50 100-150 200 250 I
I I
i 1
l l
8 100 2
^
g..
x m
8 g
3.
3 3
0
'\\'
120
3.0 100 2.5 5
s2 zg E
gg v,,
j;, -
n 7
60 1.5 e b#
1.0 "
3, d
3 M
5 0
0 f
E i
i i
i i
i i
g 180 2#
2 12 s
AI - 2@
l#
180 1
9 120 - UNNUCRB 160 1#
==
s, 100 120 ^
c m
8~
5 atAoun 10
.Y A0 1.13 X 10" n/an 2 E
5
~
E M
d.
AO W
M 3
0 0
-200
-100 0
100 200 300 500 BPERATURE (* F )
Figure 5-4 Charpy V-Notch Impact Properties for Braidwood Unit 2 Reactor Vessel Weld Heat-Affected-Zone (HAZ) Metal 5-21
1 FL57 FL58 FL59 FL52 FL55
.4
,.Y *?
j i,
y FL49 FL54 FL51.
FL48 FL60..
l FOO
~ FL46
~ FL47 ~
FL56'
~
FL53 I
i i
Figure 5-5 Charpy impact Specimen Fracture Surfaces for Braidwood Unit 2 Reactor Vessel Lower j
Shell Forging (Tangential Orientation)
I 5-22
l FT59 FT49 Fr54 Fr46 FF60
_ Fr52
_. Fr55
. Fr56_
. _. _ Fr48 FT51 Fr47 FT53 FT58 Fr57 Fr50 I
l l
\\
i Figure 5-6 Charpy Impact Specimen Fracture Surfaces for Braidwood Unit 2 Reactor Vessel Lower She!! Forging (Axial Orientation) 5-23 l
1 i
l l
... w D:, n _ _ =
w I
~.f.,7 A
8 l
se ;
r J $,
,g,-
s_.
. ). I
,;7 A,
,,,. ~...
I-gl 8 C g '. Z,;.. :
1::~ti&*l1
[-
.r-*Cru o,.
3--
FW53 FW52 FW47 ~~ '
FW45 FW54 i
i t
9..'J :.f l
{
j 4 -
/
9.(.(
i i
1, 1 e i
i
. y' !
I.
> f I
.~
FW56 FW50 FW59
, FW57 FW55 4h.n2 ea f;zC.k',
b?
l t '.
sd..%'1 1
i l
- q.;qr < g; 1
i, FW48
~FW46 FW60' FW58 FW51
~
~
1 l
l l
i
.l I
j Figure 5-7 Charpy Impact Specimen Fracture Surfaces for Braidwood Unit 2 Reactor Vessel Surveillance Weld Metal l
5-24 i
i 1
_.l ;.(;=;,4:
',.m :: 1
(,.('. -; -
...,a 3
- ;.a-
- .; x.,
t
[;5
^
~,f, w' i-
' ; ;.f
., ', L, i
[
.,.'Njt.
'J-% '
FH57 FH46 FH47 FH51 FH60 FH48 FH58 FH59 FH49 FH55 FH50 FH54 FHS[
FH52 FH56 l
Figure 5-8 Charpy impact Specimen Fracture Surfaces for Braidwood Unit 2 Reactor Vessel Weld Heat-Affected-Zone (HAZ) Metal 5-25
(c-)
I 0
50 100 150 200 250 300 120 l
l I
I I
I I_ g 11 0 100
~
LUMA 1E USLE STRENG1H m
m W
U 2
7
~
g=
g a
500 "
l n
N m
g
[
l m moS -
i i
i i
i E
g CODE:
OPEN PONTS - lNRRMXAE CLOSED PONTS - RRADIAD TO 1.126 X 10"n/cm' l
I I
I I
i 1
n e
~
even u m 33 o
nt amn 2
I l
l 8
2 A
^
i 20 2
10
/
i tamW EM110N 0
0 100 200 300 400 500 600 TEMPERATURE (* F )
Figure 5-9 Tensile Properties for Braidwood Unit 2 Reactor Vessel Lower Shell Forging (Tangential Orientation) l l
5-26 l
( C) 0 50 100 150 200 250 300 15 I
i i
i i
l_
ggg 110 700
_,, 100 m
ULEATE BEE S1BG1H
{
CE I
g,
{80 i
500 ~
u) l 1
60
-2 0
g 50 0.2% 1 eld 51RDGi u
u u
e r
E g
CODE:
1 OPEN PONTS - lNRRADIAE l
CLOSED PONTS - RRADIAE TO 1.126 X 10 "n/an '
I I
I I
I I
I A
e 60 euen a m 3
50
?!
- l F:g 30 I IA' "U""'I" L~
3
~
a 2
\\
M
=
10 5-
=
2 UNNW EONGAIM 0
O 100 200 300 500 600 TEMPERATURE (* F )
Figure 5-10 Tensile Properties for Braidwood Unit 2 Reactor Vessel Lower Shell Forging (Axial Orientation) 5-27
og )
0 50 100 150 200 250 300 120 20 110
^
^ 12 2
ULINATE TENSEI STRENG1H
- D j
600 g g
80 -N gi 500 ~
m 3
60 os ED S1RENG1H g
50 n
i u
u i
E g
O 100 200 300 500 600 CODE:
OPEN POM15 - UIERRADIAE CLOSED POM15 - ERADIAD 101.126 X 10'n/cin' E
1 i
i i
i i
i y
).
\\._
g 50 REDucn0N N AREA g#
30 unt aOnanaN 1
/
20
-a
.t--r e.
,0 5
O O
100 200 300 500 600 TEMPERATURE (* F )
- Olt.Y GE DATA P(BIT AVAUitE Figure 5-11 Tensile Properties for Braidwood Unit 2 Reactor Vessel Surveillance Weld Metal 5-28
i i
I I
i f
1 i
i 1
m, ', p m...,. y.,n :n,,3.,..p.....~,, m...n.,,,
_w,.,
^
.J e
e, !:Tm.*.
( l.f, '? ',.,~..,
R;* i(D T^ YM.,'g"*;'T (l'. C
' "l[
Q s,.k,.
- s 3, '.WL1' e N.?;gD... f$ < :. d.& -M"nnsy n,,
r-
v
// '1Nc: C.305 A, l
,+, Mi%
?N< $
- ~~
n te
~
r v
w.
,c.
u
. -~
v N'
e :u. y m e n e m w n. s e x; g s
+
,m 9 4 4N M iMy.,..i,y.,.. f Q;f,k%Wtp::t +"*"Mm#?)%%+ r n f:h. y J;2 M.. e iM 1
/;.
m % 3 & %s. -t 3,,.
.f.2 a
- 2
,'y:'.. u s+
s O
e w
- M ' 7 7'", r cv,,;n.,f, tQ N -Q Q,.
2..m
.a. ; v,.RMiQff C'_
a yt ge... w,
,ss__...o ;, r _. m.m.
g
.+
i s.
- p.
u
-m Specimen FL10 Tested at 359 4
4 7-:-,c,yw-emy,, %.f'i 7'4pEEh@R,; 57'N m
.m,g 4 M - r 4M - - p-P T- %.m. p M.i t.
f}ly'qJ MW'e,$y,Ag s,4:tpr r L*r '.. :: 7. h:
5
~~ ; :;
gg y;~
- t. ;q" L,
.u. n
.;~~
n pe-y;gget w
.a m,u -
- u... - +
n e.,,4,. -
n
+y
.x.-
.c r
s' y; s.
e w
u r
a. 5 % st',o..
+ ^
l t
aa.
%:ytrtr_
VE i
e-
!"qD j
73 y
i w
i er a e, eA A
NNNkN.
E NNf h hk h
2 m, w,.4wn n m
a>=
.e em s a nnn.m ; e.ww =a.,~s.
(e w _a. &
n
.. u vm
..._.m._
fmm w.:.. _..,,w,x, a.
,t j
Specunen FL11 Tested at 1259 i
i r.2,., --.u k
- Q vn r -
p _-
., 6., ;pD -"
.,R :4 u
- 4p %, 7,-*-??%..m;sycawsyn, nn.y 97
.f.y "
w, i
~ es
.en. y.
is 4 :: q;w.+.w.,a, n%s?
- m.,T,~
W un. ;p....%nN,w y
,~
@m-2 i
.g I,.
g
.g/.
s m_
i 1
-.e i
l V
t
.. px m#%.w wg JYW,,ictsE
@ @4. 5.;m. m. D l
be, e,;.c a r.
- .eegcc.1..~m >, m
,y a e
n OOMh,#.. w.
ay&l&.,n.].q%'.YkhHM.Nb. h&w.. Nc.e..,n.g.
, v_:e,:.m ;.e v. m.a,.., 5.~,;4hlk.n
,4x 3.,
i w
f
' $% Lf %'
n'd x:-
u wc r
m ze c w
n
._.n.~,..,m.
~._ ~.,,,
~ ~,
..m
,.m..
I Mae$UdStE32lC-^
- +.an.Joesisd."M.w%4 s S.e%2g,"T,mg. 'D.***l**"**Q l
Specimen FL12 Tested at 550P i
l 4
i I
4 l
i 1
4 1
1 i
I Figure 5-12 Fractured Tensile Specimens from Braidwood Unit 2 Reactor Vessel Lower Shell Forging 4
(Tangential Orientation) t i
e j
5-29
l 1
1 i
I
,, n.
)
w.._,,,...,._~._-.m.;w.. n,_..
..e -
.p -.u
.,g, M..
r.
+,
.g 3
- :ns, Ytu; w'.cao"snMW by%y;n;,j, Q,.
7
_t-7 g
gy.
p
- m s n:V M n. 1 4 @ e,B %,
A;fMr e.(
. fax rsn&h2:JM%
M CfM lI@djieIs}AD$ #
D W'~
.h 9
1-.
I i
r v.
t" A
- 9Wh, I
w 2,.v,, %. n.;.;n,,.y w w,. %,.w g a's.
n w
m@,s ;~e. u,,n,4m E
.. ~
/, W7 %n a
. 'yG n.n ep.c?;Rw+.-*1 M M.~L..gy p.
.-s
. __' M:\\ ;z M= j6 qiyV. u,QQ i{G.:pQ??%y q u j ~: M V.
. -vMe,,e
...~
w.
v t
e
. s.m_._-
m~.#.,
m 3g f.7...g.
3 i
~
Specimen FTIO Tested at 75*F t
.. = m..,r, ~.,,_.- w
,y, ', t,..
7"*;'*,,. * * *. *,
.y4
..(;_. ;
n {4,%#wyg, 4,% 4 g.4 u-
%23pse?c e g g. t ;
e n p@nfd W s m.ed+yb nlW
%m &,W;;E 4
W f
,g m
4.g
- a. v.4, a n,
,.y ~pgm., s.
a r*d
,/$sf.
gh
{
J>d,'MW Mt,
t:.
- p. y c;
a.
.V:
t
.,d;;,;o,:N 4 M MW 9 M M,g m p*;,;;
Ni _
M,n,W%,f(. 4 i
PWwr r
k y: m g wgew c
u
\\
. p.q.~.-..,m 4 m.g g gm w. e,,..g e.n n..
e,.y ma.
%g ; wse+pg:;A--
ng 5.;
v y1 m,9 m x %,n; g,
,.#gs m.,y m=w
~
Specimen FT11 Tested at 175'F r
t i
q~ H 4,
p%g.; % m.___._~ m
.cp.,.;g @ m - - - - -.
kj, ggr - d k f j.9
', Q x Q.,, %. m.,1p @ 3 %p % 4 g p.s.%. p#..M..
j'd r
,, fi y
$.. pL
- f_.t 4
y m
C1 l
l %%Q.
~
b[
?g l
@sL e
~n s
>b.
G,'.; Q p,y 3lf0?b$h. *$'hhhh$,$ 1?@EW *;h?!^ IR
)
v
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m c.
p=
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it.m Mars 2.smw2 :mw;wth e::..4 g
.pe r*<.es -
n 1
Specimen FT12 Tested at 550"F r
l i
t i
l i
I Figure 5-13 Fractured Tensile Specimens from Braidwood Unit 2 Reactor Vessel Lower Shell Forging (Axial Orientation) 5-30
r_..._.,_,_
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i
~,, m....,r _#.-
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..:~-,-.-,.m Specimen FW11 Tested at 100P 1
.I l
z... w &<.w aa n. we a e s,.s, s.e -m.e" z 3 4 -
o c_. ;;.~. ;~e
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Specimen FW12 Tested at 550P d
I i
1
- l 1
I Figure 5-14 Fractured Tensile Specimens from Braidwood Unit 2 Reactor Vessel Suneillance Weld i
Metal i
i
'l F
l 5-31 I
- - - - - - ^ - - - - -
ie
=
100.00 7
A
's 90.00-80.00-70.00-co*
60.00-(fI 50.00-xy 40.00-30.00-FL10 20.00-35 F 10.00-0.00 0.00 0.10 0.20 0.30 STRAIN, IN/lN 100.00 90.00-80.00-70.00-(O 60.00-(fi 50.00-zy 40.00-30.00-FL 11 20.00-10.00-125 F 0.00 0.00 0.10 0.20 0.30 STRAIN, IN/IN Figure 5-15 Engineering Stress-Strain Curves for Forging Tensile Specimens FL10 and FL11 (Tangential Orientation)
L 5-32 I
. y; I
i t
i l
1 i
i t
.}
i 100.00
}
90.00-i r
I 80.00-i 70.00-i u) 60.00-vi M
50.00-i W
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-t 20.00-5$0 F 10.00 0.00 0.00 0.10 0.20 STRAIN, IN/IN t
i l;
i t
Figure 5-16 Engineering Stress-Strain Curve for Forging Tensile Specimen FL12 (Tangential
,l Orientation)
.i f
5-33 t
f er
?
t
,I
.i i
100.00 i
i 90.00,,
80.00-I 70.00-
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?
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t i
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70.00-i e 60.00-M of 50.00-f m
W i
x 40.00-H m
4 30.00-
~
FT 11 20.00 10.00-175 F I
0.00 0.00 0.10 0.20 i
STRAIN, IN/lN i
f Figure 5-17 Engineering Stress-Strain Curves for Forging Tensile Specimens FTIO and FT11 (Axial Orientation) l 5-34 t
e t
1
)
i i
I e
h i
l 8
l i
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?
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20.00-550 F 10.00-t
\\
0.00
'I 0.00 0.10 0.20 l
STRAIN, IN/IN l
[
k 1
i t
i f
(
i Figuie 5-18 Engineering Stress-Strain Curve for Forging Tensile Specimen FT12 (Axial Orientation)
{
i
'f r
l r
5-35 i
)
i 100.00 d
90.00 80.00-70.00-i m
60.00-i 50.00-e H
40.00-M i
30.00-FW 10 l
20.00-25 F 10.00-O.00 O.00 0.10 0.20
~
STRAIN. IN/IN l
100.00 90.00-80.00-70.00-
[
m r
80.00-
[
d I
50.00-C 40.00-l 30.00-FW 11 l
20.00-t l
10.00 100 F 0.00 0.00 O.10 O.5!O STRAIN, IN/IN l
Figure 5-19 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens FW10 and FW11.
i l
I 5-36 r
h
i r..-
f-i i
I i
l t
n i
90.00
- i 80.00-70.00-l E 80.00-x vi 50.00-mw e 40.00-1 t
v) 30.00-FW 12 e
20.00-550 F 10.00-o.00 -
0.00 0.04 0.08 0.12 0.16' t
STRAIN, IN/IN i
Figure 5-20 Engineering Stress-Strain Curve for Weld Metal Tensile Specimen FW12.
i 5-37
~
SECTION 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6.1 Introduction l
i Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule l
geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two i
reasons. First, in order to interpret the neutron radiation induced material property changes observed i
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 j
specimens to the present and future condition of the reactor vessel, a relationship must be established i
between the neutron environment at various positions within the pressure vessel and that experienced I
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 contamed in each of the surveillance capsules. The latter information is generally derived solely from i
analysis.
The use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the material 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 f
years, however, it has been suggested that an exposure model that accounts for differences in neutron l
energy spectra between surveillance capsule locations and positions within the vessel wall could lead i
to an improvement in the uncertainties associated with damage trend curves as well as to a more l
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 i
function for data correlation, ASTM Standard Practice E853, " Analysis and laterpretation of Light Water Reactor Surveillance Results,""U recommends reportag displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a database for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693 "Charactermng Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom."UM 'Ihe application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel l
wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, " Radiation Embrittlement j
of Reactor Vessel Materials."
f 6-1 f
Dis section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance Capsule X, withdrawn near the end of the fomth fuel cycle. Also included is an updated evaluation of the dosimetry contained in Capsule U, withdrawn at the conclusion of cycle one. His update is based on current state-of-the-art methadalogy and nuclear data; and, together with the Capsule X results, provides a consistent up-to-date database for use in evaluating the material properties of the Braidwood Unit 2 reactor vesset 1
In each of the dosimetry evaluations, fast neutron exposure parameters in tenns of neutron fluence (E > 1.0 MeV), neutron fluence (E > 0.1 MeV), and iron atom disp 1meements (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 wall.
Also, uncertainties associated with the derived exposure parameters at the survedlance capsules and with the projected exposure of the pressure vessel are provided.
6.2 Discrete Ordiameae Analysis t
A plan view of the reactor geometry at the core midplane is shown in Figure 4-1. Six irradiation I
capsules attachert to the neutron pad 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 axis as shown in Figure 4-1. A plan view of a dual surveillance capsule holder anached to the neutron pad is shown in Figure 6-1. The stamless steel specimen containers are 1.182 by 1-inch and approximately 56 inchec in height. De containers are positioned axially such that the test sparin=a= are matered on the core midplane, thus spanning the central five feet of the 12-foot high reactor core.
From a neutronic standpoint, the survedience capsules and associated support structures are significant.
The presence of these materials has a marked effect on both the spatial distribution of neutron flux and the neutron energy spewuru in the water annulus between the neutron pad and the reactor vessel. In order to determine the neutron environment at the test specimen location, the capadae themselves must be included in the analytical model.
In performing the fast neutron exposure evaluations lor the survedlance cap =idae and reactor vessel, i
two distinct sets of transport calculations were canied out. The first, a single comptttation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions l
6-2 m
throughout the' reactor geometry as well as to establish relative ra&al distributions of exposure parameters {$ (E > 1.0 MeV), & (E > 0.1 MeV), and dpa/sec} through the vessel wall. 'Ihe neutron spectral information was required for the interpretanon of neutron dosimetry withdrawn from the surveillance capsules as well as for the determination of exposure p-mr ratios; i.e.,
[dpa/sec) / [$ (E > 1.0 MeV)), within the pressure vessel geometry. The relative rad al gradient information was required to permit the projecuan of measured exposure per T_.; to locations interior to the pressure vessel wall; i.e., the 1/4T,1/2T, and 3/4T locanons.
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 at several azimuthal locations on the pressure vessel inner radius to neutron source distributions within the reactor core. The source importance functions generated from these adjoint analyses provided the basis for all absolute exposure calculahans and comparison with measurement. These i.T.pmi.r.:e functions, when combmed with fuel cyclupecific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for each cycle of irradiation. They also established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles. It is important to note that the cyclupecific neutron source distributions utilized in these analyses included not only spanal variations of fission rates within the reactor core but also accounted for the effects of varying neutron yield per fission and fission spectrum introduced by the build-up of plutonium as the burnup of individual fuel assemblies increased The absolute cycle-specific data from the adjoint evaluations'together with the relative neutron energy spectra and radial distribution information from the reference forward calculation provided the means to:
1-Evaluate neutron dosimetry obtamed from surveillance capsules.
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 summarued in Figures 4-1 and 6-1 was camed out in R,0 geometry using the DOT two-dimensional discrete ordinates code
- and the SAILOR cross-6-3
section librarynu. De SAILOR library is a 47-neutron-energy-group ENDF/B.IV based data set produced specifically for light water reactor applications. In these analyses anisotropic scattermg was j
treated with a P expansion of the uttering cross-sections and the angular discretization was modeled 3
with an S order of angular quadrature.
1he core power distribution utilized in the reference forward transport calculation 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 p.Wy. Furthermore, for the peripheral fuel assembhes, the neutron source was meressed by a 20 margm derived from the statistical evaluation of plans 40 plant and cycle 4e cycle variations in peripheral power. Since it is unlikely that any single reactor would exhibit power levels l
on the core periphery at the nonunal +2o value for a large number of fuel cycles, the use of this l
reference distribution is expected to yield somewhat conservative results.
All adjoint calculations were also camed out using an S, order of angular quadrature and the P cross-3 section approximation from the SAILOR library. Adjoint source locations were chosen at several azimuthal locations along the pressure vessel inner radius as well as at the geometnc center of each surveillance capsule. Again, these calculations were run in R,0 seometry to provide neutron source distribution importance functions for the exposure parameter of interest, in this case $ (E > 1.0 MeV).
Having the importance functions and appropnate core source distributions, the response of interest could be calculated as:
R(r,0) = [ [ [ 1(r,0,E) S(r,0,E) r dr dB dE r ez where: R(r,0) = $ (E > 1.0 MeV) at radius r and azimuthal angle 6.
I(r,0,E)= Adjoint source is.yer-i.ce function at radms r, azunuthal angle 6, and neutron source energy E.
S(r,0,E)= Neutron source strength at core location r,0 and energy E.
Although the adjoint importance functions used in this analysis were based on a response funcuon defined by the threshold neutron flux $ (E > 1.0 MeV), prior calculationst223 have shown that, while the implementation of low leakage loadmg panems significantly impacts both the magnitude and spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Dus, for a given location the ratio of [dpa/sec) / [$ (E > 1.0 MeV)] is insensitive to changing 64 i
E
.. ~ _,, _. ~. _..
...m
.- _j
7
.l core source distributions. In the applicarian of these adjoint importance functions to the Braidwood Unit 2 reactor, therefore, the iron atom displacernant rates (dpa/sec) and the neutron flux
$_(E > 0.1 MeV) were computed on a cycle specific basis by using [dpa/sec) / [$ (E > 1.0 MeV)] and
[$ (E > 0.1 MeV)] / [$ (E > 1.0 MeV)] ratios from the forward analysis in conjunction with the.
cycle-specific $ (E > 1.0 MeV) solutions from the individual adjoint evaluations.
l t
The reactor core power distributions used in the plant-specific adjoint calculations were taken from the fuel cycle design reports for the first four operating cycles of Braidwood Unit 2r2s e-,m, j
l Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-5. The data listed in these tables establishes the means for absolute comprison of analysis and measurement for -
the capsule irradimion persods and provides the means to correlate hy results with the f
I s-ig-:- *: exposure of the pressme vessel wall.
i In TaNe 6-1, the calenistad exposure parameters [$ (E > 1.0 MeV), $ (E > 0.1 MeV), and dpa/sec) l are given at the geometric center of the two survallance capsule positions for both the reference and i
the plant-specific core power distributions. The plant-specinc data, based on the adjoint irarapwi analysis, is meant to establish the absolute comparison of measurement with analysis. The reference data derived from the forward calentation is provided' as a conservative spuwe evaluation agamst j
which plant-specific fluence calenistinne can be compared. Similar data is given in Table 6-2 for the -
pressure vessel inner radius. Again, the three perunent exposure parameters are listed for the reference l
and the cycle one through four plant-specific power distributions. It is important to note that the data for the vessel inner radius was taken at the clad / base metal interface; and, thus, represents the maximum predicted exposure levels of the vessel wall itself.
i I
Radial gradient information applicaNe to & (E > 1.0 MeV), $ (E > 0.1 MeV), and dpa/sec is given in l
'I Tables 6-3,6-4, and 6-5, respectively. The data, obtained from the reference forward neutron transport caledanan. is presented on a relative basis for each exposure parameter at several arimnthal locations.
l Exposure distributions through the vessel wall may be obtamed by nar-Ma= the calculated or projected exposure at the vessel inner radius to the gradient data listed in TaNes 6-3 through 6-5.
For example, the neutron flux $ (E > 1.0 MeV) at the 1/4T depth in the pressure vessel wall along the 45' azimuth is given by:
l 6-5
g
.i i
$,, (45*) = $ (220.27, 45*) F (225.75, 45*) -
where: $, (45')
= Pmjected neutron flux at the 1/4T position on the 45* azimuth.
$ (220.27,45')
= Projected or calculated neutron flux at the vessel inner radius on the 45' azimuth.
L F (225.75,45')
= Ratio of the neutron flux at the 1/4T position to the flux at the vessel i
inner radius for the 45' azimuth. ' Ibis data is obtained from Table 6 3.
l Similar expressions apply for exposure p i.nais expressed in terms of $ (E > 0.1 MeV) and dpa/see t
where the attenuation function F is obtained from Tables 6-4 and 6-5, respectively.
+
6.3 Neutron Dosametry
)
i i
The passive neutron sensors included in the Braidwood Unit 2 surveillance program are listed in i
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 phmu within the surveillance capsules and in
{
the subsequent determmaten of the various exposure parameters of interest [$ (E > 1.0 MeV),
j
& (E > 0.1 MeV), and dpa/sec]. The relative locations of the neutron sensors within the capsules are shown in Figure 4-2. h iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were l
t placed in holes drilled in spacers at several axial levels withm the capsules. The cadmium shielded uranium and neptunium fission monitors were accommodated withm the dosimeter block located near t
the center of the capsule.
t
'Ihe use of passive monitors such as those listed in Table 6-6 does not yield a dueet measure of the energy dependent neutron flux at the point of interest. Rather, the activation or fission process is a I
measure of the integrated effect that the time and energy Mt neutron flux has on the target f
matenal over the course of the irradiation penod. An accurate assessment of the average neutron flux i
level incident on the various monitors may be derived from the activation measurements only if the
{
t irradiation parameters are well known. In particular, the following variables are of interest-j The measured specific activity of each monitor.
The physical characteristics of each monitor.
h operating history of the reactor.
l 6-6
l The energy nesponse of each monitor.
The neutron energy spectrum at the monitor location.
The specific activity of each of the neutron monitors was determined using established ASTM proceduresr27.n " *l. Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma iay spectrometer. The irradiation history of the Braidwood Unit 2 reactor dudng cycles one through four was obtained from l
NUREG-0020, " Licensed Operating Reactors Status Summary Report," for the applicable period. The irradiation history applicable to Capsules U and X is given in Table 6-7.
t Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full power operation were determined from the following equation:
R=
P
.y N, F Y [ p C [1-e
'] [e.y']
l j
where:
"I R
Reaction rate averaged over the irradiation period and referenced to operation at a core
=
power level of P,(rps/ nucleus).
A
=
Measured specific activity (dps/gm).
No Number of target element atoms per gram of sensor.
=
F Weight fraction of the target isotope in the sensor material.
=
Y Number of product atoms produced per reaction.
=
P, Average core power level during irradiation period j (MW).
=
P, =
Maximum or reference power level of the reactor (MW)
Calculated ratio of & (E > 1.0 MeV) during irradiation period j to the time weighted C,
=
average $ (E > 1.0 MeV) over the entire irradiation period.
A Decay constant of the product isotope (1/sec).
=
Length of irradiation period j (sec).
t
=
j Decay time following irradiation period j (sec).
to
=
and the summation is carried out over the total number of monthly intervals comprising the irradiation period.
6-7 l
In the equation describing the reaction rate calculation, the ratio [P) / [P,,,] accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio C which can be calculated for each fuel cycle using the adjoint transport method p
discussed in Secdon 6.2, accounts for the change in sensor reaction rates caused by variations in flux level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single cycle irradiation C, is normally taken to be 1.0. However, for multiple cycle irradiadons, particularly those employing low leakage fuel management, the additional C; term should be employed. The impact of changing flux levels for contrant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has made the transition from non low-lenkage to Iow-leakage fuel management or for sensor sets contamed in surveillance caped *= that have been moved from one capsule location to another.
For the irradiation history of Capsules U and X, the flux level term in the reaction rate calculations was developed from the plant-spectfic analysis provided in Table 6-1. Measured and saturated reaction product specific activities as well as the derived full power reaction rates are listed in Tables 6-8 and 6-9 for Capsules U and X, respeedvely.
Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code"U. The FERRET approach used the measured reaction rate data, sensor reaction cross-sections, and a calculated trial spectrum as input and proceeded to adjust the group fluxes from the trial spectrum to produce a best fit (in a least squares sense) to the measured reaction rate data. The " measured" exposure parameters along with the associated uncertainties were then obtained from the adjusted spectrum 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 Iznearly related to the flux & by some response matrix A:
r-Egg i
s where i indexes the measured values belonging to a single data set s, g designates the energy group, and a delineates spectra that may be simultaneously adjusted. For example, R=[o &s i
4 l
l relates a set of measured reaction rates R to a single spectrum $, by the multigroup reaction cross-i 6-8
f t
section o,. The log-normal approach automatically accounts for the physical constraint of positive '
fluxes, even with large assigned uncertainties.
l In the least squares adjustment, the continuous quantities (i.e., neutron spectra and cross-sections) were I
approximated in a multi-group format consisting of 53 energy groups. The trial input spasmu was j
convened to the FERRET 53 group structure using the SAND-II code". This procedure was carned f
out by first expanding the 47 group calculated spectmm into the SAND-II 620 group structure using a SPLINE interpolation procedure in regions where group boundanes do not comeide. "Ihe 620 point spectrum was then re-collapsed into the group structure used in FERRET.
l l
The sensor set reaction cross-sections, obtamed from the ENDF/B-V dosimetry file, were also collapsed into the 53 energy group structure using the SAND-H code. In this instance, the trial l
i spectrum, as expanded to 620 groups, was employed as a weighting functon in the cross-section collapsing procedure. Reaction cross-section uncertainties in the form of a 53 x 53 covariance matrix for each sensor reaction were also constructed from the information contained on the ENDF/B V data t
files. These matrices included energy-group to energy-group uncertamty correlations for each of the individual reactions. However, correlations between cross-sections for different sensor reactions were i
i not included. The omission of this additional uncertamty information does not significantly impact the results of the adjustment.
Due to the importance of providing a trial spectrum that exhibits a relative energy distribution close to the actual spectrum at the sensor set locations, the neutron spectrum input to the FERRET evaluation i
i was taken from the center of the surveillance capsule modeled in the reference forward transport-j calculation. While the 53 x $3 group covariance matnces applicable to the sensor reaction cross-sections were developed from the ENDF/B-V data files, the covanance matnx for the input trial l
3 spectrum was constructed from the following relation:
M i = R* + R R o P i l
u
=
- s a
. where R, specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the set l
of values. The fractional uncertainties R, specify additional random uncertainties for group g that are l
conelated with a correlation matrix given by:
i P,= [1-0] 6,+ 0 e
- where:
ss as 6-9 l
4
-m.
.-e--.
.-e
~-. -
H=
2 y*
i
'Ibe first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes short range correlations over a group range y (6 specifies the strength of the lauer term). 'Ibe value of 6 is I when g = g' and 0 otherwise. For the trial spectrum used in the f
current evaluations, a short range correlation of y = 6 groups was used. This choice implies that i
neighboring groups are strongly correlarM when 6 is close to 1. Strong long range correlations (or anti correlations) were justified based on information presented by R. E. Macrker"'1 Macrker's results t
are closely duplicated when y = 6.
l s
The uncertainties associated with the measured reaction rates included both statistical (&ndag) and systematic compaaaa'e The systematic component of the overall uncertatory accounts for counter efficiency, counter calibrations, uradsation history corrections, and correcuons for compenng reacnons in the individual sensors.
\\
Results of the FERRET evaluations of the Capsules U and X dosimetry are given in Tables 6-10 and 6-11, respectively. The data summarized in these tables includes fast neutron exposure evaluations in terms of 4 (E > 1.0 McV), 4 (E > 0.1 MeV), and dpa. In general good results were achieved in the fits of the adjusted spectra to the individual measured reaction rates. The adjusted spectra from the
)
least squares evaluations are given in Tables 6-12 and 6-13 in the FERRET 53 energy group stmeture.
The results for Capsule X are consistent with results obtained from similar evaluations of dacimeny from other Westinghouse reactors.
i 6.4 Proiections of Pisswc Venet hacnrg i
Neutron exposure projections at key lentians on the pressure vessel inner miius are given in i
Table 6-15. Along with the current (4.215 EFPY) exposure, projections are also provided for sp,.m l
periods of 16 EPPY and 32 EFPY. In computing these vessel exposures, the calMatM values from Table 6 2 were scaled by the average measurement to calculation ratios (M/C) observed from the evaluations of dosimetry from Capsules U and X for each fast neutron exposure parameter (Table 6-14). This procedure resulted in bias factors of 1.08,1.08, and 1.04 being applied to the calculated values of 4 (E > 1.0 MeV), 4 (E > 0.1 MeV), and dpa, respectively. Projections for 6-10 i
i y
w r
--~
future operation were based on the assumption that the average exposure rates characteristic of the cycle one through four irradiation would continue to be applicable throughout plant life.
The overall uncertainty associated with the best esttmate exposure projections at the pressure vessel wall depends on the individual uncertamties in the measurement process, the uncertainty in the dosimetry location, and on the uncertainty in the extrapolation of results from the measurement points to the point of interest in the vessel wall. For Braidwood Unit 2, the uncertainty in each capsule derived fluence is estimated to consist of a 6% random component and a 6% systematic component.
The extrapolation uncertainty is estimated to be 5%. A statistical combination of these uncertamtics for the two capsules produces an overall uncertamty estimate in the exposure of the pressure vessel wall in the beltline region of 8% (1o) for fluence with energy above 1.0 MeV.
For Braidwood Unit 2, an extrapolation uncertainty of 5% has been combined with a 8% tmcertahity in the plant-specific measurement #@dadon bias factor derived from the two surveillance e.psrAs to produce a net uncertainty of 8% in the projected exposure of the pressure vessel wall. This 8 fo uncertainty applies at the 10 level for @ (E > 1.0 MeV).
In the calculation of exposure gradients for the Braidwood Unit 2 reactor vessel, exposure prt.jections to 16 EFPY and 32 EFPY were also employed. Data based on both a 4 (E > 1.0 MeV) slope and a plant-specific dpa slope through the vessel wall are provided in Table 616.
In order to access RT,er versus fluence curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations:
4pa(1/4D K 1/4 D - K 0 D dpa(0D
""d 4pa(3/4D K 3/4 D - K 0 D 4pa(0D Using this approach results in the dpa equivalent fluence values listed in Table 6-16. In Table 6-17, updated lead factors are listed for each of the Braidwood Unit 2 surveillance capsules. Lead factor j
data based on the accumulated fluence through cycle four is provided for each remaining capsule.
6-11
FIGURE 6-1 PLAN VIEW OF A DUAL REACTOR VESSEL SURVEILLANCE CAPSULE i
(TYPICAL)
- 61.08
- s\\
- 58.58 7
7 NNNNN's
}
NEUTRON PAD I
4 P
I 6-12 i
FIGURE 6-2 t
AXIAL DISTRIBUTION OF NEUTRON FLUENCE (E > 1.0 MEV)
ALONG THE 45 DEGREE AZIMUTH 1.9E+20 j
1.0E+19 ::
N:
e u
w
- s t'
.g i
\\
~
'/
5 i
(
1 l
\\
g 1.4E+18 1
I l
-4.215 EFPY --16 EFPY 32 EFPY 1.0E+17 0 12 3 4 5 6 78 9191112 Distance From Core Bottom (ft) 6-13
Ll
a-7:
TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE RATES AT THE SURVEILLANCE CAPSULE CENTER 2
j
$ (E >1.0 MeV) [n/cm -sec]
I CAPSULE LOCATION 29.0*
31.5* -
s CYCLE 1 9.067E+10 9.720E+10 l.
CYCLE 2 6.669E+10 7.454E+10 L
CYCLE 3 6.%7E+10 7.515E+10 L
CYCLE 4 6.883E+10 7.417E+10 i
CRSD Data
' 1.130E+11
'1.210E+11 2
$ (E > 0.1 MeV) [n/cm -sec]
CAPSULE LOCATION 29.0*
31.5*
CYCLE 1 3.921E+11 4.167E+11 CYCLE 2 2.884E+11 3.195E+11 CYCLE 3 3.013E+11 3.222E+11 CYCLE 4 2.977E+11 3.179E+11 CRSD Data
. 4.887E+11 5.187E+11 i
Iron Atoss E' ;'--
--' Rate [dpa/sec]
CAPSULE LOCATION.
29.0' 31.5*
CYCLE 1 1.777E-10 1.895E-10 CYCLE 2 1.307E-10 1.454E-10.
CYCLE 3 1.366E-10 1.465E-10 CYCLE 4 1.349E-10 1.446E-10 CRSD Data 2.215E-10 2.360E-10 CRSD - Core Radiation Source Data 6-14
TABLE 6-2 CALCULATED AZIMUTHAL VARIATION OF FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL C2.AD / BASE METAL INTERFACE l
2
$ (E > 1.0MeV) [n/cm -sec)
- 6. 0*
15.0*
25.0*
35.0" 45.0*
'CYCLEI 1.352E+10 2.107E+10 2.418E+10 1.991E+10 2.315E+10 CYCLE 2 1.040E+10 1.478E+10.
1.714E+10 1.557E+10 1.849E+10 CYCLE 3 1.114E+10 1.631E+10 1.856E+10 1.547E+10 1.798E+10
[-
CYCLE 4 1.133E+10 1.644E+10 1.842E+10 1.525E+10 1.766E+10 1
CRSD Data 1.780E+10 2.660E+10 3.010E+10 2.450E+10 2.810E+10
$ (E > 0.1MeV) [n/cm'-sec) 1
- 0. 0*
15.0*
25.0*
35.0" 45.03 CYCLE 1 2.810E+10 4.435E+10 6.603E+10 5.656E+10 5.799E+10 CYCLE 2 2.162E+10 3.112E+10 4.682E+10 4.423E+10 4.632E+10 CYCLE 3 2.315E+10 3.433E+10 5.068E+10 4.395E+10 4.505E+10 CYCLE 4 2.356E+10 3.461E+10 5.030E+10 4.332E+10 4.425E+10 CRSD Data 3.700E+10 5.600E+10 8.220E+10 6.960E+10 7.040E+10 Iron Atom Displacement Rate [dpa/sec)
- 6. 0*
15.0" 25.0*
35.0*
45.0*
CYCLE 1 2.103E-11 3.263E-11 4.048E-11 3.372E-11 3.690E-11 CYCLE 2 1.619E-11 2.290E-11 2.870E-Il 2.638E-11 2.948E-11 CYCLE 3 1.733E-11 2.525E-11 3.107E-11 2.620E-11 2.867E-11 CYCLE 4 1.764E-11 2.546E-11 3.084E-11 2.583E-11 2.816E-11 CRSD Data 2.770E-11 4.120E-11 5.040E-11 4.150E-11 4.480E-11 CRSD - Core Radiation Source Data 6-15
f TABLE 6-3 RELATIVE RADIAL DISTRIBUTION OF & (E > 1.0 MeV)
WITHIN THE PRESSURE VESSEL WALL Radius (can)
- 6. 0*
15.0*
25.0*
35.0*
45.0*
i i
220.27m 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 22431 0.653 0.653 0.657 0.655 0.648 225.63 0.551 0.550 0.554 0.552 0.543 226.95 0.462 0.460 0.465 0.463 0.452 228.28 0386 0.384 0388 0386 0375 229.60 0321 0319 0324 0.321 0.311 230.92 0.267 0.263 0.275 0.267 0.257 r
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 238.86 0.0828 0.0817 0.0846
.0835 0.0762 240.19 0.0671 0.0660 0.0689
.0679 0.0608 241.51 0.0538
-- 0.0522 0.0550 0.0545 0.0471 242.175 0.0506 0.0488 0.0518 0.0521 0.0438 NO7ES:
- 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius t
6-16
-.- =.
=
- jy TABLE 6-4 RELATIVE RADIAL DISTRIBUTION OF $ (E > 0.1 MeV)
WI7111N THE PRESSURE VESSEL WALL Radius (can)
- 6. 0*
15.0*
25.0*
35.0*
45.0*
220.27
- 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*
0.233 0.226
.237 0.223 0.188 NOTES:
- 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius 6-17
TABLE 6-5 RELATIVE RADIAL DIS 7RIBUTION OF DPA/SEC WITHIN THE PRESSURE VESSEL WALL Radius (cm)
- 6. 0*
15.0*
25.0" 35.0*
45.0" 220.27*
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 l
l 229.60 0.439 0.435 0.462 0.475 0.447 l
l 230.92 0.387 0.383 0.410 0.423 0.394 1
232.25 0.341 0.338 0.364 0.376 0.347 l
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 i
237.54 0.199 0.198 0.218 0.227 0.199 238.86 0.171 0.170 0.189 0.1%
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 1
242.17 5 0.116 0.113 0.128 0.134 0.106 NOTES:
- 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius 6-18
.=
-i 7,.
. TABLE 6-6..
I NUCLEAR PARAME7ERS USED IN THE EVALUATION OF NEUTRON SENSORS Reactmn Target Fission -
1 Monitor of '
. Weight
Response
Product
. Yield j
Mau:nal launat faam Bangg_
Half-Life fil.,
j Copper Cu'8(n,a)Co" 0.6917 E > 4.7 MeV 5.271 yrs 1ron'-
Fe"(n,p)Mn" 0.0580 E > 1.0 MeV 312.5 days Nickel:
Ni"(n.p)Co" 0.6827 E > 1.0 MeV 70.78 days
.f Uranium-238' U "(n.f)Cs
l.0 E > 0.4 MeV.
30.17 yrs 6.00 28 Neptunium-237*
Np '(n.f)Cs -
1.0 E > 0.08 MeV 30.17 yrs 6.27 Cobalt-Aluminum
- Co"(n,y)Co" 0.0015 0.4ev>E> 0.015 MeV-5.271 yrs j
Cobalt-Aluminum Co"(n,y)Co" 0.0015 E > 0.015 MeV 5.271 yrs
- Denotes that monitor is cadmium shielded.
l l
i t
i 4
t l
}
l i
~
I ia f
l i
6-19
{
TABLE 6-7 l
MONTHLY THERMAL GENERA" DON DURING THE FIRST FOUR FUEL CYCLES OF THE BRAIDWOOD UNIT 2 REACTOR f
I i
'Ihermal Generation Thermal Generation f
Year M anth (MW-hr)
' Year M a=*h (MW-hr) l
?
1988 Jan 0
Jun 2,144,813
-Feb 0
Jul 2,379,800 Mar 0
Aug 2,256,254 Apr 0
Sep 2,171,245
.j May 116,073 Oct 2,465,600 Jun 533,248 Nov 2,191,093 Jul 1,169,397 Dec 2,439,339 Aug 1,916,252 1993 Jan 2,399,465-Sep 984,844 Feb 1,930,384 l
Oct 498,112 Mar 249,733 l
Nov 1,618,294 Apr 0
i Dec 1,961,212 May 1,937,581 1989 Jan 2,244,107 Jun 2,342,521 Feb 724,667 Jul 2,415,905 Mar 303,269 Aug 2,458,798 i
Apr 2,278,398 Sep 2,391,762 May 1,675,139 Oct 1,449,593 i
Jun 1,714,433 Nov 2,408,486
{
Jul 1,857,133 Dec 2,491,223 i
Aug 2,099,116 1994 Jan 1,667,825 Sep 1,908,625 Feb 2,220,994 Oct 2,202,479 Mar 2,522,680
{
Nov 2,283,637 Apr 337,197 Dec 2,472,337 May 62,421 i
1990 Jan 1,961,079 Feb 1,251,329 Cumulative 1.260E+08 Mar 537,829 Apr 0
j May 85,374 Jun 1,472,014 l
Jul 2,415,715 l
Aug 2,451,032 s
Sep 2,318,122 Oct 2,316,935 Nov 2,262,639 Dec 2,397,002 1991 Jan 2,498,943 Feb 2,085,837 l
Mar 2,465,072 Apr 2,364,969 May 1,620,535 l
Jun 2,225,330 l
Jul 2,343,558 Aug 1,686,282 t
Sep 616,849 Oct 0
l Nov 122,919 Dec 2,015,229 l
1992 Jan 2,422,273 Feb 1,949,523 i
Mar 2,275,870 Apr 2,356,176 May 1,605,206 I
6-20 i
p
.w s
TABLE 6-8 MEASURED SENSOR ACTIVITIES, SATURATED ACTIVITIES, AND REACTION RATES SURVEILLANCE CAPSULE U MEASURED SATURATED. REACTION MONITOR AND ACTIVITY ACTIVITY RA'IE AXIAL LOCATION (dis /sec-mm)
(dis / soc-mn)
(rns/ nucleus)
Cu.43 (a,a) Code 901634 "!OP 5.410E+04 4.295E+05 6.553E-17 90-1639 MID 4.780E+04 3.795E+05 5.790E-17 901644 BOT 4.810E404 3.819E+05 5.826E-17 AVERAGES 5.000E+04 3.970E+05 6.056E-17 Fe 54 (a,p) Ma 54 901636 *IDP 1309E+06 4.029E+06 6.441E-15 90-1641 MID 1.159E+06 3.567E+06 5.703E-15 90-1646 BOT
-1.139E406 3.505E+06 5.605E-15 AVERAGES 1.202E+06 3.700E406 5.916E-15 Ni-58 (a,p) Co-58 90-1635 TOP 4.985E+06 5.782E+07 8.255E-15 90-1640 MID 4.462E+06 5.175E+07 7389E-15 90-1645 BOT 4.436E+06 5.145E+07 7.346E-15 AVERAGES 4.628E406 5367E+07 7.663E-15 Co 99 (a,y) Co40 (bere) 90-1632 TOP 1.079E+07 8.567E+07 5.589E-12 90-1637 MID 1.071E+07 8.503E+07 5.548E-12 90-1642 BOT 1.056E+07 8384E+07 5.470E-12 AVERAGES 1.069E+07 8.485E+07 5.536E-12 Co 59 (a,y) Code (Cd am) 90-1633 TOP 5.423E+06 4306E+07 2.809E-12 90-1638 MID 5.569E+06 4.422E+07 2.885E-12 901643 BOT 5.563E+06 4.417E+07 2.882E-12 AVERAGES 5.518E+06 4381E+07 2.858E-12 U-238 (a,f) Cs 137 (Cd sidelded) 90-1630 MID 1.542E+05 6.039E+06 3.980E-14 Np 237 (n.f) Cs-137 (Cd sidelded) 90-1631 MID 1387E+06 5.432E+07 3.410E-13 6-21
. _ =.
h
?
TABLE 6-9 MEASURED SENSOR ACTIVITIES, SATURATED ACTIVITIES, AND REACTION RATES SURVEILLANCE CAPSULE 'X MEASURED SATURATED REACTION i
MONITOR AND ACTIVITY ACITVITY RATE AX1AL LOCATION (ddacc-m)
(ddanc-mn)
(ma/ nucleus) t' Cu43 (a,s) Co40 i
94-1681 TDP 1350E+05 3.665E+05 5.591E-17 94-1687 MID 1.230E+05 3339E405 5.094E-17 94-1693 BOT 1.200E+05 3.258E+05 4.970E-17 i
AVERAGES 1.260E+05 3.421E405 5.218E-17 Fe-54 (a,p) Ma-54 l
94-1679 TOP 1.710E406 -
3.204E+06 5.123E-15 94-1685 MID 1.530E+06 2.867E+06 4.584E-15 l
94-1691 BOT 1.520E+06 2.848E+06 4.554E-15 AVERAGES 1.587E406 2.973E+06 4.754E-15 Ni-58 (a,p) Co-58 l
r I
94-16801DP 9.390E406 4.876E+07 6.%2E-15 94-1686 MID 8.48(E+06 4.403E407 6.287E-15 94-1692 BOT 8.430E+06 4377E+07 6.250E-15 AVERAGES 8.767E+06 4.552E+07 6.500E-15 j
Co 59 (a,y) Co40 (here) i 94-1678 TOP 2.270E+07 6.163E407 4.021E-12 94-1684 MID 2350E+07 6380E+07 4.162E-12 94-1690 BOT 2320E+07 6.298E+07 4.109E-12 AVERAGES 2313E+07 6.280E+07 4.097E-12 Co-59 (a,y) Code (Cd shielded) 94-1677 TOP 1.200E+07 3.258E+07 2.125E-12
[
94-1683 MID 1.240E+07 3366E+07 2.196E-12 i
941689 BOT 1.260E+07 3.421E+07 2.232E-12 f
AVERAGES 1.233E+07 3348E+07 2.184E-12 U 238 (a,f) Cs-137 (Cd shielded) l 94-1676 MID 4.840E+05 5390EkV>
3.552E-14 i
Np 237 (a.f) Cs-137 (Cd shielded)
{
94-1675 MID 3310E+06 3.686E+07 2314E-13 l
6-22 i
o TABLE 6-10
SUMMARY
OF NEUTRON DOSIMETRY RESULTS SURVFrILANCE CAPSULES U AND X h
+
Calculation of Measured Fluence for Capsule U Hux Time Fluence Uncertainty i
Measured Fluence < 0.414 MeV 1.096E+11 3.621E+07 3.969E+18 221 %
l Measured Fluence > 0.1 MeV 4.790E+11 3.621E+07 1.735E+19 15 %
Measured Fluence > 1.0 MeV 1.086E+11 3.621E+07 3.933E+18 28%
Measured dpa 2.076E-10 3.621E+07 7.518E-03 211%
l i
Calculation of Measured Fluence for Capsule X Flux Time Fluence Unceruunty Measured Fluence < 0.414 MeV 7.887E+10 1.330E+08 1.049E+19 221 %
i Measured Fluence > 0.1 MeV 3.485E+11 1.330E+08 4.635E+19 115 %
Measured Fluence > 1.0 MeV 8.466E+10 1.330E+08 1.126E+19 28%
Measured dpa 1.557E-10 1.330E+0S 2.071E42 211%
5 i
i t
f r
I i
l i
l J
1
8 TABLE 6-11 i
i COMPARISON OF MEASURED AND PERRET CALCULATED REACTION RATES AT' DIE SURVEILLANCE CAPSULE CENTER l
l i
SURVEILLANCE CAPSULE U ADJUSTED REACTION MEASURED CALCULATION fai.
Cu-63 (n,a) Co-60 6.06E-17 6.13E-17 1.01 I
i Fe-54 (n.p) Mn 54 5.92E-15 5.84E-15 0.99 Ni-58 (n.p) Co-58 7.66E-15 7.77E-15 1.01 j
U-238 (n,f) Cs-137 (Cd) 3.46E-14 3.32E-14 0.96 j
{
Np-237 (n,f) Cs-137 (Cd)
' 3.41E-13 3.43E-13 1.01 Co-59 (n,y) Co-60 5.54E-12 5.49E-12 0.99 Co 59 (n,y) Co-60 (Cd) 2.86E-12 2.87E-12 1.00 l
SURVEILLANCE Capsule X j
ADJUS7ED l
i REACTION MEASURED CALCULATION fai.
Cu-63 (n,a) Co-60 5.22E-17 5.25E-17 1.01 j
Fe-54 (n.p) Mn-54 4.75E 15 4.79E-15 1.01 Ni-58 (n.p) Co-58 6.50E-15 6.51E-15 1.00 i
U-238 (n,f) Cs-137 (Cd) 2.99E-14 2.69E-14 0.90 i
Np-237 (n.f) Cs-137 (Cd) 2.31E-13 2.47E-13 1.07 Co-59 (n,y) Co-60 4.10E-12 4.06E-12 0.99 Co-59 (n,y) Co-60 (Cd) 2.18E-12 2.19E-12 1.00 l
I 6-24 i
=
TABLE 6-12 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCE CAPSULE U i
ENERGY ADJUSTED FLUX ENERGY ADJUSTED FLUX GROUP NeV)
(n/cm'-see)
GROUP MeV)
(n/cm'-see)
I 1.733E+01 8.923E+06 28 9.119E-03 2.167E+10 2
1.492E+01 1.995E+07 29 5.531E-03 2.558E+10 I
3 1350E+01 7.610E407 30 3355E-03 8.048E+09 4
1.162E+01 1.692E+08 31 2.839E-03 7.755E+09 5
1.000E+0!
3.705E+08 32 2.404E-03 7.591E+09 6
8.607E+00 6.277E+08 33 2.035E-03 2.248E+10 7
7.408E+00 1.435E+09 34 1.234E-03 2.178E+10 8
6.065E+00 2.048E+09 35 7.485E-04 1.942E+10 9
4.966E400 4.289E+09 36 4.540E-04 1.6&lE+10 10 3.679E+00 5.711E+09 37 2.754E-04 1.958E+10 11 2.865E+00 1.189E+10 38 1.670E-04 2.136E+10
{
12 2.231E+00 1.643E+10 39 1.013E-04 2.151E+10 13 1.738E400 2306E+10 40 6.144E-05 2.150E+10 1
14 1353E400 2.587E+10 41 3.727E-05 2.106E+10 15 1.108E+00 4.853E+10 42 2.260E-05 2.042E+10 j
16 8.208E-01 5386E+10 43 1371E-05 1.971E+10 17 6393E-01 5.805E+10 44 8315E-06 1.879E+10 j
18 4.979E-01 3.969E+10 45 5.043E-06 1.757E+10 i
19 3.877E-01 5.725E+10 46 3.059E-06 1.641E+10 1
20 3.020E-01 6.221E+10 47 1.855E-06 1.488E+10 21 1.832E-01 5.816E+10 48 1.125E-06 1.136E+10 1
22 1.IllE-01 4.410E+10 49 6.826E 07 1.443E+10 23 6.738E-02 3.489E+10 50 4.140E-07 1.841E+10 j
24 4.087E-02 1.745E+10 51 2.511E-07 1.849E+10 25 2.554E-02 2367E+10 52 1.523E-07 1.768E+10 26 1.989E-02 1.199E+10 53 9.237E-08 5.501E+10 27 1.503E-02 1.776E+10 l
Note: Tabulated energy levels represent the upper energy in each group.
6-25 1
.j TABLE 6-13 i
ADJUSTED NEtJIRON ENERGY SPECIRUM AT THE CENIER OF SURVEILLANCE CAPSULE X ENERGY ADJUSTED FLUX ENERGY ADJUSTED FLUX r
GROUP MW (n/cm'-sec)
GROUP MW (nkzn'-sec) 1 1.733E+01 7.458E+06 28 9.119F 03 1.635E+10 2
1.492E+01 1.680E+07 29 5.531E 03 1.942E+10
[
3 1350E+01 6.453E+07 30 3355E-03 6.141E+09 i
4 1.162E+01 1.443E+08 31 2.839E 03 5.936FA09 5
1.00(E+01 3.170E+08 32 2.404F 03 5.82(E+09 6
8.607E+00 5362E+08 33 2.035E03 1.723E+10 7
7.408E+00 1.22(E409 34 1.234E-03 1.668E+10 8
6.065E+00 1.721E+09 35 7.485E-04 1.487E+10 9
4.966E+00 3.558E+09 36 4340E-04 1.284E+10 10 3.679E+00 4.674E+09 37 2.754E-04 1.490E+10 11 2.865E+00 9.629E+09 38 1.670E-04 1.631E+10 l
12 2.231E+00 1311E+10 39 1.013E-04 1.641E+10 13 1.738E+00 1.798E+10 40 6.144E 05 1.627E+10 14 1353FA00 1.950E+10 41 3.727E-05 1.605E+10 15 1.108E400 3.565E+10 42 2.260E 05 1.563E+10 16 8.208E-01 3.888E+10 43 1371E 05 1.513E+10 17 6393E 01 4.139E+10 44 8315E-06 1.444E+10 18 4.979F,01 2.812E+10 45 5.043E-06 1354E+10 19 3.877E-01 4.037E+10 46 3.059E-06 1.267E+10 20 3.020E-01 4386E+10 47 1.855E06 1.149E+10 21 1.832E.01 4.116E+10 48 1.125E 06 8.779E+09 22 1.111E 01 3.143E+10 49 6.826E-07 1.102E+10 23 6.738F-02 2.509E+10 50 4.14(E 07 1389E+10 24 4.087E-02 1.269E+10 51 2.511E 07 1372E+10 25 2.554E-02 1.740E+10 52 1523Fe07 1.293E+10 26 1.989E 02 8.909E409 53 9.237E-08 3.833E+10 27 1.503E 02 1332E+10 Note: Tabulated energy levels represent the upper energy in each group.
j 6-26
'I r
o TABLE 6-14 i
COMPARISON OF CALCULA~IED AND MEASURED NEUTRON EXPOSURE i
LEVELS FOR BRAIDWOOD UNIT 2 SURVEILLANCE CAPSULES U AND X I
i Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE for Capsule U Calculated Measured C/M Fluence (E > 1.0 MeV) [n/cm'-sec]
3.520E+18 3.933E+18 0.895 l
Fluence (E > 0.1 MeV) [n/cm'-sec]
1.509E+19 1.735E+19 0.870 dpa 6.863E-03 7.518E-03 0.913 i
i Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE for Capsule X
[
Calculated Measured C/M Fluence (E > 1.0 MeV) [n/cm'-sec) 1.075E+19 1.126E+19 0.954 2
Fluence (E > 0.1 MeV) [n/cm -sec) 4.607E+19 4.635E+19 0.994 dpa 2.095E-02 2.071E-02 1.012 t
?
1 I
i t
I t
l 6-27
.~
e
-2
. TABLE 6-15 '
l NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS 1
_.ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE i
if BEST ES11 MATE EXPOSURE (4.215 EPPY) AT THE PRESSURE VESSEL INNER RADIUS -
ODEG-15 DEG 25 DEG 35 DEG 45 DEG i
. E > 1.0 1.711E+18 2.534E+18 2.896E+18 2.452E+18
' 2.864E+18
-l E > 0.1 3.54IE+18 5.311E+18 7.875E+18 6.934E+18 7.14ml8
'dpa 2.562E-03 3.777E43 4.667E-03 3.996E 03 4.393E 03 f
i BEST ESTIMATE EXTRAPOLATION PLUX AT THE PRESSURE VESSEL INNER RADIUS l
l 0DEG 15 DEG 25 DEG 35 DEG 45 DEG
.j E > 1.0 1.286E+10 1.905E+10 2.178E+10 1.843E+10 2.153E+10
'l i
E > 0.1 2.662E+10 3.99m10 5.921E+10 5.214E+10 5.370E+10 dpa 1.926E-11 2.840E-11 3.509E-11 3.005E-11 3.303E-11 t
l BEST ESTIMATE EXPOSURE (16.0 EPPY) AT THE PRESSURE VESSEL INNER RADIUS ODEG 15 DEG 25 DEG 35 DEG 45 DEG E > 1.0 6.495E+18 9.620E+18 1.100E+19 9.307E+18 1.087E+19 E > 0.1 1.344E+19 2.016E+19 2.990E+19 2.633E+19 2.712E+19 dpa 9.727E-03 1.434E 02 1.772E-02 1.517E-02 1.668E-02
~
i BEST ESTIMATE EXPOSURE (32.0 EPPY) AT THE PRESSURE VESSEL INNER RADIUS 0DEG 15 DEG 25 DEG 35 DEG 45 DEG 1
l
?
E > 1.0 1.299E+19 1.924E+19 2.199E+19 1.861E+19 2.174E+19 E > 0.1 2.689E+19 4.033E+19 5.979E+19 5.265E+19 5.423E+19 dpa 1.945E-02 2.868E-02 3.543E-02 3.034E-02 3.336E 02
)
i 6-28 i
7, A
. TABLE 6-16 NEUTRON EXPOSURE VALUES FOR THE BRAIDWOOD UNIT 2 REACIDR VESSEL FLUENG BASED ON E > 1.0 MeV SLOPE ODEG 15 DEG
- 25 DEG 35 DEG 45 DEG l'
16 EFPY FLUENG SURFAG 6.495E+18 9.620E+18 1.100E+19 9.307E+18 1.087E+19 1/4T 3.527E+18 5.204E+18 6.003E+18 5.082E+18 5.794E+18 3/4T 7.535E+17 1.097E+18 1.297E+18 1.089E+18 1.174E+18 32 EFPY FLUENCE SURFACE 1.299E+19 1.924E+19 2.199E+19 1.861E+19 2.174E+19 1/4T 7.054E+18 1.041E+19 1.201E+19 1.016E+19 1.159E+19 3/4T.
1.507E+18 2.193E+18 2.595E+18 2.178E+18 2.348E+18 FLUENG BASED ON dpa SLOPE ODEG 15 DEG 25 DEG 35 DEG 45 DEG 16 EFPY FLUENCE SURFACE 6.495E+18 9.620E+18 1.100E+19 9.307E+18 1.087E+19 1/4T 4.099E+18 6.022E+18 7.147E+18 6.180E+18 6.936E+18 3/4T 1.422E+18 2.087E+18 2.628E+18 2.318E+18 2.370E+18 32 EFPY FLUENCE SURFAG 1.299E+19 1.924E+19 2.199E+19 1.861E+19 2.174E+19 1/4T 8.197E+18 1.204E+19 1.429E+19 1.236E+19 1.387E+19 3/4T 2.845E+18 4.175E+18 5.256E+18 4.635E+18 4.740E+18 1
6-29
i TABLE 6-17.
UPDATED LEAD FACTORS FOR BRAIDWOOD UNIT 2 SURVETILANCE CAPSULES l
i CAPSULE ISAD FACTOR U
4.00*
V 3.70 W
4.02 X
4.02**
I Y
3.70 Z
4.02 I
- Withdrawn EOC 1
- Withdrawn Near EOC 4, Basis for this Analysis L
t I
h 6-30 I
p...i........,..
SECTION
8.0 REFERENCES
1, P
d T =*nry Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vesse! Materials, U.S.
Nuclear Regulatory Commission, May 1988.
2.
Code of Federal Reda'ians,10 CFR Pat 50, Appandir G, Fracture Toughness Requirements, U.S. Nuclear Reed =*ary rwmmacion, Waehinataa D.C.
3.
WCAP-11188, Commonwealth Edison Company Braidwood Station Unit No. 2 Reactor Vessel Radiation Surveillance Program, L. R. Singer, December 1986.
t 4.
Section XI of the ASME Boiler and Pressure Vessel Code, A=adir G. Protection Against Nonductile Failure.
5.
ASTM E208, Standard Test Methodfor Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature ofFerritic Steels.
1 6.
WCAP-12845, Analysis of Capsule Ufrom the Commonwealth Edison Canpany Braidnood Unit 2 Reactor Vessel Radiarion Surveillance Program, E. Terek, et al., March 1991.
7.
Maternal Test Results, Upper Shell, 'Ibe Japan Steel Works, Ltd., JSW Job No. FN3-4274, IR No. 4274-309 (1), December 13, 1974.
8.
Material Test Results, Lower Shell, 'Ibe Japan Steel Works, IJd., JSW Job No. FN3-4274, IR No. 4274-410 (1), May 30,1975.
9.
Babcock & Wilcox Co., Record of Filler Wire Qualification Test, Test No. WF-562,562-1, &
562 2, Revised date 5/2/78.
10.
Analytical Request #15482, "Braidwood Nuclear Plant, Unit 2. Irradiated low Alloy Steel Reactor Surveillance Dosimetry". L. Kardos, dated October 27,1994.
l 8-1
\\
\\
l l
__-_-______-______-___a
7 i
11.
Code of Federal Reed %,10CFR50, Appendix H Reactor VesselMaterialSurveillance i
Progran Requirements U.S. Nuclear Regulatory Conunission, Washington, D.C.
12.
ASTM E185-82, Standard Practicefor Conducting Surveillance Testsfor Light-Water Cooled l
Nuclear Power Reactor Vessels.
i 13.
ASTM E23-93a, Test Methodsfor Notched Bar impact Testing of Metallic Materials.
l
)
14.
ASTM A370-92. Standard Test Methods and Definitionsfor Mechanical Testing ofStee!
Products.
l 15.
ASTM E8-93, Test Methods of Tension Testing of Metallic Materials.
1 i
16.
ASTM E21-92, Standard Practicefor Elevated Temperature Tension Tests of Metallic i
Materials.
l 17.
ASTM E83-93, Practicefor Venfication and Classification of Extensometers.
l8.
ASTM Designation E853-87, Standard Practicefor Analysis and Interpretation of l
Light. Water Reactor Surveillance Results, in ASTM Standards, Section 12, American Society for Testing and Materials, PhilaMphia, PA,1993.
I
)
19.
ASTM Designation E693-79, Standard Practicefor Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa), in ASTM Standards, Section 12 American Society for Testing and Materials, Philadelphia, PA,1993.
20.
WANL PR(LL)-034, Vol. 5, Nuclear Rocket Shielding Methods, Modification, Updating and input Data Preparation. Vol. 5--IWo-Dimensional Discrete Ordinates Transport Technique, R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler, August 1970.
21.
ORNL RSCI Data Library Collection Dir-76 SAILOR Coupled Sef-Shielded. 47 Neutron. 20
}
l Gamma-Ray, P3, Cross Section Libraryfor Light Water Reactors.
I 8-2
7 y
=.
.22..
Nuclear Science and Faciamine Volume 94, Accountingfor Changing Source Distributions in Ught Water Reactor Surveillance Dosimetry Analysis, R. E. Maerker, et al., Pages 291-308,.
.I 1986.
I 23.
J. V, Alexander, K. E. Bahr and E. F. Pulver, " Core Physics Parameters and Plant Operations i
Datafor the Braidwood Generating Station Unit 2 Cycle 1,* WCAP-11475, June 1987.~
l 24.
M. J. M=i===W
- Nuclear Design Reportfor Braidwood Unit 2, Cycle 2 " NPSR-0080. Rev.0, l
(Commanweakh Edison Propnetary), May 1990.
i i
~l 25.
P. D. Cichanai, A. W. Wong and S. K. Kujak, ' Nuclear Design Reportfor Braiducod Unit 2, Cycle 3," NPSR-0094. Rev.0, (Commonweahh Edison Propnetary) November 1991.
l i
26.
T. L. Stevens, et al, " Nuclear Design Reportfor Braidwood Unit 2, Cycle 4," NPSR-0102.
Rev.0, (C9mmonwealth Edison Ptopnetary), April 1993.
i i
27.
ASTM Designation E482-89, Standard Guidefor Application ofNeutron Transpon Methods for Reactor VesselSurveillance, in ASTM Standards, Section 12, American Society for
-l Testing and Matenals, Philadelphia, PA,1993.
i 28.
AS*IM Designation E560-84, Standard Recommended Practicefor Enrapolating Reactor Vessel Surveillance Dosimetry Results, in ASTM Standards, Secuan 12 American Society for i
Testing and Materials, Pluladelphia, PA,1993.
29.
ASTM Designation E693-19, Standard Practicefor Characterizing Neu,ron Exposures in t
Ferritic Steels in Terms ofDisplacements per Atom (dpa), in AS*IM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
t 30.
ASTM Designation E706-87, Standard Master Matrixfor Light-Water Reactor Pressure Vessel l
Surveillance Standard, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
{
t i
83
~
f I
31.
ASTM Designation E853-87, Standard Practicefor Analysis and interpretation of l
Light. Water Reactor Surveillance Results, in ASTM Standards, Section 12, American Society
{
for Testing and Matenals, Philadelphia, PA,1993.
?
b 32.
ASTM Designanon E261-90,' Standard Methodpr Determining Neutron Flux, Fluence, and Spectra by ? di~=M!vation Techniques,in ASTM Standards, Section 12, American Society for Testing and Matenals, Philadelphia, PA,1993.
l i
33.
ASTM Designation E262-86, Standard Methodfor Measuring Thermal Neutron Flux by Fadi~=Mmtion Techniques, in ASTM Standards, Section 12, Amencan Society for Testing f
and Materials, Philadelphia, PA,1993.
'j F
34.
ASTM Designation E263-88, Standard Methodfor Determuung Fast-Neutron Flux Density by Rad 3~=Mwarion offron,in ASTM Standards, Section 12, Amencan Society for Testag and Matenals, Philadelphia, PA,1993.
l 35.
ASTM Designation E264-92, Standard Methodpr Determming Fast-Neutron Flux Density by a
Radumetimtion ofNickel, in ASTM Standards, Section 12 Amencan Society for Testing and j
Materials, Philadelphia, PA,1993.'
36.
ASTM Designation E481-92, Standard Methodfor Measunng Neutron Flux Density by Radmacrimtion of Cobalt and Silver, in ASTM Standards, SIeenan 12, Amencan Society for Testing and Matenals, Philadelphia, PA,1993, i
i f
37.
ASTM Designation ES23-87, Standard Methodfor Determining Fast-Neutron Flux Density by Radmactivation of Copper, in ASTM Standards, Section 12, American Society for Testing and Matenals, Philadelphia, PA,1993.
38.
ASTM Designation E704-90, Standard Methodfor Measuring Reacnon Rates by
?ae~>Mmtion of Uranium-238, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
j i
8-4 4
b 39.
- AS'IM Designation E705-90, Standard Methodfor Measuring Fast-Neutron Flut Density by Radioactivation ofNeptunium 237,in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelpha, PA,1993.
l 40.
ASM Designauon E1005-84, Standard Methodfor Application and Analysis of Radiometric Monitorsfor Reactor Vessel Surveillance, in ASTM Standards, Section 12, Amencan Society I
i for Testing and Materials, Philadelphia, PA,1993.
41.
HEDL-TME 79-40, FERRETData Analysis Core, F. A. Schmittroth, Hanfold Engmeermg Development Laboratory, Richland, WA, September 1979.
I i
42.
AFWL-TR-7-41, Vol. I-IV, A Computer-Automated iterative Method ofNeutron Flux Spectra Determined by Foil Activation, W. N. McElmy, S. Berg and T. Crocket, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1%7.
43.
EPRI-NP-2188, Development and Demonstration of an Advanced Methodologyfor LWR Dosimetry Applications, R. E. Maetker, et al.,1981.
ti b
I 8-5 a
E, I -
APPENDIX A LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS I
i A-0 l
w Cune 784472 aft 3 Wi Wp e
-PM = Maximim Load P = Fast Fracture Load u
p P
General GY = Yield Load l
u la I
3 l
P = Fast Fracture p
u i
Arrest Load I
I I
I I
i I
I I
I i
I I
~
l l
l 1
1 I
I I
I i
i 4--tgy tM tp 2
Time W, = Fracture Initiation Region tGY = Time to GeneralYielding Wp = Fracture Propagation Region t
l M = Time to Maximum Load t
= Time to Fast (Brittle) Fracture Start p
Fig. A-1-Idealizedload-time record
. - ~
(
r i
BRAIDuCOD 82
- X*
FL5?
i i
i 4
a 54 S *a-T p
w i
$a
- f. -
.s 1.a a.4 3.s 4.s
- 6. o TIE
< MSEC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FL57 MATERIAL l
CAPSULE
- BRAIDWOOD #2 l
l sRAIDuGOD et
'X" FLS8 i
e i
a
\\
l
- e-l 1e a
n h
j n
v
~. _
h--
i.
l
.D
- 1. *.
a.4 3.6 4.5 6.0 l
i TIE
( MSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FL58 MATERIAL f
CAPSULE
- BRAIDWOOD #2 f
Figure A-2. Load-time records for Specimens FL57 and FL58.
i A-2 l
l
i
.L l
SECTION 7.0 i
SURVm LANCE CAPSULE WITHDRAWAL SCHEDULE I
f I
The following withdrawal schedule meets AS'IM E185-82 and is recommended for future capsules to be removed from the Braidwood Unit 2 seactor vessel:
l i
i TABLE 7-1 Recommended Surveillance Capsule Withdrawal Schedule i
i Capsule Location Withdrawal Fluence Capsule (Degree)
l 2
U *'
58.5 4.00 1.15 3.933 x 10
X *)
238.5 4.02 4.215 1.126 x 10" W
121.5 4.02 8
2.210 x 10" Z
301.5 4.02 12 3.315 x 10'*
I V
61.0 3.70 Standby -
i l
Y 241.0 3.70 Standby 301T5 I
(a) Effective Full Power Years (EFPY) fmm plant startup.
(b) Plant specific evaluation.
(c) Equal to the neutron fluence not less than once or greater than twice the maximum EOL (32 EFPY) inner vessel wall fluence.
h
.i F
t 7-1
_o:
nn::,ue::o.a x-
- ria, j
i i
e 0.-
s4 v
2 S "*-
w l
w-a l
=-
e -_--_____
.c
.a a.4 3.6 4.s s.o TIE -
( PtSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FT49 MATERIAL CAPSULE
- BRAIDWOOD #2
)
t Specimen Alignment Ermr - Data is not valid Figure A-In Load-time records for Specimens FT49 and FT54.
A-10
n t
9PAIDuc00 #2 ***
F.53
[
2 i
4 i
L o
i a
n ag i
ga o
g e
i
.t 1.2 2.4 3.6 4.8 6.0 TIE
( MSEC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FL53 MATERIAL i
CAPSULE
- BRAIDWOOD #2 enaroucoo e2 x-ns, 4
n s
l E "~
)
S t-l a
i w
W em l
g4 i
5 i
l q
i 1
'l W~~7-
.9 1.2 2.4 3.6 4.s' 6.0 TIE
< Mstc )
)
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
':FT59 MATERIAL CAPSULE
- BRAIDWOOD #2 Figure A 9. Load. time records for Specimens FL53 and FT59.
A-9 4
p
. e.f
'e.
fre:~ C *2
' ' c' r.4,*
f S4 7
)
l S 1-M
~
l l
a g-
.o
- 1. a a.4 3.s 4.s s.o TIE
( MSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FL47 MATERIAL I
~~
CAPSULE
- BRAIDWOOD #2 anatmsmo se x-rtss s
s s
a j 0-a 3 t-m w
N-
.o 1.a a.4 '
3.s 4.s ~
- 6. 0 TIE
( MSEC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FL56 MATERIAL CAPSULE
- BRAIDWOOD #2 l
L Figure A-8. Load-time records for Specimens FL47 and FL56.
A-8 l
l-l
)
aac.:: :::o.a
-c r' n j
4 4
I, I
(
E. 4 A
59-a w
a y_
l f
i.,-
.6 6..
T!ft
( ftSEC 3 BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FL50 MATERIAL CAPSULE
- BRAIDWOOD #2 snarouoao se x-rL46 a
e s
l 15-e a
S 9-n w
l I ~.
l
.D 1.2 2.4 3.6 4.0 6.0 TIfE
( ftSEC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FL46 MATERIAL CAPSULE
- BRAIDWOOD #2 Figure A-7. Load-time records for Specimens FL50 and FL46.
I A-7 l
.a.
ERAIDWCCD e2
- X" TL48 i
4 j
'L 4 A
S 9-n w
v a
q_
3.6' 4.8
- 6. 0
.0 1.2 2.4 TIE
( MSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FL48 MATERIAL CAPSULE
- BRAIDWOOD #2 BRAIOLEED #2
- X" FL60 i
a i
a g
~
z a
S 9-n N. -
1 i
- D 1.2 2.4 3.6 4.8
- 6. 0 TIE C MSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FL60 MATERIAL CAPSULE
- BRAIDWOOD #2 Figure A-6. Load-time records for Specimens FL48 and FL60.
A-6
e i
su:cac::o.z
-x-n.s.
e i
S 9-n w
4 v.
8d a
u-k
_____m
.D 1.2 2.4 3.6 4.8
- s. 0 TIPE C ft3EC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FL54 MATERIAL CAPSULE
- BRAIDWOOD #2 aneranno ea x-rtsi ji i
r.
a1._
i I
i i-
= n-e r
w a
N-
_-___-____w___
o
.t s.a e.4 3.s 4.s s.o TIPE"
( pistC )
BRAIDWOOD #2 "X"
i SPECIMEN NUMBER
- FLS1 MATERIAL
)
i CAPSULE
- BRAIDWOOD #2 Figure A-5. Load time records for Specimens FL54 and FL51.
P h
A-5
s 1
I i
era?.. O e2 x-rL55 s
n s
4 i
l
&4 i
~
i a
t S 9-n w
t I
i w,
4 I
k ~.
i
.D 1.2 2.4 3.6 4.8 6.0 T!!C
( MSEC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FL55 MATERIAL 3
CAPSULE
- BRAIDWOOD #2 i
DRAIDWIXD #2
- x" FL49 i
g E
$ ~~
f
^
5 9-I f t
e s w
w-4
+
u-t r
.c 1.2 e.4 3.s 4.e
- s. o fire
( MSEC )
j BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FL49 r
MATERIAL l
CAPSULE
- BRAIDWOOD #2 l
l i
Figure A-4. Load-time records for Specimens FL55 and FL49.
l
+
i k
A-4
[7 e
i
$#a!Cm e2 x"
rL59 4
s i
7 e-E#
[
l t
j 3 "._
i, w
I
=
a i
~-
\\
r A_
o A.
t
.9 1.2 2.4 3.6 4.8 6.0, i
TIPE
( PtSCC )
BRAIDWOOD #2 "X"
i SPECIMEN NUMBER
- FL59 MATERIAL i
CAPSULE
- BRAIDWOOD #2 Saarm eno sa x-rtse g
i i
i e
i i
I" e-I e
z S *. -
w N
=_
i H
f o
a i
.9 1.2 2.4 3.6 4.8 6.0 TIPE
( PISCC >
i BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FL52 MATERIAL i
CAPSULE
- BRAIDWOOD #2 i
i i
i Figure A-3. Load-time records for Specimens FL59 and FL52.
i t
l A-3 i
s,.
etarcu::::o.a w-risa i
j 7
e-E. 4 i
A 51-M w
u-f
= - -
.o s.e e.4 2.s 4.s 6.o TIE
( ftSEC )
i BRAIDWOOD #2-
"X" SP'CIMEN NUMBER
- FT52 r
MA'ERIAL CA/SULE
- BRAIDWOOD #2 saatouaco ea x-rTss j
i e
a a
)"
~
m a
51-a l
w
.D 1.2 2.4 3.6 4.9 6.0 TIE
( PtSEC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FT55 MATERIAL CAPSULE
- BRAIDWOOD #2 Figure A-12. Load-time records for Specimens FT52 and FT55.
A-12
~
k.
BP4:01ClD e2 "x"
rT46 I
e S
s'
~
a i
5 *-
n t
v e
4 t
N. -
t m_
.t 1.2 2.4 3.6 4.8 6.O I
T!!C
( ftSCc )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FT46 MATERIAL CAPSULE
- BRAIDWOOD #2 BRAIDLEED e2
- x*
FT60 i
l 1
J at-
- ) * -
n l
l *.
i
=-
C.-
i
.s 1.2 2.4 3.6 4.s 6.o Tric
< nste >
l BRAIDWOOD #2 "X"
1 SPECIMEN NUMBER
- FT60 MATERIAL CAPSULE
- BRAIDWOOD #2 s
t Figure A 11. Load-time records for Specimens FT46 and FT60.
A-11 i
t
i.'
==
sear:wc= ea x-rr::
a s
s 3s
?._
&4 7
A
. g_
+
n w
Rd a
y_
f
- =
.c 1.2 a.4 3.6 4.s s.o TIE C MSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FT51 MATERIAL CAPSULE
- BRAIDWOOD #2 l
bra!D e et
- x*
FT47 s
a e
s i
4 i
? e_
&4 I
i 59-n I
r
.N. -
l
.c 1.7 2.4 3.6 4.s 6.0 TIE
( MSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FT47 MATERIAL CAPSULE
- BRAIDWOOD #2 Figure A-14. Load-time records for Specimens FT51 and FT47.
A-14
e.
l E#AIDWDDD 't "x" _
rT56 l
j
.s;
~
'l
~
5 "*-
w 1
I
$"t-a i
w
\\
~
k
.9 1.2 2.4 3.6 4.s 6.0 7tpE
( MSEC >
(
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FT56 MATERIAL i
CAPSULE
- BRAIDWOOD #2 anato m se x-rT4e l
7 m-5-
~
i o
h 5 *-
a v
P
- v. -
N r
i a-i
's
.c t.e e.4 3.6 4.s 6.o Tret
< nste )
l BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FT48 MATERIAL CAPSULE
- BRAIDWOOD #2 l
~
i Figure A-13. Load-time records for Specimens FT56 and FT48.
A-13
)
l
~,,
i, e
sn.t ~m:0.a
-e ner 4
4 j }
7=-
8' A
e e-
-' ci w
$a y-n i.e' e.4'
- 3. 6 '
'4.s' 6.e TIE C ftSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FT57 MATERIAL CAPSULE
- BRAIDWOOD #2 snarouomo se x-nso i
$7 s
A 5 *-
M w
v-
$a q-
.3 s.e e.4 3.6 4.s 6.o T!E
( ftSEC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FT50 MATERIAL CAPSULE
- BRAIDWOOD #2 i
I Figure A-16. Load-time records for Specimens FT57 and FT50.
A-16 1
I
a
=
GRa!3.tDD ea x-rT53 J
4 7 m.
&4 e
S 9-m w
w-N. -
- 8 1.2 2.4 3.6 4.8 6.0 TIfC
( MSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FT53 MATERIAL CAPSULE
- BRAIDWOOD #2 nearoucoo na x-rise j
~
z a
51-n w
a q-o n.e' e.4
- 3. 6 '
4.s'
- s. o T!fE
( MSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FT58 MATERIAL CAPSULE
- BRAIDWOOD #2 Figure A-15. Load-time records for Specimens FT53 and FT58.
A-15
5 spa.!CacD s2 "4"
rus7 utt.c i
4 4
4 j
0 e-E. '
A n
w t-
$a n-
.5 1.2 2.4 3.6 4.s
- 6. 0 Titt C MSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMEER
- FW47 MATERIAL
- WELD CAPSULE
- BRAIDWOOD #2 BRAttuXX) e2
- x*
FM54 MA2 i
?
e-E. 4 m
w I
u-t i.e' z.4'
- s. 6 '
4.s' 6.o T!PC
( MSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FW54 MATERIAL
- WELD CAPSULE
- BRAIDWOOD #2 Figure A*18. Load-time records for Specimens FW47 and FW54.
A-18
e.
ERA!D e a2 ax' ruS3 3
uCLD s
s
&4 "e
^
39-m i
i N-1
-r
.D 1.2 '
2.4'
- 3. 6 '
4.8' 6.0' TIE
( ttstC >
BRAIDWOOD #2 "X"
t SPECIMEN NUMBER
- FW53 MATERIAL
- WELD CAPSULE
- BRAIDWOOD #2
. __ ORAID M s2
- x" rW52 g
i m
i e
i k"
e..
e
,a n
S 9-n w
v 4
~
\\
.9 1.2 '
2.4'
- 3. 6 '
4.8' TIE
( ftSCC )
6.0 BRAIDWOOD #2 "X"
P SPECIMEN NUMBER
- FW52 MATERIAL
- WELD i
CAPSULE
- BRAIDWOOD #2 r
Figure A-17. Load-time records for Specimens FW53 and FW52.
1 f
I A-17 i
t
.e.
pa:ou::co.a x-ruso wcto I
j.
7e 6
I., #
S 9-e w
f v_
o
.D 1.2 2.4 3."
4.8 6.0 TIE C PtSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FW50 MATERIAL
- WELD CAPSULE
- BRAIDWOOD #2 searouaco se x-russ utto j
i S 9-a w
t g4 N-
.t s.e a.e 3.s 4.s
- 6. o TIE
< ftSEC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FW59 MATERIAL
~
CAPSULE
- BRAIDWOOD #2 Figure A-20. Load-time records for Specimens FW50 and FW59.
A-20
e BPAIDuCll0 s2 "X"
rud9 ugLD i
e s
a
&4 7
n w
v-
$ ~.
a q-N z
.t s.a a.4 3.6 4.s 6.o TIE
( PISEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FW49
!'ATEP.IAL
- WELD snarouano sa x-rus6 utto 4
mke
~
7 S *. -
n v
I 1
i e-1 gd
~
I i
9-
~
i i
.t
- 1. 2 E.4 3.6 4.s 6.0 T!PE C ftSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FW56 MATERIAL
- WELD CAPSULE
- BRAIDWOOD #2 Figure A-19. Load time records for Specimens FW49 and FW56.
f A-19
i g.
j I
l v.
l BRAIDuCDD e2 M-rude utLD s
s s
7._
g4 m
a 4_
n i
w a
=_
l i
.3 s.a a.4 s.6 4.s 6.o Tilt
( PtSEC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FW48 MATERIAL
- WELD CAPSULE
- BRAIDWOOD #2 enarouano se x-rue 6 uCLD 4
s s
e a
s
~
n a, * -
n v
.5 1.2 2.4 3.6 4.8
- 6. 0 TIfE
< ftSCC 3 BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FW46 MATERIAL
- WELD CAPSULE
- BRAIDWOOD #2 Figure A-22. Load-time records for Specimens FW48 and FW46.
1-A-22
____________-_____m.-_____m_______
r
+
e B*AIOWCCD at x*
rW57 WELD j
i 4
a i
c a4 7
a
, Y es M
uup t
h y-
.9 1.2 2.4 3.6 4.8 6.0 T:E
( MSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FW57 MATERIAL
~
CAPSULE I
- BRAIDWOOD #2 i
mentoWoon se
-x-runs utto a
s 50-
~
m t
5 *-
n W
I y-
.D 1.2 2.4 S.6 4.8
- 6. 0 TIE
( F4C )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FW55 MATERIAL
- WELD CAPSULE
- BRAIDWOOD #2 I
i I
Figure A-21. Load-time records for Specimens FW57 and FW55.
A-21
[J:
e b
- . g -
s U. air,i C.::3 42 rc:
- t
- .c
,g' 4
4 y e_
S4 7-n 5 *-
c w
a y_
o
.9 1.2 2.4 3.6 4.8
- 6. 0 7!E
( NSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FW51 MATERIAL
- WELD CAPSULE
- BRAIDWOOD #2 ORAIDutIE #2
- x" FM57 He2
,g i
a i
a
=_
8#
a e
e_
w,;
v
~_
4 o
.9 1.2 2.4 3.6 4.0
- 6. 0 TIE
( MSCC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH57 MATERIAL
- HAZ CAPSULE
- BRAIDWOOD #2 Figure A-24. Load-time records for Specimens FW51 and FH57.
A-24
c, BRAIDuD00 82 "x"
rW60 WCLD e
s 5#
m a
S 9-m w
v
- n. -
e i
i
.5 1.2 2.4 3.6 4.8
- 6. 0 Titt
( MSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FW60 MATERIAL
- WELD CAPSULE
- BRAIDWOOD #2 nearDuaco et x*
ruS8 WCLD 4
s 1
7 e-h S 9-n w
J N. -
.5 1.2 2.4 3.6 4.8 6.0 TIPC
( fCEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FW58 MATERIAL
- WELD CAPSULE
- BRAIDWOOD #2 Figure A-23. Load-time records for Specimens FW60 and FW58.
A-23
_J
BPAIOu::::0 #2 "x"
rH51 HAZ a
e i
s
- 8. #
A 5 *-
M w
j-i
=_
4
.c 1.e z.4 3.5 4.s s.o TIE
( MSEC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH51 MATERIAL
- HAZ CAPSULE
- BRAIDWOOD #2 snarounao se x-rwse Hn2
~~
~
a S *-
n w
= -
.1 1.2 2.4 3.6 4.8 6.0 TIE
( MSEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH60 MATERIAL
- HAZ CAPSULE
- BRAIDWOOD #2 Figure A-26. Load-time records for Specimens FH51 and FH60.
1 A-26 L.
4 BPA!D @ D e2 "x"
N6 HAZ j
4 6
0-E4 n
S 9-n I
v a
.n. -
t
.5 1.2 2.4 3.6 4.8 6.0 TIMC
( fCCC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH46 MATERIAL
- HAZ CAPSULE
- BRAIDWOOD #2 BRAIDuuuD #2
- X*
D447 HA2 j
i a
a i
0-h S 1-m w
w-ga a
N. -
.D 1.2 2.4 3.6 4.8 6.0 TINC
( MSCC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH47 MATERIAL
- HAZ CAPSULE
- BRAIDWOOD #2 Figure A.25. Load-time records for Specimens FH46 and FH47.
A-25 J
s:
15-::
2 e
r-n c.:
j 4
0 m_
E4 7
s S *-
a w
g =.
a o
.A
.0 1.2 2.4 3.6 4.t 6.0 TIPE
< Mste >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH59 MATERIAL
- HAZ CAPSULE
- BRAIDWOOD #2 BRAID S s2 "x"
rwe9 w
J
+
i i
7e S4
~
E
^
i 5 *. -
i n
w e
lN i
a N
l k
e i
=0 1.2 2.4 3.6 4.8 6.0 i
T!!C
( MSEC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH49 MATERIAL
- HAZ
~
CAPSULE
- BRAIDWOOD #2 Figure A-28. Load. time records for Specimens FH59 and FH49.
i A-28 t
s L-.
par:e,::o.2
-x-ruas saz j
O e-h 3 *, -
n I
w a
q-l o
a e' z.4
- 3. 6 '
4.s'
- 6. o TIPC
( fCCC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH48 MATERIAL
- HAZ CAPSULE
- BRAIDWOOD #2 sRarouaco sa x-rwse saz 0.-
4
~
n v
a N. -
m - _ _
.V 3.2 2.4 3.6 4.8
- 6. 0 TIMC
( fCCC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH58 MATERIAL
- HAZ CAPSULE
- BRAIDWOOD #2 Figure A-27. Load-time records for Specimens FH48 and FH58.
A-27 i
d
l l
ist.:0 *C e2 a'<*
t rw!4 M
j e
i I
7.-
&4 f
I a
i l
$ 1-m i
d l
me t
N_
t 1
.D 1.2 2.4 3.6.
4.8 6.0 TIE
< MstC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH54 MATERIAL
- HAZ CAPSULE
- BRAIDWOOD #2 saatoucoo e2 x-nes w
j g
i s
e 6
i i
?._
?
a S 9-i t
y_
t
.D 1.2 2.4 3.6 4.8 6.0 TIE
< MSEC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH53 i
MATERIAL
- HAZ
^
CAPSULE
- BRAIDWOOD #2 t
t t
Figure A-30. Load time records for Specimens FH54 and FH53.
t A-30 i
1 r
Pra::C e2
"<a rwn A:
?=-
h A
= +-
m w
w-q-
~
.0 1.2 2.4 3.6 4.8 6.0 TIME
< PCEC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH55 MATERIAL
- HAZ CAPSULE
- BRAIDWOOD #2 eRAIDWDDD s2
- x*
nc0 HRZ j
i h
a v
a q -
.c
.2 2.4
- 3. 6 4.s 6.o TIPC
( PCEC )
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH50 MATERIAL
- HAZ CAPSULE
- BRAIDWOOD #2 Figure A-29. Load-time records for Specimens FH55 and FH50.
A-29
p.'s-V
-. if " e2
~A*
rAE ut.2
.m si n
s-gg a
- 9. -
e
.O 1.2 2.4 3.6 4.4 6.0 T!!T
( ftSEC >
BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH52 i
MATERIAL
- HAZ CAPSULE
- BRAIDWOOD #2 searoucoo se x-rwss Haz
$ "~
~
7 e g n
w a
1 N-
~
e
.O
- .a r.4 3.6 4.s 6.o T!fC
( ft3CC )
i BRAIDWOOD #2 "X"
SPECIMEN NUMBER
- FH56 MATERIAL
- HAZ CAPSULE
- BRAIDWOOD #2 i
Figure A-31. Load-time records for Specimens FH52 and FH56.
- A-31 L.
- - - - - - - - - - -