ML17310B247
| ML17310B247 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 02/28/1994 |
| From: | LIPPINCOTT E P, MADEYSKI A, TEREK E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML17310B245 | List: |
| References | |
| WCAP-13935, NUDOCS 9405020098 | |
| Download: ML17310B247 (174) | |
Text
W CAP-13935 WESTINGHOUSE CLASS 3 (Non-Prorietary)
',2405020098
'.940415'DR""."ADOCK,0500052'P.P,."'PD~~Analysis of the 137'apsule from the Arizona Public Service Company Palo Verde Unit No.2 Reactor Vessel Radiation Surveillance Program E.Terek E.P.Lippincott A.Madeyski February 1994 Work Performed Under Shop Order MFYP-106 Prepared by Westinghouse Electric Corporation for the Arizona Public Service Company Approved by T.A.Meye, Manag r Structunl Reliability and Plant Life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O.Box 355 Pittsburgh, Pennsylvania 15230-0355
-1994 Westinghouse Electric Corporation 0 e PREFACE This report has been technically reviewed and verified.Reviewer: Sections 1 through 5, 7, 8, and Appendix A Section 6 J.M.Chicots J.<(.CIVCc G.N.Wnghts TABLE OF CONTENTS Section Title~Pa e 1.0
SUMMARY
OF RESULTS
2.0 INTRODUCTION
2-1 3.0 BACKGROUND 3-1
4.0 DESCRIPTION
OF PROGRAM 5.0 TESTING OF SPECIMENS FROM CAPSULE W137 5.1 Overview 5.2 Charpy V-Notch Impact Test Results 5.3 Precracked Charpy Specimen Test Results 5.4 Tension Test Results 5-1 5-1 5-5 5-7 5-8 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6.1 Introduction
6.2 Discrete
Ordinates Analysis 6.3 Neutron Dosimetry 6.4 Projections of Pressure Vessel Exposure 6-1 6-1 6-2 6-6 6-11 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1
8.0 REFERENCES
8-1 APPENDIX A: Load-Time Records for Charpy Specimen Tests and Comparisons of Data for Unirradiated and Irradiated Becracked Charpy Specimens LIST OF TABLES Table Title Pa~e 4-1 Chemical Composition (wt%)of the Palo Verde Unit 2 Reactor Vessel Surveillance Materials 4-2 Heat Treatment of the Palo Verde Unit 2 Reactor Vessel Surveillance Materials 4-3 Summary of Unirradiated Surveillance Material Data Arrangement of Encapsulated Test Specimens by Code Number within the Palo Verde Unit 2 137'apsule 5-1 Charpy V-notch Data for the Palo Verde Unit 2 Lower Shell Plate F-773-1 5-9 Irradiated at 550'F to a Fluence of 4.071 x 10" n/cm (E>1.0 MeV)(Longitudinal Orientation) 5-2 Charpy V-notch Data for the Palo Verde Unit 2 Lower Shell Plate F-773-1 Irradiated at 550'F to a Fluence of 4.071 x 10" n/cm (E>1.0 MeV)(Transverse Orientation) 5-10 5-3 Charpy V-notch Data for the Palo Verde Unit 2 Surveillance Weld Metal Irradiated at 55Q F to a Fluence of 4.Q71 X 10'/cm (E>1.0 MeV)5-11 Charpy V-notch Data for the Palo Verde Unit 2 Heat-Affected-Zone (HAZ)5-12 Metal Irradiated at 550'F to a Fluence of 4.071 X 10" n/cm (E>1.0 MeV)5-5 Charpy V-notch Data for the Palo Verde Unit 2 Correlation Monitor Standard 5-13 Reference Material HSST 01MY Irradiated at 550'F to a Fluence of 4.071 x 10" n/cm'E>1.0 MeV)(longitudinal Orientation) 5-6 Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Lower 5-14 Shell Plate F-773-1 Irradiated at 550'F to a Fluence of 4.071 x 10" n/cm~(E>1.0 MeV)(Longitudinal Orientation)
LIST OF TABLES (continued)
Table Title~Pa e 5-7 Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Lower 5-15 Shell Plate F-773-1 Irradiated at 550'F to a Fluence of 4.071 x 10" n/cm~(E>1.0 MeV)(Transverse Orientation) 5-8 Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Surveillance Weld Metal Irradiated at 550'F to a Fluence of 4.071 x 10" n/cm'E>1.0 MeV)5-16 5-9 Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Surveillance Heat-Affected-Zone (HAZ)Metal Irradiated at 550'F to a Fluence of 4.071 x 10'/cm (E>1.0 MeV)5-17 5-10 Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Surveillance Standard Reference Material HSST 01MY Irradiated at 550'F to a Fluence of 4.071 x 10" n/cm~(E>1.0 MeV)(Longitudinal Orientation) 5-18 , 5-11 Effect of 550'F Irradiation to 4.071 x 10" n/cm (E>1.0 MeV)on the Notch Toughness Properties of the Palo Verde Unit 2 Reactor Vessel Surveillance Materials 5-19 5-12 Comparison of the Palo Verde Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions 5-20 5-13 Precracked Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Lower Shell Plate F-773-1 Irradiated at 550'F to a Fluence of 4.071 x 10" n/cm (E>1.0 MeV)(Longitudinal Orientation) 5-21 5-14 Precracked Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Lower Shell Plate F-773-1 Irradiated at 550'F to a Fluence of 4.071 x 10" n/cm'E>1.0 MeV)(Transverse Orientation)
LIST OF TABLES (continued)
Table Title Pace 5-15 Precracked Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Surveillance Weld Metal Irradiated at 550'F to a Fluence of 4.071 x 10" n/cm (E>1.0 MeV)5-23 5-16 Tensile Properties of the Palo Verde Unit 2 Reactor Vessel Surveillance Materials Irradiated at 550'F to 4.071 x 10" n/cm'E>1.0 MeV)5-24 6-1 Calculated Fast Neutron Exposure Rates at the Surveillance Capsule Center 6-13 6-2 Calculated Azimuthal Variation of Fast Neutron Exposure Rates at the Pressure Vessel Clad/Base Metal Interface 6-14 6-3 Relative Radial Distribution of)(E>1.0 MeV)within the Pressure Vessel Wall 6-15 Relative Radial Distribution of$(E>0.1 MeV)within the Pressure Vessel Wall 6-16 6-5 Relative Radial Distribution of dpa/sec within the Pressure Vessel Wall 6-17 6-6 Nuclear Parameters used in the Evaluation of Neutron Sensors 6-18 6-7 Monthly Thermal Generation During the First Four Fuel Cycles of the Palo Verde Unit 2 Reactor 6-19 6-8 Measured Sensor Activities and Reaction Rates Surveillance Capsule W137 6-20 6-9 Summary of Neutron Dosimetry Results Surveillance Capsule W137 6-21 6-10 Comparison of Measured and Ferret Calculated Reaction Rates at the Surveillance Capsule Center Surveillance Capsule W137 6-21 LIST OF TABLES (continued)
Table Title 6-11 Adjusted Neutron Energy Spectrum at the Center of Surveillance Capsule W137 6-22 6-12 Comparison of Calculated and Measured Neutron Exposure Levels for Palo 6-23 Verde Unit 2 Surveillance Capsule W137 6-13 Neutron Exposure Projections at Key Locations on the Pressure Vessel Clad/Base Metal Interface 6-24 6-14 Neutron Exposure Values 6-25 6-15 Updated Lead Factors for Palo Verde Unit 2 Surveillance Capsules 6-26 7-1 Palo Verde Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule 7-1 LIST OF ILLUSTRATIONS
~Fi ure Title~Pa e 41 Arrangement of Surveillance Capsules in the Palo Verde Unit 2 Reactor Vessel 4-2 Typical Palo Verde Unit 2 Surveillance Capsule Assembly 43 Typical Palo Verde Unit 2 Surveillance Capsule Charpy Impact Compartment 4-9 Assembly Typical Palo Verde Unit 2 Surveillance Capsule Tensile-Monitor Compartment Assembly 4-10 5-1 Palo Verde Unit 2 Capsule W-137 Thermal Monitors 5-25 5-2 Charpy V-Notch Impact Properties for Palo Verde Unit 2 Reactor Vessel Lower Shell Plate F-773-1 (Longitudinal Orientation) 5-26 5-3 Charpy V-Notch Impact Properties for Palo Verde Unit 2 Reactor Vessel Lower Shell Plate F-773-1 (Transverse Orientation) 5-27 Charpy V-Notch Impact Properties for Palo Verde Unit 2 Reactor Vessel Surveillance Weld Metal (F-773-2/F-773-3) 5-28 5-5 Charpy V-Notch Impact Properties for Palo Verde Unit 2 Reactor Vessel Weld Heat-Affected-Zone Metal 5-29 5-6 Charpy V-notch Impact Properties for Palo Verde Unit 2 SRM HSST 01MY 5-30 (Longitudinal Orientation) 5-7 Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Reactor Vessel Lower Shell Plate F-773-1 (Longitudinal Orientation) 5-31 LIST OF ILLUSTRATIONS
~Fi ure Title 0 5-8 Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Reactor 5-32 Vessel Lower Shell Plate F-773-1 (Transverse Orientation) 5-9 Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Reactor Vessel Surveillance Weld Metal 5-33 5-10 Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Reactor Vessel Heat-Affected-Zone Metal 5-34 5-11 Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 SRM HSST 5-35 01MY (Longitudinal Orientation) 5-12 Comparison of Unirradiated and Irradiated Dynamic Fracture Toughness Values Determined by Testing of Precracked Charpy Specimens from Lower Shell Plate F-773-1 (Longitudinal Orientation) 5-36 5-13 Comparison of Unirradiated and Irradiated Dynamic Fracture Toughness Values Determined by Testing of Precracked Charpy Specimens from Lower Shell Plate F-773-1 (Transverse Orientation) 5-37 5-14 Comparison of Unirradiated and Irradiated Dynamic Fracture Toughness Values Determined by Testing of Precracked Charpy Specimens from the Palo Verde Unit 2 Surveillance Weld Metal 5-38 5-15 Precracked Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Lower Shell Plate F-773-1 (Longitudinal Orientation) 5-39 5-16 Precracked Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Lower Shell Plate F-773-1 (Transverse Orientation) 540 e vnI LIST OF ILLUSTRATIONS
~Fi ure Title~Pa e 5-17 Precracked Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Reactor Vessel Surveillance Weld Metal'I 5<1 5-18 Tensile Properties for Palo Verde Unit 2 Reactor Vessel Lower Shell Plate F-773-1 (Transverse Orientation) 5<2 5-19 Tensile Properties for Palo Verde Unit 2 Reactor Vessel Surveillance Weld Metal 5<3 5-20 Fractured Tensile Specimens from Palo Verde Unit 2 Reactor Vessel Lower Shell Plate F-773-1 (Transverse Orientation) 5-21 Fractured Tensile Specimens from Palo Verde Unit 2 Reactor Vessel Surveillance Weld Metal 5-45 5-22 Engineering Stress-Strain Curves for Lower Shell Plate F-773-1 Tensile Specimens 1B2J2 and 1B2K1 (Transverse Orientation) 5-46 5-23 Engineering Stress-Strain Curve for Lower Shell Plate F-773-1 Tensile Specimen 1B2J3 (Transverse Orientation) 5<7 5-24~Engineering Stress-Strain Curves for Weld Metal Tensile Specimens 1B3J7 5-48 and 1B3JY 5-25 Engineering Stress-Strain Curve for Weld Metal Tensile Specimen 1B3J5 5<9 6-1 Palo Verde Reactor Model Showing a 45 Degree (R,S)Sector 6-27 6-2 Azimuthal Variation of Neutron Flux (E)1.0 MeV)at the Reactor Vessel 6-28 Inner Radius 6-3 Axial Distribution of Reactor Power ix 6-29
SECTION 1.0
SUMMARY
OF RESULTS The analysis of the reactor vessel materials contained in the surveillance capsule removed from the 137'ocation, the first capsule to be removed from the Palo Verde Unit 2 reactor pressure vessel, led to the following conclusions:
o The capsule received an average fast neutron fluence (E>1.0 MeV)of 4.071 x 10" n/cm after 4.54 effective full power years (EFPY)of plant operation.
o Irradiation of the reactor vessel lower shell plate F-773-1 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal orientation), to 4.071 x 10" n/cm (E>1.0 MeV)resulted in a 30 ft-lb transition temperature increase of 10'F and a 50 ft-lb transition temperature increase of 25'F.This results in an irradiated 30 ft-lb transition temperature of 10'F and an irradiated 50 ft-lb transition temperature of 60'F for the longitudinally oriented specimens.
o Irradiation of the reactor vessel lower shell plate F-773-1 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major rolling direction of the plate (transverse orientation), to 4.071 x 10" n/cm'E>1.0 MeV)resulted in a 30 ft-lb transition temperature increase of 19'F and a 50 ft-lb transition temperature increase of 25'F.This results in an irradiated 30 ft-lb transition temperature of 15'F and an irradiated 50 ft-lb transition temperature of 55'F for transversely oriented specimens.
o Irradiation of the weld metal Charpy specimens to 4.071 x 10" n/cm'E>1.0 MeV)resulted in a 30 and 50 ft-lb transition temperature increase of 15'F.This results in an irradiated 30 ft-lb transition temperature of-28'F and an irradiated 50 ft-lb transition temperature of 4'F.o Irradiation of the weld Heat-Affected-Zone (HAZ)metal Charpy specimens to 4.071 x 10" n/cm'E>1.0 MeV)resulted in a 30 ft-lb transition temperature increase of 57'F and a 50 ft-lb transition temperature increase of 18'F.This results in an irradiated 30 ft-lb transition temperature of 45'F and an irradiated 50 ft-lb transition temperature of 75'F.1-1 o The average upper shelf energy of the lower shell plate F-773-1 Charpy specimens (longitudinal orientation) resulted in an average energy increase of 6 ft-lbs after irradiation to 4.071 x 10" n/cm (E>1.0 MeV).This results in an irradiated average upper shelf energy of 118 ft-lbs for the longitudinally oriented specimens.
o The average upper shelf energy of the lower shell plate F-773-1 Charpy specimens (transverse orientation) resulted in an average energy decrease of 21.5 ft-Ibs after irradiation to 4.071 x 10" n/cm (E>1.0 MeV).This results in an irradiated average upper shelf energy of 115 ft-lbs for the transversely oriented specimens.
o The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of 1 ft-lb after irradiation to 4.071 x 10" n/cm (E>1.0 MeV).This results in an irradiated average upper shelf energy of 108 ft-lbs for the weld metal specimens.
o The average upper shelf energy of the weld HAZ metal Charpy specimens increased 29 ft-lbs after irradiation to 4.071 x 10" n/cm (E>1.0 MeV).This results in an irradiated average upper shelf energy of 113 ft-lbs for the weld HAZ metal.o A comparison of the Palo Verde Unit 2 surveillance material 30 ft-lb transition temperature shifts and upper shelf energy decreases with Regulatory Guide 1.99, Revision 2"', lead to the following conclusions:
The 30 ft-lb transition temperature increases of the surveillance weld metal and the longitudinally oriented lower shell plate F-773-1 Charpy test results are less than the Regulatory Guide 1.99, Revision 2"', predictions.
The 30 ft-Ib transition temperature in'me and average upper shelf energy decrease of transversely oriented lower shell plate F-773-1 Charpy test results are in good agreement with Regulatory Guide 1.99, Revision 2"', predictions.
The measured average upper shelf energy decrease of the weld metal and lower shell plate F-773-1 longitudinally oriented Charpy test results are less than the Regulatory Guide 1.99, Revision 2'", predictions.
1-2 o The precracked Charpy specimen test results are in good agreement with the unirradiated test results"'.
The data are bounded by the KR curve, which provides a lower bound estimate for the fracture toughness.
o The calculated end-of-life (EOL)32 effective full power years (EFPY)maximum neutron fluence (E>1.0 MeV)for the Palo Verde Unit 2 reactor vessel is as follows: Vessel inner radius'2.047 x 10" n/cm Vessel 1/4 thickness=1.087 x 10'/cm Vessel 3/4 thickness=2.157 x 10" n/cm*Clad/base metal interface 1-3 0'
SECTION
2.0 INTRODUCTION
This report presents the results of the examination of the Palo Verde Unit 2 surveillance capsule removed from the 137'ocation.
This is the first capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Arizona Public Service Company Palo Verde Unit 2 reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Arizona Public Service Company Palo Verde Unit 2 reactor pressure vessel materials was designed and recommended by ABB Combustion Engineering.
A description of the preirradiation mechanical properties of the reactor vessel materials is presented in TR-V-MCM-013,"Arizona Public Service Company Palo Verde Unit 2 Evaluation for Baseline Specimens Reactor Vessel Materials Irradiation Surveillance Program""'.
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".The 137'apsule was removed from the reactor after less than 5 EFPY of exposure and shipped to the Westinghouse Science and Technology Center Hot Cell Facility, where the postirradiation mechanical testing of the Charpy V-notch impact, tensile and precracked Charpy V-notch surveillance specimens was performed.
This report summarizes the testing of and the post irradiation data obtained from the surveillance capsule removed from the 137'ocation of the Arizona Public Service Company Palo Verde Unit 2 reactor vessel and discusses the analysis of the data.2-1 j
SECTION 3.0 BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry.The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment.
The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA533 Grade B Class 1 (base material of the Palo Verde Unit 2 reactor pressure vessel)are well documented in the literatutu.
Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness during high-energy irradiation.
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 III of the ASME Boiler and Pressure Vessel Code'".The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RT~).RTur is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208"')or the temperature 60'F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the plate.The RT~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 toughness results obtained from several heats of pressure vessel steel.When a given material is indexed to the K~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.RT~~and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties.
The changes in mechanical properties of a given reactor pressure vessel steel, due to irradiation, can be monitored by a reactor vessel surveillance program, such as the Palo Verde Unit 2 reactor vessel materials irradiation surveillance program"', in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested.3-1 The increase in the average Charpy V-notch 30 ft-lb temperature (bRTNDr)due to irradiation is added to the initial RTNDr to adjust the RT~(ART)for radiation embrittlement.
The ART (RTNDr initial+dRT~)is used to index the material to the~curve and, in turn, to set operating limits for the nuclear power plant that take into account the effects of irradiation on the reactor vessel materials.
3-2 SECTION
4.0 DESCRIPTION
OF PROGRAM, Six surveillance capsules for monitoring the effects of neutron exposure on the Palo Verde Unit 2 reactor pressure vessel core region materials were inserted in the reactor vessel prior to initial plant start-up.The six capsules were positioned in the reactor vessel between the core support barrel and the vessel wall as shown in Figure 4-1.The vertical center of the capsules is opposite the vertical center of the core.The capsule, removed from the 137'ocation, consisted of three compartments (Figure 4-2).Each compartment consisted of two sections attached by a connecting spacer.The top and bottom compartments of the capsule (Figure 4-3)contained Charpy V-notch and precracked Charpy V-notch specimens along with flux monitors.The middle compartment of the capsule (Figure 4-4)contained tension and Charpy V-notch specimens along with flux and temperature monitors.The test specimens contained in the capsule were made from lower shell plate F-773-1 and submerged arc weld metal fabricated with Mil BP weld filler wire and are representative of the reactor vessel beltline region materials.
The capsule was removed after 4.54 EFPY of plant operation.
This capsule.contained Charpy V-notch, tensile, and precracked Charpy V-notch specimens made from lower shell plate F-773-1 and submerged arc weld metal representative of the reactor vessel beltline welds.In addition, this capsule contained Charpy V-notch specimens from the Heavy Section Steel Technology (HSST)plate 01MY and the weld HAZ metal from lower shell plate F-773-1.The Palo Verde Unit 2 reactor vessel lower shell plate F-773-1 was fabricated from steel plate produced according to ASME Specification SA-533 Grade B Class 1 mechanical properties.
The Palo Verde Unit 2 surveillance plate material was taken from sections of lower shell plate F-773-1.Weld metal material was fabricated by welding together lower shell plates F-773-2 and F-773-3.Weld HAZ test material was fabricated by welding together lower shell plates F-773-1 and F-773-2.Test specimens were machined from approximately the 1/4 thickness (1/4T)location.Specimens from the weld metal were machined from a weldment joining lower shell plate F-773-2 and adjacent lower shell plate F-773-3.All heat-affected-zone specimens were obtained from the weld heat-affected-zone of lower shell plate F-773-1.The Palo Verde Unit 2 surveillance capsule also contained Charpy V-notch specimens from a standard heat of ASTM A 533 Grade B Class 1 manganese-molybdenum-nickel steel made available by the NRC sponsored HSST Program.This reference material has been fully processed and characterized and was used for Charpy impact~specimen correlation monitors.
Charpy V-notch impact and tension specimens were machined from lower shell plate F-773-1 in both the longitudinal orientation (longitudinal axis of the specimen parallel to the major rolling direction of the plate)and transverse orientation (longitudinal axis of the specimen normal to the major rolling direction of the plate).Charpy V-notch and tensile specimens from the weld metal were oriented such that the long dimension of the specimen was normal to the welding direction.
The notch of the weld metal Charpy specimens was machined such that the direction of crack propagation in the specimen is in the welding direction.
Precracked Charpy V-notch test specimens from lower shell plate F-773-1 were machined in both the longitudinal and transverse orientations and precracked Charpy V-notch specimens from the weld metal were machined such that the simulated crack in the specimen would propagate in the direction of welding.The chemical composition and heat treatment of the surveillance material is presented in Tables 4-1 and 4-2.The chemical analysis reported in Table 4-1 was obtained from unirradiated material used in the surveillance program"'.
The capsule contained flux monitors made of sulfur, titanium, iron, nickel (cadmium shielded), copper (cadmium shielded), cobalt (cadmium shielded and unshielded) and uranium (cadmium shielded and unshielded).
The capsule contained thermal monitors made from four low-melting-point eutectic alloys sealed in glass tubes.These thermal monitors were used to define the maximum temperature attained by the test specimens during irradiation.
The composition of the four eutectic alloys and their melting points are as follows: 80%Au,20%Sn 90%Pb, 5%Sn, 5%Ag 2.5%Ag, 97.5%Pb 1.75%Ag, 0.75%Sn, 97.5%Pb Melting Point: 536'F (280'C)Melting Point: 558'F (292'C)Melting Point: 580'F (304'C)Melting Point: 590'F (310'C)4-2 TABLE 4-1 Chemical Composition (wt%)of the Palo Verde Unit 2 Reactor Vessel Surveillance Materials"'lement
.Si Cr Ni Mo Cb B Co As Sn Zf Sb Plate F-773-1 0.21 0.009 0.006 1.54 0.24 0.03 0.68 0.52 0.003<0.01<0.001 0.016 0.015 0.010<0.01 0.021 0.003<0.001 0.014 0.0018 Weld Metal F-773-2/F-773-3 0.47 0.011 0.010 1.46 0.12 0.10 0.09 0.51 0.005<0.01<0.001 0.010 0.07 0.005 0.010<0.01 0.010 0.004<0.001 0.011 0.0114 TABLE 4-2 Heat Treatment of the Palo Verde Unit 2 Reactor Vessel Surveillance Materials"'aterial Surveillance Program Test Plate F-773-1 Weldment Temperature
('F)Austenitizing:
1600+25 Tempered: 1225+25 Stress Relief: 1150+50 Stress Relief: 1125+25 Time (hr)40 44 hr.Ec 48 min.Coolant Water quenched Air cooled Furnace cooled to 600'F Furnace cooled to 600'F TABLE 4-3 Summary of Unirradiated Surveillance Material Data"'aterial and Code CUpper Shelf (ft-lb)30 ft-lb Index ('F)50 ft-lb Index ('F)35 Mils Lat.Exp.Index ('F)NDTT ('F)RT ('F)Static Dynamic RT Yield Strength (ksi)Base Metal Plate F-773-1 (WR)(Transverse)
Base Metal Plate F-773-1 (RW)(Longitudinal) 136.5 112 30 35 12+0-40-10 10 61.7 64.1 107.7 91.4 Weld Metal F-773-2/F-773-3 HAZ Metal F-773-1/F-773-2 SRM HSST Plate 01MY (RW)(Longitudinal) 109 84 136-43-12 57 34-41 14-70-30-50 50 63.3/62.2 85.0 TABLE4-4 Arrangement of Encapsulated Test Specimens by Code Number within the Palo Verde Unit 2 137'apsule'" Compartment Position Compartment Number 18312 18322-1 18332-4 1834D-3 18353 18363-1 Specimen Numbers 1823L 1823P 18243 1826D 1822E 182AE 1824J 1823M 18267 1822M 18224 182AM 18255 1822U 1821U 18278 18213 1823T 18226 1826C 18245 18275 1825Y 1822T 1814J 1811K 18116 18141 1813M 18148 18151 18113 1812L 182KI 182J3 182J2 1844D 18438 1843K 18444 18432 1843Y 18418 1844L 1843T 18411 18456 1843M IBD43 IBD35 IBD52 IBD4P IBD2D IBD3E IBD2M IBD3Y IBD28 183J5 183JY 183J7 1835E 1834U 183AL 1834E 1835P 1835M 183A3 1833E 183AD 18318 18316 1832A 18342 18326 18312 1833Y 1836M 1833J 1837A 1831U 183AT 1831A 1831K 18323 1811M 1811E 18124 1814Y 18145 1815D 1814D 18128 18136 Material IB IXX 182XX 183XX 184XX IBDXX Lower Shell Plate F-773-1 (Longitudinal Orientation)
Lower Shell Plate F-773-1 (Transverse Orientation)
Weld Metal Heat-Affected-Zone Material SRM HSST 01MY Material (Longitudinal Orientation) 4-6 VESSEL 142o VESSEL i3~'ESSEL zaao Sao'OUTLETIIOZELE IIP I O A 4 C E 0 B.8 VESSEL 43 VESSEL aa'ORE SHROUD COBE EUPPOBT BARREL REACTOR VESSEL VESSEL ato'NLET i NOZZLE I I I CORE MIDPLANE\l//I I REACTOR VESSEL VESSEL CAPSULE ASSEMBLY CORE SUPPORT BARREL ENLARGED PLAN VIEW I I I oo ELEVATION VIEW Lock Assembly Wedge Coupling Assembly Charpy and Flux Compartment Ass embly or Charpy, Flux, and Compact Tension Compartment Ass embly Connecting Spacer Temperatu re, Flux, Tension and Charpg Compartment Ass embly Charpy and Ftux Compartment Ass embly or Charpy, Flux, and Compact Tens ion Compa rtment Ass embly Figure 4-2.Typical Palo Verde Unit 2 Surveillance Capsule Assembly Wedge Coupling-fnd Cap I l I l I)1 l)Charpy Impact Specimens Flux Monitor Housing Precracked Charpy and/or Charpy Impact Specimens Connecting Spacer Spacers Rectangul ar Tubing Vledge Coupling-End Cap Figure 4-3.Typical Palo Verde Unit 2 Surveillance Capsule Charpy Impact Compartment Assembly 49 Vfedge Coupling-End Cap 0 Tension Specimens and Tension Spec>men Housing Charpy Impact Specimens Connecting Spacer Flux Monitor Housing Stainless Steel Tubing Threshold Detector Flux Spectrum Monitor Temperature Monitor Temperature Monitor Housing Flux Spectrum Monitor Cadmium Shielded Stainless Steel Tubing Cadmium Shield Threshold Detector Quartz Tubing Vfeight l mv Melting Alloy Chary Impact Specimens Rectangular Tubing Tension Specimens and Tension Specimen Housing Vledge Coupling-End Cap Figure 4-4.Typical Palo Verde Unit 2 Surveillance Capsule Tensile-Monitor Compartment Assembly 4-10 SECTION 5.0 TESTING OF SPECIMENS FROM CAPSULE W137 5.1 Overview The post-irradiation mechanical testing of the Chaipy V-notch impact, precracked Charpy and tensile specimens was performed in the Remote Metalograhpic Facility (RMF)at the Westinghouse Science and Technology Center.Testing was performed in accordance with 10CFR50, Appendices G and EP and ASTM Specification E185-82"'nd Westinghouse Procedure RMF 8402, Revision 2 as modified by Westinghouse RMF Procedure 8102, Revision 1, and 8103, Revision 1.Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in Reference 5.No discrepancies were found.Examination of the four low-melting point 536'F (280'C), 558'F (292'C), 580'F (304'C)and 590'F (310'C)eutectic alloys indicated that the two thermal monitors with melting points of 536'F (280'C)and 558'F (292'C)melted (Figure 5-1).Based on this examination, the maximum temperature to which the test specimens were exposed was less than 580'F (304'C).The Charpy impact tests were performed per ASTM Specification E23-92'" and RMF Procedure 8103, Revision 1, and NSMT Procedure 9306, on a Tinius-Olsen Model 74, 358J machine.The tup (striker)of the Charpy impact test machine is instrumented with a GRC 830-I instrumentation system, feeding information into an IBM XT Computer.With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED).From the load-time curve, the load of general yielding (PG), the time to general yielding (tG), the maximum load (PQ, and the time to maximum load (t)can be determined.
Under some test conditions, a sharp drop in load indicative of fast fracture was observed.The load at which fast fracture was initiated is identified as the fast fracture load (P), and the load at which fast fracture terminated is identified as the arrest load (P).The energy at maximum load (Eg was determined by comparing the energy-time record and the load-time record.The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen.Therefore, the propagation energy for the crack (EP is the difference between the total energy to fracture (EQ and the energy at maximum load (E).5-1 The yield stress (o)was calculated from the three-point bend formula having the following expression:
a~=[PG*L]/[B*(W-a)'C]where: L=distance between the specimen supports in the impact machine B=the width of the specimen measured parallel to the notch W=height of the specimen, measured perpendicularly to the notch a=notch depth The constant C is dependent on the notch flank angle (g), notch root radius (p)and the type of loading (ie.pure bending or three-point bending).In three-point bending, for a Charpy specimen in which g=45'nd p=0.010", Equation 1 is valid with C=1.21.Therefore, (for L=4W), a=[PG*L]/[B'W-a)'1.21]=[3.3*PG*W]/[B*(W-a)'](2)For the Charpy specimen, B=0.394", W=0.394" and a=0.079" Equation 2 then reduces to: ar=33.3*Por (3)6 where ais in units of psi and Pois in units of lbs.The flow stress was calculated fmm the average of the yield and maximum loads, also using the three-point bend formula.Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-92"'.
The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.
In addition to the standard Charpy test specimens, the capsule also contained precracked Charpy specimens.
Testing of the precracked Charpy specimens provides estimates of the dynamic fracture toughness of the irradiated materials contained in the capsule using small specimens, rather than much larger ones used in Fracture Mechanics.
Although the Charpy specimens are too small to allow valid determinations of the fracture toughness, the testing of sub-sized specimens, makes it feasible to test multiple irradiated specimens.
5-2 The precracked Charpy test offers the further advantage of being simple to perform.The test requires an instrumented Charpy impact machine and the ability to adjust the drop height of the impact hammer.The load-time data for each test is recorded using high speed data acquisition equipment.
The tests in this program were conducted using the same system as for standard Charpy specimens, ie., a Tinius-Olsen Model 74 Charpy impact machine equipped with a Dynatup Products Model 830-1 data acquisition system.The adjustable drop height capability is required to allow proper analysis of the test records.The early portion of the test record is dominated by oscillations in the load signal caused by the inertial loading that occurs when the hammer impacts the specimen.The Charpy hammer must be lowered to reduce the inertial effects and to increase the length of the test.In general, the primary points of analysis (general yielding, etc.)should occur at least 100 msec after the initial impact.The load time records must be analyzed to determine fracture toughness values.The initial velocity of the Charpy hammer was determined using the Dynatup instrumentation system.The velocity measurement was then used to interpret the load-time record in terms of load and displacement.
This data was then analyzed to provide an energy versus time curve.The analysis of the instrumented data was performed using the standard Dynatup system software.At low temperatures, the specimens fail in a brittle manner, with no evidence of yielding in the test record.Specimens that failed in a brittle manner were analyzed using standard linear-elastic techniques to determine a dynamic fracture toughness, K.The determination of Krequires only a knowledge of the precrack length, which was determined from post test photos, and the maximum load, which was determined from the test record.At higher temperatures, the test records indicate that general yielding of the specimens occurs prior to failure.Elastic-plastic analysis was required to estimate a dynamic fracture toughness value, K, in the higher temperature specimens.
In small specimens, maximum load generally occurs at the onset of crack growth.The determination of Krequires a knowledge of the energy absorbed in the specimen at maximum.load, and the crack length.The energy calculated by the instrumentation system includes both the energy absorbed in the specimen and the energy absorbed by the elastic deformation of the Charpy system.The total system compliance was determined and the Charpy specimen compliance was calculated to allow correction of the measured energy values.The conected value of energy absorbed at maximum load, Q, was then used to calculate Kaccording to the formula: K=[(2*+*E)/(b*B)]'+
(4)5-3 where, E=Young's Modulus b=Remaining ligament (specimen depth less crack length)B=Specimen thickness The test records were also analyzed to determine the dynamic yield strength, s,.The general formula for the determination of general yielding for a member in three point bending is (in analogy to equation 2): s=K3.3'*W)/(b*B)'](5)where, W=Specimen depth P~=Load at general yielding Three sets of precracked Charpy specimens were contained in the surveillance capsule.These sets included specimens fiom plate F-773-1 (both transverse and longitudinal orientations) and ftom the surveillance weld metal.Tensile tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115)per ASTM Specification E8-91"'nd E21-79(1988)"", and RMF Procedure 8102, Revision 1.All pull rods, grips, and pins were made of Inconel 718.The upper pull rod was connected through a universal joint to improve axiality of loading.The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.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 failure.The extensometer gage length was 1.00 inch.The extensometer is rated as Class B-2 per ASTM E83-92'"'levated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone.All tests were conducted in air.Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperatures.
Chromel-alumel thermocouples were positioned at 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 from room temperature to 550'F (288'C).During the actual testing, the grip temperatures were used to obtain desired specimen temperatures.
Experiments indicated that this method is accurate to+2'F.The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve.The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area.The final diameter and final gage length were determined from post-fracture photographs.
The fracture area used to calculate the fracture stress (true stress at fracture)and percent reduction in area was computed using the final diameter measurement.
5.2 Ch V-Notch Im act Test Results The results of the Charpy V-notch impact tests performed on the various materials contained in the capsule, which was irradiated to 4.071 x 10" n/cm (E>1.0 MeV), are presented in Tables 5-1 through 5-10 and are compared with unirradiated results"'s shown in Figures 5-2 through 5-6.The transition temperature increases and upper shelf energy changes for the surveillance materials are summarized in Table 5-11.Irradiation of the reactor vessel lower shell plate F-773-1 Charpy specimens oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal orientation) to 4.071 x 10" n/cm'E>1.0 MeV)at 550'F (Figure 5-2)resulted in a 30 ft-Ib transition temperature increase of 10'F and a 50 ft-lb transition temperature increase of 25'F.This resulted in an irradiated 30 ft-Ib transition temperature of 10'F and an irradiated 50 ft-lb transition temperature of 60'F (longitudinal orientation).
The average upper shelf energy (USE)of the lower shell plate F-773-1 Charpy specimens (longitudinal orientation) resulted in an energy increase of 6 ft-lb after irradiation to 4.071 x 10" n/cm'E>1.0 MeV)at 550'F.This results in an irradiated average USE of 118 ft-lb (Figure 5-2).Irradiation of the reactor vessel lower shell plate F-773-1 Charpy specimens oriented with the longitudinal axis of the specimen normal to the major rolling direction of the plate (transverse 5-5 orientation) to 4.071 x 10" n/cm (E>1.0 MeV)at 550'F (Figure 5-3)resulted in a 30 ft-lb transition temperature increase of 19'F and a 50 ft-Ib transition temperature increase of 25'F.This results in an irradiated 30 ft-lb transition temperature of 15'F and an irradiated 50 ft-lb transition temperature of 55'F (transverse orientation).
The average USE of the lower shell plate F-773-1 Charpy specimens (transverse orientation) resulted in an average energy decrease of 21.5 ft-lbs after irradiation to 4.071 x 10" n/cm (E>1.0 MeV)at 550'F.This results in an irradiated average USE of 115 ft-lb (Figure 5-3).Irradiation of the surveillance weld metal Charpy specimens to 4.071 x 10" n/cm (E>1.0 MeV)at 550'F (Figure 5%)resulted in a 30 and 50 ft-lb transition temperature increase of 15'F.This results in an irradiated 30 ft-lb transition temperature of-28'F and an irradiated 50 ft-lb transition temperature of 4'F.The average USE of the surveillance weld metal resulted in an energy decrease of 1 ft-lb after irradiation to 4.071 x 10" n/cm (E>1.0 MeV)at 550'F.This resulted in an irradiated average USE of 108 ft-lb (Figure 5-4).Irradiation of the reactor vessel weld HAZ metal Charpy specimens to 4.071 x 10" n/cm (E>1.0 MeV)at 550'F (Figure 5-5)resulted in a 30 ft-lb transition temperature increase of 57'F and a 50 ft-lb transition temperature increase of 18'F.This results in an irradiated 30 ft-lb transition temperature of 45'F and an irradiated 50 ft-lb transition temperature of 75'F.0 The average USE of the weld HAZ metal resulted in an energy increase of 29 ft-lbs after irradiation to 4.071 x 10" n/cm'E>1.0 MeV)at 550'F.This resulted in an irradiated average USE of 113 ft-lb (Figure 5-5).Irradiation of the HSST plate 01MY Chatpy specimens to 4.071 x 10" n/cm (E>1.0 MeV)at 550'F (Figure 5-6)resulted in a 30 ft-lb transition temperature increase of 117'F and a 50 ft-lb transition temperature increase of 151'F.This results in an irradiated 30 ft-lb transition temperature of 120'F and an irradiated 50 ft-lb transition temperature of 185'F.The average USE of the HSST plate 01MY Charpy specimens resulted in an energy decrease of 31 ft-lbs after irradiation to 4.071 x 10" n/cm'E>1.0 MeV)at 550'F.This resulted in an irradiated average USE of 105 ft-lb (Figure 5-6)5-6 The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-7 through 5-11 and show an increasingly ductile or tougher appearance with increasing test temperature.
A comparison of the 30 ft-lb transition temperature increases and upper shelf energy decreases for the various Palo Verde Unit 2 surveillance materials with predicted values using the methods of NRC Regulatory Guide 1.99, Revision 2'" is presented in Table 5-12 and led to the following conclusions:
o The 30 ft-lb transition temperature increases for the surveillance weld metal and the longitudinally oriented lower shell plate F-773-1 Charpy test results are less than the Regulatory Guide 1.99, Revision 2, predictions.
o The 30 ft-lb transition temperature increase and average upper shelf energy decrease of transversely oriented lower shell plate F-773-1 Charpy test results are in good agreement with Regulatory Guide 1.99, Revision 2, predictions.
o The measured average USE decrease of the weld metal and lower shell plate F-773-1 longitudinally oriented Charpy test results are less than the Regulatory Guide 1.99, Revision 2, predictions.
The load-time records for the Charpy impact tests are provided in Figures A-2 through A-31 in Appendix A.5.3 Precracked Cha S cimen Test Results The results of the precracked Charpy specimen tests are reported in Tables 5-13 through 5-15 and in Figures 5-12 through 5-14.Data for the unirradiated materials was reported in the original ABB-Combustion Engineering reporP'.The unirradiated and irradiated precrack Charpy data are both included in Figures 5-12 through 5-14.The data is plotted on the basis of the RT>>r value to eliminate the effects of the relatively small shifts in the ductile-to-brittle temperature.
These figures indicate good agreement between the irradiated and unirradiated test results.The K, reference (Kz)curve is also shown in Figures 5-12 through 5-14.The data is bounded by the Kii, curve, which should provide a lower bound estimate for the fracture toughness.
The low temperature 5-7 unirradiated and irradiated data, which was determined using linear-elastic procedures (K,g, approaches the bounding curve.The low fracture toughness values may be attributed to the sub-sized specimens, which do not meet standard validity requirements.
The load-time records and comparisons of data for the unirradiated and irradiated precracked Charpy specimen tests are provided in Figures A-32 through A-56 in Appendix A.5.4 Tension Test Results The results of the tension tests performed on the various materials contained in the capsule irradiated to 4.071 x 10" n/cm (E>1.0 MeV)are presented in Table 5-16 and ate compared with unirradiated resultP as shown in Figures 5-18 and 5-19.The results of the tension tests performed on the lower shell plate F-773-1 (transverse orientation) indicated that irradiation to 4.071 x 10" n/cm (E>1.0 MeV)at 550'F caused a 0 to 4 ksi increase in the 0.2 percent offset yield strength and a 0 to 3 ksi increase in the ultimate tensile strength when compared to unirradiated data"'Figure 5-18).The results of the tension tests performed on the surveillance weld metal indicated that irradiation to 4.071 x 10" n/cm'E>1.0 MeV)at 550'F caused a 5 to 8 ksi increase in the 0.2 percent offset yield strength and a 4 to 5 ksi increase in the ultimate tensile strength when compared to unirradiated data"'Figure 5-19).The fractured tension specimens for the lower shell plate F-773-1 material are shown in Figure 5-20, while the fractured specimens for the surveillance weld metal are shown in Figure 5-21;The engineering Stress-strain curves for the tension tests are shown in Figures 5-22 through 5-25.5-8 TABLE 5-1 Chatpy V-notch Data for the Palo Verde Unit 2 Lower Shell Plate F-773-1 Irradiated at 550'F to a Fluence of 4.071 X 10" n/cm'E)1.0 MeV)(Longitudinal Orientation)
Sample Number Temperature
('C)Impact Energy (ft-lb)Lateral Expansion Shear (mils)(mm)(%)1B11E 1B124 1BI IM-5 10 25-21-12 28 27 47 38 37 30 26 43 0.76 0.66 1.09 10 20 25 1B15D 50 10 33 45 33 0.84 30 1B 14Y 1B 136'1B145 1B14D 1B12B 75 115 150 250 24 46 66 93 121 58 82 92 108 127 79 146 172 54 70 77 79 94 1.37 1.78 1.96 2.01 2.39 50 70 85 5-9 TABLE 5-2 Charpy V-notch Data for the Palo Verde Unit 2 Lower Shell Plate F-773-1 Irradiated at 550'F to a Fluence of 4.071 X 10>s n/cd (E>1 0 MeV)(Transverse Orientation)
Sample Number Temperature Impact Energy ('C)(ft-lb)(mils)(mm)(%)Lateral Expansion Shear 1B255 1B23L 1B2AE 1B224 1B267 1B23P 1B2AM 1B26D 1B21U 1B22U 1B22E IB24J 1B243 1B22M 1B23M 15 35 50 65 75 85 125 150 250-32-18 10 18 29 38 52 66 93 121 149 16 24 30 62 42 58 43 63 67 73 105 116 115 115 22 33 41 84 57 58 79 58 85 91 99 142 157 156 156 15 24 29 54 38 45 50 41 55 57 67 89 83 90 92 0.38 0.61 0.74 1.37 0.97 1.14 1.27 1.45 1.70 2.26 2.11 2.29 2.34 10 20 30 40 40 50 60 70 95 95 5-10 TABLE 5-3 Chatpy V-notch Data for the Palo Verde Unit 2 Surveillance Weld Metal Irradiated at 550'F to a Fluence of 4.071 X 10" n/cm (E)1.0 MeV)Sample Temperature Impact Energy Lateral Expansion Shear Number ('F)1B3AD ('C)-68 (ft-lb)13 (mils)16 (mm)0.41 10 1B316 1B35E 1B342 1B34U 1B312 1B34E-80-50-30-20-10-62-29-23-18 38 12 28 33 53 50 52 16 38 72 68 38 15 36 33 51 46 0.97 0.38 0.91 0.84 1.30 1.17 15 15 25 30 35 35 1B35P 25 1B 326 50 1B3A3 75 1B3AL 100 1B 35M 125 1B33E 150 1B31B 200 1B32A 250 4 10 24 52 66 93 121 57 88 75 93 92 106 107 112 77 119 102 125 126 145 152 56 78 70 81 85 95 86 95 1.42 1.98 1.78 2.06 2.16 2.41 2.18 2.41 50 70 80 90 95 5-11 TABLE 5P Charpy V-notch Data for the Palo Verde Unit 2 Heat-Affected-Zone (HAZ)Metal Irradiated at 550'F to a Fluence of 4.071 X 10" n/cm'E)1.0 MeV)Sample Number ('F)('C)Temperature Impact Energy (ft-lb)(mils)(mm)Lateral Expansion Shear (%)1B411 1B44D 1B43B 1B43Y 1B432 1B43T 1B444 1B43M IB43K 1B44L 1B456 1B41B-35 25 35 50 60 75 120 155 250-37-18 10 16 24 38 68 93 121 16 20 20 28 18 49 71 86 108 77 133 92 22 27 27 38 66 96 117-146 180 125 16 24 15 29 24 47 59 71 82 70 87 79 0.41 0.61 0.38 0.74 0.61 1.19 1.50 1.80 2.08 1.78 2.21 2.01 10 15 20 30 50 65 80 95 5-12 TABLE 5-5 Charpy V-notch Data for the Palo Verde Unit 2 Correlation Monitor Standard Reference Material HSST 01MY Irradiated at 550'F to a Fluence of 4.071 X 10" n/cm (E>1.0 MeV)(Longitudinal Orientation)
Sample Number 1BD3Y Temperature Impact Energy ('F)('C)(ft-lb)'18 Lateral Expansion Shear (mils)(mm)(%)0.20 1BD43 1BD4P 1BD35 1BD2D 1BD52 1BD3E 1BD2M 1BD2B 50 150 275 350 10 38 66 93 93 107 135 177 17 32 45 35 57 88 106 23 43 61 47 77 119 141 20 30 46 32 53 75 84 87 0.51 0.76 1.17 0.81 1.35 1.91 2.13 2.21 15 20 40 35 60 85 100 5-13 TABLE 5-6 Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Lower Shell Plate F-773-1 Irradiated at 550'F to a Fluence of 4.071 X 10" n/cm (E>1.0 MeV)(Longitudinal Orientation)
Normalized Energies (ft-Ib fjn~)Sample No.Test Temp.('F)Charpy Energy En (ft-lb)Chatty En/A Max.EJA Prop.Ep/A Yield Load PGY (Ibs)Time to Yield tm (ij sec)Max.Load Phl (lbs)Time to Max.4<(p sec)Fast Fract.Load PF (lbs)Arrest Load PA (lbs)Yield Stress Gr (ksi)Flow Stress (ksi)1B11E-5 28 225 182 43 3189 0.17 3845 0.48 3839 397 106 117 1B 124 10 27 217 155 62 3096 0.16 3719 0.43 3700 630 103 113 1B 11M 25 1B15D 50 1B14Y 75 47 33 58 378 266 467 278 101 3080 154 112 3019 280 187 2931 0.17 0.16 0.14 3959 0.68 3936 3750 0.42 3740 3969 0.68 3931 902 1276 1707 102 117 100 112 97 115 1B 136 115 82 1B 145 150 92 1B14D 200 108 1B12B 250 127 660 741 870 1023 343 317 2828 272 469 2743 245 624 2427 322 700 2499 0.15 0.19 0.14 0.14 3936 0.82 3643 3724 0.72 2897 3569 0.68 3731 0.82 2115 1628 94 112 91 107 81 100 83'03*Fully ductile fracture.
TABLE 5-7 Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Lower Shell Plate F-773-1 Irradiated at 550'F to a Fluence of 4.071 X 10" n/cm (E)1.0 MeV)(Transverse Orientation)
Normalized Energies (ft-lb/in')
Sample No.Test Temp.('F)Charpy Energy ED (ft-Ib)Charpy EG/A Max.EJA Prop.EF/A Yield Load PGY (lbs)Time to Yield tGY (psec)Max.Load Phk (lbs)Time to Max.(p sec)Fast Fract.Load PF (lbs)Arrest Load (lbs)Yield Stress aY (ksi)Flow Stress (ksi)1B255-25 1B23L 0 1B2AE 15 1B224 25 1B267 35 16 24 30 62 42 129 193 242 499 338 89 40 3350 147 46 3342 165 77 3136 359 104 3083 252 86 2987 0.17 0.17 0.16 0.16 0.16 3648 0.28 3632 3869 0.40 3865 3877 OA4 3864 4094 0.82 3988 3881 0.63 3872 141 241 1116 1022 99 114 111 116 120 104 , 116 102 119 1B23P 50 43 346.217 130 2931 0.17 3805 0.57 3792 1538 97 112 1B2AM 65 1B26D 75 1B21U 85 58 43 63 467 346 507 351 116 2948 197 149 2951 357 151 2922 0.16 0.16 0.14 4036 0.82 4023 3873 0.52 3867 4062 0.82 4053 1511 1900 2286 98 97 113 116 98 116 IB22U 100 67 1B22E 125 73 1B24J 150 105 540 588 845 279 261 2914 261 327 2639 260 586 2615 0.16 0.14 0.15 3986 0.69 3893 3704 0.69 3663 3693 0.69 2660 2155 2285 1858 97 88 87 115 105 105 1B243 200 116 1B22M 250 115 934 926 313 621 2606 247 679 2371 0.16 0.14 3774 0.80 3572 0.69 79 99 87 106 1B23M 300 115 926 279 647 2242 0.15 3393 0.80 74 94~Fully ductile fracture.
TABLE 5-8 Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Surveillance Weld Metal Irradiated at 550'F to a Fluence of 4.071 X 10" n/cm g,>1.0 MeV)Normalized Energies (ft-Ib/jn~)
Sample No.Test Temp.('F)Charpy Energy ED (ft-Ib)Charpy ED/A Max.E JA Prop.EF/A Yield Load PGY (lbs)Time to Yield tGY (psec)Max.Load Phk (lbs)Time to Max.th~(p sec)Fast Fract.Load PF (lbs)Arrest Load (Ibs)Yield Stress aY (ksi)Flow Stress (ksi)1B316-80 1B3AD-90 13 38 105 63 42 306 210 96 3630 3237 0.17 3681 0.21 0.17 3810 0.54 3668 3804 561 121 121 1644 108 117 1B35E 1B342-50-30 12 28 97 43 225 149 54 76 3142 3268 0.14 3463 0.18 0.16 3767 0.41 3457 3748 415 104 110 1384 109 117 1B34U-20 266 172 94 3301 0.18 3862 0.46 3846 1556 110 119 1B312-10 1B34E 0 1B35P 25 1B326 50 1B3A3 75 1B3AL 100 1B35M 125 1B33E 150 53 50 57 88 75 92 93 106 427 268 159 403 269 134 459 261 198 709 258 451 604 274 330 741 275 466 749 269 480 854 312 541 2961 3169 3042 2758 2974 2922 2820 2586 0.15 3734 0.67 0.16 3878 0.65 0.16 3759 0.65 0.14 3620 0.68 0.16 3847 0.67 0.15 3816 0.69 0.15 3751 0.69 0.14 3535 0.82 3673 3843 3746 3277 3443 97 112 94 109 86 102 1915 98 111 2864 105.117 2622 101 113 2550 92 106 2246 99 113 1B31B 200 1B32A 250 107 112 862 301 561 902 293 609 2400 2318 0.16 3451 0.82 0.16 3353 0.83 80 77 97 94 Fully ductile fracture.
TABLE 5-9 Instrumented Charpy Impact Test Results for the Palo Vede Unit 2 Surveillance Heat-Affected-Zone (HAZ)Metal Irradiated at 550'F to a Fluence of 4.071 X 10" n/cm'E>1.0 MeV)Normalized Energies (ft-lb/in')
Sample No.Test Temp.('F)Charpy Energy Charpy Ep EJA (ft-lb)Max.EJA Prop.Ep/A Yield Load PGY (lbs)Time to Yield tm (p sec)Max.Load Phf (lbs)Time to Max.(psec)Fast Fract.Load PF (lbs)Arrest Load (lbs)Yield Stress Gy (ksi)Flow Stress (ksi)1B411 IB44D 1B43B 1B43Y 1B432 1B43T 1B444 1B43M IB43K 1B44L 1B456 1B41B-35 25 35 50 60 75 100 120 155 200 250.16 129 89 20 161 105 20 161 105 28 225 155 18 145 69 49 395 270 71 572 284 86 692 284 108 870 377 77 620 255 133 1071 345 92 741 249 40 56 56 70 76 288 409 492 365 726 492 3356 0.17 3645 3331 0.16 3732 3136 0.16 3491 3248 0.17 3797 3057 0.16 3250 3122 0.16 3919 3041 0.14 4024 3057 0.16 4014 2997 0.25 4086 2807 0.15 3720 2836 0.16 3833 2650 0.17 3554 0.28 0.31 0.32 0.42 0.25 0.65 0.67 0.69 0.94 0.65 0.86 0.68 3641 170 111 3729 536 111 3487 928 104 3768 1136 108 3244 1287 102 3909 1724 104 3905 2022 101 3189 1540 102 3006 1940 100 2775 1772 93 94 88 , 116 117 110 117 105 117 117 117 118 108 103*Fully ductile fracture TABLE 5-10 Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Surveillance Standard Reference Material HSST 01MY Irradiated at 550'F to a Fluence of 4.071 X 10" n/cm (E>1.0 MeV)(Longitudinal Orientation)
Normalized Energies (ft-lb fin~)Sample No.Test Temp.(F)Charpy Energy ED (ft-lb)Charpy ED/A Max.E JA Prop.EF/A Yield Load PGY (lbs)Time to Yield tGY (psec)Max.Load Phf (lbs)Time to Max.thi (psec)Fast FlacL Load PF (lbs)Arrest Load PA (lbs)Yield Stress<r (ksi)Flow Stress (ksi)1BD3Y 0 1BD43 50 1BD4P 100 17 32 1BD35 150 45 56 137 258 362 29 96 195 281 27 3186 41 3405 62 3263 81 3096 0.14 0.17 0.17 0.16 3289 0.15 3282 3783 0.29 3777 4115 0.49 4096 4172 0.65 4156 176 521 1112 106 108 113 119 108 123 103 121 1BD3E 225 88 1BD2D 200 35 1BD52 200 57 282 459 709 208 74 3090 278 181 2965 289 419 2966 0.16 0.15 0.16 4095 0.52 4085 4115 0.66 3994 4145 0.68 3485 845 1804 2521 103 119 98 118 99 118 1BD2M 275 104 837 278 559 2900 0.17 4050 0.68 96 115 1BD2B 350 106 854 267 587 2724 0.16 3898 0.67 90 110*Fully ductile fracture.
TABLE 5-11 Effect of 550'F Irradiation to 4.071 X 10" n/cm'E>1.0 MeV)on the Notch Toughness Properties of the Palo Verde Unit 2 Reactor Vessel Surveillance Materials Material Average 30 (ft-lb)'ransition Temperature
('F)Average 35 mil Lateral'xpansion Temperature
('F)Average 50 ft-lb'ransition Temperature
('F)Average Energy Absorption
" at Full Shear (ft-lb)Unirradiated Irradiated bT Unirradiated Irradiated hT Unirradiated Irradiated Unirradiated Irradiated b.Plate F-773-1 (longitudinal)
Plate F-773-1 (transverse) 4 10 15 10 19 12 25 13 20 35 30 60 55 25 25 112 136.5 1.18 115.-21.5 Weld Metal HAZ Metal SRM 01MY-43-12-28 45 15 57 120 117-41 14-33 45 155 43 141-11 57 34 75 185 15 18 151 109 84 136 108 113 105+29-31 (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 5-5)
TABLE 5-12 Comparison of the Palo Verde Unit 2 Surveillance Material 30 ft-Ib Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions Material Fluence (E)1.0 MeV)(X 10" n/cm')Predicted" (F)Measured ('F)30 ft-lb Transition Temperature Shift Predicted" (%)Measured (%)Upper Shelf Energy Decrease Lower Shell Plate F-773-1 (longitudinal)
Lower Shell Plate F-773-1 (transverse)
Weld Metal HAZ Metal SRM HSST Plate 01MY (longitudinal) 4.071 4.071 4.071 4.071 4.071 19.5 19.5 31.5 10 19 15 57 117 15.5 15.5 17 16 23 (a)Based on Regulatory Guide 1.99, Revision 2 methodology using wt.%values of Cu and Ni from Reference 2.
TABLE 5-13 Precracked Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Lower Shell Plate F-773-1 Irradiated at 550'F to a Fluence of 4.071 X 10" n/cm (E>1.0 MeV)(Longitudinal Orientation)
Sample No.Test Temp.('F)Specimem Comp.Initial Velocity (in/sec)Tllne to Yield (psec)Yield Load (lbs)Time to Max.(psec)Max.Load (Ibs)Energy at Initiation (in-lbs)Available Energy (in-lbs)Crack Length (in.)K,~(ksi l in)K,~(ksi l in)Yield Stress (ksi)1B151 1B14B 1B116 1B14J 1B113 1B11K 1B13M 1B12L 1B141 25 50 71 76 100 150 200 250 45.6 47.3 45.6 66A 51.6 50.1 51.6 49.6 47.8 57.4 57.5 71.7 71.6 71.6 90.6 90.9 90.8 90.8 157 150 135 140 145 105 170 1047 180 1172 170 1235 1040 300 1066 1120 860 1285 1200 560 1381 1100 1035 1341 1170 1135 1499 940 1105 1375 15.5 64.4 54.2 101.6 122.1 108.0 256 256 399 398 398 637 642 640 640 0.176 OA47 38.2 0.180 0.457 44.0 0.176 0.447 45.0 0.213 0.541 0.189 0.480 0.186 0.472 0.189 0.480 0.185 0.470 0.181 0.459 102.9 210.9 187.4 263.6 285.0 266.0 72.2 83.9 83.3 104.8 87.9 91.5 86.4 88.4 68.4 TABLE 5-14 Precracked Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Lower Shell Plate F-773-1 Irradiated at 550'F to a Fluence of 4.071 X 10" n/cm (E)1.0 MeV)(Transverse Orientation)
Sample No.Test Temp.('F)Specimem Comp.Initial Velocity (in/sec)Time to Yield (psec)Yield Load (ibs)Time to Max.(p sec)Max.Load (lbs)Energy at Initiation (in-lbs)Available Energy (in-lbs)Crack Length (in.)Kr~(ksi)in)Kr~(ksi l in)Yield Stress (ksi)1B213 1B26C 1B23T 1B27B 25 50 76 100 45.2 50.6 44.4 51.6 57.4 71.9 71.5 90.7 155 110 1200 1060 225 1395 185 1306 455 1354 295 1186 32.1 22.6 401 397 639 0.175 0.444 50.4 0.187 0.475 51.8 0.173 0.439 0.189 0.480 137.1 118.1 95.0 94.7 81.1 83.2 1B22T 1B245 1B226 150 250 50.1 54.2 49.2 90.6 150 90.5 112 90.6 113 1260 1080 880 560 1434 54.4 950 1365 92.1 1145 1398 112.4 638 636 637 0.186 0.472 0.194 0.492 0.184 0.467 185.7 254.2 271.6 96.1 89.1 65.9 TABLE 5-15 Precracked Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 Surveillance Weld Metal Irradiated at 550'F to a Fluence of 4.071 X 10" n/cm (E>1.0 MeV)Sample No.1B3AT 1B31U 1B323 Test Temp.('F)-50-25 Specimem Comp.52.6 48.2 50.1 Initial Velocity (in/sec)90.9 57.5 71.5 Time to Yield (p sec)Yield Load (lbs)Time to Max.(psec)115 175 155 Max.Load (lbs)995 1085 1215 Energy at Initiation (in-lbs)Available Energy (in-lbs)642 257 397 Crack Length (in.)0.191 0.182 0.186 Kra (ksi1in)0.485 40.7 0.462 41.4 0.472 47.8 Kr~(ksi l in)Yield Stress (ksi)79.3 75.6 92.3 1B33J 50 1B31A 71 1B33Y 100 1B36M 125 1B 31K 150 1B37A 200 47.3 49.2 51.6 50.6 50.1 71.5 150 1240 71.8 150 1240 90.5 140 1200 203.0 60 950 90.9 110 1000 90.6 110 1000 550 1035 720 300 860 715 1390 42.0 1456 1286 1223 86.2 70.0 56.4 1319 67.6 1267 80.0 397 400 636 3201 642 637 0.180 0.174 0.184 0.189 0.187 0.186 0.457 0.442 0.467 0.480 0.475 0.472 163.5 236.4 215.5 191.7 233.8 211.8 89.4 85.4 89.8 74.6 77.0 76.3 TABLE 5-16 Tensile Properties of the Palo Verde Unit 2 Reactor Vessel Surveillance Materials Irradiated at 550'F to 4.071 X 10" n/cm'E)1.0 MeV)Material Plate F-773-1 (transverse)
Plate F-773-1 (transverse)
Plate F-773-1 (transverse)
Weld Metal Weld Metal Weld Metal Sample Number 1B2J2 1B2K1 1B2J3 1B3J7 1B3JY 1B3J5 Test Temp.('F)75 200 550 125 550 0.2%Yield Strength (ksi)63.9 63.2 54.0 73.3 70.3 62.6 Ultimate Strength (ksi)84.5 80.5 81.5 90.7 83.5 82.5 Fracture Load (kip)3.65 2.55 2.80 2.89 2.75 2.75 Fracture Stress (ksi)206.5 126.8 130.9 194.6 141.2 160.9 Fracture Strength (ksi)74.4 51.9 57.0 58.9 56.0 56.0 Uniform Elongation
(%)13.5 11.4 12.3 13.5 10.8 9.9 Total Elongation
(%)29.1 25.4 24.8 27.9 23.4 21.0 Reduction in Area (%)64 59 56 70 60 65 Figure 5-1 Palo Verde Unit 2 Capsule W-137 Thermal Monitors 5-25 (OC)-150-100-50 0 50 100 150 200 250 100 80 60 cn 40 20 2~0~0 100 80 Z 60 OC 40 20 0 BF 2,5 2.0 1.5 LO 0,5 160 140 120 100 80 60 LLI 40 20 0 8'F 200 160 120 80 40-200-100 0 100 200 300 400 500 TEMPERATURE
('F)0 MRMIAIE9~DMIIATB Ci50 0, FLlBKE 407i x 10 n/cn~K>10 HeV)Figure 5-2 Charpy V-Notch Impact Properties for Palo Vedre Unit 2 Reactor Vessel Lower Shell Plate F-773-1 (Longitudinal Ori'entation) 5-26 (C)-150-100-50 0 50 100 150 200 250 100 P 80 60 40 20 100 80 60>C 40 20 0 160 140 120 100 80 0 w 60 4J 40 2Q 0 20'F (0~0 2fF-19F o 2,5 2,0 1,5 1,0 0.5 200 160 120 80 0-200-100 0 100 200 300 400 500 TEMPERATURE
('F)0 MRRASIATE9
~IRMIATEO 659 0, flUENCE Cj71 x 1tj n/cn2 K>10 Ht.V)Figure 5-3 Charpy V-Notch Impact Properties for Palo Verde Unit 2 Reactor Vessel Lower Shell Plate F-773-1 (Transverse Orientation) 5-27 (C)-150-100-50 0 50 100 150 200 250 100 80~60@40 100 80 60 40~~20 0 160 140 120 100 80 0 5 60 LJ 40 0 8'F 0~2,5 2,0 1.5 1,0 0,5 200 160 120 80 40-200-100 0 100 200 300 400 500 TEHPERATURE
('F)0 MRRAIIIATB
~EMIATED Ci50 Fi, FLUEHCE 487l x 10 n/m~K>N ReV)Figure 5-4 Charpy V-Notch Impact Properties for Palo Verde Unit 2 Reactor Vessel Surveillance Weld Metal (F-773-2/F-773-3) 5-28 (C)-150-100-50 0 50 100 150 200 P50 100 80 60 40 20 2+0 0 0 100 80 60 OC 40~~20 0 160 140 120 100 80 60 LaJ 40 20 o o$3 F 00~~0 4>0 0 0 0 sr 0 0 2,5 2.0 1,5 1.0~0,5 200 160 120 80 40-200-100 0 100 200 300 400 500 TEMPERATURE
('F)0 MRRA9IATE9
~NA9IAIE9 650 F>, FUBKE 4971 x 19 n/cn2 tE>19 HeV)Figure 5-5 Chaq>y V-Notch Impact Properties for Palo Verde Unit 2 Reactor Vessel Weld Heat-Affected-Zone Metal 5-29 (C)-150-100-50 0 50 100 150 200 250 100 80 60~40 20 2+100 N 80 2 60 OC 40 20 0 160 140 120 100 80 60 5 40 20 0 Ill F tst F 2.5 2.0 1.5 1,0 0,5 200 160 120 80 40-200"100 0 100 200 300 400 500 TEMPERATURE (F)0 IIIRRA9IATEil
~IRMIATE9 55tl F), FlUDttE 4)71 x ill a/cn2 6>Ul HeV)Figure 5-6 Chepy V-Notch Impact Properties for Palo Verde Unit 2 SRM HSST 01MY (Longitudinal Orientation) 5-30 L.',1B15D;" 1B14Y t;.-1B136, l I f 4~'y'B145.1B14D 1B128 Figure 5-7 Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Reactor Vessel Lower Shell Plate F-773-1.(Longitudinal Orientation) 5-31 I 18255'823L'82AE,,',;;;18224',18267 i.f rg~g 0 1823P 182AM 1826D 1821U 1822U ,(1$C~Q~'*t g!,"-"~A g>>g W$~-P j W 1B22E 1824J 18243 1822M 1823M I Figure 5-8 Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Reactor Vessel Lower Shell Plate F-773-1 (Transverse Orientation) 5-32 183AD 18316 1835E 18342 1834U 1B312 1834E 1835P 18326 183A3 J i'I 1B3AL 1835M 1833E 18318 1832A Figure 5-9 Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Reactor Vessel Surveillance Weld Metal 5-33 t:18411,.f F." 1844D, 1B438 1843Y k F 18432.1843T'8444 1843M I t f*~1843K 1844L', 18456 1B41B Figure 5-10 Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Reactor Vessel Heat-Affected-Zone Metal 5-34 1BD3Y 1BD43 i 1BD4P'tt 1 P 1BD35 1BD2D 1BD52 , lBD3E 1BD2M 1BD2B Figure 5-11 Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 SRM HSST 01MY (Longitudinal Orientation) 5-35 300 200 P 150 hC 100 L 0 KIR Unirradiated tmtdiated-100 T-RTNDT (F)100 200 Figure 5-12 Comparison of Unirradiated and Irradiated Dynamic Fracture Toughness Values Determined by Testing of Precracked Charpy Specimens from Lower Shell Plate F-773-1 (Longitudinal Orientation) 300 200 KIR 100 UnirradiaIed.
Irradiated
-100 T-RTNDT (F)100 200 Figure 5-13 Comparison of Unirradiated and Irradiated Dynamic Fracture Toughness Values Determined by Testing of Precracked Charpy Specimens from Lower Shell Plate F-773-1 (Transverse Orientation) 300 200 F g 150~k K IR~UnirradiaIed Irradiated
-100 T-RTNDT (F)200 Figure 5-14 Comparison of Unirradiated and Irradiated Dynamic Fracture Toughness Values Determined by Testing of Precracked Charpy Specimens from the Palo Verde Unit 2 Surveillance Weld Metal
.1B151'1B14B 18116 1 ,kr>>I t 1B14X 1B113 1BllK~*l F)i\~~',1B13M',,'.,',.'":.
'8126 1B141 Figure 5-15 Precracked Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Lower Shell Plate F-773-1 (Longitudinal Orientation) 5-39 IB26C*1B23T!1B27B I g~*'I 1B22T'B245'B226'igure 5-16 Precracked Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Lower Shell Plate F-773-1 (Transverse Orientation) 5%0 183AT 1831U 18323'1833J'I I 1831A'1833Y 1831K',~';-f I 1837A Figure 5-17 Precracked Chatpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Reactor Vessel Surveillance Weld Metal 5%1 120 110 100 90 g 80 cn 70 60 50 40 80 70 60 8 50 0~40 50 100 ('C)150 200 250 300 800 700 600 500 400 300 20 10 100 200 300 400 TENPERATURE
('F)500 600 4 0 lNNMIATE9<~IRM1ATEO AT 5%'f FllKNCE 4)71 x 10ln/cn~K>10 Rev)Figure 5-18 Tensile Properties for Palo Verde Unit 2 Reactor Vessel Lower Shell Plate F-773-1 (Transverse Orientation) 5<2 120 110 100 90 80 cn 70 60 50 40 80 ('C>50 100 150 200 250 300 800 700 600 500 400 300 70 60 8 50 0~40~a30 10 100 200 300 400 TEHPERATURE
('F)500 600 4 Q UNIRRQIATE3 4~IRM)IATE9 AT 5SO'F, FLUENCE C)71 x Io tv'cnR (E>19 HeV)Figure 5-19 Tensile Properties for Palo Verde Unit 2 Reactor Vessel Surveillance Weld Metal 5<3 Specimen 1B2J2 75'F 4 ,n Specimen 1B2K1 200'F Qg~.,i, r C C Specimen 1B2J3 550 F Figure 5-20 Fractured Tensile Specimens from Palo Verde Unit 2 Reactor Vessel Lower Shell Plate F-773-l (Transverse Orientation)
Specimen 1B3J7 5'F Specimen 1B3JY 125'Specimen 1B3J5 550'Figure 5-21 Fractured Tensile Specimens from Palo Verde Unit 2 Reactor Vessel Surveillance Weld Metal 5-45 100.00 90.00 80.00 70.00 80.00 50.00 s 40.00 30.00 20.00 10.00 0.00 0.0 STRESS-STRAIN CURVE PALO VERDE UNIT 2 0.10 0.20 STRAIN, IN/IN 1B 2J2 75 F 0.30 100.00 90.00 80.00 70.00 CO eO.00 CO 50.00 40.00 30.00 20.00 10.00 STRESS-STRAIN CURVE PALO VERDE UNIT 2 1B 2K1 200 F 0.00 0.0 0.10 0.20 STRAIN, IN/IN 0.30 Figure 5-22 Engineering Stress-Strain Curves for Lower Shell Plate F-773-1 Tensile Specimens 1B2J2 and 1B2K1 (Transverse Orientation) 546 100.00 90.00 80.00 70.00 CO 8O.OO V)50.00 CC 40.00 30.00 20.00 10.00 STRESS-STRAIN CURVE, PALO VERDE UNIT 2 1B 2J3 550 F 0.00 0.0 0.10 0.20 STRAIN, IN/IN 0.30 Figure 5-23 Engineering Stress-Strain Curve for Lower Shell Plate F-773-1 Tensile Specimen 1B2J3 (Transverse Orientation) 5<7 100.00 90.00 80.00 70.00 8O.OO 50.00 IM 40.00 30.00 20.00 10.00 STRESS-STRAIN CURVE PALO VERDE UNIT 2 1B3J7 0.00 0.00 0.10 0.20 STRAIN, IN/IN 0.30 100.00 90.00 80.00 70.00 8O.OO CO 50.00 4 40.00 30.00 20.00 10.00 0.00 0.0 STRESS-STRAIN CURVE PALO VERDE UNIT 2 0.10 STRAIN, IN/IN 1B 3JY 125 F 0.20 Figure 5-24 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens 1B3J7 and 1B3JY 548 100.00 90.00 80.00 70.00 CO eo.oo CO 50.00 (X 40.00 30.00 20.00 10.00 0.00 0.0 STRESS-STRAIN CuRVE PALO VERDE UNIT 2 0.10 STRAIN, IN/IN 1B3J5 550 F 0.20 Figure 5-25 Engineering Stress-Strain Curve for Weld Metal Tensile Specimen 1B3J5 5<9
SECTION 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6.1 Introduction Knowledge of the neutron environment within the reactor pressure vessel and surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons.First, in order to interpret the neutron radiation induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence)to which the test specimens were exposed must be known.Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the pressure vessel and that experienced by the test specimens.
The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules.The latter information is generally derived solely from 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 years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall.Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853,"Analysis and Interpretation of Light Water Reactor Surveillance Results,"<'
recommends reporting displacements per iron atom (dpa)along with fluence (E>1.0 MeV)to provide a data base for future reference.
The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693,"Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom."t"l The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99,"Radiation Damage to Reactor Vessel Materials." 6-1 This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance capsule%137, withdrawn at the end of the fourth fuel cycle.This analysis is based on current state-of-the-art methodology and nuclear data, and is carried out in accordance with applicable ASTM standards"~~"t.The results provide a consistent fluence evaluation for use in determining the material properties of the Palo Verde Unit 2 reactor vessel.In the dosimetry evaluation, fast neutron exposure parameters in terms of neutron fluence (E>1.0 MeV), neutron fluence (E>0.1 MeV), and iron atom displacements (dpa)are established for the capsule irradiation history.The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel wall.Also, uncertainties associated with the derived exposure parameters at the surveillance capsules and with the projected exposure of the pressure vessel are provided.6.2 Discrete Ordinates Anal sis A plan view of the reactor geometry at the core midplane is shown in Figure 4-1.Six irradiation capsules attached to the reactor vessel wall are included in the reactor design to constitute the reactor vessel surveillance program.The capsules are located at azimuthal angles of 38', 43', 137', 142', 230', and 310'elative to the core cardinal axis as shown in Figure 4-1.A view of a surveillance capsule shown in Figure 4-2.The stainless steel specimen containers are 1.5 by 0.75-inch and approximately 96 inches in height.The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 8 feet of the 12.5 foot high reactor core.From a neutronic standpoint, the surveillance 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 spectrum in the region near the location of each capsule.In order to determine the neutron environment at the test specimen location, the capsules themselves must therefore be included in the analytical model.A plan view of the 1/8 core model is shown in Figure 6-1.In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, two distinct sets of transport calculations were carried out.The first set, a computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions 6-2 ,
throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters Q(E>1.0 MeV), gE>0.1 MeV), and dpa/sec)through the vessel wall.The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsules as well as for the determination of exposure parameter ratios;i.e.,[dpa/sec]/[gE
>1.0 MeV)], within the pressure vessel geometry.The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel wall;i.e., the 1/4T, 1/2T, and 3/4T locations.
As shown in Figure 4-1, Palo Verde has four 45 degree octants with no surveillance capsules, two with one surveillance capsule at 40', and two with two surveillance capsules at 38'nd 43'.The forward calculational model is for the octant geometry with two surveillance capsules.Since the capsules are located adjacent to the vessel wall near the azimuthal maximum flux points, it was necessary to calculate the maximum vessel exposure using a second model with no surveillance capsules present.A comparison of the flux level at the reactor vessel inner radius with and without the surveillance capsules present at 38 and 43 degrees is shown in Figure 6-2.The calculation indicates that the maximum vessel fluence occurs near an angle of 40 degrees.The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux, gE>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.These calculations used separate models for each of the three octant types as appropriate to determine the fluence in each surveillance capsule position and at vessel positions without capsules.The source importance functions generated from these adjoint analyses provided the basis for all absolute exposure calculations and comparison with measurement.
These importance functions, when combined with fuel cycle speciflic 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 cycle speciiflic neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core but also accounted for the effects of varying neutron yield per fission and fission spectrum introduced by the build-up of plutonium as the burnup of individual fuel assemblies increased.
6-3 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: 0 1-Evaluate neutron dosimetry obtained 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 summarized in Figure 6-1 was carried out in R,O geometry using the DOT two-dimensional discrete ordinates code'-'l and the SAILOR cross-section libraryt't.The SAILOR library is a 47 energy group ENDF/B-IV based data set produced specifically for light water reactor applications.
In these analyses anisotropic scattering was treated with a P~expansion of the scattering cross-sections and the angular discretization was modeled with an S8 order of angular quadrature.
The core power distribution utilized in the forward transport calculations was taken as an average of the first four cycles of operation for Palo Verde Unit 2.The neutron spectrum used was based on the burnup of the outer assemblies and utilized ENDF/B-V fission spectra for the contributing uranium and plutonium isotopes.The fuel power distributions were supplied by the Palo Verde staff in the form of beginning-of-cycle and end-of-cycle fuel pin and assembly burnups, and axial power shapes.All adjoint calculations were also carried out using an S8 order of angular quadrature and the P~cross-section approximation from the SAILOR library.Adjoint source locations were chosen at four azimuthal locations along the pressure vessel inner radius (0, 15, 30, 40, and 45 degrees)as well as at the geometric center of each surveillance capsule.Again, these calculations were run in R,O geometry to provide neutron source distribution importance functions for the exposure parameter of interest, in this case gE>1.0 MeV).Haying the importance functions and appropriate core source distributions, the response of interest is calculated as: 6-4 where: rR(r,O)I(r',O',E)
=S(r',O',E)
=gE>1.0 MeV)at radius r and azimuthal angle O.Adjoint source importance function at radius r', azimuthal angle O', and neutron source energy E for the flux (E>1 MeV)at location r, O.Neutron source strength at core location r',O'nd energy E.Although the adjoint importance functions used in this analysis were based on a response function defined by the threshold neutron flux gE>1.0 MeV), prior calculations
'ave shown that, while the variation in fuel loading patterns significantly impacts both the magnitude and spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order.Thus, for a given location the ratio of[dpa/sec]/[ATE
>1.0 MeV)]is insensitive to changing core source distributions.
In the application of these adjoint importance functions to the Palo Verde Unit 2 reactor, therefore, the iron atom displacement rates (dpa/sec)and the neutron flux gE>0.1 MeV)were computed on a cycle specific basis by using[dpa/sec]/[ATE
>1.0 MeV)]and[gE>0.1 MeV)]/[gE>1.0 MeV)]ratios from the forward analysis in conjunction with the cycle specific gE>1.0 MeV)solutions from the individual adjoint evaluations.
The reactor core power distributions used in the plant specific adjoint calculations were taken from the fuel cycle design data supplied for the first four operating cycles of Palo Verde Unit 2.Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-S.The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation periods and provide the means to correlate dosimetry results with the corresponding exposure of the pressure vessel wall.In Table 6-1, the calculated exposure parameters
[gE>1.0 MeV), gE>0.1 MeV), and dpa/sec]are given at the geometric center of the three surveillance capsule positions for the plant specific core power distributions.
The plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis.Similar data are given in Table 6-2 for the pressure vessel inner radius.Again, the three pertinent exposure parameters are listed for the cycle one through four plant specific power distributions.
It is important to note that the data for the 6-5 vessel inner radius were taken at the clad/base metal interface; and, thus, represent the maximum predicted exposure levels of the vessel wall itself at the axial midplane.Radial gradient information applicable to g(E>1.0 MeV), gE>0.1 MeV), and dpa/sec is given in Tables 6-3, 6-4, and 6-5, respectively.
The data, obtained from the forward neutron transport calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations.
Exposure distributions through the vessel wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data listed in Tables 6-3 through 6-5.Note that these distributions are developed for the case with no surveillance capsules present.The effect of the capsules is to slightly reduce the relative flux at the inside of the vessel at angles within about+2 degrees of the capsule location.An example of the derivation of the neutron flux q+>1.0 MeV)at the 1/4T depth in the pressure vessel wall along the 45'zimuth is given by: (f)t/4/45)=(f)(233.756, 45)F(239.511, 45)where:$4/45')@233.756,45')
F(239.511,45')
Projected neutron flux at the 1/4T position on the 45'zimuth.
Projected or calculated neutron flux at the vessel inner radius on the 45'zimuth.
Ratio of the neutron flux at the 1/4T position to the flux at the vessel inner radius for the 45'zimuth.
This ratio is obtained by interpolation from Table 6-3.Similar expressions apply for exposure parameters expressed in terms of gE>0.1 MeV)and dpa/sec where the attenuation function F is obtained from Tables 6-4 and 6-5, respectively.
6.3 Neutron
Dosimetr The passive neutron sensors included in the Palo Verde Unit 2 surveillance program are listed in Table 6-6.Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the surveillance capsules and in the subsequent determination of the various exposure parameters of interest tgE>1.0 MeV), gE>0.1 MeV), dpa/sec].The relative locations of the neutron sensors within the capsules are shown in Figure 4-2.Since the dosimeters are all located very close to the same distance from the core at the radial 6-6 0 center of the capsule, no gradient corrections were necessary.
The iron, copper, titanium, and uranium (bare and covered)were each placed at three axial locations in the capsule near the top, middle, and bottom, respectively.
The cobalt-aluminum monitors (bare and covered), as well as the nickel and sulfur, were only placed in the middle location in space provided by an extra dosimetry holder.The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent neutron flux at the point of interest.Rather, the activation or fission process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the course of the irradiation period.An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known.In particular, the following variables are of interest: The measured specific activity of each monitor.The physical characteristics of each monitor.The operating history of the reactor.The energy response 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 procedurest
~+l.Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer.
The irradiation history of the Palo Verde Unit 2 reactor during cycles one through four was supplied by NUREG-0020,"Licensed Operating Reactors Status Summary Report," for the applicable period.The irradiation history applicable to capsule W137'is given in Table 6-7.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: where: R N~F F P-J C>[1-e g[e'J ref Reaction rate averaged over the irradiation period and referenced to operation at a core power level of P, (rps/nucleus).
6-7 A=Measured specific activity (dps/gm).N~=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-J Pr C.J Average core power level during irradiation period j (MW).Maximum or reference power level of the reactor (MW).Calculated ratio of gE>1.0 MeV)during irradiation period j to the time weighted average$(E>1.0 MeV)over the entire irradiation period.Decay constant of the product isotope (1/sec).t;=Length of irradiation period j (sec).t~=Decay time following irradiation period j (sec).and the summation is carried out over the total number of monthly intervals comprising the irradiation period.In the equation describing the reaction rate calculation, the ratio[P;J/fP,)
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 was calculated for e:ich fuel cycle using the adjoint transport technology discussed in Section 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 irradiations, particularly those employing low leakage fuel management, the additional C;term can be a significant correction.
For the irradiation history of capsule W137, the Aux level term in the reaction rate calculations was developed from the plant specific 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 Table 6-8 for capsule W137.Reactions that are cadmium shielded are denoted in this table by an asterisk (').Measured activities are given as corrected to 8/8/93.The table contains averages for each dosimeter that were used in the flux derivation except for U-238 which was corrected for U-235 fissions to give a corrected value of 1.53E-14 reactions per atom per second.The average value was used to derive a average value of flux for the capsule since the variation with axial position was found to be small and showed no correlation with the calculated axi il shape (Figure 6-3).6-8 Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment codet~l.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 linearly related to the flux Q by some response matrix A: Msgr)V~(s)~(c)S where i indexes the measured values belonging to a single data set s, g designates the energy group, and ct delineates spectra that may be simultaneously adjusted.For example, R,.=Q a,relates a set of measured reaction rates R;to a single spectrum Q, by the multigroup reaction cross-section q~.The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with large assigned uncertainties.
In the least squares adjustment, the continuous quantities (i.e., neutron spectra and cross-sections) were approximated in a multi-group format consisting of 53 energy groups.The trial input spectrum was converted to the FERRET 53 group structure using the SAND-II codet'l.This procedure was carried out by first expanding the 47 group calculated spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure in regions where group boundaries do not coincide.The 620 point spectrum was then re-collapsed into the group structure used in FERRET.The sensor set reaction cross-sections, obtained from the ENDF/B-V dosimetry file, were also collapsed into the 53 energy group structure using the SAND-II code.In this instance, the trial spectrum, as expanded to 620 groups, was employed as a weighting function 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 files.These matrices included energy group to energy group uncertainty correlations for each of the individual reactions.
However, correlations between cross-sections for different sensor reactions were 6-9 not included.The omission of this additional uncertainty 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 was taken from the center of the surveillance capsule modeled in the forward transport calculation with a source distribution averaged over the first four cycles.While the 53 x 53 group covariance matrices applicable to the sensor reaction cross-sections were developed from the ENDF/B-V data files, the covariance matrix for the input trial spectrum was constructed from the following relation: M I=R+R RIP where R.specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the set of values.The fractional uncertainties R specify additional random uncertainties for group g that are correlated with a correlation matrix given by: P>=[1-8]5 I+0 e" where: (g g)2 7'he first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes short range correlations over a group range y (e specifies the strength of the latter term).The value of 5 is 1 when g=g'nd 0 otherwise.
For the trial spectrum used in the current evaluations, a short range correlation of y=6 groups was used.This choice implies that neighboring groups are strongly correlated when e is close to 1.Strong long range correlations (or anti-correlations) were justified based on information presented by R.E.Maerkerp'l.
Maerker's results are closely duplicated when y=6.The uncertainties associated with the measured reaction rates included both statistical (counting) and systematic components.
The systematic component of the overall uncertainty accounts for counter efficiency, counter calibrations, irradiation history corrections, and corrections for competing reactions in the individual sensors.Uncertainty estimates for the non-fission dosimeter results were taken to be 5%based on consistency studies of capsule dosimetry"".
6-10 The U-238 sensors in the Palo Verde Unit 2 capsules are inserted both bare and cadmium covered.The bare sensors contain a significant contribution from U-235 impurity in the U-238 and also a contribution from plutonium production in the dosimeter.
The difference between the bare and covered dosimeters can thus provide an estimated correction for determining the U-238 reaction rate in the covered dosimeters.
This correction was found to be quite small.However, comparison of the bare U-238 dosimeter result with the cobalt results indicates that these results are probably inconsistent.
In addition, the analysis of the cadmium covered U-238 dosimeters was hampered by incomplete recovery of the U-238 and mixing of the U-238 and cadmium.This necessitated a larger uncertainty assignment for the U-238 result.The inconsistency of the cobalt results also indicated that a larger uncertainty should be.assigned to the low energy flux.Results of the FERRET evaluations of the capsule W137 dosimetry are given in Tables 6-9 through 6-12.The data summarized in Table 6-9 include fast neutron exposure evaluations in terms of C(E>1.0 MeV), 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 (except for the U-238 which was assigned a large uncertainty as discussed above)as shown in Table 6-10.The adjusted spectra from the least squares evaluations are given in Table 6-11 in the FERRET 53 energy group structure.
Table 6-12 compares the flux and fluence results derived from the dosimeter measurements (Table 6-9)with the calculated values{Table 6-1).The results for capsule W137 are the first results for a capsule from Palo Verde and thus cannot be compared with other similar capsules to check for consistency.
However, the good agreement between calculated and measured values supports the adequacy of the analysis.6.4 Pro'ections of Pressure Vessel Ex osure Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-13.Along with the current{454 EFPY)exposure, projections are also provided for exposure periods of 15 EFPY and 32 EFPY.In computing these vessel exposures, the calculated values from Table 6-2 were scaled by the average measurement/calculation ratios (M/C)observed from the evaluations of dosimetry from capsule W137 for each fast neutron exposure parameter.
This procedure resulted in bias factors of 1.07, 1.16, and 1.14 being applied to the calculated values of 4(E>1.0 MeV), 4{E>0.1 MeV), and dpa, respectively.
Projections for 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.This is expected to be conservative since the fuel loading patterns employed since the first cycle have led to lower fluence than this average.6-11 The overall uncertainty associated with the best estimate exposure projections at the pressure vessel wall depends on the individual uncertainties in the measurement process, the uncertainty in the dosimetry location, and on the uncertainty in the extrapolation of results from the measurement point to the point(s)of interest in the vessel wall.For Palo Verde Unit 2, the estimated extrapolation uncertainty is 5%and the uncertainty in the plant specific measurement/calculation bias factor derived from the surveillance capsule measurement is 11%as derived by the least squares process.These uncertainties are independent and so the total uncertainty is 12%as calculated by the square root of the sum of the squares of the individual uncertainty contributors.
This 12%uncertainty in the projected exposure of the pressure vessel wall is a 1cr estimate for 4(E>1.0 MeV).0 Exposure projections through the vessel at 15 EFPY and,32 EFPY are provided in Table 6-14 for use in the development of heatup and cooldown curves for Palo Verde Unit 2.Data are calculated based on both a 4(E>1.0 MeV)slope and a plant specific dpa slope through the vessel wall.The dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions are defined by the relations:
4(>/4A=4(08 dpa(01)0(3/4Q=4(08 dpa(0T)In Table 6-15 updated lead factors are listed for each of the Palo Verde Unit 2 surveillance capsules Lead factor data based on the accumulated fluence through cycle four are provided for each capsule.6-12 TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE RATES AT THE SURVEILLANCE CAPSULE CENTER Capsule Location 38 40')1.0 MeV n/cm--sec 434 Cycle 1 Cycle 2 Cycle 3 Cycle 4 Average 3.376E+10 2.361E+10 2.470E+10 2.461E+10 2.672E+10 3.428E+10.2.374E+10 2.520E+10 2.460E+10 2.701E+10 3.355E+10 2.328E+10 2.506E+10 2.401E+10 2.653E+10 E)0.1 MeV n/cm--sec Cycle 1 Cycle 2 Cycle 3 Cycle 4 Average 6.117E+10 4.277E+10 4.475E+10 4.458E+10 4.840E+10 6.201E+10 4.294E+10 4.559E+10 4.450E+10 4.886E+10 6.060E+10 4.206E+10 4.526E+10 4.338E+10 4.792E+10 Iron Atom Dis lacement Rate d a sec Cycle 1 Cycle 2 Cycle 3 Cycle 4 Average 4.908E-11 3.432E-11 3.591E-11 3.578E-11 3.884E-11 4.985E-11 3.452E-11 3.665E-11 3.577E-11 3.927E-11 4.879E-11 3.386E-11 3.644E-11 3.493E-11 3.859E-11 6-13 TABLE 6-2 CALCULATED AZIMUTHAL VARIATION OF FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD/BASE METAL INTERFACE 00 15'0'E>1.0 MeV n/cm--sec 40'5'ycle 1 1.418E+10 2.010E+10 2.026E+10 2.370E+10 2.299E+10 Cycle 2 1.021E+10 1.422E+10 1.525E+10 1.662E+10 1.613E+10 Cycle 3 1.147E+10 1.538E+10 1.538E+10 1.786E+10 1.733E+10 Cycle 4 1.272E+10 1.606E+10 1.641E+10 1.715E+10 1.664E+10 Average 1.217E+10 1.647E+10 1.684E+10 1.887E+10 1.831E+10 E>0.1 MeV n/cm--sec Cycle 1 2.930E+10 4.208E+10 4.244E+10 4.966E+10 4.849E+10 Cycle 2 2.111E+10 2.978E+10 3.195E+10 3.484E+10 3.402E+10 Cycle 3 2.370E+10 3.220E+10 3.222E+10 3.743E+10 3.655E+10 Cycle 4 2.628E+10 3.363E+10 3.438E+10 3.595E+10 3.510E+10 Average 2.515E+10 3 449E+10 3.528E+10 3.955E+10 3.980E+10 Iron Atom Dis lacement Rate d a sec Cycle 1 2.194E-11 3.099E-11 3.112E-11 3.626E-11 3525E-11 Cycle 2 1.580E-11 2.193E-11 2.343E-11 2.544E-11 2.473E-11 Cycle 3 1.775E-11 2.371E-11 2.363E-11 2.733E-11 2.657E-11 Cycle 4 1.968E-11 2.477E-11 2.521E-11 2.625E-11 2.552E-11 Average 1.883E-11 2.540E-11 2587E-11 2.888E-11 2.893E-11 6-14 TABLE 6-3 RELATIVE RADIAL DISTRIBUTION OF$(E>1.0 MeV)WITHIN THE PRESSURE VESSEL WALL Radius~em 233.756t't 234.006 234.631 235.506 236.631 237.923 239.409 241.196 243.204 245.062 246.477 247.78 249.191 250.715 252.055, 253.098 254.181 255.181 255.994 256.775+po 1.0000 0.9854 0.9368 0.8595 0.7571 0.6460 0.5335 0.4203 0.3194 0.2460 0.2003 0.1656 0.1346 0.1076 0.0877 0.0744 0.0625 0.0527 0.0452 0.0391 15'.0000 0.9851 0.9365 0.8580 0.7549 0.6430 0.5292 0.4159 0.3154 0.2431 0.1974 0.1619 0.1311 0.1041 0.0848 0.0713 0.0592 0.0494 0.0416 0.0353 30'.0000 0.9853 0.9369 0.8587 0.7560 0.6445 0.5310 0.4178 0.3173 0.2450 0.1992'.1635 0.1326 0.1054 0.0860 0.0723 0.0601 0.0501 0.0422 0.0358 40'.0000 0.9856 0.9366 0.8572 0.7529 0.6405 0.5273 0.4143 0.3143 0.2425 0.1976 0.1631 0.1322 0.1049 0.0854 0.0720 0.0599 0.0497 0.0416 0.0351 45'.0000 0.9854 0.9370 0.8591 0.7554 0.6434 0.5303 0.4171 0.3164 0.2433 0.1981 0.1640 0.1333 0.1062 0.0861 0.0727 0.0604 0.0502 0.0421 0.0356 NOTES: 1)Base Metal Inner Radius 2)Base Metal Outer Radius 6-15 TABLE 6-4 RELATIVE RADIAL DISTRIBUTION OF@E)0.1 MeV)WITHIN THE PRESSURE VESSEL WALL Radius~em po 15'0'0'5'33.756t'>
234.006 234.631 235.506 236.631 237.923 239.409 241.196 243.204 245.062 246.477 247.780 249.191 250.715 252.055 253.098 254.181 255.181 255.994 256.775<~1.000 1.009 1.014 1.004 0.974 0.927 0.867 0.792 0.708 0.632 0.575 0.523 0.470 0.414 0.367 0.330 0.293 0.258 0.228 0.203 1.000 1.008 1.011 0.997 0.962 0.912 0.848 0.769 0.683 0.606 0.547 0.495 0.441 0.385 0.338 0.300 0.262 0.227 0.196 0.171 1.000 1.009 1.013 1.000 0.967 0.918 0.856 0.778 0.693 0.616 0557 0.504 0.450 0.393 0.345 0.307 0.268 0.232 0.199 0.173 1.000 1.008 1.011 0.996 0.961 0.910 0.846 0.768 0.682 0.605 0547 0.495 0.441 0.384 0.336 0.299 0.260 0.223 0.191 0.164 1.000 1.009 1.012 1.000 0.966 0.917 0.854 0.776 0.690 0.612 0554 0.502 0.447 0.390 0.341 0.303 0.264 0.227 0.194 0.167 NOTES: 1)Base Metal Inner Radius 2)Base Metal Outer Radius 6-16 TABLE 6-5 RELATIVE RADIAL DISTRIBUTION OF dpa/sec WITHIN THE PRESSURE VESSEL WALL Radius~cm 00 15'0'0'5'33.756<'>
234.006 234.631 235.506 236.631 237.923 239.409 241.196 243.204 245.062 246.477 247.780 249.191 250.715 252.055 253.098 254.181 255.181 255.994 256.775<~1.0000 0.9868 0.9455 0.8816 0.7981 0.7073 0.6141 0.5179 0.4283 0.3592 0.3134 0.2761 0.2403 0.2061 0.1788 0.1588 0.1394 0.1223 0.1087 0.0971 1.0000 0.9865 0.9449 0.8797 0.7951 0.7030 0.6080 0.5106 0.4203 0.3513 0.3047, 0.2662 0.2301 0.1955 0.1684 0.1478 0.1278 0.1103 0.0956 0.0834 1.0000 0.9869 0.9459 0.8816 0.7978 0.7066 0.6124 0.5156 0.4257 0.3567 0.3100 0.2712 0.2348 0.1998 0.1721 0.1911 0.1307 0.1125 0.0973 0.0847 1.0000 0.9867 0.94500.8796 0.7944 0.7024 0.6080 0.5109 0.4209 0.3520 0.3062 0.2682 0.2317 0.1967 0.1689 0.1482 0.1278 0.1094 0.0939 0.0812 1.0000 0.9870 0.9461 0.8822 0.7980 0.7065 0.6127 0.5159 0.4255 0.3557 0.3095 0.2718 0.2354 0.2000 0.1716 0.1507 0.1299 0.1112 0.0955 0.0826 NOTES: 1)Base Metal Inner Radius 2)Base Metal Outer Radius 6-17 TABLE 6-6 NUCLEAR PARAMETERS USED IN THE EVALUATION OF NEUTRON SENSORSMonitor Material Reaction of iatereat Target Weight Fraction Response Ranee Fission Product Yield Half-Life~5o Copper*Iron Nickel Titanium Uranium-238'obalt-Aluminum'obalt-Aluminum Cu"(n,a)Co Fe~(n,p)Mn~
Ni"(n,p)Co" Ti{n,p)Sc+U-'n,I)Cs'3'o"(n,y)
Co~Co"(n,y)Co~0.6917 0.0580 0.6827 0.0810 1.0 0.0017 0.0017 E>5 MeV E>2MeV E>2MeV E>2MeV E>1MeV 0.4ev>E>0.015 MeV E (0.015 MeV 5.271 yrs 312.5 days 70.78 days 83.83 days.30.17 yrs 6.00 5.271 yrs 5.271 yrs'Denotes that monitor is cadmium shielded.Both bare and cadmium shielded U-238 monitors were included.Note: The capsule also contained a sulfur dosimeter but this could not be analyzed due to decay of the P-32 which has a 14.28 day half-life.
O 6-18 TABLE 6-7 MONTHLY THERMAL GENERATION DURING THE FIRST FOUR FUEL CYCLES OF THE PALO VERDE UNIT 2 REACTOR Thermal Generation Year Month MW-hr Thermal Generation Year Month MW-hr 1986 1987 1988 1989 1990 5 6 7 8 9 10 ll 12 1 2 3 4 5 6 7 8 9 10 11 12 1 2 3 5 6 7 8 9 10 11 12 1 2 3 5 6 7 8 9 10 ll 12 1 2 3 5 6 7 8 178458 748794 179055 789746 441921 1785522 2573938 2217984 816605 0 843281 2523504 2217619 2370470 2667944 2807364 2670655 2794687 2400712 2799320 2796621 1711979 0 0 0 510519 2794833 2713036 2702001 2814560 1909336 2766725 2800843 1383148 1303549 0 0 20794 1657660 2576391 1206913 1205545 0 2546733 2806069 2064631 0 0 0 0 558801 2659419 1990 1991 1992 1993 9 10 11 12 1 2 3 4 5 6 7 8 9 10 ll 12 1 2 3 4 5 6 7 8 9 10 11 12 1 2 3 2724764 2746944 2722657 2809763 2810976 2550144 2797277 2733538 2826434 2734422 2825239 1760151 2731841 1454795 0 0 1548257 2642730 2150742 2730948 2824993 2735197 2826315 2823534 2701162 2797104 2490116 2819640 2825987 2549505 1182454 (Shutdown 3/14/93)6-19 TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCE CAPSULE W137 MONITOR AND AXIAL LOCATION Cu-63 n.c Co-60'EASURED ACTIVITY~ilis/sec.util SATURATED ACTIVITY~dls/sec-ltl REACTION RATE~>s/nucleus 93-3144 TOP 93-3153 MID 93-3158 BOT Averages Fe-54 n.Mn-54 1.22E+05 1.19E+05 1.19E+05 1.20E+05 3.204E+05 3.125E+05 3.125E+05 3.152E+05 4.808E-17 93-3142 TOP 93-3147 MID 93-3156 BOT Averages 1.29E+06 1.28E+06 1.24E+06 1.27E+06 2.334E+06 2.316E+06 2.243E+06 2.298E+06 3.674E-15 93-3152 MID 6.81E+06 3.278E+07 4.680E-15 Ti-46 n Sc-46 93-3141 TOP 93-3146 MID 93-3155 BOT Averages 2.01E+05 1.95E+05 1.90E+05 1.95E+05 7.786E+05 7553E+05 7.360E+05 7.566E+05 7.128E-16 93-3143 TOP 93-3151 MID 93-3157 BOT Averages U-238 n Cs-137 250E+05 2.08E+05 2.16E+05 2.25E+05 2.607E+06 2.169E+06 2.252E+06 2.343E+06 1.544E-14 93-3140 TOP 93-3157 BOT Averages Co-59 n Co-60 3.25E+06 2.79E+05 3.02E+05 3.389E+06 2.909E+06 3.149E+06 2.075 E-14 93-3148 MID 6.67E+07 1.752E+08 1.143E-11 Co-59 n Co-60'3-3150 MID 7.98E+06 2.096E+07 1.367E-12 6-20 TABLE 6-9
SUMMARY
OF NEUTRON DOSIMETRY RESULTS SURVEILLANCE CAPSULE W137 Calculation of Measured Fluence Meas Fluence<0.414 ev Meas Fluence>0.1 Mev Meas Fluence>1.0 Mev Flux Time Fluencc 3.177E+11 1.433Ew08 4.551E+19 Uncertainty
~22%5.575E+11 1.433E%08 7.987E+18<18%2.842E+10 1.433E+08 4.071E+18 F11%dpa 4.379E-11 1.433E+08 6.273 E-03 TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACI'ION RATES AT THE SURVEILLANCE CAPSULE CENTER SURVEILLANCE CAPSULE W137 REACTION Cu43 (n,ct)Co-60 Fe-54 (n,p)Mn-54 Ni-58 (n,p)Co-58 Ti-46 (n,p)Sc-46 U-238 (n,f)Cs-137 (Cd)Co-59 (n,y)Co-60 Co-59 (n,y)Co-60 (Cd)MEASURED 4.81E-17 3.67E-15 4.68E-15 7.13E-16 1.53E-14 1.01E-11 1.21E-12 3.68E-15 4.71E-15 7.11E-16 1.26E-14 1.00E-11 1.19E-12 1.00 1.01 1.00 0.83 0.99 0.99 ADJUSTED CALCULATION 4.85E-17 1.01 6-21 TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCE CAPSULE W137 GROUP ENERGY ADJUSTED FLUX GROUP ENERGY ADJUSTED FLUX 10 12 13 14 15 16 17 18 19 20 21 22 24 26 27~MeV 17.33 14.92 13.50 11.62 10.00 8.607 7.408 6.065 4.966 3.679 2.865 2.231 1.738 1.353 1.108 8.208E-01 6.393E-01 4.979E-01 3.877E-01 3.020E-01 1.832E-01 1.111E-01 6.738E-02 4.087E-02 2554E-02 1.989E-02 1.503 E-02~a/cn>-ecc 5.60E+06 1.30E+07 5.48E+07 1.28E+08 2.94E+08 5.06E+08 1.23E+09 1.70E+09 2.94E+09 2.82E+09 4.67E+09 4.43E+09 4.51E+09 3.47E+09 4.83E+09 4.42E+09 3.95E+09 2.87E+09 3.24E+09 4.84E+09 4.12E+09 3.39E+09 2.85E+09 2.14E+09 1.49E+09 1.24E+09 2.27E+09 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53~Me V 9.119E-03 5.531E-03 3.355E-03 2.839E-03 2.404E-03 2.035E-03 1.234E-03 7.485 E-04 4.540E-04 2.754E-04 1.670E-04 1.013E-04 6.144E-OS 3.727E-OS 2.260E-05 1.371E-05 8.315E-06 5.043E-06 3.059E-06 1.855E-06 1.125 E-06 6.826E-07 4.140E-07 2.511E-07 1.523E-07 9.237E-08~e/cmc-cec 2.45E+09 2.77E+09 9.33E+08 9.57E+08 1.01E+09 3.27E+09 3.61E+09 4.08E+09 4.60E+09 5.28E+09 9.97E+09 5.60E+09 5.15E+09 4.59E+09 4.06E+09 3.63E+09 3.33Et09 3.12E+09 2.97E+09 2.82E+09 2.63E+09 3.14E+09 4.11E+09 1.61E+10 3.90E+10 259E+11 Note: Tabulated energy levels represent tite upper energy in each group.6-22 TABLE 6-12 COMPARISON OF CALCULATED AND MEASURED NEUTRON EXPOSURE LEVELS FOR PALO VERDE UNIT 2 SURVEILLANCE CAPSULE W137 Comparison of Calculated and Mcasurcd INTEGRATED Neutron EXPOSURE for Capsule W137 Calculated Measured C/M Fluence (E>1.0 Mcv)[n/cm2-sccj Fluence (E>0.1 Mcv)[n/cm2-sec[
dpa 3.801E+18 6.865E+18 5.528E-03 4.071E+18 0.934 7.987E+18 0.860 6.273 E-03 0.881 Comparison of Calculated and Mcasurcd Neutron EXPOSURE RATE for Capsule W137 Flux (E>1.0 Mcv)[n/ctn2-sec[
Hux (E>0.1 Mev)[n/cnt2-sec[
dpa/s 2.653E+10 4.792E+10 3.859E-11 2.842E+10 5575E+10 4.379E-11 Calculated Measured~CM 0.934 0.860 0.881 6-23 TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE PRESSURE VESSEL CLAD/BASE METAL INTERFACE BEST ESTIMATE EXPOSURE (4.540 EFPY)AT THE PRESSURE VESSEL INNER RADIUS 0 DEG"'5 DEG 30 DEG~40 DEG'5 DEG E>1.0 1.873E+18 2.535E+18 2.592E+18 2.904E+18 2.818E+18 E>0.1 4.203E+1S 5.764E+18 5.897E+18 6.610E+18 6.453E+18 dpa 3.070E-03 4.141E-03 4.219E-03 4.708E-03 4.577E-03 BEST ESTIMATE FLUENCE RATE AT THE PRESSURE VESSEL INNER RADIUS 0 DEG"'5 DEG 30 DEG~40 DEG'5 DEG E>1.0 1.307E+10 E>0.1 2.934E+10 dpa 2.143E-11 1.770E+10 4.023E+10 2.891E-11 1.S09E+10 4.116E+10 2.945E-11 2.027E+10 1.967E+10 4.614E+10 4.505 E+10 3.287E-11 3.195 E-11 BEST ESTIMATE EXPOSURE (15.0 EFPY)AT THE PRESSURE VESSEL INNER RADIUS 0 DEG" 15 DEG E>1.0 6.187E+18 S.376E+18 E>0.1 1.389E+19 1.905E+19 dpa 1.014E-02 1.36SE-02 30 DEGAS'0 DEGt'5 DEG 8.565E+18 9.595E+18 9.310E+18 1.949E+19 2.184E+19 2.132E+19 1.394 E-02 1.556E-02 1.512E-02 BEST ESTIMATE EXPOSURE (32.0 EFPY)AT THE PRESSURE VESSEL INNER RADIUS 0 DEG" 15 DEG 30 DEG+~40 DEG'5 DEG E>1.0 E>0.1 dpa 1.320E+19 1.787E+19 1.827E+19 2.963E+19 4.063E+19 4.157E+19 2.164E-02 2.919E-02 2.974 E-02 2.047E+19 1.986 E+19 465 9E+19 4.549E+19 3.319E-02 3.226E-02 (a)Applies to axial weld at 90'ocation.(b)Applies to axial weld at 210'nd 330'ocations.(c)Maximum Ouence point.6-24 TABLE 6-14 NEUTRON EXPOSURE VALUES FOR USE IN THE GENERATION OF HEATUP/COOLDOWN CURVES FLUENCE BASED ON E>1.0 MeV SLOPE 0 DEG~" 15 DEG 30 DEG~'0 DEGt" 45 DEG 15 EFPY FLUENCE SURFACE 6.187E+18 8.376E+18 8.565E+18 9.595E+18 9.301E+18 1/4T 3.256E+18 4.373 E+18 4.486E+18 5.025E+18 4.842E+18 3/4T 6.352E+17 8.324E+17 8.621E+17 9.658E+17 9.323E+17 32 EFPY FLUENCE SURFACE 1.320E+19 1.787E+19 1.827E+19 2.047E+19 1.986E+19 1/4T 6.946E+18 9.328 E+18 3/4T 1.355 E+18 1.776 E+18 9.570E+18 1.072E+19 1.033E+19 1.839E+18 2.060E+18 1.989E+18 FLUENCE BASED ON dpa SLOPE 0 DEGAS')15 DEG 30 DEG~i 40 DEG" 45 DEG 15 EFPY FLUENCE SURFACE 6.187E+18 8.376E+18 8.565E+18 9.595 E+18 9.301E+18 1/4T 3.763E+18 5.042E+18 3/4T 1.235 E+18 1.583E+18 5.194E+18 5.776E+18 5.604E+18 1.654E+18 1.823E+18 1.769E+18 32 EFPY FLUENCE SURFACE 1.320E+19 1.787E+19 1.827E+19 2.047E+19 1.986E+19 1/4T 8.028E+18 3/4T 2.634E+18 1.076E+19 1.108E+19 1.232E+19 1.195E+19 3.377E+18 3.528E+18 3.889E+18 3.773E+18 (a)Applies to axial weld at 90'ocation.(b)Applies to axial weld at 210'nd 330'ocations.(c)Maximum Qucncc point.6-25 TABLE 6-15 UPDATED LEAD FACTORS FOR PALO VERDE UNIT 2 SURVEILLANCE CAPSULES CAPSULE W38 W43 W137 W142 W230 W310 LEAD FACTOR 1.41 1.40 1.40'.41 1.43 1.43~WITHDRAWN EOC 4, BASIS FOR THIS ANALYSIS 6-26 FIGURE 6-1 PALO VERDE REACTOR MODEL SHOWING A 45 DEGREE (R,Q)SECTOR CON CI~SHIELD 4'~tF 0 C~INSUIAlTON REACI'OR CAVITY SHROUD FUEL BYPASS WA'IKR INlEI'ATER BARREL 133 233 X CAl 383 6-27 FIGURE 6-2 AZIMUTHAL VARIATION OF NEUTRON FLUX (E)1.0 MEV)AT THE REACTOR VESSEL INNER RADIUS20 19 18 E V C 17 16~w 15 14~~g~SN~r Lower Curve Shows Effect of Surveillance Capsules I 12 0 10 20 30 50 Azimuthal Angle (Degrees)6-28 FIGURE 6-3 AXIAL DISTRIBUTION OF REACTOR POWER 1.2 o 0.9 C4~~0.8 0.7 0.6 0.5 0 50 Axial Distance&om Core Bottom (inches)150 6-29 0 i f SECTION 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following surveillance capsule removal schedule meets the requirements of ASTM E185-82 and is recommended for future capsules to be removed from the Palo Verde Unit 2 reactor vessel: Table 7-1 Palo Verde Unit 2 Reactor Vessel Surveillance Capsule Withdraw Schedule Location Lead Factor Removal Time (EFPY)<'~Fluence (n/cm~)137'30'10 38'30 142 1.40 1.43 1.43 1.41 1.40 1.41 4.540 15 EOL Stand-by Stand-by Stand-By 4.071 x 10'~1.37 x 10'.93 x 10'~(a)Effective Full Power Years (EFPY)from plant startup.(b)Actual measured neutron fluence 7-1 0 0 SECTION
8.0 REFERENCES
1.Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.Nuclear Regulatory Commission, May, 1988.2.Chang, B.C., Arizona Public Service Company Palo Verde Unit 2 Evaluation for Baseline Specimens Reactor Vessel Materials Irradiation Surveillance Program, ABB Combustion Engineering Report TR-V-MCM-013, November 5, 1992.3.Section III of the ASME Boiler and Fissure Vessel Code, Appendix G, Protection Against Nonductile Failure.4.ASTM E208, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.5.Data package supplied to Westinghouse by the Arizona Public Service Company (File API'-106/13 Capsule W137)).6.Code of Federal Regulations, 10CFR50, Appendix G, Fracture Toughness Requirements, and Appendix H, Reactor Vessel Material Surveillance Program Requirements, U.S.Nuclear Regulatory Commission, Washington, D.C.7.ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF), in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.8.'STM E23-92, Standard Test Methods for Notched Bar Impact Testing of Metallic Materials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1992.9.ASTM A370-92, Standard Test Methods and Definitions for Mechanical Testing of Steel Products, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.8-1 10.ASTM E8-91, Standard Test Methods of Tension Testing of Metallic Materials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1991.11.ASTM E21-79(1988), Standard Practice for Elevated Temperature Tension Tests of Metallic Materials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1989.12.ASTM E83-92, Standard Practice for Verificarion and Classification of Extensometers, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1992.13.ASTM Designation E853-87, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.14.ASTM Designation E693-79, Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa), in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.15.ASTM Designation E706-87, Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.16.ASTM Designation E482-89, Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.17.ASTM Designation E560-84, Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.18.ASTM Designation E261-90, Standard Method for Determining Neutron Flux, Fluence, and Spectra by Radioactivation Techniques, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.8-2 19.ASTM Designation E262-86, Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.20.R.G.Soltesz, R.K.Disney, J.Jedruch, and S.L.Ziegler, Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation.
Vol.5-Two-Dimensional Discrete Ordinates Transport Technique, WANL-PR(LL)-034, Vol.5, August 1970.21.ORNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light 1Vater Reactors.22.R.E.Maerker, et al, Accounting for Changing Source Distributions in Light lVater Reactor Surveillance Dosimetry Analysis, Nuclear Science and Engineering, Volume 94, Pages 291-308, 1986.23.ASTM Designation E1005-84, Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.24.ASTM Designation E263-88, Standard Method for Determining Fast-Neutron Reaction Rates by Radioactivation of Iron, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.25.ASTM Designation E264-92, Standard Method for Determining Fast-Neutron Reaction Rates by Radioactivation of Nickel, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.26.ASTM Designation E481-86, Standard Method for Measuring Neutron Fluence Rate by Radioactivation of Cobalt and Silver, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.27.ASTM Designation E58-92, Standard Method for Determining Fast-Neutron Reaction Rates by Radioactivation of Copper, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.28.ASTM Designation E526-92, Standard Method for Determining Fast-Neutron Reaction Rates by Radioactivation of Titanium, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.8-3 29.ASTM Designation E704-90, Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.o i 30.F.A.Schmittroth, FERRET Data Analysis Core, HEDL-TME 7940, Hanford Engineering Development Laboratory, Richland, WA, September 1979.31.W.N.McElroy, S.Berg and T.Crocket, A Computer-Automated iterative Method of Neutron Flux Spectra Determined by Foil Activation, AFWL-TR-7<1, Vol.I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.32.EPRI-NP-2188, Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications, R.E.Maerker, et aL, 1981.33.WCAP-13390, Westinghouse Fast Neutron Exposure Methodology for Pressure Vessel Fluence Determination and Dosimetry Evaluation, S.L.Anderson, May 1992.
APPENDIX A Load-Time Records for Charpy Specimen Tests and Comparisons of Data for Unirtadiated and Irradiated Precarcked Chatpy Specimens 0 0 PGY GEh/ERAL YIE LO LOAD~GYQ I I I I I I I I I I I't I t-P~~MAXIMIJM LOAO FAST P~FRACTURE LOAD F f AS T FR ACT U R E'A~ARREST LOAD 7!M 6 W=Fracture initiation region I W~Fracture yroyagation region P t+>~Time to general yielding tM Time to mmcimum load tF~one to fast{br'.ttle) fracture start Pigure A-I.Idealized load-time record PALO VERDE NZ 1811E CO a N~D loZ 1B11E Zo4 3.6 TItK (ICiZC)PALO VERDE g2 4e8 6+0 18184 C5 D IeZ 1B3.24 Z.4 3+6 Tlm C nSCC)PALO VERDE P2 4 8 6.0 Figure A-2.Load-time records for Specimens 1BllE and 1B124 A-2 1B11M 2.4 3+6 TINE<NSKC)PALO VERDE g2 4+8 6.0 18150 1815D 2+4 3m 6 TItK (NSKC)PALO VERDE g2 4+8 6 0 Figure A-3.Load-time records for Specimens 1BllM and 1B15D IBI4Y g~o W g CU o~D Isa 1B14Y Ro4 3.6 TlllE C NSEC)PALO VERDE g2 4oa 6.0 18136 g~o 5 Cl C9 Q ol O o.D I~8 1B136 eo4 3o6 TIME (NSKC>PALO VERDE g2 4m 8 6.0 Figure AA.Load-time records for Specimens 1B14Y and 18136 18145 1B145 eo4 3.6 TItC C!CEC>PALO VERDE g2 4ee 6.0 1814/ea4 3.6 TIIK<ICXC)PALO VERDE g2 BatCh11B14gl 4+8 6o0 Figure A-5.Load-time records for Specimens 1B145 and 1B14D 18188 1B12B eo4 3o6 TBK C ICKC>PALO VERDE g2 4ee 6.0 1B255 ee4 3 6 TItK<NXC>PALO VERDE g2 4ee 6m 0 Figure A-6.Load-time records for Specimens 1B12B and 1B255 Ol o W Cl C9 g Ol O~D I 2 1B23L L4 3.6 T1%C CEC)PALO VERDE jNI2 4os 6m 0 CU eD Ice 1B2AE 2.4 3+6 TIlK C NSKC>PALO VERDE g2 4.8 6o0 Figure A-7.Load-time records for Specimens 1B23L and 1B2AE 18224 CU~D 1+2 1B224 ee4 3e6 TIlK C CKC)~PALO VERDE S2 4oe 6m 0 CU~V oD 1e2 1B267'o4 3.6 TIJOU C ICKC)PALO VERDE g2 4.8 6e0 Figure A-8.Load-time records for Specimens 18224 and IB267 PALO VKROE I2~D Ioe 1B23P 2,4 3o6 TIttE C ttSEC)PALO VERDE g2 4ee 6.0 C9 g~o\4~D i+2 2e4 36 TlttE C tCEC>PALO VERDE f2 1B2 5AM 4oe 6.0 Figure A-9.Load-time records for Specimens 1B23P and 1B2AM PALO VERDE 48 CI~D 1B26D 8+4 3.6 TItK (ICEC)PALO VERDE g2.4o8 6.0 182IU.D 1B21U 8.4 3.6 TIE C tCZC)PALO VERDE g2 4.8 6 0 Figure A-10.Load-time records for Specimens 1B26D and 1B21U A-10 PALO VERGE 82 i+2 1B22U 2,4 3 6 TlttE C ICKC>PALO VERDE g2 4s8 6+0 1+2 1B22E 2.4 3+6 TIttE (t5EC)PALO VERDE g2 4I 8 6m 0 Figure A-11.Load-time records for S pecimens 1B22U and 1B22E 1884 J 1B24 J en 4 3o6 Tlirt<NSEC>PALO VERDE g2 4,8 6m 0 lsa43 1B243 8+4 3 6 Tlirt C NSEC)PALO VERDE P2 4e8 6,o Figure A-12.Load-time records for Specimens 1B24J and 1B243 A-12 I+2 1B22M 2e4 3.6 TIIC (ttSEC)PALO VERDE g2 4+8 6+0 i+2 1B23M 2 4 3,6 TIttE C ttSEC>PALO VERDE P2 4I8 6e0'Figure A-13.Load-time.
records for Specimens 1B22M and 1B23M A-13 PALO VERGE<<2 1B3AO g~R a~D le2 1B3AD 2i4 3,6 TIttE C tCKC>~PALO VERDE g2 4,8 6.0 1831b g 0 o CI~D le2 1B316 2+4 3.6 TIttE C tCEC>PALO VERDE g2 4oe 6+0 Figure A-14.Load-time records for Specimens 1B3AD and 1B316 A-14 PALO VEROE NR g~o I ol~D 1B35E eo4 3.6 TlliE<NSKC)PALO VERDE f2 4e8 6o0 m 1 8 CU O.D 1B342 8.4 3+6 TItK C tiSEC>PALO VERDE g2 4oe 6.0 Figure A-15.Load-time records for Specimens 1B35E and 1B342 A-I5 PALO VEROE 02 Ie2 1B34U 2e4 3,6 TINE (tCXC)PALO VERDE g2 4oe 6m 0 18312 Ioe 1B312 2+4 3+6 TINE<NSEC)PALO VERDE g2 4+8 6+0 Figure A-16.Load-time records for Specimens 1B34U and 1B312 A-16 PALO VERtK 08 f 834C LB34E 2+4 3o6 TIN (ICKC)PALO VERDE g2 4+8 6.0 f.2 1B35P em 4 3.6 Tilg C NSCC)PALO VERDE g2 4e8 6e0 Figure A-17.Load-time records for Specimens 1B34E and 1B35P A-17 IER6 w~o o~D 1B326 a+4 3.6 TitK C NSEC)PALO VERDE g2 4o8 6e0 o o oD 1.2 1B3A3 a+4 3+6 TitK (lCXC)PALO VERDE if2 4i8 6e0 Figure A-18.Load-time records for Specimens 1B326 and 1B3A3 A-18 1B3AL em 4 3+6 TINE<tCEC>PALO VERDE g2 4e8 6m 0 1B35M ee4 3+6 TIJOU C INC>PALO VERDE g2 4+8 6+0 Figure A-19.Load-time records for Specimens IB3AL and IB35M PALO VERDE 62 l833E CO o I Ol Ol 04 O~D li2 1B33E 2o4 3+6 TIJOU C 1CEC>PALO VERDE g2 4+8 6.0 l83l8 CO o CO C9 I Ol Ol CI ,D la2 1B31B 2,4 3,6 TIIC<tCXC>PALO VERDE g2 4ee 6io Figure A-20.Load-time records for Specimens 1833E and 1B31B A-20 CD Cl o oD so 4 3o6 TIIK (NSEC>PALO VERDE g2 Batch:1B32A 4o8 6.0 18411 CD 5 4 D CD CD I DC CD oD 1B411 a+4 3+6 TIIK (t5EC>PALO VERDE g2 4+8 6.0 Figure A-21.Load-time records for Specimens 1B32A and 1B411 A-21 PALO VEROE Ie IB440 yg~a r W CI~0 los 1B44D Bo4 3o6 TItK C ICXC>PALO VERDE g2 4es 6+0 g~ou)C9 I CIJ CI~0 1B43B Be 4 3+6 TItK C NSEC)PALO VERDE g2 4.8 6m 0 Figure A-22.Load-time records for Specimens 1B44D and 1B43B A-22 I Al CI~0 1B43Y eo4 3+6 TIklE C!CEC>PALO VERDE g2 4oe 6e0 C0 o Q ol D O eo 1B432 e 4 3+6 TnlE (llSEC)PALO VERDE I2 408 6+0 Figure A-23.Load-time records for Specimens 1B43Y and 1B432 A-23 PALD VERDE 08 1843T 1B43T 84 3.6 TIlK C lCKC)PALO VERDE t2 4 8 6.0 g 0 a 8 CU CI oo 1B444 ee4 3+6 TNE (llsEC)PALO VERDE g2 4,8 6+0 Hgure A-24.Load-time records for Specimens 1B43T and 1B444 A.24 184311 m o W CA C9 g Al Al~D 1o2 1B43M 2o4 3o6 TIIK C lCKC)PALO VERDE P2 4os 6o0 o 0 AI 5 Al o.D 1o2 1B43K 2.4 3.6 TIIK<CXC)PALO VERDE g2 4os 6.0 Figure A-25.Load-time records for Specimens 1B43M and 1B43K A-26 1844L 1.2 2,4 3.6 TINE C N3EC)PALO VERDE g2 Batch:1B44L 4os 6.0 le2 1B456 2+4 3.6 TINE C NSEC>PALO VERDE g2 4+8 6m 0 Figure A-26.Load-time records for Specimens 1B44L and 1B456 A'-26 PALO VEROE NR 18418 m 1oe P 1B41B Ro4 3.6 TItK C!ISEC)PALO VERDE g2 4,8 6.0 m I ol CI~9 1 R 1BD3Y Ro4 3o6 TItK (INC)PALO VERDE g2 4as 6o0 Figure A-27.Load-time records for Specimens 1B41B and 1BD3Y A 27 18043 g~a Ct~D 1,2 1BD43 2o4 3+6 TIIK (NSEC>PALO VERDE 82 4os 6.0 CO 4)C9'D 1.2 1BD4P 2.4 3.6 fI K (tSEC)PALO VERDE g2 4.8 6+0 Figure A-28.Load-time records for Specimens 1BD43 and 1BD4P PALO VERDE Ne CO I CU~D 1BD35 8.4 3.6 TIIK C IISEC>PALO VERDE g2 4.8 6.0 mg 0 R g CU CCC O.D 1BD2D ee4 3.6 TIttL<NSEC>PALO VERDE g2 4o8 6+0 Figure A-29.Load-time records for Specimens 1BD35 and 1BD2D A-29 0)u)~D lo2 1BD52 2+4 3.6 Tlm (eXC)PALO VERDE f2 6 0 O oD le2 1BD3E 2e4 3+6 TllK C tCKC)PALO VERDE g2 4oe 6o0 Figure A-30.Load-time records for Specimens 1BD52 and 1BD3E A<0 g~o CI.D I~2 1BD2S 2j4 3j6 TIE<NSEC)PALO VERDE g2 4je 6j0 I cU CQ jD.Ioe 1BD2M 2j4 3,6 TIlK (ICKC>PALO VERDE g2 4e8 600 Figure A-31.Load-time records for Specimens 1BD2M and 1BD2B 1B151 2 3 4 5 6 7 8 Time (msec)9 10 Specimen Temperature Available Energy initial Velocity 1B'i51 0 F 256 in-lbs 57.4 in/sec Time to Yield ield Load Time to Maximum Maximum Load 0.165 mSec'040 lbs*0.170 mSec 1047 lbs Energy at Max.Load 3.7 in-lbs Crack Length 0.176 in 0.447 KID 38.2 ksi-in"1/2 Yield Stress 72.2 ksi*Specimen Compliance 45.6 Machine Compliance 63.7*No General Yielding Figure A-32.Load-time record and data for precracked specimen 1B 151 A42 1B14B 1200 600 200 2 3 4 5 6 7.8 9 10 Time (msec)0 Specimen Temperature Available Energy Initial Velocity 1B'14B 25 F 256 in-lbs 57.5 in/sec mme to Yield Yield Load mme to Maximum Maximum Load Energy at Max.Load 0.180 mSec'165 Ibs*0.180 mSec 1172 Ibs 5.1 in-Ibs Crack Length 0.180 in 0.457 KID 44.0 ksi-in"1/2 Yield Stress 83.9 ksi*Specimen Compliance 47.3 Machine Compliance 57.1*No General Yielding Figure A-33.Load-time record and data for precracked specimen IB14B 0 XB116 1200 1'.2 3 4 5 6 Time{msec)7 8 9 10 Specimen Temperature Available Energy initial Velocity 18116 50 F 399 in-lbs 71.7 in/sec Time to Yield iefd Load Time to Maximum Maximum Load 0.145 mSec'200 Ibs*',170 mSec 1235 lbs Energy at Max.Load 7.3 in-Ibs Crack Length 0.176 in OA47 KID 45,0 ksi-in"1/2 Yield Stress 83.3 ksi*Specimen Compliance 45.6 Machine Compliance 53.1'o General Yielding Figure A-34.Load-time record and data for precracked specimen 1B116 132.4<7 1600 1200 0 1 2 3 4 5 6 7 6 Time (msec)9 10 Specimen Temperature Available Energy initial Velocity 71 F 398 in-lbs 71.6 inlsec ime to Yield ield Load Time to Maximum Maximum Load Energy at Max.Load 0.157 mSec 1040 Ibs 0.300 mSec 1066 Ibs 15,5 in-Ibs Crack Length 0.213 in 0.541 KJD 102.9 ksi-in"1/2 Yield Stress 104.8 ksi Specimen Compliance 66,4 Machine Compliance 58.4 Figure A-35.Load-time record and data for precracked specimen 1B14J 0 1600 0 1 2 3 4 5 6 7 Time (msec)8 9 10 Specimen Temperature Available Energy Initial Velocity 1B113 76 F 398 in-lbs 71.6 in/sec ime to Yield ield Load Time to Maximum Maximum Load Energy at Max.Load 0.150 mSec 1120 Ibs 0,860 mSec 1285 Ibs 64.4 in-lbs Crack Length Specimen Compliance Machine Compliance 0.189 in 0.480 51.6 60.7 KJD 210.9 ksi-in"1/2 Yield Stress 87.9 ksi Pigure A-36.Load-time record and data for precraeked specimen IB113 1B11K 2 3 4 5 6 Time (msec)7 8 9 10 0 Specimen Temperature Available Energy Initial Velocity 1B11K 100 F 637 in-Ibs 90.6 in/sec mme to Yield ietd Load Time to Maximum Maximum Load Energy at Max.Load 0.135 mSec 1200 Ibs 0.560 mSec 1381 Ibs 54.2 in-Ibs Crack Length 0.186 in 0.472 KJD Yield Stress 187,4 ksi-in"1/2 91.5 ksi Specimen Compliance 50.1 Machine Compliance 71.6 Figure A-37.Load-time record and data for precracked specimen 1BllK 0 1600 2 3 4 5 6 Time (tnsec)7 8 9 10 Specimen Temperature Available Energy Initial Velocity 1B13M 150 F 642 in-lbs 90.9 in/sec ime to Yield ield Load Time to Maximum Maximum Load Energy at Max.Load 0.140 mSec 1100 Ibs 1.035 mSec 1341 Ibs 101.6 in-Ibs Crack Length Specimen Compliance Machine Compliance 0.189 in 0.480 51.6 83.7 KJD Yield Stress 263.6 ksi-in"1/2 86.4 ksi Figure A-38.Load-time record and data for precracked specimen 1B13M A-38 1B12L 1200 0 1 2 3 4 5 6 7 Time (msec)8 9 10 0 Specimen Temperature Available Energy initial Velocity 1B12L 200 F 640 in-Ibs 90.8 in/sec Time to Yield Yield Load Time to Maximum Maximum Load Energy at Max.Load 0.145 mSec 1170 Ibs 1.135 mSec 1499 lbs 122,1 in-Ibs Crack Length 0.185 in 0,470 Yield Stress 285.0 ksi-in"1/2 88.4 ksi Specimen Compliance 49,6 Machine Compliance 81.0 Figure A-39.Load-time record and data for precracked specimen 1B12L 0 A49 1B141 1200 2 3 4 5 6 7 Time (msec)8 9 10 Specimen Temperature Available Energy initial Velocity 1B141 250 F 640 in-lbs 90.8 in/sec ime to Yield Yield Load Time to Maximum Maximum Load 0.105 mSec 940 lbs 1.105 mSec 1375 Ibs Energy at Max.Load 108.0 in-Ibs Crack Length Specimen Compliance Machine Compliance 0,181 in 0.459 47;8 70.0 KJD Yield Stress 266.0 ksi-in"1/2 68.4 ksi Figure A-40.Load-time record and data for precracked specimen 1B141 AAO 1B213 800 O 1 2 3 4 5 6 7 8 9 10 Time (msec)Specimen Temperature Available Energy Initial Velocity 18213 25 F 256 in-lbs 57.4 in/sec Time to Yield ield Load Time to Maximum Maximum Load 0.218 mSec*1380 Ibs*0.225 mSec 1395 Ibs Energy at Max.Load 8.7 in-Ibs Crack Length 0.175 in 0.444 KID 50.4 ksi-in*1/2 Yield Stress 95.0 ksi*Specimen Compliance 45.2 Machine Compliance 61.4*No General Yielding Figure A<1.Load-time record and data for precracked specimen 13213 i 132 6C 1200 800 O 0 1 2 3 4 5 6 Time (msec)7 8 9 10 Specimen Temperature Available Energy Initial Velocity 1826C 50 F 401 in-lbs 71.9 inlsec Time to Yield Yield Load Time to Maximum Maximum Load Energy at Max.Load 0.155 mSec'230 Ibs'.185 mSec 1306 Ibs 8.1 in-lbs Crack Length 0.187 in 0,475 KID Yield Stress 51.8 ksi-in"1/2 94.7 ksi*Specimen Compliance 50.6 Machine Compliance 55.2*Mo General Yielding Figure A-42.Load-time record and data for precracked specimen 1826C A<2 XB23T 1400 1200 1000 0 1 2 3 4 5 6 7 Time (msec)9$0 Specimen emperature Available Energy Initial Velocity 1B23T 76 F 397 in-Ibs 71.5 in/sec Time to Yield Yield Load Time to Maximum Maximum Load 0.155 mSec 1200 Ibs 0.455 mSec 1354 Ibs Energy at Max.Load 32.1 in-lbs Crack Length 0.173 in 0.439 KJD Yield Stress 137.1 ksi-in"1/2 81.1 ksi Specimen Compliance 44.4 Machine Compliance.
60.5 Pigure A<3.Load-time record and data for precracked specimen 1B23T A43 1B27B 1600 1400 1200 1000 Xl 800 0 1 2 3 4 5 6 7 8 Time (msec)9 10 Specimen Temperature Available Energy initial Velocity 18278 100 F 639 in-Ibs 90.7 in/sec Time to Yield Yield Load Time to Maximum Maximum Load Energy at Max.Load 0.110 mSec 1060 lbs 0.295 mSec 1186 Ibs 22.6 in-Ibs Crack Length 0.189 in 0.480 KJD Yield Stress 118.1 ksi-in'1/2 83.2 ksi Specimen Compliance 51.6 Machine Compliance 61.1 Pigure A~.Load-time record and data for precracked specimen 1B27B 1B22T 1600 1200 800 o 0 1 2 3 4 5 6 7'tlme (msec)8 9 10 Specimen emperature Available Energy Initial Velocity 150 F 638 in-Ibs 90.6 io/sec Time to Yield Yield Load Time to Maximum Maximum Load Energy at Max.Load 0.150 mSec 1260 Ibs 0,560 mSec 1434 Ibs 54.4 in-lbs Crack Length 0.186 in 0.472 KJD Yield Stress 185.7 ksi-in"1/2 96.1 ksi Specimen Compliance 50.1 Machine Compliance 73.5 Pigure A<5.Load-time record and data for precracked specimen 1B22T 1B245 1600 1200 8N O 0 1 2 3 4 5'7 8 9 10 Time (msec)Specimen Temperature Available Energy Initial Velocity 1B245 200 F 636 in-lbs 90.5 in/sec Time to Yield Yield Load Time to Maximum Maximum Load 0.112 mSec 1080 lbs 0.950 mSec 1365 lbs Energy at Max.Load 92.1 in-lbs Crack Length Specimen Compliance Machine Compliance 0.194 in 0.492 54.2 64,1 KJD 254.2 ksi-in"1/2 Yield Stress 89.1 ksi Figure A<6.Load-time record and data for precracked specimeri 1B245 A46 1B226 1400$200 800 o 0 i 2 3 4 5 6 Time (msec)7 8 9 10 0 Specimen Temperature Available Energy Initial Velocity 1B226 250 F 637 in-lbs 90.6 in/sec ime to Yield Yield Load ime to Maximum Maximum Load Energy at Max.Load 0.113 mSec 880 lbs 1.'145 mSec 1398 Ibs 112.4 in-lbs Crack Length Specimen Compliance Machine Compliance 0.184 in 0.467 49.2 86.3 KJD 271.6 ksi-in"1/2 Yield Stress 65.9 ksi Figure A-47.Load-time record and data for precracked specimen 1B226 0 1B3AT 1200 800 O 2 3 4 5 6 7 8 Time (msec)9$0 Specimen Temperature Available Energy Initial Velocity-183AT-50 F 642 in-lbs 90.9 in/sec Time to Yield Yield Load Time to Maximum Maximum Load Energy at Max, Load 0.111 mSec*990 Ibs" 0.115 mSec 995 lbs 4.2 in-lbs Crack Length 0.191 in OA85 KID Yield Stress 40.7 ksi-in"1/2 79.3 ksi*Specimen Compliance 52,6 Machine Compliance 68.8'o General Yielding Hgure A<8.Load-time record and data for precracked specimen lB3AT 1B31U 1200 2 3 4 5 6 7 8 9 10 Time (msec)Specimen Temperature Available Energy Initial Velocity 1831U-25 F 257 in-Ibs 57.5 in/sec ime to Yield ield Load ime to Maximum Maximum Load Energy at Max.Load 0.157 mSec*1030 lbs*0.175 mSec 1085 Ibs 4.7 in-lbs Crack Length 0.182 in 0.462 KID 41.4 ksi-in"1/2 Yield Stress 75.6 ksi*Specimen Compliance 48.2 Machine Compliance 56.2'o General Yielding Figure A<9.Load-time record and data for precracked specimen 1B31U A49 2.B323 1600 1200 800 O 1 2 3 4 5 6%me (msec)7 8 9'0 Specimen Temperature Available Energy Initial Velocity'IB323 0 F 397 in-lbs 71.5 in/sec ime to Yield ield Load Time to Maximum Maximum Load 0.154 mSec*1210 Ibs*0.155 mSec 1215 Ibs Energy at Max.Load 5.5 in-lbs Crack Length 0.186 in 0.472 KlD 47.8 ksi-in"1/2 Yield Stress 92.3 ksi*Specimen Compliance 50.1 Machine Compliance 56.6*No General Yielding Pigure A-50.Load-time record and data for precracked specim'en 1B323 A40 XB33iX 1200 0 1 2 3 4 5 6 7 8 9 10 Time (msec)0 Specimen Temperature Available Energy Initial Velocity 1B33J 50 F 397 in-Ibs 71.5 in/sec Time to Yield Yield Load Time to Maximum Maximum Load 0.150 mSec 1240 Ibs 0.550 mSec 1390 Ibs Energy at Max.Load 42.0 in-lbs Crack Length Specimen Compliance Machine Compliance 0.180 in 0.457 47.3 54.0 KJD 163.5 ksi-in"1/2 Yield Stress 89.4 ksi Figure A-51.Load-time record and data for precracked specimen 1'B33J 1B3 ZA 0 1 2 3 4 5 6 7 Time (msec)8 9 10 Specimen emperature Availabte Energy Initial Velocity 1B31A 71 F 400 in-Ibs 71.8 in/sec Time to Yield ield Load Time to Maximum Maximum Load Energy at Max.Load 0.150 mSec 1240 Ibs 1.035 mSec 1456 lbs 86,2 in-Ibs Crack Length Specimen Compliance Machine Compliance 0.174 in 0.442 44.8 56.6 KJD 236.4 ksi-in"1/2 Yield Stress 84.5 ksi Figure A-52.Load-time record and data for precracked specimen 1B31A A-62 1B33Y 1200 600 O 0 1 2 3 4 5 6 Time (msec)7 8 9 10Specimen Temperature Available Energy initial Velocity IB33Y 636 in-Ibs 90.5 in/sec ime to Yield Yield Load ime to Maximum Maximum Load Energy at Max.Load 0.140 mSec 1200 Ibs 0.720 mSec 1286 Ibs 70.0 in-lbs Crack Length Specimen Compliance Machine Compliance 0.184 in 0,467 49.2 76.0 KJD 215.5 ksi-in"1/2 Yield Stress 89.8 ksi Figure A-53.Load-time record and data for precracked specimen 1B33Y A43 1B36H 1600'2 3 4 5 6 7 8 9 10 Time (msec)Specimen Temperature Available Energy initial Yelocity 1836M 125 F 3201 in-Ibs 203.0 in/sec Time to Yield Yield Load Time to Maximum Maximum Load Energy at Max, Load 0.060 mSec.950 Ibs 0.300 mSec 1223 lbs 56.4 in-Ibs Crack Length Specimen Compliance Machine Compliance 0.189 in 0.480 51.6 98.5 KJD 191.7 ksi-in"1/2 Yield Stress 74.6 ksi Figure A-54.Load-time record and data for pre'cracked specimen LB36M 1B3 1K 1600 1400 200 0 2 3 4 5 6 7 Time (msec)8 9 10 Specimen Temperature Available Energy Initial Velocity 1B31K 150 F 642 in-Ibs S0.9 in/sec Time to Yield Yield Load Time to Maximum Maximum Load Energy at Max.Load 0.110 mSec 1000 Ibs 0.860 mSec 1267 Ibs 80.0 in-Ibs Crack Length Specimen Compliance Machine Compliance 0.187 in 0,475 50.6 64.3-KJD 233.8 ksi-in"1/2 Yield Stress 77.0 ksi Figure A-55.Load-time re'cord and data for precracked specimen lB31K A@6 1B37A f200 0 1 2 3 4 5 6 7 Time (msec)8 9 10 Specimen Temperature Available Energy Initial Velocity 1837A 200 F 637 in-lbs 90,6 in/sec Time to Yield Yield Load Time to Maximum Maximum Load Energy at Max.Load 0,110 mSec 1000 Ibs 0.715 mSec 1319 lbs 67.6 in-Ibs Crack Length Specimen Compliance Machine Compliance 0,186 in 0.472 50.1 63.4 KJD Yield Stress 211.8 ksi-in"1/2 76.3 ksiFigure A-56.Load-time record and data for precracked specimen 1B37A 0'