ML17310B247

Nonproprietary Analysis of 137 Degree Capsule from APS Palo Verde Unit 2 Reactor Vessel Radiation Surveillance Program.

ML17310B247
Person / Time
Site:	Palo Verde
Issue date:	02/28/1994
From:	Lippincott E, Madeyski A, Terek E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML17310B245	List: ML17310B244 ML17310B247
References
WCAP-13935, NUDOCS 9405020098
Download: ML17310B247 (174)

Text

WESTINGHOUSE CLASS 3 (Non-Prorietary)

W CAP-13935

'2405020098 '.940415

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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 J. M. Chicots J.<(. CIVCc Section 6 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 5.1 Overview 5-1 5.2 Charpy V-Notch Impact Test Results 5-5 5.3 Precracked Charpy Specimen Test Results 5-7 5.4 Tension Test Results 5-8 6.0 RADIATIONANALYSIS AND NEUTRON DOSIMETRY 6-1 6.1 Introduction 6-1 6.2 Discrete Ordinates Analysis 6-2 6.3 Neutron Dosimetry 6-6 6.4 Projections of Pressure Vessel Exposure 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 5-10 Irradiated at 550'F to a Fluence of 4.071 x 10" n/cm (E > 1.0 MeV)

(Transverse Orientation) 5-3 Charpy V-notch Data for the Palo Verde Unit 2 Surveillance Weld Metal 5-11 Irradiated at 55Q F to a Fluence of 4.Q71 X 10'/cm (E > 1.0 MeV)

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 5-16 Surveillance Weld Metal Irradiated at 550'F to a Fluence of 4.071 x 10" n/cm'E > 1.0 MeV) 5-9 Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 5-17 Surveillance Heat-Affected-Zone (HAZ) Metal Irradiated at 550'F to a Fluence of 4.071 x 10'/cm (E > 1.0 MeV) 5-10 Instrumented Charpy Impact Test Results for the Palo Verde Unit 2 5-18 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) 5-11 Effect of 550'F Irradiation to 4.071 x 10" n/cm (E > 1.0 MeV) on the 5-19 Notch Toughness Properties of the Palo Verde Unit 2 Reactor Vessel Surveillance Materials 5-12 Comparison of the Palo Verde Unit 2 Surveillance Material 30 ft-lb 5-20 Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions 5-13 Precracked Instrumented Charpy Impact Test Results for the Palo Verde 5-21 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) 5-14 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 5-23 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 Tensile Properties of the Palo Verde Unit 2 Reactor Vessel Surveillance 5-24 Materials Irradiated at 550'F to 4.071 x 10" n/cm'E > 1.0 MeV) 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 6-14 Pressure Vessel Clad/Base Metal Interface 6-3 Relative Radial Distribution of )(E > 1.0 MeV) within the Pressure Vessel 6-15 Wall Relative Radial Distribution of $ (E > 0.1 MeV) within the Pressure Vessel 6-16 Wall 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 6-19 Palo Verde Unit 2 Reactor 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 6-21 Surveillance Capsule Center Surveillance Capsule W137

LIST OF TABLES (continued)

Table Title 6-11 Adjusted Neutron Energy Spectrum at the Center of Surveillance Capsule 6-22 W137 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 6-24 Clad/Base Metal Interface 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 4-10 Compartment Assembly 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 5-26 Lower Shell Plate F-773-1 (Longitudinal Orientation) 5-3 Charpy V-Notch Impact Properties for Palo Verde Unit 2 Reactor Vessel 5-27 Lower Shell Plate F-773-1 (Transverse Orientation)

Charpy V-Notch Impact Properties for Palo Verde Unit 2 Reactor Vessel 5-28 Surveillance Weld Metal (F-773-2/F-773-3) 5-5 Charpy V-Notch Impact Properties for Palo Verde Unit 2 Reactor Vessel 5-29 Weld Heat-Affected-Zone Metal 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 5-31 Vessel Lower Shell Plate F-773-1 (Longitudinal Orientation)

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 5-33 Vessel Surveillance Weld Metal 5-10 Charpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Reactor 5-34 Vessel Heat-Affected-Zone Metal 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 5-36 Values Determined by Testing of Precracked Charpy Specimens from Lower Shell Plate F-773-1 (Longitudinal Orientation) 5-13 Comparison of Unirradiated and Irradiated Dynamic Fracture Toughness 5-37 Values Determined by Testing of Precracked Charpy Specimens from Lower Shell Plate F-773-1 (Transverse Orientation) 5-14 Comparison of Unirradiated and Irradiated Dynamic Fracture Toughness 5-38 Values Determined by Testing of Precracked Charpy Specimens from the Palo Verde Unit 2 Surveillance Weld Metal 5-15 Precracked Charpy Impact Specimen Fracture Surfaces for Palo Verde 5-39 Unit 2 Lower Shell Plate F-773-1 (Longitudinal Orientation) 5-16 Precracked Charpy Impact Specimen Fracture Surfaces for Palo Verde 540 Unit 2 Lower Shell Plate F-773-1 (Transverse Orientation) e vnI

LIST OF ILLUSTRATIONS

~Fi ure Title ~Pa e 5-17 Precracked Charpy Impact Specimen Fracture Surfaces for Palo Verde 5<1 Unit 2 Reactor

'I Vessel Surveillance Weld Metal 5-18 Tensile Properties for Palo Verde Unit 2 Reactor Vessel Lower Shell 5<2 Plate F-773-1 (Transverse Orientation) 5-19 Tensile Properties for Palo Verde Unit 2 Reactor Vessel Surveillance 5<3 Weld Metal 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 5-45 Surveillance Weld Metal 5-22 Engineering Stress-Strain Curves for Lower Shell Plate F-773-1 Tensile 5-46 Specimens 1B2J2 and 1B2K1 (Transverse Orientation) 5-23 Engineering Stress-Strain Curve for Lower Shell Plate F-773-1 Tensile 5<7 Specimen 1B2J3 (Transverse Orientation) 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 6-29 ix

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 Melting Point: 536'F (280'C) 90% Pb, 5% Sn, 5% Ag Melting Point: 558'F (292'C) 2.5% Ag, 97.5% Pb Melting Point: 580'F (304'C) 1.75% Ag, 0.75% Sn, 97.5% Pb 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 Weld Metal

. Plate F-773-1 F-773-2/F-773-3 Si 0.21 0.47 0.009 0.011 0.006 0.010 1.54 1.46 0.24 0.12 Cr 0.03 0.10 Ni 0.68 0.09 Mo 0.52 0.51 0.003 0.005 Cb <0.01 <0.01 B <0.001 <0.001 Co 0.016 0.010 0.07 0.015 0.005 0.010 0.010

<0.01 <0.01 As 0.021 0.010 Sn 0.003 0.004 Zf <0.001 <0.001 0.014 0.011 Sb 0.0018 0.0114

TABLE 4-2 Heat Treatment of the Palo Verde Unit 2 Reactor Vessel Surveillance Materials"'aterial Temperature ('F) Time (hr) Coolant Surveillance Austenitizing: Water quenched Program 1600 + 25 Test Plate F-773-1 Tempered: Air cooled 1225 + 25 Stress Relief: 40 Furnace cooled 1150+ 50 to 600'F Weldment Stress Relief: 44 hr. Ec 48 min. Furnace cooled 1125 + 25 to 600'F

TABLE 4-3 Summary of Unirradiated Surveillance Material Data"'aterial RT Yield CUpper 30 ft-lb 50 ft-lb 35 Mils Lat. Strength (ksi) and Code Shelf Index Index Exp. Index NDTT RT

('F) ('F) ('F) ('F) ('F) Static Dynamic (ft-lb)

Base Metal Plate 136.5 30 +0 -10 61.7 107.7 F-773-1 (WR)

(Transverse)

Base Metal Plate 112 35 12 -40 10 64.1 91.4 F-773-1 (RW)

(Longitudinal)

Weld Metal 109 -43 -41 -70 -50 63.3/62.2 85.0 F-773-2/F-773-3 HAZ Metal 84 -12 57 -30 50 F-773- 1/F-773-2 SRM HSST 136 34 14 Plate 01MY (RW)

(Longitudinal)

TABLE4-4 Arrangement of Encapsulated Test Specimens by Code Number within the Palo Verde Unit 2 137'apsule'"

Compartment Compartment Specimen Position Number Numbers 18312 1823L 1823P 18243 1826D 1822E 182AE 1824J 1823M 18267 1822M 18224 182AM 18255 1822U 1821U 18322-1 18278 18213 1823T 18226 1826C 18245 18275 1825Y 1822T 1814J 1811K 18116 18141 1813M 18148 18151 18113 1812L 18332-4 182KI 182J3 182J2 1844D 18438 1843K 18444 18432 1843Y 18418 1844L 1843T 18411 18456 1843M 1834D-3 IBD43 IBD35 IBD52 IBD4P IBD2D IBD3E IBD2M IBD3Y IBD28 183J5 183JY 183J7 18353 1835E 1834U 183AL 1834E 1835P 1835M 183A3 1833E 183AD 18318 18316 1832A 18342 18326 18312 18363-1 1833Y 1836M 1833J 1837A 1831U 183AT 1831A 1831K 18323 1811M 1811E 18124 1814Y 18145 1815D 1814D 18128 18136 Material IB IXX Lower Shell Plate F-773-1 (Longitudinal Orientation) 182XX Lower Shell Plate F-773-1 (Transverse Orientation) 183XX Weld Metal 184XX Heat-Affected-Zone Material IBDXX SRM HSST 01MY Material (Longitudinal Orientation) 4-6

VESSEL Sao'OUTLETIIOZELE i3~'ESSEL VESSEL 142o I

IIP zaao i NOZZLE SHROUD O I COBE EUPPOBT BARREL I I

REACTOR VESSEL CORE VESSEL A MIDPLANE CAPSULE 4 ato'NLET ASSEMBLY C

aa'ORE \

l E

0 /

I/

I CORE SUPPORT BARREL B.

REACTOR VESSEL VESSEL VESSEL 43 VESSEL 8

ENLARGED PLAN VIEW ELEVATION VIEW I I I

oo

Lock Assembly Wedge Coupling Assembly Charpy and Flux Compartment Ass embly or Charpy, Flux, and Connecting Spacer Compact Tension Compartment Ass embly 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

I l

) 1 l

)

Charpy Impact Specimens Flux Monitor Housing Connecting Spacer Precracked Charpy and/or Charpy Impact Specimens 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 Spectrum Monitor Cadmium Shielded Flux Monitor Housing Stainless Steel Tubing Cadmium Shield Stainless Steel Tubing Threshold Detector Threshold Detector Quartz Tubing Flux Spectrum Monitor Temperature Monitor Vfeight Temperature Monitor Housing l mv Melting Alloy Chary Impact Specimens Tension Specimens and Tension Specimen Housing Rectangular Tubing 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 a is in units of psi and Po is 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 0

(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.

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 Temperature Impact Energy Lateral Expansion Shear Number ('C) (ft-lb) (mils) (mm) (%)

1B11E -5 -21 28 38 30 0.76 10 1B124 10 -12 27 37 26 0.66 20 1BI IM 25 47 43 1.09 25 1B15D 50 10 33 45 33 0.84 30 1B 14Y 75 24 58 79 54 1.37 50 1B 136 115 46 82 70 1.78 70

'1B145 150 66 92 77 1.96 85 1B14D 93 108 146 79 2.01 1B12B 250 121 127 172 94 2.39 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 Temperature Impact Energy Lateral Expansion Shear Number ('C) (ft-lb) (mils) (mm) (%)

1B255 -32 16 22 15 0.38 1B23L -18 24 33 24 0.61 10 1B2AE 15 30 41 29 0.74 20 1B224 62 84 54 1.37 1B267 35 42 57 38 0.97 30 1B23P 50 10 58 45 1.14 40 1B2AM 65 18 58 79 50 1.27 40 1B26D 75 43 58 41 1B21U 85 29 63 85 55 50 1B22U 38 67 91 57 1.45 60 1B22E 125 52 73 99 67 1.70 70 IB24J 150 66 105 142 89 2.26 95 1B243 93 116 157 83 2.11 95 1B22M 250 121 115 156 90 2.29 1B23M 149 115 156 92 2.34 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) ('C) (ft-lb) (mils) (mm) 1B3AD -68 13 16 0.41 10 1B316 -80 -62 38 52 38 0.97 15 1B35E -50 12 16 15 0.38 15 1B342 -30 28 38 36 0.91 25 1B34U -20 -29 33 33 0.84 30 1B312 -10 -23 53 72 51 1.30 35 1B34E -18 50 68 46 1.17 35 1B35P 25 4 57 77 56 1.42 50 1B 326 50 10 88 119 78 1.98 70 1B3A3 75 24 75 102 70 1.78 80 1B3AL 100 93 125 81 2.06 90 1B 35M 125 52 92 126 85 2.16 95 1B33E 150 66 106 95 2.41 1B31B 200 93 107 145 86 2.18 1B32A 250 121 112 152 95 2.41 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 Temperature Impact Energy Lateral Expansion Shear Number ('F) ('C) (ft-lb) (mils) (mm) (%)

1B411 -35 -37 16 22 16 0.41 1B44D -18 20 27 24 0.61 10 1B43B 25 20 27 15 0.38 15 1B43Y 35 28 38 29 0.74 20 1B432 50 10 18 24 0.61 30 1B43T 60 16 49 66 47 1.19 1B444 75 24 71 96 59 1.50 50 1B43M 38 86 117- 71 1.80 65 IB43K 120 108 146 82 2.08 80 1B44L 155 68 77 70 1.78 95 1B456 93 133 180 87 2.21 1B41B 250 121 92 125 79 2.01 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 Temperature Impact Energy Lateral Expansion Shear Number ('F) ('C) (ft-lb) (mils) (mm) (%)

1BD3Y '18 0.20 1BD43 50 10 17 23 20 0.51 15 1BD4P 38 32 43 30 0.76 20 1BD35 150 66 45 61 46 1.17 40 1BD2D 93 35 47 32 0.81 35 1BD52 93 57 77 53 1.35 60 1BD3E 107 88 119 75 1.91 85 1BD2M 275 135 141 84 2.13 100 1BD2B 350 177 106 87 2.21 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-Ibfjn~)

Time Time Fast Charpy Yield to Max. to Fract. Arrest Yield Flow Test Energy Chatty Max. Prop. Load Yield Load Max. Load Load Stress Stress Sample Temp. En En/A EJA Ep/A PGY tm Phl 4< PF PA Gr (ksi)

No. ('F) (ft-lb) (Ibs) (ijsec) (lbs) (p sec) (lbs) (lbs) (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 47 378 278 101 3080 0.17 3959 0.68 3936 902 102 117 1B15D 50 33 266 154 112 3019 0.16 3750 0.42 3740 1276 100 112 1B14Y 75 58 467 280 187 2931 0.14 3969 0.68 3931 1707 97 115 1B 136 115 82 660 343 317 2828 0.15 3936 0.82 3643 2115 94 112 1B 145 150 92 741 272 469 2743 0.19 3724 0.72 2897 1628 91 107 1B14D 200 108 870 245 624 2427 0.14 3569 0.68 81 100 1B12B 250 127 1023 322 700 2499 0.14 3731 0.82 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')

Time Time Fast Charpy Yield to Max. to Fract. Arrest Yield Flow Test Energy Charpy Max. Prop. Load Yield Load Max. Load Load Stress Stress Sample Temp. ED EG/A EJA EF/A PGY tGY Phk PF aY (ksi)

No. ('F) (ft-Ib) (lbs) (psec) (lbs) (p sec) (lbs) (lbs) (ksi) 1B255 -25 16 129 89 40 3350 0.17 3648 0.28 3632 141 111 116 1B23L 0 24 193 147 46 3342 0.17 3869 0.40 3865 241 120 1B2AE 15 30 242 165 77 3136 0.16 3877 OA4 3864 104 , 116 1B224 25 62 499 359 104 3083 0.16 4094 0.82 3988 1116 102 119 1B267 35 42 338 252 86 2987 0.16 3881 0.63 3872 1022 99 114 1B23P 50 43 346 . 217 130 2931 0.17 3805 0.57 3792 1538 97 112 1B2AM 65 58 467 351 116 2948 0.16 4036 0.82 4023 1511 98 116 1B26D 75 43 346 197 149 2951 0.16 3873 0.52 3867 1900 98 113 1B21U 85 63 507 357 151 2922 0.14 4062 0.82 4053 2286 97 116 IB22U 100 67 540 279 261 2914 0.16 3986 0.69 3893 2155 97 115 1B22E 125 73 588 261 327 2639 0.14 3704 0.69 3663 2285 88 105 1B24J 150 105 845 260 586 2615 0.15 3693 0.69 2660 1858 87 105 1B243 200 116 934 313 621 2606 0.16 3774 0.80 87 106 1B22M 250 115 926 247 679 2371 0.14 3572 0.69 79 99 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~)

Time Time Fast Charpy Yield to Max. to Fract. Arrest Yield Flow Test Energy Charpy Max. Prop. Load Yield Load Max. Load Load Stress Stress Sample Temp. ED ED/A E JA EF/A PGY tGY Phk th~ PF aY (ksi)

No. ('F) (ft-Ib) (lbs) (psec) (lbs) (p sec) (lbs) (Ibs) (ksi) 1B3AD -90 13 105 63 42 3630 0.17 3681 0.21 3668 561 121 121 1B316 -80 38 306 210 96 3237 0.17 3810 0.54 3804 1644 108 117 1B35E -50 12 97 43 54 3142 0.14 3463 0.18 3457 415 104 110 1B342 -30 28 225 149 76 3268 0.16 3767 0.41 3748 1384 109 117 1B34U -20 266 172 94 3301 0.18 3862 0.46 3846 1556 110 119 1B312 -10 53 427 268 159 2961 0.15 3734 0.67 3673 1915 98 111 1B34E 0 50 403 269 134 3169 0.16 3878 0.65 3843 2864 105 . 117 1B35P 25 57 459 261 198 3042 0.16 3759 0.65 3746 2622 101 113 1B326 50 88 709 258 451 2758 0.14 3620 0.68 3277 2550 92 106 1B3A3 75 75 604 274 330 2974 0.16 3847 0.67 3443 2246 99 113 1B3AL 100 92 741 275 466 2922 0.15 3816 0.69 97 112 1B35M 125 93 749 269 480 2820 0.15 3751 0.69 94 109 1B33E 150 106 854 312 541 2586 0.14 3535 0.82 86 102 1B31B 200 107 862 301 561 2400 0.16 3451 0.82 80 97 1B32A 250 112 902 293 609 2318 0.16 3353 0.83 77 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')

Time Time Fast Charpy Yield to Max. to Fract. Arrest Yield Flow Test Energy Charpy Max. Prop. Load Yield Load Max. Load Load Stress Stress Sample Temp. Ep EJA EJA Ep/A PGY tm Phf PF Gy (ksi)

No. ('F) (ft-lb) (lbs) (p sec) (lbs) (psec) (lbs) (lbs) (ksi) 1B411 -35 .16 129 89 40 3356 0.17 3645 0.28 3641 170 111 , 116 IB44D 20 161 105 56 3331 0.16 3732 0.31 3729 536 111 117 1B43B 25 20 161 105 56 3136 0.16 3491 0.32 3487 928 104 110 1B43Y 35 28 225 155 70 3248 0.17 3797 0.42 3768 1136 108 117 1B432 50 18 145 69 76 3057 0.16 3250 0.25 3244 1287 102 105 1B43T 60 49 395 270 3122 0.16 3919 0.65 3909 1724 104 117 1B444 75 71 572 284 288 3041 0.14 4024 0.67 3905 2022 101 117 1B43M 100 86 692 284 409 3057 0.16 4014 0.69 3189 1540 102 117 IB43K 120 108 870 377 492 2997 0.25 4086 0.94 3006 1940 100 118 1B44L 155 77 620 255 365 2807 0.15 3720 0.65 2775 1772 93 108 1B456 200 133 1071 345 726 2836 0.16 3833 0.86 94 1B41B 250 92 741 249 492 2650 0.17 3554 0.68 88 103

• 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-lbfin~)

Time Time Fast Charpy Yield to Max. to FlacL Arrest Yield Flow Test Energy Charpy Max. Prop. Load Yield Load Max. Load Load Stress Stress Sample Temp. ED ED/A E JA EF/A PGY tGY Phf thi PF PA <r (ksi)

No. (F) (ft-lb) (lbs) (psec) (lbs) (psec) (lbs) (lbs) (ksi) 1BD3Y 0 56 29 27 3186 0.14 3289 0.15 3282 106 108 1BD43 50 17 137 96 41 3405 0.17 3783 0.29 3777 176 113 119 1BD4P 100 32 258 195 62 3263 0.17 4115 0.49 4096 521 108 123 1BD35 150 45 362 281 81 3096 0.16 4172 0.65 4156 1112 103 121 1BD2D 200 35 282 208 74 3090 0.16 4095 0.52 4085 845 103 119 1BD52 200 57 459 278 181 2965 0.15 4115 0.66 3994 1804 98 118 1BD3E 225 88 709 289 419 2966 0.16 4145 0.68 3485 2521 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 Average 30 (ft-lb) Average 35 mil Lateral Average 50 ft-lb Average Energy Absorption "

Temperature ('F)

'ransition Temperature ('F)'xpansion Temperature ('F)

'ransition at Full Shear (ft-lb)

Material Unirradiated Irradiated bT Unirradiated Irradiated hT Unirradiated Irradiated Unirradiated Irradiated b.

Plate F-773-1 10 10 12 25 13 35 60 25 112 1.18 (longitudinal)

Plate F-773-1 4 15 19 20 30 55 25 136.5 115 . - 21.5 (transverse)

Weld Metal -43 -28 15 -41 -33 -11 15 109 108 HAZ Metal -12 45 57 45 43 57 75 18 84 113 +29 SRM 01MY 120 117 14 155 141 34 185 151 136 105 -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 30 ft-lb Transition Upper Shelf Energy Fluence Temperature Shift Decrease (E) 1.0 MeV)

Predicted " "

Measured Predicted Measured Material (X 10" n/cm')

(F) ('F) (%) (%)

Lower Shell 4.071 19.5 10 15.5 Plate F-773-1 (longitudinal)

Lower Shell 4.071 19.5 19 15.5 16 Plate F-773-1 (transverse)

Weld Metal 4.071 31.5 15 17 HAZ Metal 4.071 57 SRM HSST 4.071 117 23 Plate 01MY (longitudinal)

(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 Test Specimem Initial Tllne Yield Time Max. Energy at Available Crack K,~ K,~ Yield No. Temp. Comp. Velocity to Load to Load Initiation Energy Length (ksi l in) (ksi l in) Stress

('F) (in/sec) Yield (lbs) Max. (Ibs) (in-lbs) (in-lbs) (in.) (ksi)

(psec) (psec) 1B151 45.6 57.4 170 1047 256 0.176 OA47 38.2 72.2 1B14B 25 47.3 57.5 180 1172 256 0.180 0.457 44.0 83.9 1B116 50 45.6 71.7 170 1235 399 0.176 0.447 45.0 83.3 1B14J 71 66A 71.6 157 1040 300 1066 15.5 398 0.213 0.541 102.9 104.8 1B113 76 51.6 71.6 150 1120 860 1285 64.4 398 0.189 0.480 210.9 87.9 1B11K 100 50.1 90.6 135 1200 560 1381 54.2 637 0.186 0.472 187.4 91.5 1B13M 150 51.6 90.9 140 1100 1035 1341 101.6 642 0.189 0.480 263.6 86.4 1B12L 200 49.6 90.8 145 1170 1135 1499 122.1 640 0.185 0.470 285.0 88.4 1B141 250 47.8 90.8 105 940 1105 1375 108.0 640 0.181 0.459 266.0 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 Test Specimem Initial Time Yield Time Max. Energy at Available Crack Kr~ Kr~ Yield No. Temp. Comp. Velocity to Load to Load Initiation Energy Length (ksi )in) (ksi l in) Stress

('F) (in/sec) Yield (ibs) Max. (lbs) (in-lbs) (in-lbs) (in.) (ksi)

(psec) (p sec) 1B213 25 45.2 57.4 225 1395 0.175 0.444 50.4 95.0 1B26C 50 50.6 71.9 185 1306 401 0.187 0.475 51.8 94.7 1B23T 76 44.4 71.5 155 1200 455 1354 32.1 397 0.173 0.439 137.1 81.1 1B27B 100 51.6 90.7 110 1060 295 1186 22.6 639 0.189 0.480 118.1 83.2 1B22T 150 50.1 90.6 150 1260 560 1434 54.4 638 0.186 0.472 185.7 96.1 1B245 54.2 90.5 112 1080 950 1365 92.1 636 0.194 0.492 254.2 89.1 1B226 250 49.2 90.6 113 880 1145 1398 112.4 637 0.184 0.467 271.6 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 Test Specimem Initial Time Yield Time Max. Energy at Available Crack Kra Kr~ Yield No. Temp. Comp. Velocity to Load to Load Initiation Energy Length (ksi1in) (ksi l in) Stress

('F) (in/sec) Yield (lbs) Max. (lbs) (in-lbs) (in-lbs) (in.) (ksi)

(p sec) (psec) 1B3AT -50 52.6 90.9 115 995 642 0.191 0.485 40.7 79.3 1B31U -25 48.2 57.5 175 1085 257 0.182 0.462 41.4 75.6 1B323 50.1 71.5 155 1215 397 0.186 0.472 47.8 92.3 1B33J 50 47.3 71.5 150 1240 550 1390 42.0 397 0.180 0.457 163.5 89.4 1B31A 71 71.8 150 1240 1035 1456 86.2 400 0.174 0.442 236.4 85.4 1B33Y 100 49.2 90.5 140 1200 720 1286 70.0 636 0.184 0.467 215.5 89.8 1B36M 125 51.6 203.0 60 950 300 1223 56.4 3201 0.189 0.480 191.7 74.6 1B 31K 150 50.6 90.9 110 1000 860 1267 80.0 642 0.187 0.475 233.8 77.0 1B37A 200 50.1 90.6 110 1000 715 1319 67.6 637 0.186 0.472 211.8 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) 0.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Sample Test Temp. Strength Strength Load Stress Strength Elongation Elongation in Area Material Number ('F) (ksi) (ksi) (kip) (ksi) (ksi) (%) (%) (%)

Plate F-773-1 1B2J2 75 63.9 84.5 3.65 206.5 74.4 13.5 29.1 64 (transverse)

Plate F-773-1 1B2K1 200 63.2 80.5 2.55 126.8 51.9 11.4 25.4 59 (transverse)

Plate F-773-1 1B2J3 550 54.0 81.5 2.80 130.9 57.0 12.3 24.8 56 (transverse)

Weld Metal 1B3J7 73.3 90.7 2.89 194.6 58.9 13.5 27.9 70 Weld Metal 1B3JY 125 70.3 83.5 2.75 141.2 56.0 10.8 23.4 60 Weld Metal 1B3J5 550 62.6 82.5 2.75 160.9 56.0 9.9 21.0 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 2

~

80 0 ~

60 0

cn 40 20 100 2,5 80 2.0 Z

60 1.5 OC 40 LO BF 20 0,5 0

160 200 140 120 160 100 120 80 60 80 LLI 40 0 8'F 40 20

-200 -100 0 100 200 300 400 500 TEMPERATURE ('F) 0 MRMIAIE9

~ DMIIATBCi50 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 2,5 2Q 80 0 2,0 60 1,5

>C 40 20'F 1,0 20 0.5 0

160 o 200 140 120 160 100 120 80 0

w 60 (0 ~

80 4J 0 2fF 40 19F 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 2,5 80 2,0 60 1.5 0

40 8'F 1,0

~~ 20 0,5 0

160 200 140 120 160 100 120 80 0 0 ~

5 60 80 LJ 40 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 2+ 0 80 60 0 0

40 20 100 2,5 80 2.0 60 1,5 o o OC 40 1.0

~

$3 F

~20 00 0,5

~ ~

0 160 200 140 120 160 100 0 120 80 4> 0 0

60 0 0 80 LaJ sr 40 0

0 40 20

-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 2+

80 60

~ 40 20 100 2.5 N

80 2.0 2

60 1.5 OC 40 1,0 20 0,5 0

160 200 140 120 160 100 120 80 60 80 5 tst F 40 Ill F 40 20 0

-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 t

"  ;.-1B136,

.',1B15D; 1B14Y 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

-P C ~Q '*

$ ~

t g>>

g W$

g ~-"~A

!," 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, . ."

1844D, 1B438 1843Y f

F 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

i 1BD3Y 1BD43 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 L 0 200 P KIR 150 Unirradiated hC tmtdiated 100

-100 100 200 T-RTNDT (F)

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 UnirradiaIed.

Irradiated 100

-100 100 200 T-RTNDT (F)

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

~ k F K IR

~

g 150 UnirradiaIed Irradiated

-100 200 T-RTNDT (F)

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

I g~*

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', ~ ';- 1837A f

I Figure 5-17 Precracked Chatpy Impact Specimen Fracture Surfaces for Palo Verde Unit 2 Reactor Vessel Surveillance Weld Metal 5%1

('C) 50 100 150 200 250 300 120 800 110 700 100 90 600 g 80 500 cn 70 60 400 50 300 40 80 70 60 8 50 0

~ 40 20 10 100 200 300 400 500 600 TENPERATURE ('F) 40 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

('C>

50 100 150 200 250 300 120 800 110 700 100 90 600 80 500 cn 70 60 400 50 300 40 80 70 60 8 50 0

~ 40

~

a30 10 100 200 300 400 500 600 TEHPERATURE ('F) 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

STRESS-STRAIN CURVE PALO VERDE UNIT 2 100.00 90.00 80.00 70.00 80.00 50.00 s 40.00 30.00 1B 2J2 20.00 75 F 10.00 0.00 0.0 0.10 0.20 0.30 STRAIN, IN/IN STRESS-STRAIN CURVE PALO VERDE UNIT 2 100.00 90.00 80.00 70.00 CO eO.00 CO 50.00 40.00 30.00 20.00 1B 2K1 10.00 200 F 0.00 0.0 0.10 0.20 0.30 STRAIN, IN/IN Figure 5-22 Engineering Stress-Strain Curves for Lower Shell Plate F-773-1 Tensile Specimens 1B2J2 and 1B2K1 (Transverse Orientation) 546

STRESS-STRAIN CURVE, PALO VERDE UNIT 2 100.00 90.00 80.00 70.00 CO 8O.OO V) 50.00 CC 40.00 30.00 1B 2J3 20.00 550 F 10.00 0.00 0.0 0.10 0.20 0.30 STRAIN, IN/IN Figure 5-23 Engineering Stress-Strain Curve for Lower Shell Plate F-773-1 Tensile Specimen 1B2J3 (Transverse Orientation) 5<7

STRESS-STRAIN CURVE PALO VERDE UNIT 2 100.00 90.00 80.00 70.00 8O.OO 50.00 IM 40.00 30.00 1B3J7 20.00 10.00 0.00 0.00 0.10 0.20 0.30 STRAIN, IN/IN STRESS-STRAIN CURVE PALO VERDE UNIT 2 100.00 90.00 80.00 70.00 8O.OO CO 50.00 4 40.00 30.00 1B 3JY 20.00 125 F 10.00 0.00 0.0 0.10 0.20 STRAIN, IN/IN Figure 5-24 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens 1B3J7 and 1B3JY 548

STRESS-STRAIN CuRVE PALO VERDE UNIT 2 100.00 90.00 80.00 70.00 CO eo.oo CO 50.00 (X

40.00 30.00 1B3J5 20.00 550 F 10.00 0.00 0.0 0.10 0.20 STRAIN, IN/IN Figure 5-25 Engineering Stress-Strain Curve for Weld Metal Tensile Specimen 1B3J5 5<9

SECTION 6.0 RADIATIONANALYSIS 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

r

where:

R(r,O) gE > 1.0 MeV) at radius r and azimuthal angle O.

I(r',O',E) = 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.

S(r',O',E) = 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') Projected neutron flux at the 1/4T position on the 45'zimuth.

@233.756,45') Projected or calculated neutron flux at the vessel inner radius on the 45'zimuth.

F(239.511,45') 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:

N~ FF P J ref C> [1-e g [e 'J where:

R 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 Average core power level during irradiation period j (MW).

Pr Maximum or reference power level of the reactor (MW).

C.

J 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 0

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).

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' 434

)1.0 MeV n/cm--sec Cycle 1 3.376E+10 3.428E+10 . 3.355E+10 Cycle 2 2.361E+10 2.374E+10 2.328E+10 Cycle 3 2.470E+10 2.520E+10 2.506E+10 Cycle 4 2.461E+10 2.460E+10 2.401E+10 Average 2.672E+10 2.701E+10 2.653E+10 E ) 0.1 MeV n/cm--sec Cycle 1 6.117E+10 6.201E+10 6.060E+10 Cycle 2 4.277E+10 4.294E+10 4.206E+10 Cycle 3 4.475E+10 4.559E+10 4.526E+10 Cycle 4 4.458E+10 4.450E+10 4.338E+ 10 Average 4.840E+10 4.886E+10 4.792E+10 Iron Atom Dis lacement Rate d a sec Cycle 1 4.908E-11 4.985E-11 4.879E-11 Cycle 2 3.432E-11 3.452E-11 3.386E-11 Cycle 3 3.591E-11 3.665E-11 3.644E-11 Cycle 4 3.578E-11 3.577E-11 3.493E-11 Average 3.884E-11 3.927E-11 3.859E-11 6-13

TABLE 6-2 CALCULATED AZIMUTHALVARIATIONOF FAST NEUTRON EXPOSURE RATES AT THE PRESSURE VESSEL CLAD/BASE METAL INTERFACE 00 15'0' n/cm--sec E >1.0 MeV 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 RADIALDISTRIBUTION OF $ (E > 1.0 MeV)

WITHIN THE PRESSURE VESSEL WALL Radius

~em po 15'.0000 30'.0000 40'.0000 45'.0000 233.756t't 1.0000 234.006 0.9854 0.9851 0.9853 0.9856 0.9854 234.631 0.9368 0.9365 0.9369 0.9366 0.9370 235.506 0.8595 0.8580 0.8587 0.8572 0.8591 236.631 0.7571 0.7549 0.7560 0.7529 0.7554 237.923 0.6460 0.6430 0.6445 0.6405 0.6434 239.409 0.5335 0.5292 0.5310 0.5273 0.5303 241.196 0.4203 0.4159 0.4178 0.4143 0.4171 243.204 0.3194 0.3154 0.3173 0.3143 0.3164 245.062 0.2460 0.2431 0.2450 0.2425 0.2433 246.477 0.2003 0.1974 0.1992 0.1976 0.1981 247.78 0.1656 0.1619 '.1635 0.1631 0.1640 249.191 0.1346 0.1311 0.1326 0.1322 0.1333 250.715 0.1076 0.1041 0.1054 0.1049 0.1062 252.055, 0.0877 0.0848 0.0860 0.0854 0.0861 253.098 0.0744 0.0713 0.0723 0.0720 0.0727 254.181 0.0625 0.0592 0.0601 0.0599 0.0604 255.181 0.0527 0.0494 0.0501 0.0497 0.0502 255.994 0.0452 0.0416 0.0422 0.0416 0.0421 256.775+ 0.0391 0.0353 0.0358 0.0351 0.0356 NOTES: 1) Base Metal Inner Radius

2) Base Metal Outer Radius 6-15

TABLE 6-4 RELATIVE RADIALDISTRIBUTION OF @E ) 0.1 MeV)

WITHIN THE PRESSURE VESSEL WALL Radius

~em po 15'0'0'5'33.756t'>

1.000 1.000 1.000 1.000 1.000 234.006 1.009 1.008 1.009 1.008 1.009 234.631 1.014 1.011 1.013 1.011 1.012 235.506 1.004 0.997 1.000 0.996 1.000 236.631 0.974 0.962 0.967 0.961 0.966 237.923 0.927 0.912 0.918 0.910 0.917 239.409 0.867 0.848 0.856 0.846 0.854 241.196 0.792 0.769 0.778 0.768 0.776 243.204 0.708 0.683 0.693 0.682 0.690 245.062 0.632 0.606 0.616 0.605 0.612 246.477 0.575 0.547 0557 0547 0554 247.780 0.523 0.495 0.504 0.495 0.502 249.191 0.470 0.441 0.450 0.441 0.447 250.715 0.414 0.385 0.393 0.384 0.390 252.055 0.367 0.338 0.345 0.336 0.341 253.098 0.330 0.300 0.307 0.299 0.303 254.181 0.293 0.262 0.268 0.260 0.264 255.181 0.258 0.227 0.232 0.223 0.227 255.994 0.228 0.196 0.199 0.191 0.194 256.775<~ 0.203 0.171 0.173 0.164 0.167 NOTES: 1) Base Metal Inner Radius

2) Base Metal Outer Radius 6-16

TABLE 6-5 RELATIVE RADIALDISTRIBUTION OF dpa/sec WITHIN THE PRESSURE VESSEL WALL Radius

~cm 00 15'0'0'5'33.756<'>

1.0000 1.0000 1.0000 1.0000 1.0000 234.006 0.9868 0.9865 0.9869 0.9867 0.9870 234.631 0.9455 0.9449 0.9459 0.9450 0.9461 235.506 0.8816 0.8797 0.8816 0.8796 0.8822 236.631 0.7981 0.7951 0.7978 0.7944 0.7980 237.923 0.7073 0.7030 0.7066 0.7024 0.7065 239.409 0.6141 0.6080 0.6124 0.6080 0.6127 241.196 0.5179 0.5106 0.5156 0.5109 0.5159 243.204 0.4283 0.4203 0.4257 0.4209 0.4255 245.062 0.3592 0.3513 0.3567 0.3520 0.3557 246.477 0.3134 0.3047, 0.3100 0.3062 0.3095 247.780 0.2761 0.2662 0.2712 0.2682 0.2718 249.191 0.2403 0.2301 0.2348 0.2317 0.2354 250.715 0.2061 0.1955 0.1998 0.1967 0.2000 252.055 0.1788 0.1684 0.1721 0.1689 0.1716 253.098 0.1588 0.1478 0.1911 0.1482 0.1507 254.181 0.1394 0.1278 0.1307 0.1278 0.1299 255.181 0.1223 0.1103 0.1125 0.1094 0.1112 255.994 0.1087 0.0956 0.0973 0.0939 0.0955 256.775<~ 0.0971 0.0834 0.0847 0.0812 0.0826 NOTES: 1) Base Metal Inner Radius

2) Base Metal Outer Radius 6-17

TABLE 6-6 NUCLEAR PARAMETERS USED IN THE EVALUATIONOF NEUTRON SENSORS Reaction Target Fission Monitor of Weight Response Product Yield Material iatereat Fraction Ranee Half-Life ~5o Copper* Cu"(n,a) Co 0.6917 E>5 MeV 5.271 yrs Iron Fe~(n,p)Mn~ 0.0580 E>2MeV 312.5 days Nickel Ni"(n,p)Co" 0.6827 E>2MeV 70.78 days Titanium Ti {n,p)Sc+ 0.0810 E>2MeV 83.83 days .

U-'n,I)Cs'3'o"(n,y) 1.0 E> 1MeV 30.17 yrs 6.00 Uranium-238'obalt-Aluminum'obalt-Aluminum Co~ 0.0017 0.4ev>E> 0.015 MeV 5.271 yrs Co"(n,y) Co~ 0.0017 E ( 0.015 MeV 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 MONTHLYTHERMAL GENERATION DURING THE FIRST FOUR FUEL CYCLES OF THE PALO VERDE UNIT 2 REACTOR Thermal Generation Thermal Generation Year Month MW-hr Year Month MW-hr 1986 5 178458 1990 9 2724764 6 748794 10 2746944 7 179055 11 2722657 8 789746 12 2809763 9 441921 1991 1 2810976 10 1785522 2550144 ll 12 2573938 2217984 2

3 4

2797277 2733538 1987 1 816605 5 2826434 2 0 6 2734422 3 843281 7 2825239 4 2523504 8 1760151 5 2217619 9 2731841 6 2370470 10 1454795 7

8 2667944 2807364 ll 12 0

0 9 2670655 1992 1 1548257 10 2794687 2 2642730 11 2400712 3 2150742 12 2799320 4 2730948 1988 1 2796621 5 2824993 2 1711979 6 2735197 3 0 7 2826315 0 8 2823534 5 0 9 2701162 6 510519 10 2797104 7 2794833 11 2490116 8 2713036 12 2819640 9 2702001 1993 1 2825987 10 2814560 2 2549505 11 1909336 3 1182454 12 2766725 1989 1 2800843 (Shutdown 3/14/93) 2 1383148 3 1303549 0

5 0 6 20794 7 1657660 8 2576391 9 1206913 10 1205545 ll 12 2546733 0

1990 1 2806069 2 2064631 3 0 0

5 0 6 0 7 558801 8 2659419 6-19

TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCE CAPSULE W137 SATURATED REACTION MONITOR AND ACTIVITY ACTIVITY RATE AXIALLOCATION Co-60'EASURED ~ilis/sec.util ~dls/sec- ltl ~>s/nucleus Cu-63 n.c 93-3144 TOP 1.22E+05 3.204E+05 93-3153 MID 1.19E+05 3.125E+05 93-3158 BOT 1.19E+05 3.125E+05 Averages 1.20E+05 3.152E+05 4.808E-17 Fe-54 n. Mn-54 93-3142 TOP 1.29E+06 2.334E+06 93-3147 MID 1.28E+06 2.316E+06 93-3156 BOT 1.24E+06 2.243E+06 Averages 1.27E+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 2.01E+05 7.786E+05 93-3146 MID 1.95E+05 7553E+05 93-3155 BOT 1.90E+05 7.360E+05 Averages 1.95E+05 7.566E+05 7.128E-16 93-3143 TOP 250E+05 2.607E+06 93-3151 MID 2.08E+05 2.169E+06 93-3157 BOT 2.16E+05 2.252E+06 Averages 2.25E+05 2.343E+06 1.544E-14 U-238 n Cs-137 93-3140 TOP 3.25E+06 3.389E+06 93-3157 BOT 2.79E+05 2.909E+06 Averages 3.02E+05 3.149E+06 2.075 E-14 Co-59 n Co-60 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 Flux Time Fluencc Uncertainty Meas Fluence < 0.414 ev 3.177E+11 1.433Ew08 4.551E+19 ~22%

Meas Fluence > 0.1 Mev 5.575E+11 1.433E%08 7.987E+18 <18%

Meas Fluence > 1.0 Mev 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 ADJUSTED REACTION MEASURED CALCULATION Cu43 (n,ct) Co-60 4.81E-17 4.85E-17 1.01 Fe-54 (n,p) Mn-54 3.67E-15 3.68E-15 1.00 Ni-58 (n,p) Co-58 4.68E-15 4.71E-15 1.01 Ti-46 (n,p) Sc-46 7.13E-16 7.11E-16 1.00 U-238 (n,f) Cs-137 (Cd) 1.53E-14 1.26E-14 0.83 Co-59 (n,y) Co-60 1.01E-11 1.00E-11 0.99 Co-59 (n,y) Co-60 (Cd) 1.21E-12 1.19E-12 0.99 6-21

TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCE CAPSULE W137 GROUP ENERGY ADJUSTED FLUX GROUP ENERGY ADJUSTED FLUX

~MeV ~a/cn> -ecc ~Me V ~e/cmc-cec 17.33 5.60E+06 28 9.119E-03 2.45E+09 14.92 1.30E+07 29 5.531E-03 2.77E+09 13.50 5.48E+07 30 3.355E-03 9.33E+08 11.62 1.28E+08 31 2.839E-03 9.57E+08 10.00 2.94E+08 32 2.404E-03 1.01E+09 8.607 5.06E+08 33 2.035E-03 3.27E+09 7.408 1.23E+09 34 1.234E-03 3.61E+09 6.065 1.70E+09 35 7.485 E-04 4.08E+09 4.966 2.94E+09 36 4.540E-04 4.60E+09 10 3.679 2.82E+09 37 2.754E-04 5.28E+09 2.865 4.67E+09 38 1.670E-04 9.97E+09 12 2.231 4.43E+09 39 1.013E-04 5.60E+09 13 1.738 4.51E+09 40 6.144E-OS 5.15E+09 14 1.353 3.47E+09 41 3.727E-OS 4.59E+09 15 1.108 4.83E+09 42 2.260E-05 4.06E+09 16 8.208E-01 4.42E+09 43 1.371E-05 3.63E+09 17 6.393E-01 3.95E+09 44 8.315E-06 3.33Et09 18 4.979E-01 2.87E+09 45 5.043E-06 3.12E+09 19 3.877E-01 3.24E+09 46 3.059E-06 2.97E+09 20 3.020E-01 4.84E+09 47 1.855E-06 2.82E+09 21 1.832E-01 4.12E+09 48 1.125 E-06 2.63E+09 22 1.111E-01 3.39E+09 49 6.826E-07 3.14E+09 6.738E-02 2.85E+09 50 4.140E-07 4.11E+09 24 4.087E-02 2.14E+09 51 2.511E-07 1.61E+10 2554E-02 1.49E+09 52 1.523E-07 3.90E+10 26 1.989E-02 1.24E+09 53 9.237E-08 259E+11 27 1.503 E-02 2.27E+09 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 3.801E+18 4.071E+18 0.934 Fluence (E > 0.1 Mcv) [n/cm2-sec[ 6.865E+18 7.987E+18 0.860 dpa 5.528E-03 6.273 E-03 0.881 Comparison of Calculated and Mcasurcd Neutron EXPOSURE RATE for Capsule W137 Calculated Measured ~CM Flux (E > 1.0 Mcv) [n/ctn2-sec[ 2.653E+10 2.842E+10 0.934 Hux (E > 0.1 Mev) [n/cnt2-sec[ 4.792E+10 5575E+10 0.860 dpa/s 3.859E-11 4.379E-11 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 1.770E+10 1.S09E+10 2.027E+10 1.967E+10 E > 0.1 2.934E+10 4.023E+10 4.116E+10 4.614E+10 4.505 E+10 dpa 2.143E-11 2.891E-11 2.945E-11 3.287E-11 3.195 E-11 BEST ESTIMATE EXPOSURE (15.0 EFPY) AT THE PRESSURE VESSEL INNER RADIUS 0 DEG" 15 DEG 30 DEGAS'0 DEGt'5 DEG E > 1.0 6.187E+18 S.376E+18 8.565E+18 9.595E+18 9.310E+18 E > 0.1 1.389E+19 1.905E+19 1.949E+19 2.184E+19 2.132E+19 dpa 1.014E-02 1.36SE-02 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 1.320E+19 1.787E+19 1.827E+19 2.047E+19 1.986 E+19 E > 0.1 2.963E+19 4.063E+19 4.157E+19 465 9E+19 4.549E+19 dpa 2.164E-02 2.919E-02 2.974 E-02 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 9.570E+18 1.072E+19 1.033E+19 3/4T 1.355 E+18 1.776 E+18 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 5.194E+18 5.776E+18 5.604E+18 3/4T 1.235 E+18 1.583E+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 1.076E+19 1.108E+19 1.232E+19 1.195E+19 3/4T 2.634E+18 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 LEAD FACTOR W38 1.41 W43 1.40 W137 1.40'.41 W142 W230 1.43 W310 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 INSUIAlTON 4'~

tF 0

REACI'OR C~ CAVITY SHROUD FUEL INlEI'ATER BYPASS WA'IKR BARREL 133 233 383 X CAl 6-27

FIGURE 6-2 AZIMUTHALVARIATIONOF NEUTRON FLUX (E ) 1.0 MEV)

AT THE REACTOR VESSEL INNER RADIUS 20 19 18 ~ ~

E g ~ SN V ~ r C

17 16

~w 15 14 Lower Curve Shows Effect of Surveillance Capsules I

12 0 10 20 30 50 Azimuthal Angle (Degrees) 6-28

FIGURE 6-3 AXIALDISTRIBUTION OF REACTOR POWER 1.2 o 0.9 C4

~ ~

0.8 0.7 0.6 0.5 0 50 150 Axial Distance &om Core Bottom (inches) 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 Removal Time Fluence Location Lead Factor (EFPY)<'~ (n/cm~)

1.40 4.540 4.071 x 10'~

137'30'10 1.43 15 1.37 x 10'.93 1.43 EOL x 10'~

1.41 Stand-by 38'30 1.40 Stand-by 142 1.41 Stand-By (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 Matrixfor 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

- P~~ MAXIMIJMLOAO FAST P

~ FRACTURE LOAD F

PGY GEh/ERAL YIE LO LOAD I 0 I I

0 I

I I

fAS T FR ACT U R LOAD E'A~ARREST I

I I

I I

't I

t

~GYQ 7!M 6 WI= Fracture initiation region t+> ~ Time to general yielding WP ~ Fracture yroyagation region 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 aN

~D loZ Zo4 3.6 4e8 6+0 TItK ( ICiZC )

PALO VERDE g2 1B11E 18184 C5 D IeZ Z.4 3+6 4 8 6.0 Tlm C nSCC )

PALO VERDE P2 1B3.24 Figure A-2. Load-time records for Specimens 1BllE and 1B124 A-2

2.4 3+6 4+8 6.0 TINE < NSKC )

PALO VERDE g2 1B11M 18150 2+4 3m 6 4+8 6 0 TItK ( NSKC )

PALO VERDE g2 1815D Figure A-3. Load-time records for Specimens 1BllM and 1B15D

IBI4Y og

~

W CU g

o

~D Isa Ro4 3.6 4oa 6.0 TlllE C NSEC )

PALO VERDE g2 1B14Y 18136 og

~

5 Cl C9 Q ol O

o

.D I~ 8 eo4 3o6 4m 8 6.0 TIME ( NSKC >

PALO VERDE g2 1B136 Figure AA. Load-time records for Specimens 1B14Y and 18136

18145 eo4 3.6 4ee 6.0 TItC C !CEC >

PALO VERDE g2 1B145 1814/

ea4 3.6 4+8 6o0 TIIK < ICXC )

PALO VERDE g2 BatCh11B14gl Figure A-5. Load-time records for Specimens 1B145 and 1B14D

18188 eo4 3o6 4ee 6.0 TBK C ICKC >

PALO VERDE g2 1B12B ee4 3 6 4ee 6m 0 TItK < NXC >

PALO VERDE g2 1B255 Figure A-6. Load-time records for Specimens 1B12B and 1B255

Ol o

W Cl C9 Ol g

O

~ D I 2 L4 3.6 4os 6m 0 T1% C CEC )

PALO VERDE jNI2 1B23L CU eD Ice 2.4 3+6 4.8 6o0 TIlK C NSKC >

PALO VERDE g2 1B2AE Figure A-7. Load-time records for Specimens 1B23L and 1B2AE

18224 CU

~D 1+2 ee4 3e6 4oe 6m 0 TIlK C CKC ) ~

PALO VERDE S2 1B224 CU

~V oD 1e2 'o4 3.6 4.8 6e0 TIJOU C ICKC )

PALO VERDE g2 1B267 Figure A-8. Load-time records for Specimens 18224 and IB267

PALO VKROE I2

~ D Ioe 2,4 3o6 4ee 6.0 TIttE C ttSEC )

PALO VERDE g2 1B23P C9 o\4g ~

~ D i+2 2e4 36 4oe 6.0 TlttE C tCEC >

PALO VERDE f2 1B2 5AM Figure A-9. Load-time records for Specimens 1B23P and 1B2AM

PALO VERDE 48 CI

~ D 8+4 3.6 4o8 6.0 TItK ( ICEC )

PALO VERDE g2 .

1B26D 182IU

.D 8.4 3.6 4.8 6 0 TIE C tCZC )

PALO VERDE g2 1B21U Figure A-10. Load-time records for Specimens 1B26D and 1B21U A-10

PALO VERGE 82 i+2 2,4 3 6 4s8 6+0 TlttE C ICKC >

PALO VERDE g2 1B22U 1+2 2.4 3+6 4I 8 6m 0 TIttE ( t5EC )

PALO VERDE g2 1B22E Figure A-11. Load-time records for S pecimens 1B22U and 1B22E

1884 J en 4 3o6 4,8 6m 0 Tlirt < NSEC >

PALO VERDE g2 1B24 J lsa43 8+4 3 6 4e8 6,o Tlirt C NSEC )

PALO VERDE P2 1B243 Figure A-12. Load-time records for Specimens 1B24J and 1B243 A-12

I+2 2e4 3.6 4+8 6+0 TIIC ( ttSEC )

PALO VERDE g2 1B22M i+2 2 4 3,6 4I8 6e0 TIttE C ttSEC >

PALO VERDE P2 1B23M

'Figure A-13. Load-time. records for Specimens 1B22M and 1B23M A-13

PALO VERGE <<2 1B3AO g ~

R a

~ D le2 2i4 3,6 4,8 6.0 TIttE C tCKC > ~

PALO VERDE g2 1B3AD 1831b og 0

CI

~D le2 2+4 3.6 4oe 6+0 TIttE C tCEC >

PALO VERDE g2 1B316 Figure A-14. Load-time records for Specimens 1B3AD and 1B316 A-14

PALO VEROE NR og

~

I ol

~ D eo4 3.6 4e8 6o0 TlliE < NSKC )

PALO VERDE f2 1B35E m

1 8

CU O

.D 8.4 3+6 4oe 6.0 TItK C tiSEC >

PALO VERDE g2 1B342 Figure A-15. Load-time records for Specimens 1B35E and 1B342 A-I5

PALO VEROE 02 Ie2 2e4 3,6 4oe 6m 0 TINE ( tCXC )

PALO VERDE g2 1B34U 18312 Ioe 2+4 3+6 4+8 6+0 TINE < NSEC )

PALO VERDE g2 1B312 Figure A-16. Load-time records for Specimens 1B34U and 1B312 A-16

PALO VERtK 08 f 834C 2+4 3o6 4+8 6. 0 TIN ( ICKC )

PALO VERDE g2 LB34E f.2 em 4 3.6 4e8 6e0 Tilg C NSCC )

PALO VERDE g2 1B35P Figure A-17. Load-time records for Specimens 1B34E and 1B35P A-17

IER6

~

ow o

~ D a+4 3.6 4o8 6e0 TitK C NSEC )

PALO VERDE g2 1B326 o

o oD 1.2 a+4 3+6 4i8 6e0 TitK ( lCXC )

PALO VERDE if2 1B3A3 Figure A-18. Load-time records for Specimens 1B326 and 1B3A3 A-18

em 4 3+6 4e8 6m 0 TINE < tCEC >

PALO VERDE g2 1B3AL ee4 3+6 4+8 6+0 TIJOU C INC >

PALO VERDE g2 1B35M Figure A-19. Load-time records for Specimens IB3ALand IB35M

PALO VERDE 62 l833E CO o

I Ol Ol 04 O

~ D li2 2o4 3+6 4+8 6.0 TIJOU C 1CEC >

PALO VERDE g2 1B33E l83l8 CO o

CO C9 I Ol Ol CI

,D la2 2,4 3,6 TIIC 4ee 6io

< tCXC >

PALO VERDE g2 1B31B Figure A-20. Load-time records for Specimens 1833E and 1B31B A-20

CD Cl o

oD so 4 3o6 4o8 6.0 TIIK ( NSEC >

PALO VERDE g2 Batch:1B32A 18411 CD 5 4 D

CD CD I DC CD oD a+4 3+6 4+8 6.0 TIIK ( t5EC >

PALO VERDE g2 1B411 Figure A-21. Load-time records for Specimens 1B32A and 1B411 A-21

PALO VEROE Ie IB440 a

yg r ~

W CI

~ 0 los Bo4 3o6 4es 6+ 0 TItK C ICXC >

PALO VERDE g2 1B44D og

~

u)

C9 I CIJ CI

~ 0 Be 4 3+6 4.8 6m 0 TItK C NSEC )

PALO VERDE g2 1B43B Figure A-22. Load-time records for Specimens 1B44D and 1B43B A-22

I Al CI

~ 0 eo4 3+6 4oe 6e0 TIklE C !CEC >

PALO VERDE g2 1B43Y C0 o

Q ol D

O eo e 4 3+6 408 6+0 TnlE ( llSEC )

PALO VERDE I2 1B432 Figure A-23. Load-time records for Specimens 1B43Y and 1B432 A-23

PALD VERDE 08 1843T 84 3.6 4 8 6.0 TIlK C lCKC )

PALO VERDE t2 1B43T ag 0

8 CU CI oo ee4 3+6 4,8 6+0 TNE ( llsEC )

PALO VERDE g2 1B444 Hgure A-24. Load-time records for Specimens 1B43T and 1B444 A.24

184311 m

o W

CA C9 Al g

Al

~D 1o2 2o4 3o6 4os 6o0 TIIK C lCKC )

PALO VERDE P2 1B43M o

0 AI 5

Al o 4os 6.0

.D 1o2 2.4 3.6 TIIK < CXC )

PALO VERDE g2 1B43K Figure A-25. Load-time records for Specimens 1B43M and 1B43K A-26

1844L 1.2 2,4 3.6 4os 6.0 TINE C N3EC )

PALO VERDE g2 Batch:1B44L le2 2+4 3.6 4+8 6m 0 TINE C NSEC >

PALO VERDE g2 1B456 Figure A-26. Load-time records for Specimens 1B44L and 1B456 A'-26

PALO VEROE NR 18418 m

1oe P

Ro4 3.6 4,8 6. 0 TItK C !ISEC )

PALO VERDE g2 1B41B m

I ol CI

~ 9 1 R Ro4 3o6 4as 6o0 TItK ( INC )

PALO VERDE g2 1BD3Y Figure A-27. Load-time records for Specimens 1B41B and 1BD3Y A 27

18043 ag

~

Ct

~ D 1,2 2o4 3+6 4os 6.0 TIIK ( NSEC >

PALO VERDE 82 1BD43 CO 4)

C9'D 1.2 2.4 3.6 4.8 6+0 fI K ( tSEC )

PALO VERDE g2 1BD4P Figure A-28. Load-time records for Specimens 1BD43 and 1BD4P

PALO VERDE Ne CO I CU

~ D 8.4 3.6 4.8 6.0 TIIK C IISEC >

PALO VERDE g2 1BD35 m

g 0 R

CU g

CCC O

.D ee4 3.6 4o8 6+0 TIttL < NSEC >

PALO VERDE g2 1BD2D Figure A-29. Load-time records for Specimens 1BD35 and 1BD2D A-29

0) u) 3.6 6 0

~ D lo2 2+4 Tlm ( eXC )

PALO VERDE f2 1BD52 O

oD le2 2e4 3+6 4oe 6o0 TllK C tCKC )

PALO VERDE g2 1BD3E Figure A-30. Load-time records for Specimens 1BD52 and 1BD3E A<0

og

~

CI

.D I ~ 2 2j4 3j6 4je 6j0 TIE < NSEC )

PALO VERDE g2 1BD2S I cU CQ jD .Ioe 2j4 3,6 4e8 600 TIlK ( ICKC >

PALO VERDE g2 1BD2M Figure A-31. Load-time records for Specimens 1BD2M and 1BD2B

1B151 2 3 4 5 6 7 8 9 10 Time (msec)

Specimen 1B'i51 Time to Yield 0.165 mSec'040 Temperature 0 F ield Load lbs*

Available Energy 256 in-lbs Time to Maximum 0.170 mSec initial Velocity 57.4 in/sec Maximum Load 1047 lbs Energy at Max. Load 3.7 in-lbs Crack Length 0.176 in KID 38.2 ksi-in"1/2 0.447 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 Time (msec) 6 7 .8 9 10 0

Specimen 1B'14B mme to Yield 0.180 mSec'165 Temperature 25 F Yield Load Ibs*

Available Energy 256 in-lbs mme to Maximum 0.180 mSec Initial Velocity 57.5 in/sec Maximum Load 1172 Ibs Energy at Max. Load 5.1 in-Ibs Crack Length 0.180 in KID 44.0 ksi-in"1/2 0.457 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 7 8 9 10 Time {msec)

Specimen 18116 Time to Yield 0.145 mSec'200 Temperature 50 F iefd Load Ibs*',170 Available Energy 399 in-lbs Time to Maximum mSec initial Velocity 71.7 in/sec Maximum Load 1235 lbs Energy at Max. Load 7.3 in-Ibs Crack Length 0.176 in KID 45,0 ksi-in"1/2 OA47 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 9 10 Time (msec)

Specimen ime to Yield 0.157 mSec Temperature 71 F ield Load 1040 Ibs Available Energy 398 in-lbs Time to Maximum 0.300 mSec initial Velocity 71.6 inlsec Maximum Load 1066 Ibs Energy at Max. Load 15,5 in-Ibs Crack Length 0.213 in KJD 102.9 ksi-in"1/2 0.541 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 8 9 10 Time (msec)

Specimen 1B113 ime to Yield 0.150 mSec Temperature 76 F ield Load 1120 Ibs Available Energy 398 in-lbs Time to Maximum 0,860 mSec Initial Velocity 71.6 in/sec Maximum Load 1285 Ibs Energy at Max. Load 64.4 in-lbs Crack Length 0.189 in KJD 210.9 ksi-in"1/2 0.480 Yield Stress 87.9 ksi Specimen Compliance 51.6 Machine Compliance 60.7 Pigure A-36. Load-time record and data for precraeked specimen IB113

1B11K 2 3 4 5 Time (msec) 6 7 8 9 10 0

Specimen 1B11K mme to Yield 0.135 mSec Temperature 100 F ietd Load 1200 Ibs Available Energy 637 in-Ibs Time to Maximum 0.560 mSec Initial Velocity 90.6 in/sec Maximum Load 1381 Ibs Energy at Max. Load 54.2 in-Ibs Crack Length 0.186 in KJD 187,4 ksi-in"1/2 0.472 Yield Stress 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 7 8 9 10 Time (tnsec)

Specimen 1B13M ime to Yield 0.140 mSec Temperature 150 F ield Load 1100 Ibs Available Energy 642 in-lbs Time to Maximum 1.035 mSec Initial Velocity 90.9 in/sec Maximum Load 1341 Ibs Energy at Max. Load 101.6 in-Ibs Crack Length 0.189 in KJD 263.6 ksi-in"1/2 0.480 Yield Stress 86.4 ksi Specimen Compliance 51.6 Machine Compliance 83.7 Figure A-38. Load-time record and data for precracked specimen 1B13M A-38

1B12L 1200 0 1 2 3 4 5 Time (msec) 6 7 8 9 10 0

Specimen 1B12L Time to Yield 0.145 mSec Temperature 200 F Yield Load 1170 Ibs Available Energy 640 in-Ibs Time to Maximum 1.135 mSec initial Velocity 90.8 in/sec Maximum Load 1499 lbs Energy at Max. Load 122,1 in-Ibs Crack Length 0.185 in 285.0 ksi-in"1/2 0,470 Yield Stress 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 8 9 10 Time (msec)

Specimen 1B141 ime to Yield 0.105 mSec Temperature 250 F Yield Load 940 lbs Available Energy 640 in-lbs Time to Maximum 1.105 mSec initial Velocity 90.8 in/sec Maximum Load 1375 Ibs Energy at Max. Load 108.0 in-Ibs Crack Length 0,181 in KJD 266.0 ksi-in"1/2 0.459 Yield Stress 68.4 ksi Specimen Compliance 47;8 Machine Compliance 70.0 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 18213 Time to Yield 0.218 mSec*

Temperature 25 F ield Load 1380 Ibs*

Available Energy 256 in-lbs Time to Maximum 0.225 mSec Initial Velocity 57.4 in/sec Maximum Load 1395 Ibs Energy at Max. Load 8.7 in-Ibs Crack Length 0.175 in KID 50.4 ksi-in*1/2 0.444 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 7 8 9 10 Time (msec)

Specimen 1826C Time to Yield 0.155 mSec'230 Temperature 50 F Yield Load Ibs'.185 Available Energy 401 in-lbs Time to Maximum mSec Initial Velocity 71.9 inlsec Maximum Load 1306 Ibs Energy at Max. Load 8.1 in-lbs Crack Length 0.187 in KID 51.8 ksi-in"1/2 0,475 Yield Stress 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 9 $ 0 Time (msec)

Specimen 1B23T Time to Yield 0.155 mSec emperature 76 F Yield Load 1200 Ibs Available Energy 397 in-Ibs Time to Maximum 0.455 mSec Initial Velocity 71.5 in/sec Maximum Load 1354 Ibs Energy at Max. Load 32.1 in-lbs Crack Length 0.173 in KJD 137.1 ksi-in"1/2 0.439 Yield Stress 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 9 10 Time (msec)

Specimen 18278 Time to Yield 0.110 mSec Temperature 100 F Yield Load 1060 lbs Available Energy 639 in-Ibs Time to Maximum 0.295 mSec initial Velocity 90.7 in/sec Maximum Load 1186 Ibs Energy at Max. Load 22.6 in-Ibs Crack Length 0.189 in KJD 118.1 ksi-in'1/2 0.480 Yield Stress 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 8 9 10

'tlme (msec)

Specimen Time to Yield 0.150 mSec emperature 150 F Yield Load 1260 Ibs Available Energy 638 in-Ibs Time to Maximum 0,560 mSec Initial Velocity 90.6 io/sec Maximum Load 1434 Ibs Energy at Max. Load 54.4 in-lbs Crack Length 0.186 in KJD 185.7 ksi-in"1/2 0.472 Yield Stress 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 1B245 Time to Yield 0.112 mSec Temperature 200 F Yield Load 1080 lbs Available Energy 636 in-lbs Time to Maximum 0.950 mSec Initial Velocity 90.5 in/sec Maximum Load 1365 lbs Energy at Max. Load 92.1 in-lbs Crack Length 0.194 in KJD 254.2 ksi-in"1/2 0.492 Yield Stress 89.1 ksi Specimen Compliance 54.2 Machine Compliance 64,1 Figure A<6. Load-time record and data for precracked specimeri 1B245 A46

1B226 1400

$ 200 800 o

i 0

0 2 3 4 5 6 7 8 9 10 Time (msec)

Specimen 1B226 ime to Yield 0.113 mSec Temperature 250 F Yield Load 880 lbs Available Energy 637 in-lbs ime to Maximum 1.'145 mSec Initial Velocity 90.6 in/sec Maximum Load 1398 Ibs Energy at Max. Load 112.4 in-lbs Crack Length 0.184 in KJD 271.6 ksi-in"1/2 0.467 Yield Stress 65.9 ksi Specimen Compliance 49.2 Machine Compliance 86.3 Figure A-47. Load-time record and data for precracked specimen 1B226 0

1B3AT 1200 800 O

2 3 4 5 6 7 8 9 $0 Time (msec)

Specimen -183AT Time to Yield 0.111 mSec*

Temperature -50 F Yield Load 990 Ibs" Available Energy 642 in-lbs Time to Maximum 0.115 mSec Initial Velocity 90.9 in/sec Maximum Load 995 lbs Energy at Max, Load 4.2 in-lbs Crack Length 0.191 in KID 40.7 ksi-in"1/2 OA85 Yield Stress 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 1831U ime to Yield 0.157 mSec*

Temperature -25 F ield Load 1030 lbs*

Available Energy 257 in-Ibs ime to Maximum 0.175 mSec Initial Velocity 57.5 in/sec Maximum Load 1085 Ibs Energy at Max. Load 4.7 in-lbs Crack Length 0.182 in KID 41.4 ksi-in"1/2 0.462 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

%me (msec) 6 7 8 9 '0 Specimen 'IB323 ime to Yield 0.154 mSec

• Temperature 0 F ield Load 1210 Ibs*

Available Energy 397 in-lbs Time to Maximum 0.155 mSec Initial Velocity 71.5 in/sec Maximum Load 1215 Ibs Energy at Max. Load 5.5 in-lbs Crack Length 0.186 in KlD 47.8 ksi-in"1/2 0.472 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 Time (msec) 6 7 8 9 10 0

Specimen 1B33J Time to Yield 0.150 mSec Temperature 50 F Yield Load 1240 Ibs Available Energy 397 in-Ibs Time to Maximum 0.550 mSec Initial Velocity 71.5 in/sec Maximum Load 1390 Ibs Energy at Max. Load 42.0 in-lbs Crack Length 0.180 in KJD 163.5 ksi-in"1/2 0.457 Yield Stress 89.4 ksi Specimen Compliance 47.3 Machine Compliance 54.0 Figure A-51. Load-time record and data for precracked specimen 1'B33J

1B3 ZA 0 1 2 3 4 5 6 7 8 9 10 Time (msec)

Specimen 1B31A Time to Yield 0.150 mSec emperature 71 F ield Load 1240 Ibs Availabte Energy 400 in-Ibs Time to Maximum 1.035 mSec Initial Velocity 71.8 in/sec Maximum Load 1456 lbs Energy at Max. Load 86,2 in-Ibs Crack Length 0.174 in KJD 236.4 ksi-in"1/2 0.442 Yield Stress 84.5 ksi Specimen Compliance 44.8 Machine Compliance 56.6 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 7 8 9 10 Time (msec)

Specimen IB33Y ime to Yield 0.140 mSec Temperature Yield Load 1200 Ibs Available Energy 636 in-Ibs ime to Maximum 0.720 mSec initial Velocity 90.5 in/sec Maximum Load 1286 Ibs Energy at Max. Load 70.0 in-lbs Crack Length 0.184 in KJD 215.5 ksi-in"1/2 0,467 Yield Stress 89.8 ksi Specimen Compliance 49.2 Machine Compliance 76.0 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 1836M Time to Yield 0.060 mSec Temperature 125 F Yield Load .950 Ibs Available Energy 3201 in-Ibs Time to Maximum 0.300 mSec initial Yelocity 203.0 in/sec Maximum Load 1223 lbs Energy at Max, Load 56.4 in-Ibs Crack Length 0.189 in KJD 191.7 ksi-in"1/2 0.480 Yield Stress 74.6 ksi Specimen Compliance 51.6 Machine Compliance 98.5 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 8 9 10 Time (msec)

Specimen 1B31K Time to Yield 0.110 mSec Temperature 150 F Yield Load 1000 Ibs Available Energy 642 in-Ibs Time to Maximum 0.860 mSec Initial Velocity S0.9 in/sec Maximum Load 1267 Ibs Energy at Max. Load 80.0 in-Ibs Crack Length 0.187 in KJD 233.8 ksi-in"1/2 0,475 Yield Stress 77.0 ksi Specimen Compliance 50.6 Machine Compliance 64.3 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 8 9 10 Time (msec)

Specimen 1837A Time to Yield 0,110 mSec Temperature 200 F Yield Load 1000 Ibs Available Energy 637 in-lbs Time to Maximum 0.715 mSec Initial Velocity 90,6 in/sec Maximum Load 1319 lbs Energy at Max. Load 67.6 in-Ibs Crack Length 0,186 in KJD 211.8 ksi-in"1/2 0.472 Yield Stress 76.3 ksi Specimen Compliance 50.1 Machine Compliance 63.4 Figure A-56. Load-time record and data for precracked specimen 1B37A

0'