ML20134A879
| ML20134A879 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/31/1997 |
| From: | Boyle D WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20134A874 | List: |
| References | |
| WCAP-14779, NUDOCS 9701290146 | |
| Download: ML20134A879 (282) | |
Text
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WESTINGHOUSE NON PROPRIETARY CLASS 3 WCAP-14779
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ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND UNIT 1 4
REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM i
, T. J. Laubham
- J. F. Williams G. K. Roberts January 1997 i
Work Performed Under Shop Order NLDP-106 4
Prepared by Westinghouse Electric Corporation for Northem States Power Company Approved: .,
N D. E. Boyle, Manager [
Reactor Equipment & Materials Engineering WESTINGHOUSE ELECTRIC CORPORATION Systems and Major Projects Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355
@ 1997 Westinghouse Electric Corporation All Rights Reserved 9701290146 970115 PDR P
ADOCK 05000282 PDR
I PREFACE This report has been technically reviewed and verified.
Reviewer:
Sections 1 through 5,7,8 P. A. Grendys f. d . OrphM Appendices A, B, and C k Section 6 L- A T. M. Lloyd '
7 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM j
- . - _ _ - _ _ . - . - - _ _ - . - - - _ _ _ . - _ ..- -- _.__ - . . ~ _ - -
4 .
11 TABLE OF CONTENTS LI ST O F TABLE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii LI ST OF I LLU STRATION S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vi 1.0
SUMMARY
OF RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 2.0 I NTROD U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 3.0 BAC KG ROU N D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1
4.0 DESCRIPTION
OF PROG RAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 5.0 TESTING OF SPECIMENS FROM CAPSULE S . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1 O ve rvie w . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2 Charpy V-Notch Impact Test Results . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5.3 Tensile Test Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 5.4 Wedge Opening Loading (WOL) Specimens . . . . . . . . . . . . . . . . . . . . . . 5-6 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY . . . . . . . . . . . . . . . . . . . 6-1 6.1 I ntrod u ction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2 Discrete Ordinates Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 6.3 Neutron Dosimetry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-5 6.4 Projections of Pressure Vessel Exposure . . . . . . . . . . . . . . . . . . . . . . . 6-10 7.0 RECOMMENDED SURVEILLANCE CAPSULE REMOVAL SCHEDULE . . . . . . . 7-1 8.0 R EFER EN C ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1 APPENDIX A - LOAD-TIME RECORDS FOR CAPSULE S CHARPY IMPACT TESTS ...................................................... A-0 APPENDIX B - CHARPY V-NOTCH SHIFT RESULTS FOR EACH CAPSULE HAND-DRAWN VS. HYPERBOLIC TANGENT CURVE-FITTING METHOD (CVG RAPH VER SION 4.1) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-0 APPENDIX C - CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING HYPERBOLIC TANGENT CURVE FITTING METHOD . . . . . . . . . . . . . . . . . . . . C-0 APPENDIX D - SURVEILLANCE DATA CREDIBILTY EVALUATION . . . . . . . . C-0 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
in ;
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LIST OF TABLES Table 4-1 Heat Treatment of the Prairie Island Unit 1 Reactor Vessel Surveillance l Mate rial s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 Table 4-2 Chemical Composition (wt%) of the Unirradiated Prairie Island Unit 1 Reactor Vessel Surveillance Materials . . . . . . . . . . . . . . . . . . . . . . . . . . 4-4 l
Table 4-3 Chemical Composition of the Prairie Island Unit 1 Charpy Specimens l Removed from Surveillance Capsule S . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5 Table 4-4 Chemistry Results from the Alloy Steel NBS Certified Reference S ta n da rds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ............ 4-6 Table 4-5 Calculation of Average Cu and Ni Weight Percent Values for Beltline M ate rials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-7 Table 5-1 Charpy V-notch Data for the Prairie Island Unit 1 Intermediate Shell Forging C Irradiated to a Fluence of 4.017 x 10" n/cm" (E > 1.0 MeV)
(Tangential Orientation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 Table 5-2 Charpy V notch Data for the Prairie Island Unit 1 Intermediate Shell Forging C Irradiated to a Fluence of 4.017 x 10" n/cm2 (E > 1.0 MeV)
(Axial Orientation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-8 Table 5-3 Charpy V-notch Data for the Prairie Island Unit 1 Surveillance Weld Metal Irradiated to a Fluence of 4.017 x 10" n/cm (E > 1.0 MeV) . . . . . . 5-9 Table 5-4 Charpy V-notch Data for the Prairie Island Unit 1 Heat-Affected-Zone (HAZ) Metal Irradiated to a Fluence of 4.017 x 10" n/cm" (E > 1.0 MeV) 5-10 Table 5-5 Charpy V-notch Data for the Prairie Island Unit 1 Correlation Monitor 2
Material Irradiated to a Fluence of 4.017 x 10" n/cm (E > 1.0 MeV) . . . 5-11 Table 5-6 Instrumented Charpy impact Test Results for the Prairie Island Unit 1 Intermediate Shell Forging C Irradiated to a Fluence of 4.017 x 10" n/cm8 (E > 1.0 MeV) (Tangential Orientation) . . . . . . . . . . . . . . . . . . . . 5-12 Table 5-7 Instrumented Charpy impact Test Results for the Prairie Island Unit 1 Intermediate Shell Forging C Irradiated to a Fluence of 4.017 x 10" 2
n/cm (E > 1.0 MeV) (Axial Orientation) . . . . . . . . . . . . . . . . . . . . . . . . . 5-13
" ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
iv LIST OF TABLES (Continued)
Table 5-8 Instrumented Charpy impact Test Results for the Prairie Island Unit 1 Surveillance Weld Metal irradiated to a Fluence of 4.017 x 10 n/cm2 l ( E > 1.0 M eV) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-14 Table 5-9 Instrumented Charpy impact Test Results for the Prairie Island Unit 1 Heat-Affected-Zone (HAZ) Metal irradiated to a Fluence of 4.017 x 10
2 n/cm (E > 1.0 MeV) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-15 Table 5-10 Instrumented Charpy impact Test Results for the Prairie Island Unit 1 Correlation Monitor Material Irradiated to a Fluence of 4.017 x 10 n/cm' (E > 1.0 MeV) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-16 Table 5-11 Effect of Irradiation to 4.017 x 10 n/cm (E > 1.0 MeV) on the Notch Toughness Properties of the Prairie Island Unit 1 Capsule S Reactor Vessel Surveillance Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-17 1
1 Table 5-12 Comparison of the Prairie Island Unit 1 Surveillance Material 30 ft-Ib <
Transition Temperature Shifts and Upper Shelf Energy Decreases with
, Regulatory Guide 1.99, Revision 2, Predictions . . . . . . . . . . . . . . . . . . . 5-18 i Table 5-13 Tensile Properties of the Prairie Island Unit 1 Reactor Vessel Surveillance Materials irradiated to 4.017 x 10 n/cm' (E > 1.0 MeV) . . . 5-19 j Table 6-1 Calculated Fast Neutron Exposure Rates and Iron Atom Displacement Rates at the Surveillance Capsule Center . . . . . . . . . . . . . . . . . . . . . . . 6-14 Table 6-2 Calculated Azimuthal Variation of Fast Neutron Exposure Rates and Iron Atom Displacement Rates at the Reactor Vessel Clad / Base Metal I nte rf ace . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-16 j Table 6-3 Relative Radial Distribution of $(E > 1.0 MeV) Within the Reactor Vessel W all . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 6-18 Table 6-4 Relative Radial Distribution of Q(E > 0.1 MeV) Within the Reactor Vessel Wall.................................................. 6-19
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Table 6-5 Relative Radial Distribution of dpa/sec Within the Reactor Vessel Wall . . 6-20 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSE! A8ntATION SURVEILLANCE PROGRAM
- - v LIST OF TABLES (Continued)
Table 6-6 Nuclear Parameters Used in the Evaluation of Neutron Sensors . . . . . . 6-21 Table 6-7 Monthly Thermal Generation During the First Seventeen Fuel Cycles of the Prairie Island Unit 1 Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-22 Table 6-8 Measured Sensor Activities and Reaction Rates, Surveillance Capsule S Saturated Activities and Reaction Rates . . . . . . . . . . . . . . . . . . . . . . . . 6-26 Table 6-9 Summary of Neutron Dosimetry Results Surveillance Capsules S, R. P, andV................................................. 6-30 I l
Table 6-10 Comparison of Measured and FERRET Calculated Reaction Rates at !
the Surveillance Capsule Center . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-31 l Table 6-11 Adjusted Neutron Energy Spectrum at the Center of Surveillance C ap s ul e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-33 j Table 6-12 Comparison of Calculated and Measured integrated Neutron Exposure of Prairie Is!and Unit 1 Surveillance Capsules S, R, P, and V . . . . . . . . . 6-37 Table 6-13 Neutron Exposure Projections at Key Locations on the Reactor Vessel Clad / Base Metal Interface . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-38 Table 6-14 Neutron Exposure Values Within the Prairie Island Unit 1 Reactor Vessel ................................................ 6-39 Table 6-15 Updated Lead Factors for Prairie Island Unit 1 Surveillance Capsules . . 6-40 Table 7-1 Recommended Surveillance Capsule Removal Schedule for the Prairie Island Unit 1 Reactor Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
l vi i
LIST OF ILLUSTRATIONS 1
Figure 41 Arrangement of Surveillance Capsules in the Prairie Island Unit 1 l
R eactor Ve ssel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-8 I l
Figure 4 2 Capsule S Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-9 l
Figure 5-1 Charpy V-Notch impact Energy vs. Temperature for Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C (Tangential Orientation) . . . 5-20 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C (Tangential l
O rie ntation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-21 '
Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C (Tangential Orientation) . . . 5-22 l l
Figure 5-4 Charpy V-Notch impact Energy vs. Temperature for Prairie island Unit 1 '
Reactor Vessel intermediate Shell Forging C (Axial Orientation) . . . . . . . 5-23 l l
Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Prairie Island Unit 1 Reactor Vessel intermediate Shell Forging C (Axial Orientation) . . 5-24 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C (Axial Orientation) . . . . . . . 5-25 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for Prairie Island Unit 1 Reactor Vessel Weld Metal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-26 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for Prairie Island Unit 1 Reactor Vessel Weld Metal . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-27 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for Prairie Island Unit 1 Reactor Vessel Weld Metal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-28 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for Prairie Island Unit 1 Reacter Vessel Heat-Affected-Zone (HAZ) Metal . . . . . . . . . . . . . . . . . . 5-29 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for Prairie Island Unit 1 Reactor Vessel Heat-Affected-Zone (HAZ) Metal . . . . . . . . . . . . . 5-30 ANALYSIS CF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVE!LLANCE PROGRAM
VII LIST OF ILLUSTRATIONS (Continued) i i
Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for Prairie Island Unit 1 Reactor Vessel Heat-Affected-Zone (HAZ) Metal . . . . . . . . . . . . . . . . . . 5 31 Figure 513 Charpy V Notch Impact Energy vs. Temperature for Prairie Island Unit 1 Reactor Vessel Correlation Monitor Material . . . . . . . . . . . . . . . . . . . . . 5-32 Figure 5-14 Charpy V-Notch 1.ateral Expansion vs. Temperature for Prairie Island Unit 1 Reactor Vessel Correlation Monitor Material . . . . . . . . . . . . . . . . 5-33 Figure 5-15 I Charpy V-Notch Percent Shear vs. Temperature for Prairie Island Unit 1 Reactor Vessel Correlation Monitor Material . . . . . . . . . . . . . . . . . . . . . 5-34 Figure 5-16 Charpy impact Specimen Fracture Surfaces of the Prairie Island Unit 1 )
Reactor Vessel Intermediate Shell Forging C (Tangential Orientation) . . . 5-35 l Figure 5-17 Charpy impact Specimen Fracture Surfaces of the Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C (Axial Orientation) . . . . . . . 5-36 Figure 5-18 Charpy impact Specimen Fracture Surfaces of the Prairie Island Unit 1 80d.tcr Vessel Weld Metal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-37 Figure 5-19 Charpy impact Specimen Fracture Surfaces of the Prairie Island Unit 1 Reactor Vessel Weld Heat Affected-Zone (HAZ) Metal . . . . . . . . . . . . . 5-38 I
Figure 5-20 Charpy impact Specimen Fracture Surfaces of the Prairie Island Unit 1 l Reactor Vessel Correlation Monitor Material . . . . . . . . . . . . . . . . . . . . . 5-39 Figure 5-21 Tensile Properties for the Prairie Island Unit 1 Reactor Vessel 1 Intermediate Shell Forging C (Tangential Orientation) . . . . . . . . . . . . . . 5-40 Figure 5-22 Tensile Properties for the Prairie island Unit 1 Reactor Vessel Intermediate Shell Forging C (Axial Orientation) . . . . . . . . . . . . . . . . . . 5-41 Figure 5-23 Tensile Properties for the Prairie Island Unit 1 Reactor Vessel Weld Metal................................................. 5-42 Figure 5-24 Fractured Tensile Specimens from the Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C (Tangential Orientation) . . . . . . . . . 5-43 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
I viii J
LIST OF ILLUSTRATIONS (Continued)
Figure 5-25 Fractured Tensile Specimens from the Prairie island Unit 1 Reactor Vessel Intermediate Shell Forging C (Axial Orientation) . . . . . . . . . . . . . 5-44 Figure 5-26 Fractured Tensile Specimens from the Prairie Island Unit 1 Reactor Vessel Weld Metal ......................................5-45 l Figure 5-27 Engineering Stress-Strain Curves for Prairie Island Unit 1 Reactor j Vessel Intermediate Shell Forging C Tensile Specimens N7 and N8
- (Tangential Orientation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-46 1
i l Figure 5-28 Engineering Stress-Strain Curve for Prairie Island Unit 1 Reactor Vessel i i
intermediate Shell Forging C Tensile Specimen N9 (Tangential
- Orie ntation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-47 i Figure 5-29 Engineering Stress-Strain Curves for Prairie Island Unit 1 Reactor Vessel intermediate Shell Forging C Tensile Specimens S7 and S8
- (Axial Orientation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-48 q Figure 5-30 Engineering Stress-Strain Curve for Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C Tensile Specimen S9 (Axial Orientation) . . 5-49 J
t I
Figure 5-31 Engineering Stress-Strain Curves for Prairie Island Unit 1 Reactor i Vessel Weld Metal Tensile Specimens W7 and W8 . . . . . . . . . . . . . . . . 5-50 Figure 5-32 Engineering Stress-Strain Curve for Prairie Island Unit 1 Reactor Vessel Weld Metal Tensile Specimen W9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-51 l Figure 6-1 Plan View of a Reactor Vessel Surveillance Capsule . . . . . . . . . . . . . . . 6-13 i
1 i
a ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPAfff PRAIRIE ISLAND Unit 1 REACTOR 'ESSEL RADIATION SURVEILLANCE PROGRAM
4
+
1-1 i
l 1.0
SUMMARY
OF RESULTS The analysis of the reactor vessel materials contained in surveillance dapsule S, the fourth capsule to be removed from the Northem States Power Company Prairie Island Unit 1 reactor pressure vessel, led to the following conclusions:
The capsule received an average fast neutron fluence (E > 1.0 MeV) of 4.017 x 10
2 n/cm after 18.12 Effective Full Power Years (EFPY) of plant operation.
l
+
Irradiation of the reactor vessel Intermediate Shell Forging C Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major rolling direction (tangential orientation), to 4.017 x 10 n/cm' (E > 1.0 MeV) resulted in a 30 ft-Ib transition temperature increase of 101.46 F and a 50 ft-lb transition temperature increase of 105.15 F. This results in an irradiated 30 ft-lb transition temperature of 4 62.55 F and an irradiated 50 ft-lb transition temperature of 98.80 F for the tangentially-oriented specimens.
Irradiation of the reactor vessel Intermediate Shell Forging C Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major rolling l direction (axial orientation), to 4.017 x 10 n/cm' (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 74.27'F and a 50 ft-Ib transition temperature increase of 76.68 F. This results in an irradiated 30 ft-Ib transition temperature of 42.95'F and ;
an irradiated 50 ft-lb transition temperature of 80.63'F for the axially-oriented i specimens.
Irradiation of the weld metal Charpy specimens to 4.017 x 10 n/cm' (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 160.43 F and a 50 ft-lb transition temperature increase of 170.84*F. This results in an irradiated 30 ft-lb transition temperature of 95.98'F and an irradiated 50 ft-lb transition temperature of 143.91*F.
=
frradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 4.017 x 10 n/cm 2(E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 137.11*F and a 50 ft-Ib transition temperature increase of 98.20 F. This results in an irradiated 30 ft-lb transition temperature of -62.89 F and an irradiated 50 ft-Ib transition temperature of -26.80 F.
Irradiation of the correlation monitor material Charpy specimens to 4.017 x 10 n/cm' (E > 1.0 MeV) resulted in a 30 ft-Ib transition temperature increase of 166.08'F and a 50 ft-lb transition temperature increase of 159.58 F. This results in an irradiated 30 ft-lb transition temperature of 212.29 F and an irradiated 50 ft-Ib transition temperature of 237.98 F.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
l j- 12 j . :
i
. The average upper shelf energy of Intermediate Shell Forging C (tangential orientation) 4 resulted in an energy decrease of 15.5 ft-lb after irradiation to 4.017 x 10 n/cm' (E >
j 1.0 MeV). This results in an irradiated average upper shelf energy of 142.5 ft-lb for the tangentially-oriented specimens. ,
! . The average upper shelf energy of Intermediate Shell Forging C (axial orientation) l resulted in an energy decrease of 8 ft-Ib after irradiation to 4.017 x 10 n/cm" (E > 1.0 l MeV). This results in an irradiated average upper shelf energy of 135 ft-Ib for the axially-oriented specimens.
. The average upper shelf energy of the weld metal Charpy specimens resulted in an !
8 energy increase of 6 ft-Ib after irradiation to 4.017 x 10 n/cm (E > 1.0 MeV). This results in an irradiated upper shelf energy of 84.5 ft-lb for the weld metal specimens.
. The average upper shelf energy of the weld HAZ metal decreased 75 ft-lb after irradiation to 4.017 x 10 n/cm' (E > 1.0 MeV). This results in an irradiated upper shelf energy of 136 ft-Ib for the weld HAZ metal.
- The average upper shelf energy of the correlation monitor material decreased 41 ft-Ib !
after irradiation to 4.017 x 10 n/cm" (E > 1.0 MeV). This results in an irradiated upper shelf energy of 82.5 ft-lb for the correlation monitor material.
. The surveillance Capsule S test results indicate that all 30 ft-lb transition temperature l shifts are greater than the Regulatory Guide 1.99, Revision 2"), predictions. However, ;
the shift values are less than the two-sigma allowance required by Regulatory Guide i 1.99, Revision 2 for all of the materials except intermediate shell forging C (tangential orientation) and the weld metal.
. The surveillance Capsule S test results indicate that all average upper shelf energy decreases of the surveillance materials are less than the Regulatory Guide 1.99, Revision 2, with exceptic.7 of the correlation monitor material.
. The surveillance capsulo materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are expected to maintain an upper shelf energy of no less than 50 ft-Ib throughout the life of the vessel (35 EFPY) as required by 10 CFR Part 50, Appendix Graj ,
I l
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 ;
REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
. - . _ _ . _ _ _ . _ . . _ . _ ~ _ _ . .
13 8
The calculated 35 EFPY maximum neutron fluence (E > 1.0 MeV) for the Prairie Island Unit 1 reactor vesselis as follows:
, Vessel inner radius * = 3.07 x 10 n/cm' s
Vessel 1/4 thickness = 1.96 x 10 n/cm" t
, Vessel 3/4 thickness = 6.02 x 10 e n/cm8 j
- Clad / base metal interface i
i 1
l l
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM i
i
! .. ~
21 i
l
2.0 INTRODUCTION
This report presents the results of the examination of Capsule S, the fourth capsule to be
- removed from the reactor in the continuing surveillance program which monitors the effects of i neutron irradiation on the Northem States Power Company Prairie Island Unit 1 reactor I pressure vessel materials under actual operating conditions.
f The surveillance program for the Northem States Power Company Prairie Island Unit 1 reactor i .
pressure vessel materials was designed and recommended by the Westinghouse Electric i Corporation. A description of the surveillance program and the pre-irradiation mechanical i
properties of the reactor vessel materials is presented in WCAP-8086 entitled "Northem States l Power Co. Prairie Island Unit No.1 Reactor Vessel Radiation Surveillance Program
- The l surveillance program was planned to cover the 40-year design life of the reactor pressure l
4 vessel and was based on ASTM E185-70, " Recommended Practice for Surveillance Tests for
- l i Nuclear Reactor Vessels". Westinghouse personnel were contracted to aid in the preparation j of procedures for removing Capsule S from the reactor and its shipment to the Westinghouse Science and Technology Center Hot Cell Facility, where the oot.t-irradiation mechanical testing l of the Charpy V-notch impact and tensile surveillance specimeia was performed.
l This report summarizes the testing of and the post-irradiation data obtained from surveillance Capsule S removed from the Northem States Power Company Prairie Island Unit 1 reactor vessel and discusses the analysis of the data.
s ,
ANALYSTS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVELLLANCE PROGRAM
3-1 1
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 AS08 Class 3 (base material of the Prairie Island Unit 1 reactor pressure vessel) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensilo 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 Appendix G to Section XI of the ASME Boiler and Pressure Vessel CodeW. The method uses fracture mechanics concepts and is based on the reference nil-ductility temperature (RTxo7).
RTuo7 is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E208W) or the temperature 60 F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (tangential) to the major working direction of the forging. The RTuor of a given material is used to index that inaterial 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 j material is indexed to the K,, curve, allowable stress intensity factors can be obtained for this t material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.
RTuo7 and, in tum, the operating limits of nuclear power plants, can be adjusted to account for the effects of radiation on the reactor vessel material properties. The radiation embrittlement changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program, such as the Prairie Island Unit 1 Reactor Vessel Radiation Surveillance ProgramW, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTuoy) due to irradiation is added to the initial RTuo, to adjust the RTuoy for radiation embrittlement. This adjusted reference temperature (ART = initial RTuoy +AMuo7) is used to index the material to the K,, curve and, in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
+
. 4-1
4.0 DESCRIPTION
OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the Prairie Island Unit 1 reactor pressure vessel core region materials were laserted in the reactor vessel prior to initial plant start-up. The six capsules were positioned in the reactor vessel between the thermal shield and the vessel wall as shown in Figure 4-1. The test capsules are in baskets attached to the thermal shield. The vertical center of the capsules is opposite the vertical center of the core.
Capsule S was removed after 18.12 Effective Full Power Years (EFPY) of plant operation.
The capsule contained Charpy V-notch impact specimens made from intermediate Shell Forging C and weld metal which joined sections of material from the intermediate and lower shell rings, heat affected-zone, and ASTM correlation monitor material. Additionally, tensile and Wedge Opening Loading (WOL) specimens were included in the capsule (Figure 4-2).
Test material obtained from the Intermediate Shell Forging (heat-treated with the shell) was taken at least one forging thickness (6.692 inches) from the quenched edges of the forging.
All test specimens were machined from the 1/4-thickness location of the forging after
, performing a simulated postweld, stress-relieving treatment. Specimens were machined from weld metal and the heat-affected-zone (HAZ) metal of a stress-relieved weldment joining sections of the intermediate and lower shell forgings. All heat-affected-zone specimens were obtained from the weld heat-affected-zone of intermediate Shell Forging C. The A533 Grade B Class 1 material (HSST Plate 02) for the correlation monitor plate specirnens was supplied by the Oak Ridge National Laboratory from a 12-inch-thick plate.
Charpy V-notch impact specimens from Intermediate Shell Forging C were machined in both the axial orientation (longitudinal axis of specimen normal to major working direction) and tangential orientation (longitudinal axis of specimen parallel to major working direction). The core region weld Charpy impact specimens were machined from the weldment such that the long dimension of the Charpy was normal to the weld direction; the notch was machined such that the direction of crack propagation in the specimen was in the weld direction.
Tensile specimens were machined with the longitudinal axis of the specimen in the major working direction (tangential) and also normal to the major working direction (axial) of the shell ring forging.
WOL test spccimens were machined in a tangential direction so that the loading of the specimen would be in the major working direction of the forging with the simulated crack propagating in the axal direction. In addition, axial specimens 'were machined so that the loading of the specimens would be in the axial direction of the forging with the simulated crack propagating in the major working direction. All specimens were fatigue pre-cracked per ASTM 4
E399-70T.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIADON SURVEILLANCE PROGRAM
4-2 The heat treatment of the beltline region materials is presented in Table 4-1. The results of the chemical analyses on the unirradiated beltijne region materials are presented in Table 4-2, which were obtained from the surveillance program reporti ). Additionally, a chemical analysis ;
using inductively Coup led Plasma Spectrometry (ICPS) was performed on four irradiated l Charpy specimens, three weld metal and one base metal, and is reported in Table 4-3. The chemistry results from the NBS certified reference standards are reported in Table 4-4. The results were obtained from the Westinghouse Electric Corporation Nuclear Services Division i CMT Analytical Laboratory'l f Table 4-5 provides the calculations of the average Cu and Ni weight percent values of the reactor vessel beltline materials, which were used in the Prairie Island Unit 1 surveillance Capsule S calculations.
Capsule S contained dosimeter wires of pure copper, iron, nickel, and aluminum-0.15 weight percent cobalt wire (cadmium-shielded and unshielded). In addition, cadmium shielded dosimeters of neptunium (Np2 '7) and uranium (U23e) were placed in the capsule to measure the integrated flux ht specific neutron energy levels.
Thermal monitors made of two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsuie. These thermal monitors were used to define the maximum temperature attained by the test specimens during irradiation. The composition of the two alloys and their melting points are as follows:
2.5% Ag,97.5% Pb Melting Point: 579 F (304*C) 1.75% Ag,0.75% Sn,97.5% Pb Melting Point: 590*F (310*C) l The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in Capsule S is shown in Figure 4-2.
l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAlRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
4-3
) .
- TABLE 4-1 Heat Treatment of the Prairie Island Unit 1 Reactor Vessel Surveillance MaterialsDI Material Temperature ('F) Time (hours) Coolant 4
Heated to 1652/1715 Tempered at i **
1175/1238 Heated to 1652/1724
- 9 " "
l Intermediate Shell Tempered at Forging C 1202/1238 i
Stress Relieved at 8 Fumace-cooled 1022 Stress Relieved at 14 Furnace-cooled j 1112 Stress Relieved at 5 Fumace-cooled 1022 Weldment Stress Relieved at 7 Fumace-cooled 11'i2 1675 25 4 Air-cooled
(
1600 i 25 4 Water-quenched Correlation Monitor Material 1125 i 25 4 Fumace-cooled i "**
- 1150i25 40 to 600 F l
5 4
i d
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
4-4 TABLE 4-2 Chemical Composition (wt%) of the Unirradiated Prairie Island Unit 1 Reactor Vessel Surveillance Materials m
Element intermediate Shell Weld Metal Correlation Monitor Material F'orging C Ladie Check C 0.17 0.052 0.22 0.22 Mn 1.41 1.30 1.45 1.48 P 0.013 0.017 0.011 0.012 S 0.005 0.014 0.019 0.018 Si 0.28 0.36 0.22 0.25 Mo 0.48 0.51 0.53 0.52 4
Ni 0.72 - 0.62 0.68 Cr 0.17 0.015 - -
V <0.002 0.001 - -
Cu 0.06 0.13 - 0.14 ;
Co 0.010 0.001(* - -
Al 0.033 0.015 - -
1 N, 0.006 0.014 - -
Sn 0.007 0.007 - -
Zn 0.001 0.001'" - -
Ti 0.001t
- 0.001 - -
Zr 0.001 0.001 - -
As 0.011 0.061 - -
Sb 0.001 0.001 - -
B 0.003(* 0.003t * - -
No1 es:
(a) Not detected. The number indicates the minimum limit of detection.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
4-5 TABLE 4-3 Chemical Composition of the Prairie Island Unit 1 Charpy Specimens Removed from Surveillance Capsule S Base Metal Weld Metal Element S-25 W 18 W-22 W-23 a
Al 0.03 <0.02 <0.02 <0.02 As 0.03 0.14 0.13 0.12 8 <0.004 <0.004 <0.004 <0.004 Co 0.014 0.022 0.021 0.019 Cr 0.217 0.024 0.018 0.014 Cu 0.078 0.149 0.138 0.143 Mn 1.97 1.67 1.60 1,42 i
Mo 0.71 0.67 0.64 0.58 Ni 0.956 0.138 0.118 0.091 P 0.018 0.025 0.024 0.022 l Si 0.350 0.334 0.325 0.338 Sn <0.01 <0.01 <0.01 <0.01 Ti 0.005 <0.002 <0.002 0.017 j V <0.004 <0.004 <0.004 <0.004 Zr <0.01 <0.01 <0.01 <0.01 -
Carbon 0.182 0.067 0.072 0.060 i
l Sulfur 0.012 0.015 0.016 0.016 i
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
. . . - . , - . . _ _ _ - . . - . . . - - - - . . - . . - . . . - . . . _ . . . . . . . . . . - =. - . . ~ . - -. - . .-.
4 4-6 TABLE 44 Chemisti/ Results from the Low Alloy Steel NBS Certified Reference Standards Concentration in Weight Percent NBS-362 NBS-121d i
Eiernent Measured Certified Measured Certified i
1 At 0.06 0.09 - -
- As 0.10 0.09 - --
l B -
0.003 - -
i Co 0.32 'O.30 - --
! Cr 0.29 0.30 - -
i l Cu 0.50 0.50 - -
Mn 1.17 1.04 - -
Mo 0.069 0.068 - - ;
Ni 0.62 0.59 - -
P 0.03 0.04 - -
Si 0.410 0.39 - -
Sn 0.017 0.016 - -
Ti 0.030 0.08 - -
V 0.042 0.040 -- -
Zr 0.22 0.19 - -
NBS-362 NBS 121d C 0.161 0.160 - -
S -- -
0.013 0.013 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATI i POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRA4
,. 4-7 1
1
- TABLE 4-5 '
, Calculation of Average Cu and Ni Weight Percent Values for Beltline Materials A533 Gr. B, CL1
, Inter. to Lower Shell Correlation Monitor j
. Intermediate Shell Lower Shell Circumferential Material
{ Forging C(*) Forging D Weld *" (HSST Plate 02) l Ref. Cu % Ni % Cu % Ni % Cu % Ni % Cu % Ni %
8 0.06 0.72 2
8 0.06 0.72 9 0.07 0.66 j 9 0.065 0.66 l
- 3 0.13 -
0.14 0.68 !
6 0.13 0.09 7 0.078 0.956 0.149 0.138 7 0.138 0.118
- 7 0.143 0.091 1
Avg. 0.07 0.80 0.07 0.66 0.14 0.11 0.14 0.68 NOTES:
(a) Surveillance program material (b) The surveillance weld specimens were made of the same wire and flux as the intermediate to lower shell circular seam (Wire UM 89, Heat Number 1752, UM 89 Flux, Batch Number 1230).
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEt RADIATION SURVEILLANCE PROGRAM
4 4-8 R 270* - REACTOR P VESSEL
/[ THERMAL SHIELD I
{
10*
i CAPSULE -'
I (TYP) -'
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- __ i _F-
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T
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90*
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l Figure 4-1 Arrangement of Surveillance Capsules in the Prairie Island Unit 1 Reactor Vessel l
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPAl# PRAIRIE ISLAND Unit 1 I REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
._m . .m. _ . _ _ .. _ ._ . _ . _ . . _ _ . . . _ _ _ . . . _ _ . . _ _ . _ . . . . _ _ . - _ _ . - _. .. . _
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SPECIMEM NUMBERDeG COCE: " EP Com S
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R - ASTM CORRELATION MONITOR W- WELD METAL CAPSULE S THEPMAL SMELD H - HFM AFFECTED ZONE MATERIAL VESSEL WALL i
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TO TOP OF VESSEL TO BOTTOM OF VESSEL Figur. 42 Cat =da S Diagram Showing the Location of Spedmans. Thermal uannars and Dosaneters i i
f
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- . .- - . - - _ . - - .=
i I. 5-1 I
5.0 TESTING OF SPECIMENS FROM CAPSULE S 5.1 Overview
, The post irradiation mechanical testing of the Charpy V-notch and tensile specimens was i
performed at the Remote Metallographic Facility (RMF) at the Westinghouse Science and Technology Center (STC). Testing was performed in accordance with 10 CFR Part 50, j Appendix H D 1, ASTM Specification E185-82"1, 1 and Westinghouse Procedure RMF 8402, 4
Revision 2, as modified by Westinghouse RMF Procedures 8102, Revision 1, and 8103, i
, 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 l l
! WCAP-8086'I No discrepancies were found.
Thermal monitors made from two low-melting point eutectic alloys sealed in Pyrex tubes were included in the capsule. Examination of the two low-melting point 579 F (304'C) and 590*F (310'C) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the specimens were exposed was less than 579 F (304'C).
The Charpy impact tests were performed per ASTM Specification E23-93ap21 and RMF Procedure 8103, Revision 1, on a Tinius-Olsen Model 74,358J machine. The tup (striker) of the Charpy machine is instrumented with a GRC 830-1 instrumentation system, feeding into an IBM compatible 486 computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (Eo). From the load-time curve (Appendix A), the load of general yielding (P ay ), time to general yielding (toy), maximum ,
load (Pu), and time to maximum load (tu) can be determined. Under some test conditions, a l sharp drop in load indicative of fast fracture was observed. The load at which fast fracture I was initiated is identified as the fast fracture load (Pr ), and the load at which fast fracture l terminated is identified as the arrest load (P). l l
The energy at maximum load (Eu) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is approximately equivalent to the energy l
required to initiate a crack in the specimen. Therefore, the propagation energy for the crack l
(E,) is the difference between the total energy to fractuce (Eo) and the energy at maximum '
load (Eu).
I ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
_. _ . . _ _ _ _ _ _ _ _ _ _ _ . _ . . _ _ . _ _ ~ _ . _ . _ _ _ . _ _ . - _ . _ _ . .
- - 5-2 i
- The yield stress (o y) was calculated from the three-point bend formula having the following
] expression:
L oy-Pay (1)
B(W-a) C l l
} where L is the distarre between the specimen supports in the impact testing machine; B is !
the width of the speenSF neasured parallel to the notch; Wis the height of the specimen, i measured perpendicularly to the notch; and als the notch depth. The constant Cis
! dependent on the notch flank angle ($), notch root radius (p), and the type of loading (Le., .
i l pure bending or three-point bending).
! I is a d =1 1 The ore ( =4 5 !
? L = 3.33PavW l 4
o y= Pay (2) ,
j B(W-a)a1.21 B(W-a)
Fcr the Charpy specimens, B is 0.394 in., Wis 0.394 in., and a is 0.079 in. Equation 2 then i reduces to: -
1 2
)
of=33.3* Pay (3) a
- where ey is in units of psi and P ay is in units of Ib. The flow stress was calculated from the j average of the yield and maximum loads, also using the three-point bend formula. i k l
- Symbol A is columns 4,5, and 6 of Tables 5-6 through 5-10 is the cross-sectional area under the notch of the Charpy specimens
i A=B+(W-a)=0.1241 square inches (4) i
[ Percent shear was determined from post-fracture photographs using the ratio-of-areas
! methods in compliance with ASTM Specification A370-92"81. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.
i
. Tension tests were performed on a 20,000-pound Instron Model 1115, split-console test i machine, per / STM Specification E8-93 t ui and E2192"53, and RMF Procedure 8102, Revision
] 1. The upper pull rod of the test machine was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches i per minute throughout the test.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
5-3 1
Extension measurements were made with a linear variable displacement transducer (LVDT) extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length is 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-93W .
Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a nine-inch hot zone. All tests were conducted in air.
Because of the difficulty in remotely attaching a ther'mocouple directly to the specimen, the following procedure was used to monitor specimen temperature. Chromel-Alumel thermocouples were inserted in shallow holes in the center, each and of the gage section of a l
dummy specimen, and in each grip. In the test configuration, with a slight load on the '
specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range of room temperature to 550 F (288 C). The upper grip was used to control the fumace temperature. During the actual testing, the grip temperatures were used to obtain desired specimen temperatures. Experiments indicate that this method is accurate to i2*F. l 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 l computed using the final diameter measurement.
5.2 Charov V-Notch Impact Test Results The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule S, which was irradiated to 4.017 x 10" n/cm* (E > 1.0 MeV), are presented in Tables 5-1 through 5-10. The unirradiated and Capsule S results, as well as the results from previously tested capsules, are presented in Figures 5-1 through 5-15. These figures were generated using the hyperbolic tangent curve-fitting program CVGRAPH, Version 4.1. The transition temperature increases and upper shelf energy decreases for the Capsule S materials are sumrnarized ' - ble 5-11.
Irradiation of the reactor vessen intermediate Shell Forging C Charpy specimens oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the forging (tangential orientation) to 4.017 x 10" n/cm2 (E > 1.0 MeV) (Figure 51) resulted in a 30 ft-lb transition temperature increase of 101.46 F and'a 50 ft-Ib transition temperature increase of 105.15 F. This resulted in an irradiated 30 ft-lb transition temperature of 62.55 F and an irradiated 50 ft-lb transition temperature of 98.80'F (tangential orientation).
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
.. 3
. . l 5-4 ;
\
l i
i The average upper shelf energy (USE) of the Intermediate Shell Forging C Charpy specimens ,
(tangential orientation) resulted in a energy decrease of 15.5 ft-Ib after irradiation to 4.017 x !
10 n/cm' (E > 1.0 MeV). This results in an irradiated average USE of 142.5 ft-lb (Figure 5-1).
Irradiation of the reactor vessel Intermediate Shell Forging C Charpy specimens oriented with l the longitudinal axis of the specimen perpendicular to the major rolling direction of the forging l (axial orientation) to 4.017 x 10 n/cm' (E > 1.0 MeV) (Figure 5-4) resulted in a 30 ft-lb
- transition temperature increase of 74.27'F and a 50 ft-Ib transition temperature increase of l i 76.68'F. This resulted in an irradiated 30 ft-lb transition temperature of 42.95'F and an I irradiated 50 ft-lb transition temperature of 80.63*F (axial orientation). ;
l The average upper shelf energy (USE) of the Intemediate Shell Forging C Charpy specimens !
(axial orientation) resulted in a energy decrease of 8 ft lb after irradiation to 4.017 x 10 n/cm' !
(E > 1.0 MeV). This results in an irradiated average USE of 135 ft lb (Figure 5-4). :
Irradiation of the surveillance weld metal Charpy specimens to 4.017 x 10 n/cm' (E > 1.0 MeV) (Figure 5-7) resulted in a 30 ft-lb transition temperature shift of 160.43'F and a 50 ft-lb l
transition temperature increase of 170.84*F. This results in an irradiated 30 ft-lb transition temperature of 95.98*F and an irradiated 50 ft-lb transition temperature of 143.91*F.
The average upper shelf energy (USE) of the surveillance weld metal resulted in an energy l 2
increase of 6 ft-lb after irradiation to 4.017 x 10 n/cm (E > 1.0 MeV). This resulted in an l I
irradiated average USE of 84.5 ft-lb (Figure 5-7).
Irradiation of the reactor vessel weld HAZ metal Charpy specimens to 4.017 x 10 n/cm 8 (E > 1.0 MeV) (Figure 510) resu'ted in a 30 ft-Ib transition temperature increase of 127.11*F and a 50 ft-Ib transition temperature increase of 98.20*F. This resulted in an irradiated 30 ft-Ib transition temperature of -62.89'F and an irradiated 50 ft-Ib transition temperature of
-26.80 F.
The average upper shelf energy (USE) of the weld HAZ metal resulted in an energy decrease 8
of 75 ft-lb after irradiation to 4.017 x 10 n/cm (E > 1.0 MeV). This resulted in an irradiated !
average USE of 136 ft-lb (Figure 5-10).
Irradiation of the reactor vessel correlation monitor material Charpy specimens to 4.017 x 10
2 n/cm (E > 1.0 MeV) (Figure 5-13) resulted in a 30 ft-Ib transition temperature increase of 166.08'F and a 50 ft-lb transition temperature increase of 159.58'F. This resulted in an irrad!ated 30 ft-Ib transition temperature of 212.29*F and an irradiated 50 ft-lb transition temperature of 237.98'F.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
5-5 The average upper shelf energy (USE) of the correlation monitor material resulted in an energy decrease of 41 ft-lb after irradiation to 4.017 x 10 n/cm8 (E > 1.0 MeV). This resulted in an irradiated average USE of 82.5 ft-lb (Figure 5-13).
The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-16 through 5-20 and show an increasingly ductile or tougher appearance with increasing test temperature.
A comparison of the measured 30 ft-Ib transition temperature increases and upper shelf energy decreases for the various Prairie Island Unit 1 surveillance materials with predicted values using the methods of NRC Regulatory Guide 1.99, Revision 2M, is presented in Table 5-12 and led to the following conclusions: ;
l The surveillance Capsule S test results indicate that all 30 ft-Ib transition temperature shifts are greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift values are less than the two-sigma allowance required by Regulatory Guide 1.99, i Revision 2 for all of the materials except intermediate shell forging C (tangential orientation) and the weld metal.
The surveillance Capsule S test results indicate that all average upper shelf energy I decreases of the surveillance materials are less than the Regulatory Guide 1.99, '
Revision 2, predictions with exception of the correlation monitor material.
The Charpy V-notch property changes presented in WCAP-8086, WCAP-8916, WCAP-10102, i and WCAP-11006 are based on hand-fit Charpy curves using engineering judgement. ;
However, the results presented in this report are based on a re-plot of the capsule data using j CVGRAPH, Version 4.1, a hyperbolic tangent curve-fitting program. Hence, Appendix B l
{ contains a comparison of the Charpy V notch shift results for each surveillance material, hand-
] fit versus hyperbolic tangent curve-fitting. Additionally, Appendix C presents the CVGRAPH, j Version 4.1, Charpy V notch plots and program input data.
2
! The load-time records for the Capsule S individual instrumented Charpy specimen tests are presented in Appendix A.
- 5.3 Tensile Test Results
) The results of the tenslie tests performed on the various materials contained in Capsule S,
- irradiated to 4.017 x 10 n/cm' (E > 1.0 MeV), are presented in Table 5-13 and are compared with unirradiated results as shown in Figures 5-21 through 5-23. l
^ ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
? S-6 The results of the tensile tests performed on the Intermediate Shell Forging C (tangential orientation) indicated that irradiation to 4.017 x 10" n/cm* (E > 1.0 MeV) caused a 12 to 14 ksi increase in the 0.2 percent offset yield strength and a 9 to 10 ksi increase in the ultimate tensile strength when compared to unirradiated data (Figure 5-21).
The results of the tensile tests performed on the Intermediate Shell Forging C (axial orientation) indicated that irradiation to 4.017 x 10" n/cm" (E > 1.0 MeV) caused a 10 to 14 ksi increase in the 0.2 percent offset yield strength and a 8 to 12 kai increase in the ultimate l tensile strength when compared to unirradiated data (Figure 5-22).
The results of the tensile tests performed on the surveillance weld metal indicated that 2
irradiation to 4.017 x 10" n/cm (E > 1.0 MeV) caused a 9 to 16 ksi increase in the 0.2 percent offset yield strength and a 5 to 10 ksi increase in the ultimate tensile strength when compared to unitradiated data (Figure 5-23).
The fractured tensile specimens for the Intermediate Shell Forging C material are shown in Figures 5-24 and 5-25, while the fractured specimens for the surveillance weld metal are shown in Figure 5-26. The engineering stress-strain curves for the tensile tests are shown in Figures 5-27 through 5-32.
5.4 Wedae Openina Loadina (WOL) Specimens Per the surveillance capsule testing contract with the Northem States Power Company, WOL specimens will not be tested. The specimens will be stored at the Westinghouse Science and Tedinology Center Hot Cell.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
- I I l 5-7
s l i l
- TABLE 5-1 i Charpy V-notch Data for the Prairie Island Unit 1 Intermediate Shell Forging C '
{ Irradiated to a Flunce of 4.017 x 10" n/cm' (E > 1.0 MeV) .
I (Tangential Orientation)
Sample Temperature Impact Energy Lateral Expansion Shear k Number (*F) ('C) (ft-Ib) (J) (mils) (mm)
(%)
! N27 -25 -32 9 12 4 0.10 5 '
1
! N28 25 -4 10 14 10 0.25 5 l N32 26 -3 33 45 26 0.66 10 i
- N25 72 22 35 47 29 0.74 10 4
l N31 100 38 51 69 37 0.94 15 I
J j N29 125 52 68 92 51 1.30 30 ;
i N35 150 66 74 100 54 1.37 35 l N34 175 79 106 144 77 1.96 60 i
- N26 250 121 136 184 93 2.36 100 l
N36 300 149 150 203 80 2.03 100 i
! N30 350 177 1 51 205 90 2.29 100 l
I I
- N33 400 204 133 180 94 2.39 100 t I ;
4 i ,
I i ,
1 1
1 b
i l
)
4 a
4 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATK)N SURVEILLANCE PROGRAM
'
- 5-8 ,
1 TABLE 5-2
, Charpy V-notch Data for the Prairie Island Unit 1 Intermediate Shell Forging C 8 2 Irradiated to a Fluence of 4.017 x 10 n/cm (E > 1.0 MeV)
(Axial Orientation) 1 l
Sample Temperature impact Energy Lateral Expansion Shear i i
Number ('F) PC) (ft-lb) (J) (truL) (mm) (%)
S25 25 -32 7 9 2 0.05 0 )
S34 25 -4 22 30 14 0.36 5 S29 50 10 19 26 14 0.36 10 S28 60 16 49 66 36 0.91 20 S30 72 22 63 85 45 1.14 25 S27 100 38 59 80 46 1.17 30 S33 125 52 69 94 50 1.27 40 l S36 175 79 97 132 66 1.68 65 S26 225 107 135 183 90 2.29 100 S31* 250 121 -- -- -- - -
S35 250 121 138 187 85 2.16 100 S32 300 149 132 179 70 1.78 100
?
NOTE:
- Specimen alignment error. Data is not valid.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRA!RIE ISLAto Unit 1 REACTOR VESSEL RADIATION SURVEILi>.NCE PROGRAM
i
, 5-9 TABLE 5-3 l Charpy V notch Data for the Prairie Island Unit 1 Surveillance Weld Medal Irradiated to a Fluence of 4.017 x 10" n/cm' (E > 1.0 MeV) 3 Sample Temperature Impact Energy Lateral Expansion Shear f Number (*F) (*C) (ft-lb) (J) (mits) (mm) (%)
{ W23 -25 -32 8 11 3 0.08 10 W18 25 -4 17 23 12 0.30 10
- W22 72 22 24 33 16 0.41 40 i W21 100 38 28 38 24 0.61 60 i
! W19 150 66 45 61 37 0.94 80 t -
W24 175 79 67 91 53 1.35 90 W17 225 107 77 104 64 1.63 100 W20 300 149 92 125 71 1.80 100 i
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
5-10 I
l l TABLE 5-4 4
Charpy V-notch Data for the Prairie Island Unit 1 Heat-Affected-Zone (HAZ) Metal i
Irradiated to a Fluence of 4.017 x 10 n/cm (E > 1.0 MeV) 4 Sample Temperature Impact Energy Lateral Expansion Shear Number ('F) ('C) (ft-lb) (J) -(mils) (mm) (%)
! H17* -100 73 - - - - -
l 3 H21 -100 -73 20 27 8 0.20 10 i H24 -50 -46 32 43 16 0.41 25 H22 0 -18 82 111 50 1.27 30 i
! H2O 50 10 58 79 40 1.02 50 H23 72 22 143 194 71 1.80 60 i
i H19 175 79 149 202 74 1.88 100 l -
H18 SP 149 123 167 82 2.08 100
- Specimen alignment error. Data is not valid.
i 1
i
)
3.
1 1
i ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
. _ _ - . _ _ _ . _ _ . _ _ _ _ . _ _ _ . . . ~ . _ . _ . _ . _ - _ . . _ . _ . . _ . _ . _ . _ _ . . _ _ . _ . . _ . _ . . . . . . . _ _ _ _ . _ . _ _ _ . _ . .
5-11
~
t 4
TABLE 5-5 i
!. Charpy V-notch Data for the Prairie Island Unit 1 Correlation Monitor Material j
Irradiated to a Fluence of 4.017 x 10 n/cm' (E > 1.0 MeV) :
, Sample Temperature impact Energy Lateral Expansion Shear 4
l Number ('F) ('C) (ft-lb) (J) er (mils) (mm) (%)
R17 150 66 9 12 9 0.23 15
, R22 200 93 21 28 14 0.36 20 l
R24 206 97 19 26 13 0.33 15 !
i -
- R18 225 107 47 64 29 0.74 30 1
R21 250 121 57 77 41 1.04 55 !
e i
R23 300 149 76 103 57 1.45. 80 I
l R19 350 177: 78 106 64 1.63 95 4 R20 400 204 87 118 54 1.37 100 i
2 i
i i
i i
I a
d l
l i
i 4
i ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND LMt 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
. . _ _ _. m _ ._ _ __.. ___. ._. ._ _ _ _ _ _ . .
TABLE 5-6 mg F Instrumented Charpy impact Test Results for the Prairie Island Unit 1 Intermediate Shell Forging C Bh Irradiated to a Fluence of 4.017 x 10 n/cm' (E > 1.0 MeV) l
- g g (Tangential Orientation) n u) 3 Normalized Energies g Fractum Arnst Yield How n-Ildaz Yield Time to Men. 'I1me to gm Sample Test Charpy Imed Mez. Lead Lead Stress Stress Energy Charpy Man. Prop. Imed Yleid Number Temp h E.JA E,/A Ob) (msec) Ob) (msec) Ob) Ob) (ks0 (ks0 mm ('O (R-Ib) eda
'E 9 72 48 24 3724 0.14 3881 0.17 3881 0 124 126 yuj N27 -25 0.15 3737 0.2 3723 77 117 121 l N28 25 10 81 60 20 3520 ymD
- jg N32 N25 26 72 33 35 266 282 219 231 47 51 3361 3607 0.16 0.15 4427 4640 0.51 0.51 4427 4640 0
0 112 120 129 137 0.15 4482 0.67 4413 505 114 131 l 5 N31 100 51 411 309 101 3426 4328 0.67 4175 1172 108 126 68 548 299 249 3254 0.14 h N29 125 126 0.14 4336 0.67 4182 1732 108 N35 150 74 5% 300 296 3251 ,
4323 0.84 3222 1899 106 125 106 854 383 470 3198 0.14 y N34 175 4252 0.83 N/A N/A 107 124 j l
in N26 N36 250 300 136 150 1095 1208 371 374 724 834 3209 3118 0.17 0.16 4233 0.83 N/A N/A 104 122 279 936 2933 0.14 4096 _ 0.67 N/A N/A 97 117 ;
N30 350 151 1216 h 267 804 2556 0.14 3902 '
O.67 N/A N/A 85 107 5 N33 400 133 1071 2
b i
l m
7
F,m_4 .,3 .k. *i.*A,- --.k. _-t__A A 4 4 AL A. - _ aira_ 4.~ JA. 4--+.A_.4. r._ b _ m- ._4 & o.J _4 m bi L--- 4 -+ -e.
m> TABLE 5-7 Instrumented Charpy impact Test Results for the Prairie Island Unit 1 Intermediate Shell Forging C hh Irradiated to a Fluence of 4.017 x 10 n/cm (E > 1.0 MeV) 2 88
,1 g (Axial Orientation)
Mo P3 815 '
5 en 93
$ Normalized Energies b Sample Test Charpy it-itdn3 Yield Time to Max. Time to Fracture Arrest Yleid How Prop. Lead Yield Lead Max. Lead Load Stress Stress Number Temp Energy Charpy Max.
h Ob) (ks0 (ksi) mM ('F) (ft-Ib) eda eda EgA Ob) (msec) Ob) (msec) Ob)
E 0.13 3590 0 119 119 y"a S25 -25 7 56 29 27 3590 0.13 3590 32 3608 0.14 4225 0.36 4225 0 120 130 g$ S34 25 22 177 145 56 3458 0.14 3857 0.28 3857 175 115 121
}g S29 S28 50 60 19 49 153 395 97 316 78 3538 0.15 4552 0.67 4510 0 118 134
" 200 3431 0.14 4467 0.67 4324 101 114 131 S30 72 63 97 308 4
3363 0.15 4385 0.67 4336 931 112 129
@ S27 100 59 475 304 171 0.15 4367 0.67 4179 1045 111 128
{4
~
S33 S36 125 175 69 97 556 781 297 291 258 490 3346 3211 0.14 4256 0.67 3622 2111 107 124 S26 225 135 1087 375 712 3179 0.16 4214 0.84 N/A N/A 106 123 l - --- - - - -
in S31* 250 - - - - - -
S35 250 138 1111 377 734 3227 0.19 4320 0.83 N/A N/A 107 125 S32 300 132 1063 287 775 3046 0.14 4217 0.67 N/A N/A 101 121 o
- Speci:r.en Alignment Error. ua is not valid.
h rw
s TABLE 5-8 yR
!! Instrumented Charpy impact Test Results for the Prairic Island Unit 1 Weld Metal gE m trradiated to a Fluence of 4.017 x 10" n/cm' (E > 1.0 MeV)
Ni9
'- Y l
en z
Neriaalized Energies Time to Fracture Arrest Yleid How b* Sample Test l Charpy ittllvin8 Yield Timeto Men.
Prop. Lead Yield Lead Mez. Lead Lead Stress Stress Number Temp Energy Charpy Max.
E,/A (Ib) (msec) (Ib) (Ib) (ksi) (ksi) mg ('F) (ft.Ib) Ee/A Em/A Ob) (msec) 26 39 3273 0.13 3273 0.13 3273 561 109 109
$h W23 -25 8 16 4 0.16 3956 0.28 3956 300 124 127 gh W18 25 17 137 102 35 3719 3329 0.15 4058 0.34 4058 1322 111 123 W22 72 24 193 129 64 g 0.36 3964 1949 115 123 g W21 100 28 225 138 87 3470 0.14 3964
" 235 127 3185 0.2 3926 0.61 3926 2436 106 118 W19 150 45 362 67 540 280 259 3231 0.14 4029 0.66 3576 2497 107 121 W24 175 N/A f
m W17 W20 225 300 77 92 620 741 271 275 349 465 2998 3072 0.13 0.14 4001 3981 0.64 0.66 N/A N/A N/A 100 102 116 117 m
5 5
D
TABLE 5-9 gg Instrumented Charpy impact Test Results for the Prairie Island Unit 1 Heat-Affected Zone (HAZ) Metal o$ 2 Irradiated to a Fluence of 4.017 x 10" n/cm (E > 1.0 MeV)
$8 N9 h0 8s E,,
8,m 2 Normalized F6 Cy km m Sample Test Charpy ft-lb/int Yield Time to Man. Time to Fracture Arrest Yield How Yield Lead Max. land Lead Stress Stress charpy Max. Prop. Imad 2g Number Temp Energy E,/A (Ib) (mrec) (Ib) Ob) (ks0 (ks0 mQ M (ft-Ib) E4A Em/A (Ib) (msec) 3z - - - - - - -
g H17* -100 - - - -
0.19 4780 0.31 4780 0 150 154
{;g;( H21 -100 20 161 136 25- 4522 4640 036 4612 1321 139 146 H24 -50 32 258 159 99 4170 0.16 h 0 82 660 334 326 3870 0.16 4808 0.67 4366 0 129 144 1122 g 467 89 378 3911 0.19 3998 0.27 1405 269 130 131 H2O 50 58 0.16 4970 0.84 2823 1344 124 145 h H23 72 143 II51 437 714 3748 0.84 N/A N/A 117 135 fo H19 HI8 175 300 149 123 1200 990 409 297 791 693 3523 3269 0.14 0.14 4399 4419 0.67 N/A N/A 109 128 h
m
- Specimen Alignment Error. IData is not valid.
O r
f I
aA,
e
> TABLE 5-10 Instrumented Charpy impact Test Results for the Prairie Island Unit 1 Correlation Monitor Material h"h Irradiated to a Fluence of 4.017 x 10" n/cm* (E > 1.0 MeV) 88 i
rii % '
8o
@h i bM
~~
> us
!a gb Normalized Energies Max. Time to Fractere Arrest Yleid Row j Charpy 1R-Iblin! Yleid Time to Semple Test Load Men. Imed Imed Stress Stress Energy ICharpy Men. Prop. 14ed Yield Number Temp '
(Ib) (mecc) (th) (Ib) (ks0 (ksi) eda EnfA E,/A (Ib) (msee)
M rr, ('E) (R-Ib) m"
"* 150 9 72 40 32 3795 0.16 3795 0.16 3795 74 126 126 R17 4087 032 4087 759 117 126 200 21 169 119 50 3527 0.15 R22 117 125 3992 0.29 3992 545 b R24 206 19 153 104 49 3527 0.14 0.51 4633 1700 120 139 5 RIS 225 47 r ^: 243 135 3623 0.14 4734 0.63 4498 2494 115 132 R21 250 57 EI 293 166 3449 0.14 4518 2495 110 127 0.14 4300 0.65 3721 y R23 300 76 65 [i 295 317 323 3322 3379 0.14 4419 0.65 3253 2479 112 130
' RI9 350 78 628 305 3466 0.15 4474 0.65 N/A N/A i15 132 R20 400 87 701 306 395 iE
TABLE 5-11 gg Effect of irradiation to 4.017 x 10" n/cm2 (E > 1.0 MeV) on the Notch Toughness a$
88 Properties of the Prairie Island Unit 1 Capsule S Reactor Vessel si? Surveillance Materials (*)
O 3N Average 30 ft4b Transition Average 35-ma Lateral Expansion Ave age 50 ft-!b Transition Average Energy Absorption at g r.n Temperature (*F) fur Shear (ft4b)
Temperature (*F) Temperature (*F) 5g Material Unirr." trrad. AT Unirr." Irrad. AT Unirt." trrad. AT Unirr." trrad." AE
{O r'N'"
Z fntermediate Shen
-3131 42.95 7427 -13.05 75.02 88.07 3.95- 80.63 76.68 143 135 -8 Forging C m rn ( 80 N
No* Intermediate Shell 98 80 105.15 158 142.5 -15.5 bg Forging C -38.91 62.55 101.46 -2428 88.08 112.37 -6.35 (Tangential) f 95.98 160.43 -50.79 132.74 183.54 -26.93 143.91 170.84 78.5 84.5 6 o Weld Metal -64.44 M
N -200;00" 127.11 -152.00* -7.49 144.51 -125.00* -26.80 98.20 211 136 -75 HAZ Metal -62.89 8
% 4620 166.08 58.63 238.47 179.83 78.39 237.98 159.58 123.5 82.5 -41 a]on Monitor 21229 5 .
o h NOTES:
@ (a) All values obtained from CVGRAPH Version 4.1 results.
y (b) These values differ from those reported in WCAP-11006. Those reported in WCAP-11006 were developed g from hand-fit curves using engineering judgement while the values reported here were determined from c curves generated by CVGRAPH, Version 4.1.
$ (c) Values determined per the definition of " upper shelf energy" given in ASTM E185-82.
(d) Because the hyperbolic tangent curve (itting process did not provide a smooth S-shaped curve for the unirradiated data, these values have been retained from the original unirradiated Charpy V-notch hand fit curves (WCAP-8086).
V' G
i 5-18 i
~
\
i TABLE 5-12 Comparison of the Prairie Island Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predctions Capsule 30 ft-Ib Transition Upper Shelf Energy Fluence Temperature Shift Decrease CAPSULE (10" n/cm'.
I E>1.0MeV) Predicted'* Measured
- Predicted'* Measured MATERIAL ('F) (*F) (%) (%)
V 0.5630 36.9 24.07 16.5 0 l INTER. SHELL FORGING C (Axial Orientation) P 1.318 47.4 33.98 20.5 5 l l
l R 4.478 60.7 84.18 27 10 l l
S 4.017 59.7 7427 26.5 6 V 0.5630 36.9 56.36 16.5 9 INTER. SHELL I
FORGING C (Tangential P 1.318 47.4 23.11 20.5 10 Orientation)
R 4.478 60.7 95.85 27 8
]
1 S 4.017 59.7 101.46 265 10 V 0.5630 59.5 34.38 25 0 WELD METAL P 1.318 76.4 45.15 30 0 4.478 91.8 122.47 40 4 R
S 4.017 96.2 160.43 39 0 Heat Affected V 0.5630 - 0.00 -
(c)
Zone Material P 1318 -- 74.65 -- 32 R 4.478 - 149.69 -- 54 S 4.017 - 137.11 -- 36 V 0.5630 85.6 102.84 20 26 Correlation Monitor Material P 1.318 109.9 161.40 24.5 31 (HSST Plate 02)
(Longitudinal Onentalei) R 4.478 140 8 193.72 33 30 S 4.017 138.4 166.08 32 33 NO) t:6:
(a) Based on Regulatory Guide 1.99, Revision 2, rnethodology using the Cu and Ni weight percent and capsule fluence values.
8 (b) The Charpy data was fit using the hyperbolic tangent curve fitting program CVGRAPH Version 4.1 '3 (See Figures 5-1,5-4,5-7,5-10 and 5-13.)
(c) Upper Shett Energy not obtainable due to toughness, per WCAP-8916.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 AC&OTOQ UCCCel Aant ATION Clim\/Cill ANFC PAMAAU
_.-..__m__.__m._______
_ _ . _ _ _ _ _ . . _ . _ _ . . . . . . - _ _ _ _ . . _ _ _ _ ..~.......-._._._._m.-..- -. ...____m____..
. .i .
t m '
E'g s4=6 is 9 8o P$
Hsw TABLE 5-13 I
b3 !
w Tensile Properties of the Prairie l<Jand Unit 1 Reactor Vessel Surveillance Materials Irradiated to 4.017 x 10 n/cm'(E > 1.0 MeV)
E 31 !
.b Material Sample Test 0.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Number Temp. Strength Strength Load Stress Strength Elongatio Elongatio in Area ;
' ho
(*F) (ksi) (ksi) (idp) (ks!) (ksi) n n (%)
, jg (%) (%)
8a Inter. Shell N7 125 77.4 93.7 2.75 166.0 56.0 10.5 26.0 66 gm Forging C 3 (Tangential) N8 250 75.9 90.7 2.70 186.6 55.0 9.0 22.4 66 M
=
N9 550 69.8 90.7 3.10 178.2 63.2 9.0 21.8 65 ;
s 94.7 3.05 - 207.2 62.1 10.1 22.8 70 j g Inter. Shell S7 125 78.2 m Forgmg C (Axial) S8 200 74.4 92.7 3.95 282.9 80.5 9.3 21.5 72 h
. m ,
8 S9 550 68.8 89.6 2.95 190.3 60.1 8.6 20.3 68 [
i 91.7 3.05 210.8 62.1 12.8 26.7 71 g Surveillance W7 125 78.9 f
-. Weld-Metal W8 200 83.5 86.6 3.30 212.9 67.2 12.0 25.0 68 j WP 550 75.4 91.7 3.35 195.8 68.2 9.6 21.6 65 1
1 4
e !
e
-. . _ _ _ _ - . - - _ _ . . ..m . __m . . _ ~ __ _ .
- l. 5-20 i
l CVGRAPH 41 Hyperbolic Tangent Curve Printed at 143414 on 11-06-1996 Results Curve Fluence ISE d-lSE L'SE d-USE Te30 d-T e 3 7 e 50 d-T e 50
, 1 0 219 0 158 0 -3891 0 -G 0
( 2 0 219 0 143 -15 17.44 5636 4434 5039 ,
3 0 219 0 142 -16 -15 3 2111 1692 23 7/
4 0 2J9 0 145 -13 5633 95115 9434 10119 5 0 2J9 0 1425 -15 5 6255 10L46 968 10515 300 m 250 4
l a
x 200 X e S .
5 130 m "z g
bV (g 4
.~ 100 - .
e/g , .
g i Q .!. a .
so er ;
a / fo-
_s. Q" U
-300 -200 - 100 0 100 2'00 300 400 500 600 Temperature in Degrees F Curve tegend Ic 2C 30 4^ 5-Data Set (s) Plotted Curve Plant Capsule Ma'terial Ori. Heatl 1 PIl UNIRR NEGING SA5083 LT 21918/38566 2 Pil V RRCLNG SA5083 LT 21918/38566 3 Pil P FORCING SA5083 LT 21918/38506 4 P!1 R FORCLNG SA5083 1T 21918/38566 5 Fil S mRCING SA5083 LT 21918/38586 Figure 5-1 Charpy V-Notch impact Energy vs. Temperature for Prairie Island Unit 1 Reactor Vessel Intermediate Sheli Forging C (Tangential Orientation)
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit i REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
.- - . ._ . - ~ ~ . - - - . - . - .. . . . . ..
l 5-21 j
1 CVCRAPli 41 Hyperbolic Tangent Curve Printed at it40:31 on 11-06-1996 Paults Curve fluence USE d-USE T e II35 d-T e 1135 1 0 953 0 -2428 0 2 0 af47 -8.93 4782 7 2.11
) 3 0 88.06 -724 934 3383 4 0 795 -15 3 8030 10515
- 5 0 92.05 -325 8&OB 112.37 i
a e
<n 150 a
x W too 5 . u s
_ ~
w n . "
.%,/g$g.p l
~
2 .
e.a o (15 M
0 6/ v *
- N
/ / /
/
1 a ,
/ VC
&fh u
s 'M I g
-300 -200 -100 0 100 200 300 400 500 600 i
Temperature in Degrees F
- Cune legend
, Ic 2C 30 4^ 5-i 4
i Data Set (s) Plotted Cune Plant Capsule Material Ori Heatl i i Pil UNIPJt MRCING SA5083 LT 21918/38566 2 Pil V NRCLNG SA5083 LT 21918/38566 3 Pil P MRCING SA5083 LT 21918/38566
- 4 Pil R mRCING SA5083 LT 21918/38566 5 P!l S NRCING SA5083 LT .21918/38566 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Prairie Island Unit 1 Reactor i
Vessel Intermediate Shell Forging C (Tangential Orientation) 2 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAlRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
. . . . = . --- .-. . . . . .
.. 5-22 CVGRAPH 41 Hyperbolic Tangent Curve Printed at 114147 on 11-06-1996 Results Curve Fluence Te50zShear d-To5&AShear 1 0 35S 0 2 0 9032 54 S1 3 0 7432 2 22 4 0 13125 9534 5 0 18125 12534 100
//
gy r 8/
u 'j/? (
c5 x;
-cn o n JlI m r a e d /
0 o .i
/
/,.
lkh.l
?b
,%7 "J J u
AeM(+f, '
, i i
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve legend IC 2C 30 4^ 5-Data Set (s) Plotted Curve Plant Capsule Material Ori Heatl 1 PIl UNIRR FORGING SA500 LT 21918/31566 2 Pl1 V FORGING SA5083 LT 21918/38586 3 PIl P FORGING SA5083 LT 21918/38586 4 PIl R R)RGING S15083 LT 21918/38586 5 P11 S F0PGNG SA5083 LT 21918/38566 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C (Tangential Orientation)
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANOE PROGRAM
. - .- = - _ . - -
lj
.i .
5 23 e
CVGRAPH 41 Hyperbolic .angent Cune Printed at 163tl7 on 1141996 Paults Curve Fluence LSE d-ISE LE d-15E T e 30 d-T e 30 T e 50 d-T e 50 1 0 219 0 143 0 -3121 0 3S5 0 2 0 2 19 0 155 12 -724 24.07 2011 IE15 3 0 2.19 0 136 -7 236 33S6 5427 5022 4 0 219 0 129 -14 5237 8418 9955 95.8 5 0 219 0 135 -6 4295 7427 8043 76f8 300 t
l m 250 4 i I
c=., 2m h 0 o un _.._
s., 150 p ++
7 -
e ,,f gy - o .
e g y -
M Z
100 oj n/
v 1 > c ' 8, o , s y>, .
60 ^
V,7
- / l A _,
) O
+
ol i l i l
-300 -200 - 100 0 100 200 300 400 500 600 Temperature in Degrees F Cune kgend Ic 2C 30 4^ 5.
Data Set (s) Plotted a Curve Plant Capmle Material Ori. Heatl 1 Pil UNIRR FDRGING SA5083 TL 21918/38566
- 2 Fil V FDEING SA5083 R 21918/38'66 3 Pil P FDRGING SA50fD TL 21918/38566 4 Fil R FDRGING SA5083 TL 21918/38566 5 Pil S EDRGING SA5083 TL .21918/38566 Figure 5-4 Charpy V-Notch impact Energy vs. Temperature for Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C (Axial Orientation)
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIADN SURVEILLANCE PROGRAM
5-24 l
CVGRAPH 41 Hyperbolic Tangent Curve Printed at 16510 on !!-07-1996 1
Results Curve Fluence USE d-USE 7 e LDS d-T e LDS 1 0 96,01 0 -1105 0 2 0 7936 -16J5 18.92 3151
.3 0 61FI -34j4 1&l4 312
-4 0 8554 -10.47 8536 9821 5 0 8111 -1439 75112 8&D7 200 4
m
- 150 a
X N 100 -
es r V- r _ _ _ _ .
b es go 10
/
& /
[ v A 0 f l/ 9 "f yf r -
l Af l
l u 40 l
-300 -200 - 10 0 0 100 200 300 400 500 600 Temperature in Degrees F Curve igend I tc 2C 30 4^ 5. I Data Set (s) Plotted Curve Plant Capsule liaterial Ori Heatl l Pil 1 LINIRR MRGING SA500 TL 21918/3856E 2 Pl! V IDRGING SA500 TL 21918/38566 3 Pil , P IDRGING SA500 TL 21918/38566 4 Pil R IDRGING SA500 TL 21918/38566 5 PIl S FORCING SA500 TL 21918/38566 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Ternperature for Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging (Axial Orientation)
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
5-25 CVCRAPH 43 Hyyrbolic Tangent Curve Printed at 163757 on 1HT7-1996 Results Curve Fluenm To50xShear d-T o 50x Shear 1 0 4731 0 2 0 7537 2&O6 3 0 10112 5 i 31 4 0 140.03 9222 5 0 134.47 86.66 100 '
jg-
/ o u
- f e
0
.C .
9 m ,
- /
c
/
/
b a _ ,, / ,
y 40 a "
4 ~/ [
t/ /
20 o1 oH g 27 -
v .
U g
-300 -200 - 100 0 100 200 300 400 500 600 Temperature in Degrees F Cune legend 1C 2C 30 4^ Sv Data Set (s) Plotted Curve Plant Capsule Material Ori. Heatl 1 Pil LWIPJt MRCNG SA5083 TL 21918/38566 2 Fil V MRGING SA5083 TL 21918/38566 3 Pil P M RGNG SA5083 TL 21918/38566 4 Pil R M RC NC SA5083 TL 21918/38566 5 Pil S FORCNG SA5083 TL . '21918/38566 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Prairie Island Unit 1 Reactor Vessel Interrnediate Shell Forging C (Axial Orientation)
ANALYSIS OF CAPSULE S FROM THE HORTHERN STATES POWER COMPANY PRA!RIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
J 5-26 1
1 4
CVCPJPH 41 Hyperbolic Tangent Curve Printed at 15m53 on 11-06-1996 J
, Paults S
Curve Fluence ISE d-ISE ESE d-USE T e 30 d-T e 30 T e 50 d-T e 50 1 0 P.19 0 785 -6(44 0 0 -262 0 2 0 219 0 91 12 5 -302 34 2 2a42 47 2 3 0 219 0 83 45 -192 4115 4501 7L94 4 0 219 0 75 -35 5&O2 12147 1342 161]IB 5 0 4
' 19 0 84 5 6 95S6 160.43 14391 17034 30u m 25u
.O l
a z am h
tto L 150 D
c cza z' s ~M 1 '
D Su 89 y-ID o
7 mA<
u l
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Irgend Ic 2C -
30 4^ 5.
Data Set (s) Plotted Curve Plant C:psule Material Ori Heatl 1 Pil UNIRR WELD 1752 )
2 P!1 V IELD 1752 l 3 Pil P WELD 1752 4 Pli R VEID 1752 5 PIl S FELD 1752 I
Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for Prairie Island Unit 1 Reactor Vessel Weld Metal ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit i REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
~
\
5-27 l
CVCPJPH 4J Hyperbolic Tangent Cune Printed at 15072 on 11-06-1996 e
Results Cune fluence (JSE d-USE TELE 35 d-T e IE5 1 0 76.41 0 -50.79 0 2 0 7932 32
)
2258 7338 1 3 0 80JE 43 2555 7634 4 0 811 E68 117D4 16733 5 0 7536 -35 13174 1E54 i
i 20u I i
} :$ 150 i 5 c.
M 3
100 3
- o d O S ^ -- - '
$ 3 fo d',M~'
- a V
o f
- d 09 e o 4-- -f
& N "
c ~ ogj ~
/O /
U t j l l
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Cune legend 1c 2C -
30 4^ Sv Data Set (s) Plotted Cune Plant Capsule Material Ori. Heatl -
1 Pil UNIRR TE 1752 2 Pil V YE 1752 3 Pil P Wild 1752 4 Pll R fE 1752 l 5 Pil S IELD 1752 l
l l
Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for Prairie Island Unit 1 Reactor Vessel Weld Metal ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEll1ANCE PROGRAM
i l 5-28 l
t CVCRAPH 4J Hyperbolic Tangent Curve Printed at 15:10fTl on 11-06-1996 Results Curve Fluente T o 50x Shear d-T e 50/. Shear 1 0 '." 03
_ 0 2 0 4938 71.71 3 0 20.47 425 4 0 11108 135.11 5 0 89.06 111D9 15 o We u
- e /
7 -
c5 : 'i 0 o ,/ a
,c r4 60 n i
/
/
u C
D a a'd .
Ojo l }
o -
o # I ,
k @ i O
e-
,/,
i i j A.,
y
/
U
~J.fl l
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve lege,nd Ic 2 C)----- 30 4^ 5-Data Set (s) Plotted Curve Plant Capsule Material Ori. Heatl 1 Pil INRR YELD 1752 2 P11 V TELD 1752 3 Pil P TELD 1752 4 PIl R WELD 1752 5 Pil S WELD 1752 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for Prairie Island Unit 1 Reactor Vessel Weld Metal ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
-. .. - .. = . _ - . _ - _ . . . . - .- - - - _ _ _ . . - _ _ -
5-29 1 =
CVCPlfH 41 Hyperbolic Tangent Curve Printed at 16:44:41 on 11-07-1996 Results Curve Fluence ISE d-LSE USE d-USE T o 30 d-T e 30 T e 50 d-T o 50 1 0 0 211 0 200 0 125 0 2 0 0 <211' 0 200 0 -125 0 3 0 P.19 0 143 -68 -125.35 74.65 -88.8 36.2 4 0 U9 0 W - 11 4 -50.31- 149.69 -21.13 103.87
, 5 0 M9 0 13 6 -7a -62.89 137.11 26.8 98.2 000 to 250 a
,Q h *
/a a e o x ,/
150 l
- g a p v
c *
,.< ,/ 1
/[
- v g ,,
L, l
'" a -
g -
, y, /g[f o _ 2
/ A -
50 a ,
1 JJ, og/e U 1. 1
-300 -200 - 100 0 100 200 300 400 500 600 Temperature in Degrees F Curve legend Ic 2C 30 4^ 5-fata Set (s) Plotted Curve Plant Capsule Material Ort Heatf 1 Pli UNIRR HEAT AFFD ZONE -
2 Pil V HEAT AFFD 20h1 3 Pl! , P HEAT AFFD Z0h1 4 Pil R HEAT AFFD 20h1 5 Fil S HEAT AFFD 20h1
- Upper shelf impact energy not obtainable due to excessive toughness.
Figure 5-10 Charpy V-Notch impact Energy vs. Temperature for Prairie Island Unit 1 Reactor Vessel Heat Affected-Zone (HAZ) Metal ANALYSIS OF CAPSULE S FROM THE NORTHERN . . .TES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROuRAM
5-30 l
l CVGRAPH 41 Hyperbolic Tangent Curve Printed at 152900 on 11-Wr1996 Paults Curve Fluente i SE d-USE ToII35 d-ToIIl5 1 0 92 0 -152 0 2 0 87 -5 128 24 3 0 80.69 1131 -51.24 100.76 4 0 62.21 -29.79 -11.77 140.23 5 0 80.07 -11.93 -7.49 14431 200 4 l
I rn i
.O 150 , ;
$ i a :
x i 100 a _ l
- L DS ' '
s 2
Q) 0 G
-o, 5 c!, aig t ^ '
l d o - .
_ #/
i 2w a m uw
+y' -
, ,/ o - t ei o, 0 , ,,
.s /
0 i i t i
-300 -200 -100 0 100 200 000 400 500 600 Temperature in Degrees F Curve legend Ic 2C 30 4^ 5-I Data Set (s) Platted
)
Curve Plant Caosule !!aterial Ori. Heatl I 1 Pil (31RR HEAT AFTD ZONI 3 Pil V HEAT AFFD 20h1
. Pil P HEAT AFFD ZONE 4 P11 R HEAT AFFD Z0h1 5 Pl! S HEAT AFFD ZONI I
Figure 511 Charpy V-Notch Lateral Expansion vs. Temperature for Prairie Island Unit 1 Reactor Vessel Heat-Affected-Zone (HAZ) Metal ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISt.AND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
., 1 I
5-31
]
CCRAPH 4.1 Hypertdic Tangent Curve Printed at 152220 m 114-1996 1 Results Cune fluence T e 5t Shear d-T e 50x Shear 1 0 110 o 2 0 e g 3 0 -60.93 49.07 4 0 28.12 138.12 5 0 4135 151.85
- . _ ~ _ .
1 b; f
\ '//y7^
u 80 '
es j,,, // .
60 -
o ['
/
(( l o 7 .f m * ,
c a !
e o a 4:.
c
/
a 5 a
[, .
,)'
o /". -
20- so ,
,f i l O
W
/
u ; i i t i l
-300 -200 - 100 0 100 200 000 400 500 600 Temperature in Degrees F I Curve irgend 1C 2 C> 30 4^ 5-Data Set (s) Plotted Curve Plant Capsule Material Ori. Heatl 1 Pil l'.NIRR HEAT AFFD ZONI 2 Pli V HEAT AFFD 20h2 3 Pil P HEAT AFFD 20h2 4 Pil R HEAT AFFD 20h1 5 Pil S HEAT AFFD 20NT Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for Prairie Island Unit 1 Reactor Vessel Heat-Affected-Zone (HAZ) Metal ANALYSTS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
5-32 CVCRAPH 4j Hyperbolic Tangent Cune Printed at 152555 on 11-06-1996 Results Cune Fluena LSE d-ISE USE d-USE T o 30 d-T o 30 To50 d-T o 50 1 0 2J9 0 1235 0 4&2 0 7839 0 2 0 2.19 0 91 -325 149 2 10284 1945 11625 3 0 2.19 0 85 -385 207El 16L4 22E16 149.76 4 0 239 0 B6 -375 23933 19172 2Ba48 20206 5 0 239 0 825 -41 21229 166 2 237 2 159 2 300 en 250
.o I
a c= 2m X
t:0 6 150 0
C cza ey y"
g 100 g
,p O y Os a n -
1 50 m
a y 7 U
a . >- N-
-300 ~200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curre lagend ic 2C 3^ 4^ -
Sv Data Set (s) Plotted Cune Plant Capsule Material Ori Heati 1 Pil UNIRR SPE HSSlm LT SA53381 2 P11 V SPM IEIT LT SA533B1 ,
3 Pil P SPR HSSIV2 LT SA533B1 I 4 P11 R SPR l E ID2 LT SA533B1 j
$ Pil S SPM HSSIU2 LT SA5Il31 )
l Figure 5-13 Charpy V-Notch Irnpact Energy vs. Temperature for Prairie Island Unit 1 Reactor Vessel l l
Correlation Monitor Material ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
i* 5-33 CVCPJLPH 41 !!yperbolic Tangent Curve Printed at 152728 on Il-Wr-1996 Results I Curve Fluence USE d-USE 7eID5 d-TeIES i 1 0 8626 0 5&63 0 2 0 8064 -52 IT.96 13433 3 0 6032 -2533 21738 15 & 74 4 0 79M -719 29954 24a9 5 0 59 2 -2&74 23&47 17933 20u m
- = 150 6 i
- c. (
M N 100 l
3y k u o 4--
T
& $U
+M%.y3 ~~ ~-
O Q j g O jh '
I u i 2.AG i g
-300 -200 - 100 0 100 200 000 400 500 600 l Temperature in Degrees F Curve legend IC 2C 3^ 4^ 5.
Data Set (s) Plotted Curve Plant Capsule Material Ori Heatl 1 Pil UNIPJt SPR IEIU2 LT SA533B1 2 Pil V SPR I!SSID2 LT SA533B1 3 Pl! P SRM ESilt2 LT SA533B1 4 Pl! I, SPR I E ID2 LT SA533B1 5 Pl! S SPR llSSID2 LT SA533B1 l
i Figure 5-14 Charpy V-Notch Lateral Expansion vs. Temperature for Prairie island Unit 1 Reactor
[ Vessel Correlation Monitor Material ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
.- - ~_ . . - - _ - . . ...--- . . -.
O
. 5-34 CVCPJLPH 41 Hyperbolic Tangent Curve Printed at IE34f)4 on 11-06-1996 Results Curve Fluence 7o50xShear d-T e 50x Shear 1 0 8555 0 3 0 2171 2 131.47 3 0 22125 135.69 4 0 266.72 18116 5 0 24933 16428 fy/V u m V , '
/
C e o
.c e
" j w w .
a 8 ,
e c -
o .
% 40 of U
20 o
f
//- y 0 7E #
0 0 /4 X,
u ,
l
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve legend Ic 2C 30 4^ 5- ,
Data Set (s) Plotted Curve,. Plant Capsule Material Ort Heatl 1 Pil UNIPJt SRM ESS!12 LT SA53381 2 P11 Y SRM ESSm2 LT SA53381 3 P11, P SRM ESS!tl2 LT SA53381 4 P!1 R SPJi HSSm2 LT SA533B1 5 Pil S SRM HSSIU2 LT SA53381 Figure 5-16 Charpy V-Notch Percent Shear vs. Temperature for Prairie Island Unit 1 Reactor Vessel Correlation Monitor Material ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
_ _ _. . . . _ _ . . . . _ . _ _ _ _. ~. _
5-35 1 .
4 4
1 3
- ,/ [.: !'*d;h ;
- g, '
gg l w;y e w3 u
?
DW .J ' j.) ',,
l e .+
N27
{L N28 N32
'-)(%
.r.
.5 i
i i
j i
i MBW N31 N29 N35 N34 j N26 N36 N30 N33 1
I 1
l 1
! Figure 5-16 Charpy Impact Specimen Fracture Surfaces of the Prairie Island Unh 1 Reactor Vessel j intermediate Shell Forging C (Tangential Orientation) a 5
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM l
l
- -. ___ . - . . - - . - . . . . . _ _ - _ _- . - - . ~ . . . . _ - . _ . = _ _ _ . _ _ _ .
5-36 l
l t
i S25 S34 S29 S28 S30 S27 S33 S36 Specimen Alignment p
, ,p 2 1
- I Error ,
~, i g-* .t S26 S31 S35 S32 Figure 5-17 Charpy Impact Specimen Fracture Surfaces of the Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C (Axial Orientation) ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAtRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
- l 5 37 I
I i l l 1 8 3 .; -.y e ,l, '. W23 W18 W22 l l l f. 'i,:' . _ 2 1 - _. W21 W19 W24 Wl7 W20 Figure 5-18 Charpy impact Specimen Fracture Surfaces of the Prairie Island Unit 1 Reactor Vessal Weld Metal ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAlRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM i o 5-38 i - i j 1 Specimen J Alignment , i' d .' ), Error - 9 '4 ,;i i e , va . j '. J i L j H17 H21 H24 .;/ l m H22 H2O 23 l 1 I i 1 Mt H19 H18 i i i i 4 Figure 5-19 Charpy impact Specimen Fracture Surfaces of the Prairie Island Unit 1 Reactor Vessel
- ANALYSIS OF CAPSULE S FRG A THE NORTHERN STATES POWER COMPANY PRA!RIE ISLAND Und 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
5-39 l l l l 4 ', ji 1~ ' fl 1, , , l !l i. , ,, \ q l "*3D gg g'y42 paru m;g. L' 'rki g R17 R22 R24 M 1 - - >a nor*] ' s .t . , W6 , Vfsf t * ,pi em_ b R18 R21 R23 R19 R20 Figure 5-20 Charpy impact Specimen Fracture Surfaces of the Prairie Island Unit 1 Reactor Vessel Correlation Monitor Material ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 5-40 a j i (*C) l 0 50 100 150 200 250 300 350 { 120 l l l l l l- 800 110 ! 100 - 1 W . , ?! 90 - as - - 600 j 5 [ MATE TENEI MHy\ t , e g80 cx - = k " 2 - 500 70 - \ -1 / 60 - \ o'2 - 400 i 50 - l 40 1 4 , 80 g * %O 70 O O O O 60 - REDUCTIDN IN AREA 8 50 - T-. - 40 - T- . M c 30 - '\ TDTAL ELOCATIDN 20 - " 2 N 2 O t g 10 - 0
- 0 O UNIfDRM ELDNGATIDN I I I I I I O
O 100 200 300 400 500 600 700 TEMPERATURE (*f) TE O2 6 O UNIRRADIATED A 9 IRRADIATED AT 550*f, FLUENCE 4.017 x 100n/cm2 (E > LO Mev) Figure 5-21 Tensile Properties for the Prairie Island Unit 1 Reactor Vessel Interrnediate Shell Forging C (Tangential Orientation) ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM . 1 ! l 5-41 t i ('C)
- 0 50 100 150 200 250 300 350 10- 800 l l l l l-
- 110 00 100 I
c 90 - 600 5 ULTIMATE TDCILE STRENGTH / , w 0% = E ~ 70 E $ 3. STRENGTH
- v -
400 ! 50 - l l l l l l 00 40 i j 80
- REDUETION IN AREA 2' ;
- 70 -
O O j 0M i 60 - l j 8 50 - i > J - ] 40 b
- 30 - 2 \ 2 g N ~
TOTAL ELONGATIDN f 2
- i 0 20 -
i ,'c - 2 10 0 UNIrDRM ELONGAllDN ! I I I I 0 0 100 200 300 400 500 600 700 TEMPERATURE (*D A O UNIRRADIATED I9 A 9 IRRADIATED AT 550*f FLUENCE 4.017 x 10 n/cn2 (E > 13 MeV) Figure 5-22 Tensile Properties for the Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C (Axial Orientation) . ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM f 5-42 l (*C) 0 50 100 150 200 250 300 350 120 l l l l l 800 l l-110 i 700 l 100 m ULTIMATE TENSILE STRENGTH rn 90 - l x - 600 A O ~~ " 80 U - ^ " g ef ~ $70 - O \. 02% YIELD STRENGTH ~ \ -o 60 - o - 400 50 ~ I I I I I 40 80 70 QTSNAREA , O 60 - O 1 8 50 - l >- 1 , 40 . d 30 - 2 h - TDTAL ELDNGATIDN 20 - '7 C . \ -@ 10 UN!rDRM ELDNGATS U 0 1 0 100 200 300 400 500 600 700 TEMPERATURE (*f) N:lPC3 A O UNIRRADIATED I9 A 9 IRRADIATED AT 550'r FLUENE 4.017 x 10 n/cn2 E > 1.0 MeV) Figure 5-23 Tensile Properties for the Prairie Island Unit 1 Reactor Vessel Weld Metal ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM ! 5-43 4 I I l l Specimen N7 Tested at 125P j l h i r p, . ,t.'t , t . ; . -(_ _ ~ , ---- . . Specimen N8 Tested at 250P . . , i, Lo gY. a , a n.,L. ,c w %%%y;vm: r rr , ~74; J? ^ %?dLT: f i'di;Inogggge 4 Specimen N9 Tested at 550'F Figure 5-24 Fractured Tensile Specimens from the Prairie Island Unit 1 Reactor Vessel Intermediate
- Shell Forging C (Tangential Orientation)
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 5-44 l i 1 ! Specimen S7 Tested at 125cF t i I Specimen S8 Tested at 200'F Specimen S9 Tested at 550"F I i 1 e 5 Figure 5-25 Fractured Tensile Specimens from the Prairie Island Unit 1 Reactor Vessel intermediate Shell Forging C (Axial Orientation) ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM I 5-45 i j . i i i j l I ., ., . ), ~ ., , U"'l ., e ti { br v!t'n ibid. wit. f. ' ! M i !E..,. e 4 5 M ,i - llA
- L -
] I Specimen W7 Tested at 125cF j i 1 l l 7; j' 9 s,c36sR" "'i? - l l.. o i . a 1 i F. ?'16;' , { ! Specimen W8 Tested at 200oF i f k l Srcimen W9 Tested at 550oF i l 1 i 1 i i 1 '- Figure 5-26 Fractured Tensile Specimens from the Prairie Island Unit 1 Reactor Vessel Weld Metal l i i ANALYS:S OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM l 5-46 1 STRESS-STRAIN CURVE PRAIRIE ISLAND UNIT 1 "S" CAPSULE 100.00 90.00-j 80.00-70.00-m
- e0.00-i
@ 50.00-x g 40.00-30.00-N7
- 20.00-
- 125F 10.00-O.00 . . . . .
< 0.00 0.10 0.20 0.30 i STRAIN, IN/IN I STRESS-STRAIN CURVE PRAIRIE ISLAND UNIT 1 "S" CAPSULE 100.00 1 2 90.00-4 80.00-i 70.00- ~ i G 60.00-d l 50.00-1 g 40.00-30.00-20.00 N8 250 F 10.00-0.00 . . . 0.00 0.10 0.20 STRAIN, IN/IN i ) i i ! Figure 5-27 Engineering Stress-Strain Curves for Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C Tensile Specimens N7 and N8 (Tangential Orientation) 4 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLANO Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM _ .. _ _ _ _ _ _ ___ _ _ . . ~ _ . . _ _ _ __ , _ _ _ _ O 5-47 1 I l l f STRESS-STRAIN CURVE PRAIRIE ISLAND UNIT 1 "S" CAPSULE 100.00 90.00-80.00- _ 70.00-co
- 80.00-1 A
l $ cr: 50.00-g 40.00-30.00-20.00-10.00-0.00 . . , . 0.00 0.10 0.20 STRAIN, IN/IN I 1 l
- Figure 5-28 Engineering Stress-Strain Curve for Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C Tensile Specimen N9 (Tangential Orientation) l l
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVE!LLANCE PROGRAM 5-48 STRESS-STRAIN CURVE PRAIRIE ISLAND UNIT 1 "S" CAPSULE 100.00 90.00- , 80.00-70.00-E 80.00-vi @ 50.00-m g 40.00-30.00-S7 20.00-125 F 10.00-0.@ , . . . 0.00 0.10 0.20 STRAIN, IN/IN . STRESS-STRAIN CURVE PRAIRIE ISLAND UNIT 1 "S" CAPSULE 100.00 N 90.00-80.00- _ 70.00- $ 60.00-vi m. 50.00-g 40.00-30.00-20.00-200 F 10.00 0.00 . . . . 0.00 0.10 0.20 STRAIN, IN/IN Figure 5-29 Engineering Stress-Strain Curves for Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C Tensile Specimens S7 and S8 (Axial Orientation) ANALYSIS OF CAPSULE S FROM 1HE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 5-49 l l l l I l l STRESS-STRAIN CURVE PRAIRIE ISLAND UNIT 1 "S" CAPSULE 100.00 , l 90.00- - 80.00- j ._ 70.00- $ 60.00-u> ) @ 50.00- ! x g 40.00-30.00-20.00- S9 l 10.00- 550 F 0.00 , , , . 0.00 0.10 0.20 STRAIN, IN/IN Figure 5 30 Engineering Stress-Strain Curve for Prairie Island Unit 1 Reactor Vessel Intermediate Shell Forging C Tensile Specimen S9 (Axial Orientation) ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM - ) 5-50 I STRESS-STRAIN CURVE 1 PRAIRIE ISLAND UNIT 1 "S" CAPSULE > 100.00 3 90.00-80.00-70.00-E 80.00-50.00-m g 40.00-l 30.00- W7 20.M-
- 125 F 10.00-
] 0.00 . . . . 0.00 0.10 0.20 0.30 STRAIN, IN/IN STRESS-STRAIN CURVE PRAIRIE ISLAND UNIT 1 "S" CAPSULE 100.00 90.00-i 80.00-1. 70.00-a up I 60.00-VI 50.00-x l l g 40.00-i 30.00-3 W8 20.00-200F i 10.00-1 0.00 . . . . O.00 0.10 0.20 0.30 f' STRAIN, IN/IN Figure 5-31 Engineering Stress-Strain Curves for Prairie island Unit 1 Reactor Vessel Weld Metal Tensile Specimens W7 and W8 i a ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM .. . ..- _.-- - .- -.= .. . . ~ _ . . - . . . - . - . - . . ~ - . - . . - . - . . . . - . ..' 5-51 1 i s l t STRESS-STRAIN CURVE PRAIRIE ISLAND UNIT 1 "S" CAPSULE 100.00 l ~ 90.00- l 80.00- _. 70.00- i v> >
- eo.oo- l vi !
$ 50.00-cc g 40.00-30.00-20.00- wg 10.00- 550 F l o.00 . . , . 0.00 0.10 o.20 STRAIN, IN/lN i Figure 5-32 Engineering Stress-Strain Curve for Prairie Island Unit 1 Reactor Vessel Weld Metal Tensile Specimen W9 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 6-1 l 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6.1 Introduction Knowledge of the neutron environment within the reactor vessel and surveillance capsule geometry is required as an integral part of LWR reactor vessel surveillance programs for two reasons. First, in order to interpret the neutron radiation induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of rigorous analytical techniques and I 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 assoited with damage trend curves as well as to a more accurate evaluation of damage gradients through the reactor 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"71, " 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 E69300, " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials." This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance Capsule S, withdrawn at the end of the seventeenth fuel cycle. Also included is an update of the ; dosimetry evaluation for Capsules R, P, and V withdrawn at the end of the ninth, fifth, and first ! fuel cycles, respectively. This update is based on current state-of-the-art methodology and nuclear data including recently released neutron transport and dosimetry cross-section l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM ' 6-2 libraries derived from the ENDF/B-VI data base. This report provides a consistent up-to-date neutron exposure data base for use in evaluating the material properties of the Prairie Island Unit 1 reactor vessel. In each of the capsule dosimetry evaluations, 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 witil the projected exposure of the reactor vessel are provided. 6.2 Discrete Ordinates Analysis A plan view of the reactor geometry at the core midplane is shown in Figure 4-1. Six irradiation capsules attached to the thermal shield are included in the reactor design to constitute the reactor vessel surveillance program. The capsules are located at azimuthal angles of 57 ,67 ,77*,237 ,247', and 257 relative to the core cardinal axis as shown in Figure 4-1. A plan view of a surveillance capsule holder attached to the thermal shield is shown in Figure 6-1. The stainless steel specimen containers are approximately 1-inch square and approximately 63 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 5.25 feet of the 12-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 water annulus between the thermal shield and the reactor vessel. In order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model. In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, two distinct sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establich relative radial distributions of exposure parameters ($(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec} through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetty withdrawn from the surveillance capsules as well as for the determination of exposure parameter ratios; i.e., [dpa/secy[$(E > 1.0 MeV)], within the reactor vessel geometry. The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the reactor vessel wall; i.e., the %T, %T, and %T locations. ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 6-3 I The second set of calculations consisted of a series of adjoint analyses relating the fast i neutron flux, $(E > 1.0 MeV), at surveillance capsula positions and at several azimuthal i locations on the reactor vessel inner radius to neutron source distributions within the reactor ; core. The source importance functions generated from these adjoint analyses provided the basis for all absolute exposure calculations and comparison with measurement. These importance functions, when combined with fuel cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for each cycle of l 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 specific neutron source di ,tributions 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 bumup cf individual fuel assemblies increased. The absolute cycle-specific data from the adjoint evaluations together with the relative neutron energy spectra and radial distribution information from the reference forward calculation provided the means to: 1- Evaluate neutron dosimetry obtained from surveillance capsules, 2- Relate dosimetry results to key locations at the inner radius and through the thickness of the reactor vessel wall, 3- Enable a direct comparison of analytical prediction with measurement, and 4- Establish a mechanism for projection of reactor vessel exposure as the design of each new fuel cycle evolves. The forward transport calculation for the reactor model summarized in Figures 4-1 and 6-1 was carried out in R,0 geometry using the DORT two-dimensional discrete c;dinates code l Version 2.7.3"'I and the BUGLE-93 cross-section libraryrmi. The BUGLE-93 library is a 47 energy group ENDF/B-VI based data set produced specifically for light water reactor applications. In these analyses, anisotropic scattering was treated with a P3 expansion of the scattering cross-sections and the angular discretization was modeled with an S, order of angular quadrature. The core power distribution utilized in the reference forward transport calculation was derived from statistical studies of long-term operation of Westinghouse 2-loop plants. Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, the neutron source was increased by a 20 margin derived from the statistical evaluation of plant-to-plant and cycle-to-cycle variations in peripheral power. Since it is unlikely that any single reactor would exhibit power levels on the core periphery at the nominal + 2a value for a large number of fuel cycles, the use of this reference distribution is expected i to yield somewhat conservative results, i ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 6-4 All adjoint calculations were also carried out using an S, order of angular quadrature and the P3 cross-section approximation from the BUGLE 93 library. Adjoint source locations were chosen at several azimuthal locations along the reactor vessel inner radius as well as at the geometric center of each surveillance capsule. Again, these calculations were run in R,0 geometry to provide neutron source distribution importance functions for the exposure parameter of interest, in this case $(E > 1.0 MeV). Having the importance functions and appropriate core source distributions, the response of interest could be calculated as: a(r,e) = f f f /(r,e,s) s(r.e,s) e de de de r 0 E where: R(r,0) = $(E > 1.0 MeV) at radius r and azimuthal angle 0. l(r,0,E) = Adjoint source importance function at radius r, azimuthal I angle 0, and neutron source energy E. S(r,0,E) = Neutron source strength at core location r,0 and energy I E. I Although the adjoint importance functions used in this analysis were based on a response l function defined by the threshold neutron flux $(E > 1.0 MeV), prior calculationst2n have shown that, while the implementation of low leakage loading patterns significantly impacts both the magnitude and spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location, the ratio of [dpa/sec]/[$(E > 1.0 MeV)] is insensitive to changing core source distributions. In the application of these adjoint importance functions to the Prairie Island Unit 1 reactor, therefore, the iron atom displacement rates (dpa/sec) and the neutron flux $(E > 0.1 MeV) were computed on a cycle-specific basis l by using [dpa/sec]/[$(E > 1.0 MeV)] and [$(E > 0.1 MeV)]/[$(E > 1.0 MeV)] ratios from the , forward analysis in conjunction with the cycle specific $(E > 1.0 MeV) solutions from the ) individual adjoint evaluations. The reactor core power distributions used in the plant specific adjoint calculations were taken from the fuel cycle design reports for the first seventeen operating cycles of Prairic Island Unit 1 r22 rw sei, Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-5. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation periods and provide the means to correlate dosimetry results with the corresponding exposure of the reactor vessel wall. In Table 6-1, the calculated exposure parameters ($(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec] are given at the geometric center of the three azimuthally symmetric surveillance ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM ^ 6-5 l t capsule positions (13*,23 , and 33 ) for both the reference and the plant specific core power distributions. The plant-specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis. The reference data derived from the forward calculation are provided as a conservative exposure evaluation against which ! l plant specific fluence calculations can be compared. Similar data are given in Table 6-2 for the reactor vessel inner radius. Again, the three pertinent exposure parameters are listed for the reference and Cycles 1 through 17 plant specific power distributions. It is important to note that the data for the vessel inner radius were taken at the clad / base 1 metal interface; and, thus, represent the maximum predicted exposure levels of the vessel plates and welds. Radial gradient information applicable to $(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec is given in Tables 6-3,6-4, and 6-5, respectively. The data, obtained from the reference forward neutron transport calculation, are presented on a relative basis for each exposure parameter { at several azimuthal locations. Exposure distributions thcough the vessel wall may be 1 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. ' For example, the neutron flux $(E > 1.0 MeV) at the %T depth in the reactor vessel wall along the 0* azimuth is given by: 43g(0*) = 4(168.04, 0 ) F(172.25, 0*) where: $c( 0*) = Projected neutron flux at the %T position on the 0 azimuth. $(168.04,0*) = Projected or calculated neutron flux at the vessel inner radius on the 0 azimuth. F(172.25,0 ) = Ratio of the neutron flux at the %T position to the flux at the vessel inner radius for the 0 azimuth. This data is obtained from Table 6-3. Similar expressions apply for exposure parameters expressed in terms of $(E > 0.1 MeV) and dpa/sec where the attenuation function F is obtained from Tables 6-4 and 6 5, respectively. 6.3 Neutron Dosimetry i The passive neutron sensors included in the Prairie Island Unit 1 surveillance program are I listed in Table 6-6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the i surveillance capsules and in the subsequent determination of the various exposure ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAlRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 6-6 parameters of interest [$(E > 1.0 MeV), $(E > 0.1 MeV), dpa/sec]. The relative locations of ' the neutron sensors within the capsules are shown in Figure 4-2. The iron, nickel, copper, l and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axiallevels within the capsules. The cadmium shielded uranium and neptunium fission monitors were accommodated within the dosimeter block located near the center of the capsule. 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, and The neutron energy spectrum at the monitor location. The specific activity of each of the neutron monitors was determined using established ASTM l procedures ""* 5 55 581 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 Prairie Island Unit 1 reactor was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report," for the Cycles 1 through 17 operating period. (For the last two months of Cycle 17, this deta was obtained directly from Northern States Power Company personnel, i.e., J. E. Schaefer.) The irradiation history applicable to the exposure of Capsules S, R, P, and V 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: 1 R= A \ P N F Y{ p/ret C[1-e*1[e*#j o j i I l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 l REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 6-7 where: R = Reaction rate averaged over the . adiation period and referenced to operation at a core power level of P,, (rps/ nucleus). A = Measured specific activity (dps/gm). No = Number of target element atoms per gram of sensor. F = Weight fraction of the target isotope in the sensor material. Y = Number of product atoms produced per reaction. P3 = Average core power level during irradiation period j (MW). P,, = Maximum or reference power level of the reactor (MW). Cj = Calculated ratio of $(E > 1.0 MeV) during irradiation period j to the time weighted average $(E > 1.0 MeV) over the entire irradiation period. 1 = Decay constant of the product isotope (1/sec). t3 = 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)/[P,,) accounts for month by month variation of reactor core power level within any given fuel cycle as well as over , multiple fuel cycles. The ratio C,3 which can be calculated for each fuel cycle using the adjoint ! transport technology discussed in Section 6.2, accounts for the change in sensor reaction i 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 C3 term should be employed. The impact of changing flux levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from non-low leakage to low leakage fuel management or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another. For the irradiation history of Capsules S, R, P, and V, the flux level term in the reaction rate calculations was developed from the plant-specific analysis provided in Table 6-1. . Measured and satucated reaction product specific activities as well as the derived full power reaction rates are listed in Table 6-8. The specific activities and reaction rates of the maeU sensors provided in Table 6-8 include corrections for 25U impurities, plutonium build-in, and gamma ray induced fissions. Corrections for gamma ray induced fissions were also included in the specific activities and reaction rates for the
- 7Np sensors as well.
Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment codetsn. The FERRET approach used the ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM t 6-8 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 $ by some response matrix A: ff*'") = { Ajh*) $h") 9 where i indexes the measured values belonging to a single data set s, g designates the energy group, and a delineates spectra that may be simultaneously adjusted. For example, Ri = [ ogg 4, 9 relates a set of measured reaction rates R, to a single spectrum 4, by the multi-group reaction cross-section o,. The log-normal approach automatically accounts for the physical constraint i 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-Il code!* . This procedure was carried out by first expanding the 47 group calculated spectrum into the SAND-Il 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-VI dosimetry filet *, were also collapsed into the 53 energy group structure using the SAND-Il 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-VI 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 not included. The omission of this additional uncertainty information does not significantly impact the results of the adjustment. ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM ~ 6-9 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 reference forward transport calculation. While the 53 x 53 group covariance matrices applicable to the sensor reaction cross-sections were developed from the ENDF/B-Vi data files, the covariance matrix for the input trial spectrum was constructed from the following i relation: t l M,,1 = R,, + R, R,i P,,1 ! where R, specifies an overall fractional normalization uncertainty (i.e., complete correlation) for j the set of values. The fractional uncertainties R, specify additional random uncertainties for i group g that are correlated with.a correlation matrix given by: i l F,,, = (1 -0] 8,,, + 0 e -" where: H=(0~0Y 2 i 2y The first term in the correlation matrix equation sp : es purdly random uncertainties, while l the second term describes short range correlations over a group range y (9 specifies the strength of the latter term). The value of 8 is 1 when g = g' and 0 otherwise. For the trial l spectrum used in the current evaluations, a short range correlation of y = 6 groups was used, i This choice implies that neighboring groups are strongly correlated when e is close to 1. Stror.g long range correlations (or anti-correlations) were justified based on information l presented by R. E. Maerker". 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 senrors. Resu4s of the FERRET evaluations of the Capsules S, R, P, and V dosimetry are given in
- Table 6-9. The data summarized in this table include fast neutron exposure evaluations in terms of 4(E > 1.0 MeV), @(E > 0.1 MeV), and dpa. In general, excellent results were l achieved in the fits of the adjusted spectra to the individual measured reaction rates. The measured and FERRET adjusted reaction rates for each reaction are given in Table 6-10. An examination of Table 6-10 shows that, in cll cases, reaction rates calculated with the adjusted ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADATION SURVEILLANCE PROGRAM
[ 6-10 spectra match the measured reaction rates to better than 9%. The adjusted spectra from the least squares evaluation is given in Table 6-11 in the FERRET 53 energy group structure. In Table 6-12, absolute comparisons of the measured and calculated fluence at the center of each capsule are presented. The results for the Capsules S, R, P, and V dosimetry evaluations (M/C ratios of 0.99 for O(E > 1.0 MeV)) are consistent with results obtained froin similar evaluations of dosimetry from other reactors using methodologies based on ENDF/B-VI cross-sections. 6.4 Projections of Reactor Vessel Exposure The best estimate exposure of the Prairie Island Unit 1 reactor vessel was developed using a combination of absolute plant specific transport calculations and all available plant specific measurement data. In the case of Prairie Island Unit 1, the measurement data base consists of the four surveillance capsules discussed in this report. Combining this measurement data base with the plant-specific calculations, the best estimate vessel exposure is obtained from the following relationship:
- Best Est = K @m where: o,.,t g,,. = The best estimate fast neutron exposure at the location of interest.
K = The plant specific measurement / calculation (M/C) bias factt derived from the surveillance capsule dosimetry l data. om = The absolute calculated fast neutron exposure at the location of interest. The approach defined in the above equation is based on the premise that the measurement data represent the most accurate plant-specific information available at the locations of the dosimetry; and, further that the use of the measurement data on a plant-specific basis essentially removes biases present in the analytical approach and mitigates the uncertainties that would result from the use of analysis alone. That is, at the measurement points the uncertainty in the best estimate exposure is dominated by the uncertainties in the measurement process. At locations within the reactor vessel wall, ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 6-11 additional uncertainty is incurred due to the analytically determined relative ratios among the various measurement points and locations within the reactor vessel wall For Prairie Island Unit 1, the derived plant specific bias factors were 0.99,1.07, and 1.03 for j @(E > 1.0 MeV), @(E > 0.1 MeV), and dpa, respectively. Bias factors of this magnitude are I fully consistent with experience using the BUGLE-93 cross-section library. 1 The use of the bias factors derived from the measurement data base acts to remove l plant-specific biases associated with the definition of the core source, actual versus assumed reactor dimensions, and operational variations in water density within the reactor. As a result, the overall uncertainty in the best estimate exposure projections within the vessel wall i depends on the individual uncertainties in the measurement process, the uncertainty in the i dosimetry location, and, in the uncertainty in the calculated ratio of the neutron exposure at the point of interest to that at the measurement location. The uncertainty in the derived neutron flux for an individual measurement is obtained directly from the results of a least squares evaluation of dosimetry data. The least squares approach enmbines individual uncertainty in the calculated neutron energy spectrum, the uncertainties in dosimetry cross-sections, and the uncertainties in measured foil specific activities to produce a net uncertainty in the derived neutron flux at the measurement point. The associated uncertainty in the plant specific bias factor, K, derived from the M/C data base, in turn, depends on the total number of available measurements as well as on the uncertainty of each measurement. In developing the overall uncertainty associated with the reactor vessel exposure, the positioning uncertainties for dosimetry are taken from parametric studies of sensor position performed as part a series of analytical sensitivity studies included in the qualification of the methodology. The uncertainties in the exposure ratios relating dosimetry results to positions j within the vessel wall are again based on the analytical sensitivity studies of the vessel thickness tolerance, downcomer water density variations, and vessel inner radius tolerance. Thus, this portion of the overall uncertainty is controlled entirely by dimensional tolerances associated with the reactor design and by the operational characteristics of the reactor. l The net uncertainty in the bias factor, K, is combined with the uncertainty from the analytical ' sensitivity study to define the overall fluence uncertainty at the reactor vessel wall. In the case of Prairie Island Unit 1, the derived uncertainties in the bias factor, K, and the additional uncertainty from the analytical sensitivity studies combine to yield a net uncertainty of 12 4. l Based on this best estimate approach, neutron exposure projections at key locations on the ; reactor vessel inner radius are given in Table 6-13. Along with the current (18.12 EFPY) exposure, projections are also provided for exposure periods of 24 EFPY and 35 EFPY. Projections for future operation were based on the assumption that the average exposure ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIR!E ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 1 \ 6-12 rates averaged over the Cycles 13 through 17 irradiation period would continue to be applicable throughout plant life. ) i in the calculation of exposure gradients within the reactor vessel wall for the Prairie Island I Unit i reactor vessel, exposure projections to 24, and 35 EFPY were also employed. Data based on both a O(E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wall are provided in Table 6-14. In order to access RT, versus fluence curves, dpa equivalent fast neutron fluence levels for the %T, %T and %T positions were defined by the relations: ' 4(%7) = &(07) dpa(%7) dpa(07) \ \ &(%7) = &(07) dpa(%7) ! dpa(07) and I i l &(%7) = 4(07) dpa(%D dpa(07) Using this approach results in the dpa equivalent fluence values listed in Table 6-14. In Table 6-15 updated lead factors are listed for each of the Prairie Island Unit 1 surveillance capsules. Lead factor data based on the accumulated fluence through Cycle 17 are provided for each remaining capsule. ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM so 6-13 FIGURE 6-1 PLAN VIEW OF A REACTOR VESSEL SURVEILLANCE CAPSULE , (13*.23'.33') i CHARPY (12',22*,32') , [ SPECIMEN /p s / / 6 '//////// THERMAL SHIELD ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM i ^ 6-14 TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE RATES AND IRON ATOM DISPLACEMENT RATES AT THE SURVEILLANCE CAPSULE CENTER 2 $(E > 1.0 MeV) (n/cm -sec) Cycle No. 13* 23* 33 Reference 1.59e+11 9.35e+10 8.83e+10 1 1.421e+11 8.131e+10 7.656e+10 2 1.279e+11 7.730e+10 7.285e+10 3 1.568e+11 8.703e+10 7.931e+10 i 4 1.507e+11 8.943e+10 8.431e+10 l 5 1.586e+11 8.936e+10 8.304e+10 l 6 1.593e+11 9.117e+10 8.515e+10 { 7 1.318e+11 8.149e+10 8.422e+10 8 1.729e+11 9.813e+10 9.248e+10 9 1.256e+11 8.338e+10 8.098e+10 i 10 1.827e+11 9.315e+10 8.195e+10 l 11 1.825e+11 1.054e+11 9.506e+10 l 12 1.349e+11 9.243e+10 8.773e+10 13 1.004e+11 7.256e+10 6.842e+10 ) 14 8.159e+10 5.947e+10 5.809e+10 l 15 8.218e+10 5.828e+10 5.672e+10 16 9.478e+10 7.061e+10 6.403e+10 17 9.668e+10 - 7.086e+10 6.241e+10 2 $(E > 0.1 MeV) (n/cm -sec) i Cycle No. 13 23* 33* ! l Reference ~ I*+II 6.02e+11 1 3.22e+11 2.695e+11 j 5.384e+11 2 2.797e+11 2.564e+11 4.848e+11 3 2.659e+11 2.792e+11 5.942e+11 4 2. W e+11 2.968e+11 5.712e+11 5 .077e+11 2.923e+11 6 6 .0 % +11 2 E e+11 7 6 012e+11 036e+11 4.996e+11 .1 +11 2.We+11 8 2.803e+11 3.255e+11 6 554e+11 9 3.376e+11 2.850e+11 4.761e+11 0 2.868e+11 2.885e+11 6.926e+11 11 3.205e+11 3.346e+11 6 917e+11 12 3.62&+11 3.088e+11 5.111e+11 13 3.18&+11 2 M e+11 3.804e+11 14 3 092e+11 2 & +11 2M5e+11 15 2.046e+11 1.997e+11 16 3.115e+11 3.592e+11 225e+11 2.2Me+11 2.429e+11 2.197e+11 3.664e+11 2.438e+11 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM _ . . _ . . . . . . . . . _ _ . _ . . . . _ _ . _ . . _ . _ _ _ . _ . - _ . _ . _ _ _ . . - _ . _ m.__ ._. I' 6-15 1 4 TABLE 6-1 cont'd i I I CALCULATED FAST NEUTRON EXPOSURE RATES AND IRON ATOM j DISPLACEMENT RATES AT THE SURVEILLANCE CAPSULE CENTER
- 1 i
i Displacement Rate (dpa/sec) l Cycle No. 13' 23' 33' i j Reference 2.83e-10 1.59e-10 1.52e , 1 2.529e-10 1.382e-10 1.317e-10 i 2 2.277e-10 1.314e-10 1.253e-10 , 3 2.791e-10 1.479e-10 1.364e-10 j- 4 2.682e-10 1.520e-10 1.450e-10
- 5 2.824e-10 1.519e-10 1.428e-10 l 6 2.835e-10 1.550e-10 1.465e-10
! 7 2.346e-10 1.385e-10 1.449e-10 l 8 3.078e-10 1.668e-10 1.591e-10 1 9 2.236e-10 1.418e-10 1.393e-10
- 10 3.253e-10 1.584e-10 1.410e-10 i 11 3.249e-10 1.792e-10 1.635e-10 12 2.40le-10 1.571e-10 1.509e-10 I
13 1.787e-10 1.234e-10 1.177e-10 f 14 1.452e-10 1.011e-10 9.992e-11 15 1.463e-10 9.907e-11 9.756e-11 16 1.687e-10 1.200e-10 1.101e-10 17 1.721e-10 1.205e-10 1.073e-10 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 6-16 TABLE 6-2 CALCULATED AZIMUTHAL VARIATION OF FAST NEUTRON EXPOSURE RATES AND IRON ATOM DISPLACEMENT RATES AT THE REACTOR VESSEL CLAD / BASE METAL INTERFACE 2 $(E > 1.0 MeV) (n/cm -sec) Cycle No. 0* 15* 30 45* , Reference 5.32e+10 3.25e+10 2.22e+10 1.87e+10 1 4.827e+10 2.902e+10 1.925e+10 1.676e+10 2 4.236e+10 2.645e+10 1.838e+10 1.581e+10 3 5.266e+10 3.192e+10 2.016e+10 1.794e+10 4 4.994e+10 3.096e+10 2.123e+10 1.810e+10 5 5.322e+10 3.233e+10 2.096e+10 1.666e+10 6 5.327e+10 3.248e+10 2.146e+10 1.833e+10 7 4.103e+10 2.717e+10 2.056e+10 1.895e+10 8 5.870e+10 3.510e+10 2.325e+10 1.908e+10 9 4.436e+10 2.654e+10 2.036e+10 1.676e+10 10 6.215e+10 3.667e+10 2.096e+10 1.951 e+10 11 6.159e+10 3.720e+10 2.439e+10 1.795e+10 12 4.658e+10 2.878e+10 2.235e+10 1.741e+10 13 3.148e+10 2.196e+10 1.749e+10 1.549e+10 14 2.549e+10 1.791e+10 1.468e+10 1.366e+10 15 2.532e+10 1.794e+10 1.433e+10 1.357e+10 16 2.941e+10 2.098e+10 1.666e+10 1.384e+10 17 2.931e+10 2.135e+10 1.642e+10 1.329e+10 $(E > 0.1 MeV) (n/cm'-sec) Cycle No. 0* 15* 30* 45* Referenee 1.46e+11 9.45e+10 6.05e+10 4.91e+10 1 1.323e+11 8.444e+10 5.256e+10 4.407e+10 2 1.161e+11 7.696e+10 5.017e+10 4.158e+10 3 1.443e+11 9.289e+10 5.503e+10 4.719e+10 4 1.368e+11 9.009e+10 5.797e+10 4.761e+10 5 1.458e+11 9.408e+10 5.722e+10 4.907e+10 6 1.460e+11 9.451e+10 5.859e+10 4.821e+10 7 1.124e+11 7.907e+10 5.612e+10 4.984e+10 8 1.608e+11 1.021e+11 6.348e+10 5.019e+10 9 1.216e+11 7.723e+10 5.559e+10 4.409e+10 10 1.703e+11 1.067e+11 5.722e+10 5.131e+10 11 1.688e+11 1.082e+11 6.658e+10 4.721e+10 12 1.276e+11 8.374e+10 6.10le+10 4.579e+10 13 8.626e+10 6.390e+10 4.774e+10 4.073e+10 14 6.984e+10 5.213e+10 4.008e+10 3.593e+10 15 6.937e+10 5.220e+10 3.912e+10 3.569e+10 16 8.058e+10 6.104e+10 4.549e+10 3.640e+10 17 8.031e+10 6.212e+10 4.482e+10 3.494e+10 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILIANCE PROGRAM l 6-17 l TABLE 6-2 cont'd CALCULATED AZIMUTHAL VARIATION OF FAST NEUTRON EXPOSURE RATES AND IRON ATOM DISPLACEMENT RATES AT THE REACTOR VESSEL CLAD / BASE METAL INTERFACE Displacement Rate (dpa/sec) Cycle No. O' 15' 30 45' Reference 8.68e-11 5.46e-11 3.65e-11 3.03e-11 1 7.869e-11 4.875e-11 3.158e-11 2.715e-11 i 2 6.904e-11 4.443e-11 3.014e-11 2.561e-11 3 8.584e-11 5.363e-11 3.306e-11 2.907e-11 4 8.141e-11 5.201e-11 3.482e-11 2.932e-11 5 8.675e-11 5.431e-11 3.437e-11 3.023e-11 , 6 8.683e-11 5.456e-11 3.520e-11 2.970e-11 7 6.688e-11 4.565e-11 3.371e-11 3.070e-11 ; 8 9.568e-11 5.896e-11 3.814e-11 3.092e-11 9 7.231e-11 4.459e-11 3.339e-11 2.716e-11
- 10 1.013e-10 6.160e-Il 3.438e-Il 3.161e-11 ;
11 1.004e-10 6.249e-11 4.000e-11 2.908e-11 ; 12 7.592e-11 4.834e-11 3.665e-11 2.821e-11 : 13 5.132e-11 3.689e-11 2.868e 11 2.509e-11 - 14 4.154e-11 3.010e-11 2.408e-11 2.213e-11 15 4.127e-11 3.014e-11 2.350e-11 2.198e-11 16 4.794e-11 3.524e-11 2.733e 11 2.242e-11 17 4.778e-11 3.586e-11 2.692e-11 2.152e-11 J l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 ' REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 6 18 TABLE 6-3 RELATIVE RADIAL DISTRIBUTION OF $(E > 1.0 MeV) WITHIN THE REACTOR VESSEL WALL RADIUS AZIMUTHAL ANGLE (cm) 0* 15' 30* 45' 168.04 1.000 1.000 1.000 1.000 168.27 0.987 0.987 0.985 0.987 168.88 0.940 0.942 0.937 0.942 169.75 0.862 0.865 0.857 0.866 170.93 0.754 0.757 0.749 0.760 172.25 0.639 0.644 0.636 0.647 173.53 0.540 0.546 0.539 0.550 174.98 0.444 0.451 0.444 0.454 l 176.46 0.362 0.370 0.363 0.372 j 177.58 0.308 0.317 0.311 0.318 179.03 0.250 0.259 0.253 0.260 180.66 0.1% 0.206 0.201 0.206 181.63 0.169 0.179 0.175 0.178 182.60 0.144 0.154 0.151 0.154 184.06 0.110 0.122 0.120 0.122 184.87 0.101 0.113 0.112 0.113 Note: Base Metal Inner Radius = 168.04 cm Base Metal %T = 172.25 cm Base Metal %T = 176.46 cm Base Metal %T = 180.66 cm Base Metal Outer Radius = 184.87 cm I l l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM t 6-19 ~ i
- l. TABLE 6-4 ,
j RELATIVE RADIAL DIS'IRIBUTION OF $(E > 0.1 MeV)
- WITHIN THE REACTOR VESSEL WALL s
1 1 RADIUS AZIMUTHAL ANGLE (cm) 0' 15' 30' 45' 168.04 1.000 1.000 1.000 1.000 j 168.27 1.005 1.007 1.005 1.007 l 168.88 1.002 1.007 1.004 1.008 ! 169.75 0.980 0.990 0.985 0.992 i 170.93 0.934 0.948 0.945 0.953 172.25 0.873 0.891 0.889 0.899 ! 173.53 0.809 0.831 0.831 0.841 j 174.98 0.736 0.763 0.763 0.773 176.46 0.662 0.693 0.694 0.703 i 177.58 0.606 0.640 0.642 0.650 I 179.03 0.536 0.573 0.577 0.582 l 180.66 0.461 0.502 0.507 0.509 1 l 181.63 0.416 0.458 0.466 0.465 l 182.60 0.369 0.415 0.423 0.421 l 184.% 0.298 0.348 0.361 0.357 l 184.87 0.276 0.327 0.343 0.339 l ! 1 f Note: Base Metal Inner Radius = 168.04 cm Base Metal %T = 172.25 cm Base Metal %T = 176.46 cm Base Metal MT = 180.66 cm Base Metal Outer Radius = 184.87 cm i ANALYS$ OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM . 1 l 6-20 : ~ I l TABLE 6-5 RELATIVE RADIAL DISTRIBUTION OF dpa/sec WITHIN THE REACTOR VESSEL WALL RADIUS AZIMUTHAL ANGLE (cm) 0* 15' 30 45' 168.04 1.000 1.000 1.000 1.000 168.27 0.988 0.990 0.988 0.989 168.88 0.951 0.955 0.950 0.954 169.75 0.889 0.8% 0.889 0.857 ' 170.93 0.804 0.814 0.805 0.812 172.25 0.712 0.726 0.716 0.723 l 173.53 0.630 0.648 0.638 0.644 174.98 0.547 0.568 0.558 0.563 176.46 0.472 0.495 0.486 0.490 177.58 0.420 0.445 0.436 0.439 179.03 0.360 0.386 0.379 0.380 3 180.66 0.301 0.328 0.322 0.322 181.63 0.267 0.2 % 0.291 0.289 182.60 0.234 0.264 0.261 0.258 184.06 0.187 0.219 0.220 0.216 184.87 0.173 0.206 0.208 0.205 ; i Note: Base Metal Inner Radius = 168.04 cm Base Metal %T = 172.25 cm Base Metal %T = 176.46 cm Base Me'.al %T = 180.66 cm Base Metal Outer Radius = 184.87 cm i ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISt.AND Unit 1 , REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 6-21 TABLE 6-6 NUCLEAR PARAMETERS USED IN THE EVALUATION OF NEUTRON SENSORS l Target Fission Monitor Reaction Atom Response Product Yield Material of Interest Fraction Range Half-life (%) '3 Copper Cu (n,a) 0.6917 E > 4.7 MeV 5.271 y Iron "Fe (n,p) 0.0580 E > 1.0 MeV 312.5 d Nickel "Ni (n.p) 0.6827 E > 1.0 MeV 70.78 d Uranium-238 2"U (n,f) 0.9996 E > 0.4 MeV 30.17 y 6.00 Neptunium-237 23'Np (n,f) 1.0000 E > 0.08 MeV 30.17 y 6.27 Cobalt-Al "Co (n,y) 0.0015 E > 0.015 MeV 5.271 y Note: 2nU and 23'Np monitors are cadmium shielded. i I 4 1 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM . . _ y_ . _. _ . . . . , . - _ . _ . . _ . _ _ . _ . . . _ . _ . _ . . _ . _ _ _ . . . _ _ _ _ _ _ __. _ _ _ , . 6-22 4 TABLE 6-7 i MONTHLY THERMAL GENERATION DURING THE FIRST SEVENTEEN FUEL CYCLES 4 OF THE PRAIRIE ISLAND UNIT 1 REACTOR l Cycle 1 Cycle 2 Cycle 3 Cycle 4 Thermal Thermal Thermal Thermal : Gen. Gen. Gen. Gen. Month MWt-hr Month MWt-hr Month MWt-hr Month MWt-hr Dec-73 128000 Apr-76 . O Apr-77 0 Apr-78 375189 Jan-74 0 May-76 730614 May-77 1027843 May-78 1216949 Feb-74 385824 Jun-76 1107963 Jun-77 1093671 Jun-78 1128613 Mar-74 1044 % Jul-76 1112551 Jul-77 1202230 Jul-78 904285 Apr-74 379161 Aug-76 1167605 Aug-77 1208004 Aug-78 1065847 May-74 0 Sep 76 763875 Sep-77 1139867 Sep-78 1032140 Jun-74 0 Oct-76 1219726 Oct-77 1159500 Oct-78 1218111 Jul-74 77 % 80 Nov-76 1151899 Nov-77 1176823 Nov-78 1100717 Aug-74 933538 Dec-76 1219726 Dec-77 1204540 Dec-78 1198352 Sep-74 145829 Jan-77 1192135 Jan-78 1192991 Jan-79 1208813 l Oct-74 192888 Feb-77 1020845 Feb-78 1042857 Feb-79 1098392 Nov-74 1113715 Mar-77 583997 Mar-78 950466 Mar-79 1146048 Dec-74 1184901 Apr-79 184809 Jan-75 988566 Feb-75 886380 Mar-75 1195237 Apr-75 917380 May-75 694637 Jun-75 966751 Jul 75 966751 Aug-75 819786 Sep-75 1135530 Oct-75 1055159 Nov-75 1133234 Dec-75 1184901 Jan-76 1191790 Feb-76 831268 Mar-76 907705 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 6-23 f .
- 1
- TABLE 6-7 cont'd l MONTHLY THERMAL GENERATION DURING THE FIRST SEVENTEEN FUEL CYCLES j OF THE I'RAIRIE ISLAND UNIT 1 REACTOR Cycle 5 Cycle 6 Cycle 7 Cycle 8 Thermal Thermal Thermal Thermal Gen. Gen. Gen. Gen.
Month MWt-hr Month MWt-hr Month MWt-hr Month M Wt-hr May-79 778462 Sep.80 0 Oct-81 68509 Dec-82 455376 Jun-79 1014740 Oct-80 133518 Nov-81 1122238 Jan-83 1080444 Jul-79 118710 Nov-80 1154538 Dec-81 1194009 Feb-83 629364 Aug-79 1055832 Dec-80 1212882 Jan-82 1208146 - Mar-83 1214694 Sep-79 1116328 Jan-81 1209516 Feb-82 1088527 Apr-83 1180326 Oct-79 372109 Feb-81 1089462 Mar-82 1221195 May-83 1198584 Nov-79 533052 Mar-81 1226346 Apr-82 1151592 Jun-83. 1131996 Dec-79 1171117 Apr-81 1168002 May-82 1170085 Jul-83 1194288 Jan-80 1214492 May-81 1216248 Jun-82 -1183135 Aug-83 1198584 Feb-80 905162 Jun-81 1155660 Jul-82 1204884 Sep-83 1109442 i Mar-80 1208785 Jul-81 1216248 Aug-82 1220108 Oct-83 1218990 Apr-80 1175683 Aug-81 1203906 Sep-82 1128763 Nov-83 1172808 May-80 1201936 Sep-81 627198- Oct-82 1199446 Dec-83 38664 Jun-80 11288Fi Nov-82 493698 Jul-80 401787 Aug-80 904020 Cycle 9 Cycle 10 Cycle 11 Cycle 12 Thermal Thermal Thermal Thermal Gen. Gen. Gen. Gen. Month _M_ Wt-hr Month MWt-hr Month MWt-hr Month M Wt-hr Jan-84 1053719 Feb-85 0 Apr-86 760857 May-87 107076 Feb-84 1104603 Mar-85 764836 May-86 1222014 Jun-87 1115895 Mar-84 1212731 Apr-85 1180670 Jun-86 1181914 Jul-87 1136890 Apr-84 1175628 May-85 1171123 Jul-86 1222014 Aug-87 1219821 j May-84 1224392 Jun-85 117 % 10 Aug-86 1218848 Sep-87 1178880 : Jun-84 1177748 Jul-85 1199765 Sep-86 1168195 Oct-87 1215622 Jul-84 1212731 Aug-85 1185974 Oct-86 1224125 Nov-87 1173632 Aug-84 1148065 Sep-85 1154150 Nov-86 1179803 Dec-87 1217722 Sep-84 1139585 Oct-85 1210373 Dec-86 1182969 Jan-88 1217722 Oct-84 813081 Nov-85 117 % 10 Jan-87 1213572 Feb-88 1138990 Nov-84 890467 Dec-85 1217798 Feb-87 781962 Mar-88 1027715 Dec-84 1135345 Jan-86 1084138 Mar-87 899099 Apr-88 1177831 Jan-85 323324 Feb-86 710736 Apr-87 186785 May-88 1216672 Mar-86 82742 Jun-88 1176781 Jul-88 1050810 Aug-88 628806 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM l* 6-24 5
- TABLE 6-7 cont'd MONTHLY THERMAL GENERATION DURING THE FIRST SEVENTEEN FUEL CYCLES
, OF THE PRAIRIE ISLAND UNIT 1 REACTOR d a ! Cycle 13 Cycle 14 Cycle 15 Cycle 16 1 Thermal Thermal Thermal Thermal Gen. Gen. Gen. Gen. ; Month M Wt-hr Month MWt-hr Month MWt-br, Month M Wt-hr Sep-88 52944 Feb-90 32301 Jun-91 2079 Nov-92 0 Oct-88 1212012 Mar-90 1204506 Jul-91 1180804 Dec-92 0 Nov-88 1171334 Apr-90 1185750 Aug-91 1048795 Jan-93 736230 Dec-88 1225572 May-90 1221177 Sep-91 1183922 Feb-93 1073237 Jan-89 1226615 Jun-90 1177415 Oct-91 1222381 Mar-93 1226705 Feb-89 1106665 Jul-90 1209716 Nov-91 1164173 Apr-93 1181079 Mar-89 1222443 Aug-90 1200338 Dec-91 1222381 May-93 1224631 ; Apr-89 1179678 Sep-90 1185750 Jan-92 1221342 Jun-93 1184190 May-89 1128834 . Oct-90 1224303 Feb-92 1138187 Jul-93 1224631 Jun-89 1184893 Nov-90 1132611 Mar-92 1221342 Aug-93 1161377 . Jul-89 1181764 Dec-90 1177415 Apr-92 1182883 Sep-93 1184190 Aug-89 1224529 Jan-91 1224303 May-92 1180804 Oct-93 1223594 Sep-89 1182807 Feb-91 1102394 Jun-92 1181843 Nov-93 1171747 Oct-89 11994 % Mar-91 1182625 Jul-92 1221342 Dec-93 1223693 Nov-89 1177592 Apr-91 1175331 Aug-92 1222381 Jan-94 1223693 Dec-89 949166 May-91 934638 Sep-92 958364 Feb-94 1097176 Jan-90 349418 Oct-92 563376 Mar-94 1223693 . Apr-94 1115842 ! May-94 211554 l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 4 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM i l 1 l ,- 6-25 TABLE 6-7 cont'd MONTHLY THERMAL GENERATION DURING THE FIRST SEVENTEEN FUEL CYCLES OF THE PRAIRIE ISLAND UNIT 1 REACTOR Cycle 17 Thermal Gen. Month MWt-hr Jun-94 0 Jul-94 1085468 Aug-94 920848 Sep-94 1187328 Oct-94 1226425 Nov-94 1178068 Dec-94 1225396 Jan-95 12253 % Feb-95 1081353 Mar-95 12253 % Ap 95 1183212 May-95 12253 % Jun-95 1186299 Jul-95 1204818 - Aug-95 1225396 Sep-95 1185270 Oct-95 1182183 Nov-95 1185270 Dec-95 1134855 Jan-% 151245 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 6-26 TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCE CAPSULE S l SATURATED ACTIVITIES AND REACTION RATES 1 l Measured Saturated Reaction Activity Activity Rate Reaction (dps/gm) (dps/gm) (rps/ atom) l Cu (n,ct) *Co Top Middle 1.70e+05 2.49e+05 4.33e-17 Bottom Middle 1.83e+05 2.68e+05 4.66e-17 "Fe (n,p) "Mn Top 1.30e+06 3.03e+06 4.60e-15 Top Middle 1.13e+06 2.63e+06 4.00e-15 Middle 1.17e+06 2.73e+06 4.14e-15 Bottom Middle 1.20e+06 2.80e+06 4.25e-15 Bottom 1.28e+06 2.98e+06 4.53e-15 58 Ni (n,p) 58Co Middle 2.28e+06 3.58 +07 5.92e-15 "Co (n,y) "Co Top 3.87e+07 5.66e+07 3.55e-12 i Bottom 3.86e+07 5.65e+07 3.54e-12 i l l "Co (n,y) "Co (Cd) l Top 1.41e+07 2.06e+07 133e-12 l Bottom 1.43e+07 2.09e+07 1.35e-12 23sU (n,f) '3'Cs Middle 1.67e+06 5.21e+06 3.43e-14 23'Np (n,f) Cs Middle 1.13e+07 3.52e+07 2.21e-13 l l 1 l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 6-27 TABLE 6-8 cont'd MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCE CAPSOLE R SATURATED ACTIVITIES AND REACTION RATES Measured Saturated Reaction Activity Activity Rate Reaction (dps/gm) (dps/gm) (rps/ atom) Cu (n,n) "Co Top Middle 2.42e+05 4.29e+05 7.46e-17 Bottom Middle 2.50e+05 4.43e+05 7.70e-17 "Fe (n,p) "Mn Top 2.74e+06 6.09e+06 9.25e-15 Top Middle 2.47e+06 5.49e+06 8.34e-15 Middle 2.56e+06 5.69e+06 8.64e-15 Bottom Middle 2.59e+06 5.76e+06 8.74e-15 Bottom 2.67e+06 5.93e+06 9.01e-15 ssNi (n,p) 5 Co Middle 3.84e+06 7.58e+07 1.26e-14 "Co (n,y) "Co Top 7.41e+07 1.31e+08 8.31e-12 Bottom 8.13e+07 1.44e+08 9.12e-12 Bottom 8.16e+07 1.45e+08 9.15e-12 "Co (n,y) "Co (Cd) Top 2.96e+07 5.24e+07 3.46e-12 Bottom 3.01e+07 5.33e+07 3.51e-12 1 23: U (n,0 7Cs Middle 2.09e+06 1.20e+07 7.93e-14 23'Np (n,0 '37Cs Middle 1.41e+07 8.12e+07 5.10e.r l l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM I -._ _ _ _ . _ . . ._ _ . ._ _ _ _ . _ _ _ . , . . _ . . _ . ~ , _ . _ . . . . - _ . _ _ _ .. - . . . . . . _ _ _ _ . - . _ _ _ _ _ _ _ . . _ . t 6-28 TABLE 6-8 cont'd i MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCE CAPSULE P SATURATED ACTIVITIES AND REACTION RATES i Measured Saturated Reaction Activity Activity Rate Reaction (dps/gm) (dps/gm) (rps/ atom) Cu (n,a)
- Co ,
Top Middle 1.27e+05 3.37e+05 5.86e-17 Bottom Middle 1.18e+05 3.13e+05 5.44e-17 l "Fe (n.p) "Mn Top 1.08e+06 3.78e+06 5.74e-15 } Top Middle 8.41e+05 2.94e+06 4.47e-15 ' Bottom Middle 1.00e+06 3.50e+06 5.32e-15 Bottom 1.12e+06 3.92e+06 5.96e-15 "Ni (n.p) "Co Middle 3.77e45 4.75e47 7.86e-15 "Co (n,y) ' Co Top 2.64e+07 7.00e+07 4.34e-12 Bottom 3.15e+07 8.36e+07 5.18e-12 "Co (n,y) '"Co (Cd) Top 9.34e+06 2.48e+07 1.57e-12 Bottom 9.92e+06 2.63e+07 1.67e-12 23sU (n,0 Cs Middle 5.55e+05 5.42e+06 3.57e-14 22'Np (n,0 Cs Middle 4.28e+06 4.18e+07 2.62e-13 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM ~ I ? 6-29 ~ i ! i TABLE 6-8 cont'd MEASURED SENSOR ACTIVITIES AhT REACTION RATES
- SURVEILLANCE CAPSULE V i SATURATED ACTIVITIES AND REACTION RATES .
Measured Saturated Reaction Activity Activity Rate j Reaction (dps/gm) (dps/gm) (rps/ atom) ! "Cu (n,(x) "Co } Top Middle 5.61e+04 3.60e45 . 6.26e-17 i Bottom Middle 6.35e+04 4.07e+05 7.08e-17 i i "Fe (n,p) "Mn j Top 2.23e+06 5.03e+06 7.65e-15
- Top Middle 2.05e+06 4.63e+06 7.03e-15 i Middle 2.09e+06 4.72e+06 7.17e-15 i Bottom Middle 2.17e+06 4.90e+06 7.44e-15 Bottom 2.32e+06 5.24e+06 7.96e-15 j "Ni (n,p) "Co l Middle 1.00e+07 5.16e+07 8.54e-15 "Co (n,y) "Co Top 1.90e+07 1.22e+08 7.71e-12 Bottom 2.23e+07 1.43e+08 9.05e-12 "Co (n,y) "Co (Cd)
Top 8.25e+06 5.29e+07 3.49e-12 , Bottom 7.68e+06 4.92e+07 3.25e-12 ' 23: U (n,f) "'Cs Middle 2.44e@5 7.82e+06 5.15e-14 237Np (n,f) "'Cs Middle 1.88e+06 6.03e+07 3.78e-13 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISt.AND Unit 1 , i REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM 6-30 l TABLE 6-9
SUMMARY
OF NEUTRON DOSIMETRY RESULTS SURVEILLANCE CAPSULES S, R, P AND V 1
l Measured Flux and Fluence for Capsule S l
- Quantity Flux Quantity Fluence Uncertainty ,
2 2
, [n/cm -sec] [n/cm )
$ (E > 1.0 MeV) 7.024e+10 @ (E > 1.0 MeV) 4.017e+19 8%
$ (E > 0.1 MeV) 2.769e+11 @ (E > 0.1 MeV) 1.584e+20 ' 15 % ,
$ (E < 0.414 eV) 9.268e+10 @ (E < 0.414 eV) 5.300e+19 19% >
dpa/sec 1.278e-10 dpa 7.309e-02 11 %
Measured Flux and Fluence for Capsule R Quantity Flux Quantity Euence Uncertainty 2 2
[n/cm -sec) [n/cm )
1.645e+11 4 (E > 1.0 MeV) 4.478e+19 8%
4( ** *) 6.675e+11 @ (E > 0.1 MeV) 1.817e+20 15 %
$( > 0.1 MeV) 2.285e+11 6.221e+19
@ (E < 0.414 eV) 19 %
- I )
3.0lle-10 dpa 8.197e-02 11 %
d i
I Measured Flux and Fluence for Capsule P Quantity Flux ,! Quantity Fluence Uncertainty ;
2 2
[n/cm -sec] [n/cm j ;
$ (E > 1.0 MeV) 8.501e+10 $ (E > 1.0 MeV) 1.318e+19 8% ,
& (E > 0.1 MeV) 3.221e+11 @ (E > 0.1 MeV) 4.995e+19 15 %
$ (E < 0.414 eV) 1.294e+11 @ (E < 0.414 eV) 2.007e+19 18 %
dpa/see 1.518e-10 dpa 2.354e-02 10 %
Measured Flux and Fluence for Capsule V Quantity Flux Quantity Fluence Uncertainty 2 2
[n/cm -sec] [n/cm )
$ (E > 1.0 MeV) 1.276e+11 4 (E > 1.0 MeV) 5.630e+18 8%
$ (E > 0.1 MeV) 5.102e+11 4 (E > 0.1 MeV) 2.251e+19 15 %
$ (E < 0.414 eV) 2.122e+11 4 (E < 0.414 eV) 9.363e+18 19 %
dpa/sec 2.337e-10 dpa 1.031e-02 11 %
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM
. - . . - .-. . . _ - _ - - . . - . . _ . - - - . - - - ~ _ . - - . - . . .. - . . .-. - - -..._n...
6-31 I
TABLE 6-10 i
- COMPARISON OF MEASURED AND FERRET CALCULATED
] REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER h
l Surveillance Capsule S j.
j Reaction Rate (rps/ nucleus)
! Adjusted M/C i Measured Calc. Adiusted 63 Cu (n,a) 4.49e-17 4.38e-17 1.03
, "Fe (n,p) 4.31e-15 4.42e-15 0.98 l "Ni (n,p) 5.92e-15 6.05e-15 0.98 j 2"U (n,f) (Cd) 2.40e-14 2.28e-14 1.05 j 23'Np (n f) (Cd) 2.18e-13 2.07e-13 1.05
,. "Co (n,y) 3.54e-12 3.55e-12 1.00 i "Co (n,y) (Cd) 1.34e-12 1.35e-12 0.99 i
Surveillance Capsule R
- Reaction Rate (rps/ nucleus) i
} Adjusted l M/C l Measured Cale. Adiusted
'3 '
Cu (n,a) 7.58e-17 7.48e-17 1.01 j "Fe (n p) 8.80e-15 9.04e-15 0.97 "Ni (n.p) 1.26e-14 1.27e-14 0.99 i 2"U (n,f) (Cd) 5.61e-14 5.16e-14 1.09 j 2Np (n,f) (Cd) 5.02e-13 4.88e-13 1.03 j "Co (n,y) 8.86e-12 8.87e-12 1.00 "Co (n,y) (Cd) 3.49e-12 3.49e-12 1.00 l
i l
4 e
1 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 4
~. . _, - , _ _ , . - . . . , - . . .
_ _ _ - . _ . _ __ . . _ - _ , _ . _ _ . _ _ . . . _ . . ~ . . _ . . _ _ _ _ ._ _ . _ ._ _ _ ____ . _ . . . _
? 6-32 l
} TABLE 6-10 cont'd i COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER i
, Surveillance Capsule T i Reaction Rate (rps/ nucleus)
Adjusted M/C ;
Measured Cale. Adiusted ,
l "Cu (n,a) 5.65e-17 5.54e-17 1.02 "Fe (n p) 5.37e-15 5.57e-15 0.%
"Ni (n.p) 7.87e-15 7.85e-15 1.00 2"U (n,0 (Cd) 2.85e-14 2.79e-14 1.02 287 Np (n,0 (Cd) 2.58e-13 2.46e-13 1.05 :
"Co (n,y) 4.76e-12 4.76e-12 1.00 "Co (n,y) (Cd) 1.62e-12 1.62e-12 1.00 i
Surveillance Capsule V Reaction Rate (rps/ nucleus)
Adjusted M/C Measured Cale. Adiusted "Cu (n,a) 6.67e-17 6.60e-17 1.01 "Fe (n,p) 7.45e-15 7.62e-15 0.98 2"U (n,0 (Cd) 4.22e-14 4.08e-14 1.03 23'Np (n,0 (Cd) 3.72e-13 3.68e-13 1.01 "Co (n,y) 8.38e-12 8.39e-12 1.00 "Co (n,y) (Cd) 3.37e-12 3.37e-12 1.00 I
4 l
l I
~
ANALYSIS Cf CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM
1 .
6-33 TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCE CAPSULE i
- Capsule S Energy Flux Energy Flux Group # (MeV) (n/cm2-sec) . Group #
(MeV) (n/cm 2-sec) 1 1.73e+01 5.78e+06 28 9.12e-03 1.32e+10 2 1.49e+01 1.23e+07 29 5.53e-03 1.39e+10 3 1.35e+01 4.48e+07 30 3.35e-03 4.32e@9 4 4 1.16e+01 1.21e+08 31 2.84e-03 4.11e+09 5 1.00e+01 2.73e+08 32 2.40e-03 4.00eM9 4 6 8.61e+00 4.72e+08 33 2.03e-03 1.17e+10 l 7 7.41e+00 1.14e+09 34 1.23e-03 1.14e+10 i 8 6.07e+00 1.73e+09 35 7.49e-04 1.08e+10
, 9 4.97e+00 3.58e+09 36 4.54e-04 9.59e+09 10 3.68e+00 4.21e+09 37 2.75e-04 1.00e+10 11 2.87e+00 8.I6e+09 38 1.67e-04 9.17e+09
- 12 2.23e+00 1.08e+10 39 1.0le-04 1.02e+10 13 1.74e+00 1.47e+10 40 6.14e-05 1.03e+10 i
14 1.35e+00 1.57e+10 41 3.73e-05 1.05e+10 i 15 1.lle+00 1.70e+10 42 2.26e-05 1.04e+10 16 8.21e-01 3.01e+10 43 1.37e-05 1.02e+10 17 6.39e-01 3.21e+10 44 8.31e-06 1.02e i-10
, 18 4.98e-01 2.18e+10 45 5.04e-06 1.03e+10 19 3.88e-01 3.00e+10 46 3.06e-06 1.04e+10 ,
, 20 3.02e-01 3.66e+10 47 1.86e-06 1.04e+10 21 1.83e-01 3.31e+10 - 48 1.13e-06 9.22e+09 2
22 1.lle-01 2.50e+10 49 6.83e-07 8.06e+09
! 23 6.74e-02 2.04e+10 50 4.14e-07 1.39e+10 l
24 4.09e-02 1.23e+10 51 2.51e-07 1.51e+10 25 2.Ce-02 1.15e+10 52 1.52e-07 1.65e+10 1 26 1.99e-02 7.60e+09 53 9.24e-08 4.71e+10 27 1.50e-02 1.24e+10 Note: Tabulated energy levels represent the upper energy in each group.
l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM
.____ _ _ _ _ _ _ - . ~ . . _ _ . . _ _ . - _ _ . . _ . _ . _ - . _ _ . . . _ _ . . _ _ . _ . _ . _ _ _ ...m.__ _.
4 34 1 f
7 TABLE 6-11 cont'd 4
ADJUSTED NEUTRON ENERGY SPECTRUM AT THE i CENTER OF SURVEILLANCE CAPSULE Capsule R Energy Flux Energy Flux
?
Grouc # (MeV) (n/cm 2-sec) Group # (MeV) (n/cm:-sec) 1 1.73e+01 8.52e+06 28 9.12e-03 3.08e+ i0 2 1.49e+01 1.83e+07 29 5.53e-03 3.26e+10 3 1.35e+01 6.93e+07 30 3.35:-03 1.02e+10 4 1.16e+01 1.93e+08 31 2.84e-03 9.75e+09 5 1.00e+01 4.49e+08 32 2.40e-03 9.58e+09 6 8.61e+00 8.17e+08 33 2.03e-03 2.84e+10 7 7.41e+00 2.05e+09 34 1.23e-03 2.79e+10 i 8 6.07e+00 3.32e+09 35 7.49e-04 2.66e+10 9 4.97e+00 7.41e+09 36 4.54e-04 2.40e+10 1 10 3.68e+00 9.30e+09 37 2.75e-04 2.53e+10 ;
11 2.87e+00 1.86e+10 38 1.67e-04 2.42e+10 J 12 2.23e+00 2.55e+10 39 1.0le-04 2.60e+10 13 1.74e400 3.55e+10 40 6.14e-05 2.61e+10 14 1.35e+00 3.84e+10 41 3.73e-05 2.62e+10 i 15 1.lle+00 6.68e+10 42 2.26e-05 2.60e+10 l 16 8.21e-01 7.46e+10 43 1.37e-05 2.55e+10 17 6.39e-01 7.97e+10 44 8.31e-06 2.53e+10 18 4.98e-01 5.34e+10 45 5.04e-06 2.53e+10 19 3.88e-01 7.30e+10 46 3.06e-06 2.53e+10 20 3.02e-01 8.74e+10 47 1.86e-06 2.53e+10 21 1.83e-01 7.86e+10 48 1.13e-06 2.25e+10 22 1.lle-01 5.91e+10 49 6.83e-07 2.00e+10 23 6.74e-02 4.78e+10 50 4.14e-07 3.60e+10 24 4.09e-02 2.90e+10 51 2.51e-07 3.82e+10 25 2.55e-02 2.70e+10 52 1.52e-07 4.08e+10 26 1.99e-02 1.77e+10 53 9.24e-08 1.14e+11 27 1.50e-02 2.88e+10 Note: Tabulated energy levels represent the upper energy in each group.
)
l 7NALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
I e 6-35 '
l a :
1 1
TABLE 6-11 cont'd
-f ADJUSTED NEUTRON ENERGY SPECTRUM AT THE j CENTER OF SURVEILLANCE CAPSULE J
} Capsule P
}
Energy Flux Energy Flux Group # (MeV) (n/cm 2-sec) Group # (n/cm2 ,3,c)
(MeV) l l 1 1.73e+01 7.22e+06 28 9.12e-03 1.50e+10
) 2 1.49e+01 1.54e+07 29 5.53e-03 1.58e+10 )
i 3 1.35e+01 5.65e+07 30 3.35e-03 4.87e+09 i 4 1.16e+01 1.54e+08 SI- 2.84e-03 4.66e+09 5 1.00e+01 3.47e+08
{
, 32 2.40e-03 4.57e+09
, 6 8.61e+00 6.03e+08 33 2.03e-03 1.35e+10 7 7.41e+00 1.47e+09 34 1.23e-03 1.32e+10 I I 8 6.07e+00 2.23e+09 35 7.49e-04 1.26e+10 l 9 4.97e+00 4.54e+09 36 4.54e-04 1.13e+10
$ 10 3.68e+00 5.24e+09 37 2.75e-04 1.18e+10
] 11 2.87e+00 1.0le+10 38 1.67e-04 1.l le+10 12 2.23e+00 1.32e+10 39 1. Ole-04 1.21e+10 ,
l 13 1.74e+00 1.76e+10 40 6.14e-05 1.22e+10
- 14 1.35e+00 1.87e+10 41 3.73e-05 1.23e+10 1 15 1.11e+00 3.18e+10 42 2.26e-05 1.22e+10 l 16 8.21e-01 3.49e+10 43 1.37e-05 1.19e+10 l 17 6.39e-01 3.70e+10 44 8.31e-06 1.19e+10
- 18 4.98e-01 2.51e+10 45 5.04e-06 1.2&+10
- 19 3.88e-01 3.42e+10 46 3.06e-06 1.20e+10
- 20 3.02e-01 4.16e+10 47 1.86e-06 1.20e+10
! 21 1.83e-01 3.74e+10 48 1.13e-06 1.09e+10 22 1.11e-01 2.83e+10 49 6.33e-07 9.74e+09 23 6.74e-02 2.30e+10 50 4.14e-07 1.72e+10 i 24 4.09e 1.39e+10 51 2.51e-07 1.96e+10 3 25 2.55e-02 1.30e+10 52 1.52e-07 2.24e+10 l 26 1.99e-02 8.58e+09 53 9.24e-08 7.0le+10
- 27 1.50e-02 1.40e+10 i
Note: Tabulated energy levels represent the upper energy in each group.
)
!4 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLATA PROGRAM
i l4 6-36 TABLE 6-11 cont'd I
ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCE CAPSULE l
Capsule V Energy Flux Energy Flux Group # (MeV) 2 (n/cm -sec) Group # (MeV) (n/cm 2-sec) 1 1.73e41 7.72e+06 28 9.12e-03 2.67e+10 2 1.49e+01 1.66e+07 29 5.53e-03 2.86e+10 3 1.35e+01 6.30e+07 30 3.35e-03 8.97e+09 4 1.16e+01 1.75e+08 31 2.84e-03 8.65eM9 5 1.00e+01 4.04e+08 32 2.40e-03 8.56e+09 6 8.61e+00 7.28e+08 33 2.03e-03 2.56e+10 7 7.41e+00 1.80e+09 34 1.23e-03 2.54e+10 8 6.07e+00 2.88e+09 35 7.49e-04 2.45e+10 9 4.97e+00 6.29e+09 36 4.54e-04 2.23e+10 10 3.68e+00 7.68e+09 37 2.75e-04 2.38e+10 11 2.87e+00 1.49e+10 38 1.67e-04 2.37e+10 12 2.23e+00 1.98e+10 39 1.0le-04 2.45e+10 13 1.74e+00 2.69e+10 40 6.14e-05 2.44e+10 14 1.35e+00 2.89e+10 41 3.73e-05 2.44e+10 15 1.lle+00 4.98e+10 42 2.26e-05 2.40e+10 16 8.21e-01 5.54e+10 43 1.37e-05 2.34e+10 17 6.39e-01 5.94e+10 44 8.31e-06 2.31e+10 18 4.98e-01 4.0le+10 45 5.04e-06 2.30e+10 19 3.88e-01 5.55e+10 46 3.06e-06 2.30e+10 20 3.02e-01 6.76e+10 47 1.86e-06 2.29e+10 21 1.83e-01 6.18e+10 48 1.13e-06 2.03e+10 22 1.lle-01 4.74e+10 49 6.83e-07 1.83e+10 23 6.74e-02 3.90e+10 50 4.14e-07 3.30e+10 24 4.09e-02 2.40e+10 51 2.51e-07 3.52e+10 25 2.55e-02 2.27e+10 52 1.52e-07 3.76e+10 26 1.99e-02 1.51e+10 53 9.24e-08 1.06e+11 l, 27 1.50e-02 2.48e+10 Note: Tabulated energy levels represent the upper energy in each group.
1 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATK)N SURVEILLANCE PROGRAM
6-37 l TABLE 6-12 COMPA3!YSON OF CALCULATED AND MEASURED INTEGRATED NEUTRON EXPOSURE OF PRAIRIE ISLAND UNIT 1 SURVEILLANCE CAPSULES S, R, P. AND V i
I l CAPSULE S Calculated Measured M/C i
G(E > 1.0 MeV) [n/cm2] 4.338e+19 4.017e+19 0.93
@(E > 0.1 MeV) [n/cm2] 1.527e+20 1.584e+20 1.04 dpa 7.461e-02 7.309e-02 0.98 I
l l CAPSULE R Calculated Measured M/C l
@(E > 1.0 MeV) [n/cm2] 4.000e+19 4.478e+19 1.12 C(E > 0.1 MeV) [n/cm2] 1.516e+20 1.817e+20 1.20 dpa 7.121e-02 8.197e-02 1.15 l
CAPSULE P l
Calculated Measured M/C
$(E > 1.0 MeV) [n/cm2] 1.314e+19 1.318e+19 1.00
$(E > 0.1 MeV) [n/cm2] 4.521e+19 4.994e+19 1.10 dpa 2.234e-02 2.354e-02 1.05 l
l CAPSULE V i Calculated Measured M/C Q(E > 1.0 MeV) [n/cm2] 6.267e+18 5.630e+18 0.90
$(E > 0.1 MeV) [n/cm2] 2.375e+19 2.251e+19 0.95 dpa 1.116e-02 1.031e-02 0.92 2
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
6-38 TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS I ON THE REACTOR VESSEL CLAD / BASE METAL INTERFACE Best Estimate Exposure (18.12 EFPY) at the Reactor Vessel Inner Radius 0* 15" 30 45
$ (E > 1.0 MeV) 1.59e+19 1.13e+19 8.98e+18 7.88e+18
@ (E > 0.1 MeV) 4.74e+19 3.57e+19 2.66e+19 2.25e+19 1 dpa 2.70e-02 1.98e-02 1.53e-02 1.33e-02 Best Estimate Exposure (24 EFPY) at the Reactor Vessel Inner Radius 0 15* 30 45*
$ (E > 1.0 MeV) 2.11e+19 1.50e+19 1.19e+19 1.04e+19
@ (E > 0.1 MeV) 6.27e+19 4.73e+19 3.53e+19 2.98e+19 dpa 3.58e-02 2.62e-02 2.03e-02 1.76e-02 1
l l
Best Estimate Exposure (35 EFPY) at the Reactor Vessel Inner Radius 0 15* 30 45
@ (E > 1.0 MeV) 3.07e+19 2.18e+19 1.73e+19 1.52e+19
$ (E > 0.1 MeV) 9.15e+19 6.90e+19 5.15e+19 4.35e+19 l dpa 5.22e-02 3.82e-02 2.96e-02 2.57e-02 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
6-39 TABLE 6-14 NEUTRON EXPOSURE VALUES WITHIN THE l PRAIRIE ISLAND UNIT 1 REACTOR VESSEL 1
FLUENCE BASED ON E > 1.0 Me V SLOPE 24 EFPY @ (E > 1.0 MeV) [n/cm2 j ,
I l 0* 15* 30* 45* l l Surface 2.11e+19 1.50e+19 1.19e+19 1.04e+19
%T 1.35e+19 9.64e+18 7.56e+18 6.75e+18
%T 7.63e+18 5.54e+18 4.32e+18 3.88e+18
%T 4.13e+18 3.08e+18 2.39e+18 2.15e+18 2
35 EFPY $ (E > 1.0 MeV) (n/cm ]
0* 15* 30* 45*
Surface 3.07e+19 2.18e+19 1.73e+19 1.52e+19
%T 1.96e+19 1.41e+19 1.10e+19 9.85e+18
%T 1.1le+19 8.08e+18 6.30e+18 5.66e+18 l l %T 6.02e+18 4.50e+18 3.49e+18 3.14e+18 '
1 FLUENCE BASED ON dpa SLOPE 2
24 EFPY @ (E > 1.0 MeV) [n/cm )
0* 15* 30 45*
Surface 2.11e+19 1.50e+19 1.19e+19 1.04e+19
%T 1.50e+19 1.09e+19 8.52e+18 7.55e+18
%T 9.95e+18 7.41e+18 5.78e+18 5.12e+18
%T 6.34e+18 4.91e+18 3.83e+18 3.36e+18 2
35 EFPY $ (E > 1.0 MeV) [n/cm ]
0 15* 30* 45 Surface 3.07e+19 2.18e+19 1.73e+19 1.52e+19
%T 2.19e+19 1.58e+19 1.24e+19 1.10e+19
%T 1.45e+19 1.08e+19 8.43e+18 7.46e+18
%T 9.25e+18 7.16e+18 5.59e+18 4.90e+18 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE IStAND Unit 1 REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM
6-40 TABLE 6-15 UPDATED LEAD FACTORS FOR PRAIRIE ISLAND UNIT 1 SURVEILLANCE CAPSULES Capsule Lead Factor VI'! 2.94 P'*1 1.72 R 2.99 S'd3 1.77 N"' l.77 T*3 1.89
[a] - Withdrawn at the end of Cycle 1.
[b] - Withdrawn at the end of Cycle 5.
[c] - Withdrawn at the end of Cycle 9.
[d] - Withdrawn at the end of Cycle 17.
(e) - Capsules remaining in the reactor. J l
l l
l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 4
+
71 1
7.0 RECOMMENDED 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 Prairie Island Unit 1 reactor vessel. This recommended removal schedule is applicable to 35 EFPY.
TABLE 7-1 Recommended Surveillance Capsule Removal Schedule for the Prairie Island Unit 1 Reactor Vessel Capsule Location Withdrawal Fluence (*
f Capsule (degree) Lead EFPYN (n/cm", E > 1.0 MeV) l Factor
- V 77 2.94 1.34 5.630 x 10(4 P 247 1.72 4.60 1.318 x 10'** J R 257 2.99 8.56 4.478 x 10'H" S 57 1.77 18.12 4.017 x 10'N4 '
T 67 1.89 Standby - - i*
N 237 1.77 Standby --
- NOTES:
(a) Updated in Capsule S dosimetry analysis; see Section 6.0 of this report.
(b) Effective Full Power Years (EFPY) from plant startup.
(c) Plant-specific evaluation.
(d) Capsule T will reach the projected peak vessel fluence (at 52.5 EFPY) at approximately 27.7 EFPY. Capsule N will reach the projected peak fleunce (at 52.5 EFPY) at approximately 29.6 EFPY.
i l
l l
l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
l 8-1
8.0 REFERENCES
- 1. Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, May 1988.
- 2. 10 CFR 50, Appendix G, " Fracture Toughness Requirements", Federal Register, Volume 60, No. 243, dated December 19,1995.
- 3. WCAP-8086, " Northern States Power Co. Prairie Island Unit No.1 Reactor Vessel Radiation Surveillance Program", S. E. Yanichko and D.J. Lege, June 1973.
- 4.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, " Fracture Toughness Criteria for Protection Against Failure".
- 5. 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.
- 6. WCAP-11006, " Analysis of Capsule R from the Northem States Power Company Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program", R. S. Boggs, et al.,
February 1986.
- 7. Analysis Report No.16150," Prairie Island Unit 1", P.O. CL3DCHMLB, COTA - Analytical Laboratory, Waltz Mill Site, Approved by L. F. Becker, 10/18/96.
- 8. Societe Des Forges Et Ateliers Du Creusot Usines Schneider, Chemical Analysis Report No.17-9-2, NSP shell course C, Heat 21918/38566.
- 9. Societe Des Forges Et Ateliers Du Creusot Usines Schneider, Chemical Analysis Report No.15-8-1, NSP shell course D Heat 21887/38530.
10.10 CFR Part 50, Appendix H, " Reactor Vessel Material Surveillance Program Requirements", Federal Register, Volume 60, No. 243, dated December 19,1995.
- 11. 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.
- 12. ASTM E23-93a, " Standard Test Methods for Notched Bar impact Testing of Metallic Materials", in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1993.
l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
8-2
- 13. ASTM A370-92, " Standard Test Methods and Definitions for Mechanical Testing of Steel l Products", in ASTM Standards, Sectior. 3, American Society for Testing 'and Materials, Philadelphia, PA,1993.
- 14. ASTM E8-93, " Standard Test Methods for Tension Testing of Metallic Materials", in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1993.
15 ASTM E21-92, " Standard Test Methods for Elevated Temperature Tension Tests of Metaitic Materials", in ASTM Standards, Section 3, American Society for Testing and Mateaals, Philadelphia, PA,1993.
.16. ASTM E83-93, " Standard Practice for Verification and Classification of Extensotneters", in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1993.
- 17. ASTM Designation E853-87, Standard Practice for Analysis andInterpretation of Light Water Reactor Surveillance Results, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 18. ASTM Designation E693-79, " Standard Practice for Characterizing Neutron Exposures in l
Ferritic Steels in Terms of Displacements per Atom (dpa)", in ASTM Standards, Section I 12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 19. RSIC Computer Code Collection CCC-543, " TORT-DORT Two- and Three-Dimensional Discrete Ordinates Transport, Version 2.7.3", May 1993.
- 20. RSIC Data Library Collection DLC-175, " BUGLE-93, Production and Testing of the l
VITAMIN-B6 Fine Group and the BUGLE-93 Broad Group Neutron / Photon Cross-Section i Libraries Derived from ENDF/B-VI Nuclear Data", April 1994. I
- 21. R. E. Maerker, et al, " Accounting for Changing Source Distributions in Light Water Reactor Surveillance Dosimetry Analysis", Nuclear Science and Engineering, Volume 94, Pages 291-308,1986.
- 22. J. E. Schaefer, transmittal of Prairie Island Unit 1 Cycle 1 core inventory and average '
axial conditions, September 11,1996.
- 23. K. A. Jones, et al., "The Nuclear Design - Core Management of the Prairie Island Unit 1 Nuclear Reactor Cycle 2", WCAP-8744-R1, April 1976. [ Westinghouse Proprietary Class 2]
1 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
i 4
8-3 j
! 24. D. J. Franks, et. al., "The Nuclear Design and Core Management of the Prairie Island Unit 1 Nuclear Reactor - Cycle 3", WCAP 8956, March 1977. [ Westinghouse Proprietary
, Class 2]
}
- 25. M. F. Muenks, et. al., "The Nuclear Design and Core Management of the Prairie Island j Unit 1 Nuclear Reactor - Cycle 4", WCAP-9306, April 1978. [ Westinghouse Proprietary l Class 2]
! 26. Prairie Island Unit 1 Cycle 5 Core Data, XN-NF-78-47, November 1978, and core inventory and average axial conditions from J. E. Schaefer, September 11,1996.
i 27. Prairie Island Unit 1 Cycle 6 Core Data, XN-NF-79-104, December 1979, and core j inventory and average axial conditions from J. E. Schaefer, September 11,1996. !
t i
! 28. Prairie Island Unit 1 Cycle 7 Core Data, XN-NF-81-42, June 1981, and core inventory and
! average axial conditions from J. E. Schaefer, September 11,1996.
- 29. Prairie Island Unit 1 Cycle 8 Core Data, XN-NF-82-66, Revision 1, November 1982, and core inventory and average axial conditions from J. E. Schaefer, September 11,1996.
- 30. Prairie Island Unit 1 Cycle 9 Startup and Operations Report, NSPNAD-8403P, Revision 1, April 1984 [NSP Proprietary information], and core inventory and average axial conditions from J. E. Schaefer, September 11,1996.
- 31. Prairie Island Unit 1 Cycle 10 Startup and Operations Report, NSPNAD-8415P, Revision 1, dated February 1985 [NSP Proprietary information], and average axial conditions from J. E. Schaefer dated September 11,1996.
- 32. S. Srinilta, et al., "The Nuclear Design and Core Management of the Prairie Island Unit 1 1 Cycle 11", WCAP-11070, March 1986 [ Westinghouse Proprietary Class 2], and average axial conditions from J. E. Schaefer, September 11,1996.
- 33. Prairie Island Unit 1 Cycle 12 Final Reload Design Report, NSPNAD-8701P, January 1987 [NSP Proprietary Information], and average axial conditions from J. E. Schaefer, September 11,1996.
34.' Prairie Island Unit 1 Cycle 13 Final Reload Design Report, NSPNAD-8810P, May 1988 [NSP Proprietary Information], and average axial conditions from .
J. E. Schaefer, September 11,1996.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
8-4 l
- 35. Prairie Island Unit 1 Cycle 14 Final Reload Design Report, NSPNAD-89013, September 1989 [NSP Proprietary infortnation}, and average axial conditions from J. E. Schaefer, September 11,1996.
- 36. Prairie Island Unit 1 Cycle 15 Final Reload Design Report, NSPNAD-91006, i Revision 1, August 1991 [NSP Proprietary information], and average axial conditions from J. E. Schaefer, September 11,1996.
- 37. Prairie Island Unit 1 Cycle 16 Final Reload Design Report, NSPNAD-92007, July 1992 [NSP Proprietary Information], and average axial conditions from J. E. Schaefer, September 11,1996. i
- 38. Prairie Island Unit 1 Cycle 17 Final Reload Design Report, NSPNAD-94004, May 1994 [NSP Proprietary information), and average axial conditions from J. E. Schaefer, September 11,1996.
- 39. 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.
- 40. 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.
- 41. ASTM Designation E706-87, Standard Master Matrix for Light-Water Reactor Pressure Vessef Surveitlance Standard, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 42. ASTM Designation E261-90, Standard Practice for Detennining Neutron Fluence Rate, Fluence, and Spectra by Radioactivation Techniques, in ASTM Standards, Section 12, ;
American Society for Testing and Materials, Philadelphia, PA,1993.
- 43. ASTM Designation E262-86, Standard Method for Determining Thermal Neutron Reaction and Fluence Rates by Radioactivation Techniques, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 44. ASTM Designation E263-88, Standard Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Iron, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
8-5
- 45. ASTM Designation E264-92, Standard Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Nickel, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 46. ASTM Designation E481-92, 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.
- 47. ASTM Designation E523-87, Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Copper, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 48. ASTM Designation E704-90, Standard Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 49. ASTM Designation E705-90, Standard Test Method for Measuring Reaction Rates by Radioactivation of Neptunium-237, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 50. ASTM Designation E1005-84, Standard Test Method for Application and Analysis of l Radiometric Monitors for Reactor Vessel Surveillance, in ASTM Standards, Section 12, l American Society for Testing and Materials, Philadelphia, PA,1993.
- 51. F. A. Schmittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford Engineering '
Development Laboratory, Richland, WA, September 1979.
- 52. W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative Method of Neutron Flux Spectra Determined by Foil Activation, AFWL-TR-7-41, Vol. l-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.
- 53. RSIC Data Library Collection DLC-178, SNLRML Recommended Dosimetry Cross-Section Compendium, July 1994.
- 54. EPRl-NP-2188, Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications, R. E. Maerker, et al.,1981.
- 55. 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.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
~. . _ _ -
~
8-6
- 56. 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. ,
- 57. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATI Consulting, March 1996.
l l
l l
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRK!
% m. + 2_..a ., u . _ -w- A .~A-A-0 l I
I J
. i APPENDIX A
- 1 LOAD-TIME RECORDS FOR CAPSULE S CHARPY IMPACT TESTS 1
1 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
. I i
Cwve 784472-AF13
$h W.g Wp r lg : : :
l -Py - Maximum Load Pp- Fast Fracture Load 8!
3m
!Ey P Geneml gy Yield Load l
g s "' 1 t
c3 Em g yy g PA- Fast Fracture ;
gi 3 Arrest Load l I l-83 1
[M" i i
I !
1 I r
@ l 9 l I i 1 o 1 1 I
2 I I I ;
y I I 5 I m I I I a
i i
_ 1 g
- 4-t gy t E ty :
E :
l fp l Time W, - Fracture initiation Region tGY -Eme to GenemlMng Wp- Fracture Propagation Region t y -Eme to hxinum Load tp !
-Time to Fast (Brittle) Fracture Start ,
a Fig. A-1-idealized load-time record i
1 l
PeAIRIE 13UWC et NEP TANC
} 4 s s i e- -
M
- 1 z 2 S"
w 1
I e-4 1 u- -
A ..aa.
.a
..b AA.
___a_ _ . . ___
_m _ . _
.t 1.a a.4 3.6 4.8 6. 0 TIE 4 MSEC )
PRAIRIE ISLAND #1 SPECIMEN NUMBER :N27 MATERIAL : TANG CAPSULE : PRAIRIE ISLAND rantR E sure et Mao TaMG s
R s s s 4
M 1
7
^gk ~
S" S
A n __ A _ , m _ . p _ _ _ _ _ _ q _ _ _ -- _ _ q . - - -
n s.: a.4 s.s 4.s 6.0 TIE ( MBEC )
PRAIRIE ISLAND #1 SPECIMEN NUMBER :N28 MATERIAL : TANG
- PRAIRIE ISLAND l CAPSULE Figure A-2. Load-time records for Specimens N27 and N28. j A-2
MtA1RIC 13Use el M32 TANC
- 4 i 4 4
.i e
I S"
l 2~
q-W ha n ~ n a ,a . mm m _ ___ _ _ _ _ _ _ _ _ _ _ _
. i e i i
.n s.: a.4 s.6 4.s 6. o 717t ( MSEC 3 PRAIRIE ISLAND #1 SPECIMEN NUMBER :N32
' MATERIAL TANG CAPSULE : PRAIRIE ISLAND meArate tsuwe et ees rang i a # g e- -
e n
I e
a .- -
3 w
b l
y- -
A h AAA^u ^ - __ .__ _ __ __
. i e i i
- % 1.2 E.4 3.6 4.8 6. 0 TIPE ( ftSEC >
PRAIRIE ISLAND #1 SPECIMEN NUMBER :N25 MATERIAL : TANG CAPSULE : PRAIRIE ISLAND Figure A-3. Load-time records for Specimens N32 and N25.
A-3 '
s - - . -- - - . ~ . . . , ~ _~. ,
1 1'
MtA!RIE 13Ut0 el rt31 TsetG g 4 i 4 e
M
, 1 1
3 w
1
- l
. A 1 M-
. l 1
a a MAAmAAmAAnnamu ____ _.
4 4 i
.D 1.2 2.4 3.6 4
4.8 6.9 TIPC ( PtSEC )
' PRAIRIE ISLAND #1 SPECIMEN NUMBER :N31 1
MATERIAL : TANG
- CAPSULE
- PRAIRIE IS_ LAND MIAIRIE 13Uvc et f29 Tesc
,1 4 6 s i Y
.- - l N
l m '
1 1
eN a l w
A I
e 4 4 4 4
.9 1.2 2.4 3.6 4.8 6.0 T!PE ( PtIEC )
PRAIRIE ISLAND #1 SPECIMEN :N29 MATERIAL : TANG CAPSULE : PRAIRIE ISLAND Figure A-4. Load-time records for Specimens N31 and N29.
A-4
Pflp!RIC 13LNe et 905 Tec i
} 6 6 6
l l
4 e
z -
S" 4
- = _ - __
l 4
i ,
e . i 1.a e.* 3.6 4.s 4.s
.t TIPC ( ftSEC )
PRAIRIE ISLAND #1 SPECIMEN :N35 MATERIAL : TANG CAPSULE : PRAIRIE ISLAND PRA!RIC 13uve og toe tapg I 3 4 4
[.
T n
1 m
a 3-S" w
e- '
u-1 e a i e
.D 1.2 2.4 3.6 4.8 6.O T!fC ( ftBEC )
PRAIRIE ISLAND #1 SPECIMEN NUMBER :N34 MATERIAL : TANG CAPSULE : PRAIRIE ISLAND Figure A-5. Load-time records for Specimens N35 and N34.
A-5
-.m-_. -
Panratt Isuve et re6 Tang I
1 4 i i a 4
?
- ~ -
i ,
m I
i
~ e- -
, d 5
Y 6
l
. . . . 4
.t 1.s 3.4 a.6 4.s 6.o TIfE C Mstc 3 PRAIRIE ISLAND #1 SPECIMEN NUMBER :N26 MATERIAL : TANG i CAPSULE : PRAIRIE ISLAND
- UNIT #1 "S" PantnIC !suve et Mas fast e.
4 4 6 i l
4 M
1 z
S" w
A a
,T 4 I i f
.3 1.E 2.4 3.6 4.s 6.0 7tfC ( HsEC 3 PRAIRIE ISLAND #1 SPECIMEN NUMBER :N36 MATERIAL : TANG CAPSULE : PRAIRIE ISLAND
- UNIT #1 "S" Figure A-6. Load-time records for Specimens N26 and N36.
l A-6 I
,, Pen!!!E 13Uwe et M3c TAMG
} i i a i 4
M 1
7 S"
,d S I g- _
l i
. e i i i
.8 1.s e.4 2.s 4.s s.o T114: ( Msc > I PRAIRIE ISLAND #1 SPECIMEN :N30 MATERIAL : TANG CAPSULE : PRAIRIE ISLAND
- e e e s 4
e 1 i 3 .
5" -
l w I a
- q. _
. i i i .
.3 1.s e.4 3.s 4.s s.o 7 114: ( Mstc 3 PRAIRIE ISLAND #1 SPECIMEN NUMBER :N33 MATERIAL : TANG ,
CAPSULE : PRAIRIE ISLAND
- UNIT #1 "S" Figure A-7. Load-time records for Specimens N30 and N33. ,
i i
A-7
l b
L P
i i
1 i
j SPECIMEN NUMBERING CODE: ;
N- FORGING C (TANGENTIAL) i S- FORGING C (AXIAL) 2 j R- ASTM CORRELATION MONITOR !
W- WELD METAL 3
- . H-HEAT AFFECTED10W2 MATERIAL i
i
)
i 1 m uas wm. wm. =m. m ens wa. wa. mm. meas owns owns enams cuanevsi
- we wr as see was see was est une ass wie
- we we wr - e a er wB M gr SEs w38 833 wRt 831 wl9 W wl1 db
] 1 6 a i 4
U- >; h} - F# j 7Al .15% Co .
- s. ; (ce (j J ;,,,
w - Fe G (j !j
'i
- seca r !j l
uq n MONITon j! nj l !: sm Ih I! mourron i 1 1
l
)
- CENTER TO TOP OF VESSEL
! / R
- d
)
/
/
ss- //
l 1
- I .
4 s/
s--,
CORE
)
i i j ;
i l l CAPSULE S THERMAL SHIELD l VESSEL WALL 3
- f i
i f !
l l 1
en 834 Sas 10 5 ISS IIM 8834 ,t18 8018 4884 I.S ' IER IW 6. IIIB M18' IIN M17 IS
, , , ,,,. ,, ., ,. , m. , ,
v
. - , v
.e h
_ ,, Co - Jiea j L _ _ , = .
- 155 cal j iiIN 1i (Cd)
! U ul UU r'.
en., soo<v :!! .,.e J, l
MENITOR MONITOR jj MONITOR g l
\
EGION OF VESSEL 0\M O 14 6 l I
TO BOTTOM OF VESSEL 4
i Figure 4-2 Capsule S Diagram Showing the j
i Location of Specimens, Thermal Monitors and Dosimeters l
I I
}
i 4
._- = . . _ _ _ _ _ _ _ _ . . . . - . . . . _ .. ~ _ _ _ _ _ _ _ . _ _ _ _ _ .
o Pmat2IE ISUIMD el 325 msg !
4 4 4 4 l
e 1
a .- -
7 s" j w
e3 aa AA -A na AA --AA_ na anmaam ,a . .. . _ m _
e & 4 i A
.1 1.2 E.4 S.6 4.9 6.0 1 7IE l
( MSEC )
PRAIRIE ISLAND #1 SPECIMEN NUMBER :S25 MATERIAL : AXIAL CAPSULE : PRAIRIE ISLAND Pansatt tsune et ss4 mit
- , p 4 5 6 4
m 1
x -
S"
. i S
an A _A _ _ _ ^ _ - ._ _- - -- -
.A._ m , . . .
s.2 a.4 s.6 4.s 6.o n
TIE ( MSEC )
PRAIRIE ISLAND #1 SPECIMEN NUMBER :S34 MATERIAL : AXIAL CAPSULE : PRAIRIE ISLAND Figure A-8. Load-time records for Specimens S25 and S34.
A-8
_ . _ _ _ _ _ . _ . _ ..___m . _ ._m._ ._._.___.m __ . _ _ _ _ _.m ._. _ _ _ _ _ _ _ _ . _ _ _ . - __ _ _ _ .
MtA!RIE Iture el 529 as(Iag.
- I i 4 4 .
1 4
M
- 1 i e / -
e s '(i a
w
- 84 i
}
, u-*
4 l
3 l
- \
. .__ .__._______ l
. MAAAe.Annan-mn 4 i 4
.% 1.2 2.4 3.6 4.8 6. 0
( MIEC ) l TIE PRAIRIE ISLAND #1 I l
SPECIMEN NUMBER :S29 MATERIAL : AXIAL CAPSULE : PRAIRIE ISLAND MIAIRIE ISUWC 01 SW AstiaL i
- i 4 I i
- \
e- 1 4 .
M 1
I
- e- .
et g
4 AAAA ^& n Aa a ^^m AAa x, .. _.._ ___a .
. i e i .
.c 1.a a.4 3.s 4.s s.o TIE < MIEC >
PRAIRIE ISLAND #1 SPECIMEN NUMBER :S28 MATERIAL : AXIAL CAPSULE : PRAIRIE ISLAND Figure A-9. Load-time records for Specimens S29 and S28.
A-9
\
PanIRIC Iture 01 330
- a i i s 4
l e l 1
1 7
S"
- i
)
l a
.- - l 4 \
l l
I 1
Am A A m AA
- m AA -AA. __ _. _ _.
l
. i e i i
.3 1.3 3.4 3.6 4.s 6.e I I
TIE ( MSC )
PRAIRIE ISLAND #1 SPECIMEN NUMBER :S3o MATERIAL : AXIAL CAPSULE : PRAIRIE ISLAND .
PantRIE ISUWO el 327 ANIAL
- i i i i 4
.- 1 4
e 1
s -
a .-
e 5 a
v s-A-__ _-e __- _ _ _ _ _
, e i i i
.3 1.3 2.4 3.6 4.3 6. e TIE ( MSEC )
PRAIRIE ISLAND #1 SPECIMEN NUMBER :S27 MATERIAL : AXIAL CAPSULE : PRAIRIE ISLAND Figure A-10. Load-time records for Specimens S30 and S27.
A-10
. - . _ .- - .. _. .. .~ . _ . . - _ . . -
. . . _ - ..- __ = - .~
l Pan!R!c 13Uwe el 333 MIN. ~
- 4 4 4 i
- 1 I
.- i 4
e <
1 s
a' S <
1 w l A
~
! 9. "
e l
l
^ _
. n o s .
.D 1.2 2.4 3.s 4.s s. e TIE C Mste >
PRAIRIE ISIAND #1 l SPECIMEN :S33 MATERIAL : AXIAL CAPSULE : PRAIRIE ISLAND enntmic Isuire et sas mim. l
- i
- i i 4 I
n
!E
~ . - -
a d ed w
a q- -
. a i .
.D .a e.4 3.s 4.s s. e TIE ( MIEC )
l PRAIRIE ISLAND #1 SPECIMEN NUMBER :S36 MATERIAL : AXIAL CAPSULE : PRAIRIE ISLAND Figure A-11. Load-time records for Specimens S33 and S36.
A-11 s
. . . - . . . ._ ..-n .. . . ~ . . -. ..-.. . - _-_
Mla!RIC 13Uwe el 386 aggat i
- . 4 6 6 4
(
e i
e j 1
~ e- -
S" ,('
w j ,- -
at k
u- -
l
. i e s
. 11 1.5 2.4 3.6 4.0 6.0 T!PC ( MSCC )
PRAIRIE ISLAND #1 SPECIMEN NUMBER :S26 MATERIAL : AXIAL CAPSULE : PRAIRIE ISLAND i
Specimen Alignment Error - Data not valid.
1 I
Figure A-12. Load-time records for Specimens S26 and S31.
A-12
FRAIREC 11Upc et 335 artA j i 6 4 4 M
1 a .- -
3 w
a q- -
l
. i > > i
.D 1.2 2.4 3.6 4.8 6. 0 TIE ( MBEC )
PRAIRIE ISLAND #1 SPECIMEN :S35 MATERIAL : AXIAL CAPSULE : PRAIRIE ISLAND Pantatt Isuem es sat axist g i 4 4 4 4
M e
a e- -
S" w
4 g- -
. 4 i i i
.D 1.2 2.4 3.6 4.0 6. 0 T!E ( MBEC >
PRAIRIE ISLAND #1 SPECIMEN NUMBER :S32 MATERIAL : AXIAL CAPSULE : PRAIRIE ISLAND Figure A-13. Load-time records for Specimens S35 and S32.
A-13
. _ _ . .. . _ _ . _ _ _ _ _ . . _ _ _ _ _ ~ . . _ _ _ . _ _ . . . _ . . . _ _ . _ _ _ _ . _ _ . _ _ _ .
3 4
PRA!RIC ISLAND 88 W23 L41D 8, a i 6 6 M
1 7
m ,_ -
d 3 .
i 1
g4 l
I AnaM. A,_ _a_ _ m. ..___.__. ______
, 6 6 6 i
.3 1.a a.4 3.s 4.s 6. o
( ttre < nste >
PRAIRIE ISIAND #1 SPECIMEN :W23 l MATERIAL : WELD CAPSULE : PRAIRIE ISIAND PRAIRIE ISUWC et ute lELD
\
} 6 i i ;
l 1
l l l
M 1
I l
^-
3 w
a
=- -
I AA^^^_a -n,. ..
. , ..,...__m,_ _ __ ___ l
- D 1.2 2.4 3.6 4.8 6.0 l TIPC ( nstc )
i PRAIRIE ISLAND #1 SPECIMEN NUMBER :W18 MATERIAL : WELD
- CAPSULE
- PRAIRIE ISLAND l Figure A-14. Load-time records for Specimens W23 and W18.
A-14
. .- - . . - . . ., - ~. . . . . . . . - . - - ~ _ _ - - _ - . .- - . _ . - - - - _ . -- . _ . _ _ . . -
! . j 1
l l
. i Mte:RIC ISUWe et u22 talD s a a i l
e~ ~
M 1
2 a A es v
a 1 n~ -
l I e a a 4 .
. 3.2 2.4 3.6 4.8 6.0 TInc ( ftSEC ) I i '
PRAIRIE ISIAND #1 l
SPECIMEN NUMBER :W22 MATERIAL : WELD CAPSULE : PRAIRIE ISLAND I
cantaic tsuWe et set isto '
i a e
1 4
e z
S" a
q- -
. . i i .
l . .a a.4 a.s 4.s s.s 7 rc c nace >
i l
PRAIRIE ISLAND #1 SPECIMEN NUMBER :W21
' MATERIAL : WELD CAPSULE : PRAIRIE ISLAND l.
i Figure A-15. Load. time records for Specimens W22 and W21.
A-15 1
, _ . _ - . _ __ . ._. _ _ . _ _ . - . .- - .m _m.._._ . - - . _ . _ . - _ . _ _ _ _ . _ _ _ . _ _ . . _ . . . . ____m.__._
4 e
MIA!RIC 13UWC 01  631,0 l I e
i 3 i a l
l R~
~
l n i 1 l E
~e-5" w
1 1
! l
! N l
u- .
- i l
.t 1.2 2.4 3.6 4.S 6. 0 l TIE < MstC )
PRAIRIE ISLAND #1 SPECIMEN :W19 i MATERIAL : WELD CAPSULE :PRAIRI,E ISLAND MteIRIE !$Uuc et net LELD l 9
e 6 8 4 4 4
e 1
a
$ i i
- l eJ l
l l
~
i 1 i l
l
.9 1.2 2.4 3.6 4.8 6.0 TIE C MstC >
PRAIRIE ISLAND #1 l SPECIMEN NUMBER :W24 MATERIAL : WELD CAPSULE : PRAIRIE ISLAND
! Figure A-16. Load-time records for Specimens W19 and W24.
A-16
T 4
e !
Pen!RIC ISLMO et W17 m 4 6 s 4 6 M
1 r
a .- -
A 3
v 4
L u- -
l
. A i i 4;
.9 1.2 2.4 3.6 4.0 6. 0 TIE C2)
PRAIRIE ISLAND #1 SPECIMEN NUMBER W17 MATERIAL : WELD CAPSULE : PRAIRIE ISLAND Pen!RIE tsunc et LEO nals
} e 6 e i l
4 m
1 m
3 w
4
. s a a s-
.9 1.2 2. 4 3.6 4.8 6. 0 TIE ( MSEC >
PRAIRIE ISLAND #1 SPECIMEN NUMBER :W20 MATERIAL : WELD CAPSULE : PRAIRIE ISLAND Figure A-17. Load-time records for Specimens W17 and W20.
A-17
Specimen Alignment Error - Data not valid.
enatarc 2sure es ne anz n s i s e e
M 1
7 3?
=
q.
- n. e A A .. A A A a A A _AAnAA.& 3 . . n ._^_m m a _ m a I
. i e a i i
.3 s.: a.4 s.s 4.s s.o )
TIfE ( MSEC ) l PRAIRIE ISLAND #1 l SPECIMEN :H21 MATERIAL :HAZ CAPSULE : PRAIRIE ISLAND Figure A-18. Load-time records for Specimens H17 and H21.
A-18
MtAIRIC 13Ur0 01 Mt4 HAZ
} 6 e i 4 e, - -
w M
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a .- -
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.% 1.2 E.4 3.6 4.0 6.0 TIsE ( MEEC )
'JRAIRIE ISLAND #1 SPECIMEN NUMBER :H24 MATERIAL :EAZ CAPSULE : PRAIRIE ISLAND PRAIRIE ISUWO 81 W2 HAZ I i
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.9 1.2 a.4 3. 6 4.0 6. 0 TIPE ( FIEEC )
PRAIRIE ISLAND #1 SPECIMEN NUMBER :H22 MATERIAL :HAZ CAPSULE : PRAIRIE ISLAND Figure A-19. Load-time records for Specimens H24 and H22.
A-19
~ .- . _ . . - . _ . - - - . . _ - _ . . . - - . . . _ _ . . _ .-_ _ _
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, TIE ( MSEC >
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PRAIRIE ISLAND #1 SPECIMEN NUMBER :H23 MATERIAL :HAZ CAPSULE : PRAIRIE ISLAND Figure A-20. Load-time records for Specimens H20 and H23.
A-20
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PRAIRIE ISIAND #1 SPECIMEN NUMBER :H19 MATERIAL :HAZ CAPSULE
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l 5"
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l PRAIRIE ISIAND #1
! SPECIHEN NUMBER :HIS j MATE 7dAL :HAZ
! CAidULE : PRAIRIE ISLAND Figure A-21. Load-time records for Specimens H19 and H18. ,
A-21
. . _ . _ . .- - . - - . .. . . - ~ - .
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.9 1.E 2.4 3.6 4.0 6.0 l TIE < ft3EC ) l PRAIRIE ISLAND #1 l SPECIMEN NUMBER :R17 -
MATERIAL : CORR CAPSULE : PRAIRIE ISLAND Mta!RIC ISUwe et Ret asut
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.9 1.3 2.4 3.6 4.0 6. 0 TIE ( ftSEC 3 PRAIRIE ISLAND #1 SPECIMEN NUMBER :R22 MATERIAL : CORR CAPSULE : PRAIRIE ISLAND Figure A-22. Load-time records for Specimens R17 and R22.
A-22
. - -_ . - - ~ . ~ , .. .-_ .=_- ._ . . . . . . . . _ . . ..
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.D 1.4 2.4 3.6 4.0 6. 0 TIE ( MSEC D PRAIRIE ISLAND #1 SPECIMEN NUMBER :R24 MATERIAL : CORR CAPSULE : PRAIRIE ISLAND ena:Ric sume en ese aust
- I 6 4 e 6 M
a s .- -
3 w
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e i a 6 a
.D 1.2 2.4 3.6 4.0 6. 0 7:MC ( MstC )
PRAIRIE ISLAND #1 SPECIMEN :R18 MATERIAL : CORR CAPSULE : PRAIRIE ISLAND Figure A-23. Load-time records for Specimens R24 and R18.
A-23
d Ati Ct3Wt MtAIRIC ISLarc et 6 6
- 6 6 e
- ~
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PRAIRIE ISLAND #1 SPECIMEN NUMBER :R21 MATERIAL : CORR CAPSULE : PRAIRIE ISLAND RSS CIgut PRA1RIC ISune el 6
} l a 6 9"
w M
1 w
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8 ad a a i 6 2.4 3.6 4.8 6.0
.3 1.8 TIE C MIEC 3 PRAIRIE ISLAND #1 SPECIMEN NUMBER :R23 MATERIAL : CORR CAPSULE : PRAIRIE ISLAND Figure A-24. Load-time records for Specimens R21 and R23.
A-24
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\
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u
~
i l
I
.3 :.2 L4 3.s 4.s s. o 71,E' ( MscC )
PRAIRIE ISLAND #1 SPECIMEN NUMBER :R19 MATERIAL : CORR CAPSULE : PRAIRIE ISLAND PRA1RIC 13UuS #1 R20 CIBut
- , i 6 4 8 e ,
i
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PRAIRIE ISLAND #1 SPECIMEN NUMBER :R20 l MATERIAL : CORR l CAPSULE : PRAIRIE ISLAND Figure A-25. Load-time records for Specimens R19 and R20.
A-25
4 6
B-0 APPENDIX B CHARPY V-NOTCH SHIFT RESULTS FOR EACH CAPSULE HAND-DRAWN VS. HYPERBOLIC TANGENT CURVE-FITTING METHOD (CVGRAPH VERSION 4.1) I r
l I
i ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGPAM
B-1 Table B 1 30 ft-lb Transition Temperature Shifts ( F) for Intermediate Shell Forging C (Tangential)
Hand-Fit Plots CVGRAPH Plots Capsule Unirradiated Irradiated AT, Unirradiated Irradiated AT, V -25 '13 38 -38.91 17.44 56.36 a
P -25 -5 20 -38.91
, 15.8 23.11 R -25 55 80 -38.91 56.93 95.85 S - - -
-38.91 62.55 101.46 i
Table B-2 50 ft-Ib Transition Temperature Shifts ( F) for Intermediate Shell Forging C (Tangential)
Hand-Fit Plots CVGRAPH Plots l
Capsule Unirradiated Irradiated Unitradiated ATw Irradiated AT.
V -5 34 39 -6.35 44.34 50.69 !
P -5 20 25 -6.35 16.92 23.27 R -5 90 95 -6.35 94.84 101.19 S - - -
-6.35 98.8 105.15 Table B-3 35-mil Lateral Expansion Temperature Shifts ( F) for Intermediate Shell Forging C (Tangential)
Hand-Fit Plots CVGRAPH Plots Capsule Unirradiated irradiated AT3 Unirradiated irradiated ATu V -25 38' 63 -24.28 47.82 72.11 P 25 7 32 -24.28 9.34 33.63 R 25 70 95 -24.28 80.86 105.15 S - - -
24.28 88.06 112.37
- Extracted from plot in WCAP-8916.
ANALYSIS OF CAPSL.c S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSr.L F?JTION SURVEILLANCE PROGRAM
1
, B-2 Table B-4 Upper Shelf Energy Shifts (ft-lb) for Intermediate Shell Forging C (Tangential)
Hand-Fit Plots CVGRAPH P'-ts Capsule Unirradiated Irradiated AE Unitradiated irradiated AE V 158 143 15 158 143 15 P 158 142 -16 158 142 16 R 158 145 -13 158 145 13 S - - - 158 142.5 15.5 Table B-5 30 ft-lb Transition Temperature Shifts (*F) for Intermediate Shell Forging C (Axial)
Hand-Fit Plots CVGRAPH Plots Capsule Unirradiated irradiated AT, Unirradiated Irradiated AT, V -27 -3 24 -31.31 7.24 24.07 P -27 10 37 -31.31 2.66 33.98 R -27 60 87 -31.31 52.87 84.18 i
S - - -
-31.31 42.95 74.27 Table B-6 50 ft-Ib Transition Temperature Shifts ( F) for Intermediate Shell Forging C (Axial) l
! Hand-Fit Plots CVGRAPH Plots Capsule Unirradiated irradiated AT, Unirradiated Irradiated AT, V 4 19 15 3.95 20.11 16.15 i i P 4 55 51 3.95 54.27 50.32 i
R 4 100 96 3.95 99.55 95.6 i S - - -
3.95 80.63 76.68 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATON SURVEILLANCE,Pr.OGRAM
B-3 Table B-7 35-mil 1.ateral Expansion Temperature Shifts (*F) for lidoiii-dl ate Shell Forging C (Axial)
Hand-Fit Plots CVGRAPH Plots Capsule Unirradiated Irradiated AT. Unirradiated irradiated AT.
V -12.5 17 29.5 13.05 18.92 31.97 P -12 28 40 -13.05 18.14 31.2 R 12 85 97 13.05 85.16 98.21 S - - -
13.05 75.02 88.07 Table B-8 Upper Shelf Energy Shifts (ft-ib) for Intermediate Shell Forging C (Axial)
Hand-Fit Plots CVGRAPH Plots Capsule Unirradiated Irradiated AE Unirradiated Irradiated AE V 143 155 12 143 155 12 P 143 136 -7 143 136 -7 i R l 143 129 14 143 129 -14 S -- - -
143 135 -8 Table B-9 30 ft-Ib Transition Temperature Shifts ( F) for the Surveillance Weld Material Capsule Hand-Fit Plots CVGRAPH Plots Unitradiated Irradiated AT. Unitradiated Irradiated AT.
V -57 32 25 -64.44 -30.05 34.38 P -57 -15 42 -64.44 19.28 45.15 R -57 60 117 -64.44 58.02 122.47 S - - -
-64.44 95.98 160.43 ANALYSIS OF CAPSULE S FROM THE NORTHliRN STATES POWER COMPANY PRAIRIE ISt.AND Unit 1 REACTOR VESSEL RADIATION SUR'/EILLANGi PROGRAM
. ..__. _. _ _ _ _ _ _ . . _ _ _ _ ~ . _ _ . . _ _ _ _ _ . _
i f B-4 t
j Table B-10 50 ft-Ib Transition Temperature Shifts (*F) for the Surveillance Weld Material Hand-Fit Plots CVGRAPH Pbts i
Cgsk Unirradiated irradiated AT, Unirradiated irradiated AT.
V 14 18 32 -26.93 20.42 47.35 P 14 46 60 -26.93 45.01 71.94 i
R 14 115 129 -26.93 134.95 161.88 S - - -
-26.93 143.91 170.84 l
i l Table B-11 35-mil Lateral Expansion Temperature Shifts ( F) for the Surveillance Weld Material i
j Hand-Fit Plots CVGRAPH Plots I
{ Capsule Unirradiated Irradiated AT. Unitradiated Irradiated AT, i
i
- V -45 14 59 -50.79 22.58 73.38
( P -45 25 70 -50.79 25.55 76.34
! R -45 90 135 -50.79 117.04 167.83 S - - -
-50.79 132.74 183.53 i,
1 Table B 12 Upper Shelf Energy Shifts (ft-D) for the Surveillance Weld Material Hand-Fit Plots CVGRAPH Plots Capsule Unirradiated irradiated AE Unirradiated irradiated AE V 78.5 91 12.5 78.5 91 12.5-P 78.5 83 4.5 78.5 83 4.5 R 78.5 75 -3.5 78.5 75 -3.5 5 - - -
78.5 84.5 6 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
B-5 Table B-13 30 ft-Ib Transition Temperature Shifts ( F) for the Weld Heat Affected-Zone (HAZ)
Material Hand-Fit Plots CVGRAPH Plots Capsule Unirradiated Irradiated AT, Unirradiated* irradiated AT, V -200 -200 0 -260.00 -200.00* O P -200 130 70 200.00 125.35 74.65 R -200 -60 140 -200.00 -50.31 149.69 S - - -
-200.00 -62.89 137.11 Table B-14 50 ft-lb Transition Temperature Shifts ('F) for the Weld Heat-Affected-Zone (HAZ)
Material Hand-Fit Plots CVGRAPH Plots Capsule Unirradiated irradiated AT, Unirradiated* 1rradiated AT, V -125 -125 0 -125.00 -125.00* 0 P 170 105 65 125.00 -88.8 36.20 i R -170 -5 165 -125.00 -21.13 103.87 S - - -
125.00 -26.8 98.20 Table B 15 35-mil Lateral Expansion Temperature Shifts ( F) for the Weld Heat-Affected-Zone (HAZ)
Material Hand-Fit Plots CVGRAPH Plots Capsule Unitradiated Irradiated AT, Unirradiated* Irradiated AT, i V 15'. < -128 24 152.00 128.00 24.00 P 175 -65 110 152.00 -51.24 100.76 R 175 0 175 152.00 11.77 140.23 i
S - - -
152.00 7.49 144.51 Because the hyperbolic tangent curve fitting process did not provide a smooth S-shaped curve for the unirradiated and Capsule V data, these values have been retained from the original Charpy V notch hand fit curves documented in WCAP-8086 and WCAP-8916.
l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPAN [4AIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM i
l
- i. . . B-6 l
l Table B-16 Upper Sheff Energy Shifts (ft-lb) for the Weld Heat-Affected-Zone (HAZ) Material Hand-Fit Plots CVGRAPH Plots Capsule Unirradiated Irradiated AE Unirradiated irradiated AE V 211 211 P 211 143 -68 211 143 -68 R 211 97 114 211 97 114 i
! S - - -
211 136 -75
- Upper shelf impact energy not obtainable due to excessive toughness.
Table B-17 30 ft-lb Transition Temperature Shifts ( F) for the Correlation Monitor HSST Plate 02 Hand-Fit Plots CVGRAPH Plots Capsule Unirradiated irradiated AT, Unitradiated irradiated AT, V 49 159 110 46 2 149.05 102.84 P 49 205 156 46.2 207.61 161.4 R 49 235 186 46.2 239.93 193.72 S - - -
46 2 212.29 166.08 Table B-18 50 ft-lb Transition Temperature Shifts (*F) for the Correlation Monitor HSST Plate 02 Hand-Fit Plots CVGRAPH Plots Capsule Unirradiated Irradiated AT, Unitradated Irradiated AT, V 81 194 113 78.39 194.65 116.25 P 81 232 151 78.39 228.16 149.76 R 81 285 204 78.39 280.48 202.08 S - - -
78.39 237.98 159.58 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
B-7 Table B 19 35-mil Lateral Expansion Temperature Shifts (*F) forthe Correlation Monitor HSST Plate 02 Hand-Fit Plots CVGRAPH Plots Capsule Unirradiated irradiated AT, Unirradiated Irradiated AT.
V 53 88 35 58.63 192.96 134.33 P 53 218 165 58.63 217.38 158.74 R 53 280 227 58.63 299.54 240.9 58.63 238.47 179.82 Table B-20 Upper Shelf Energy Shifts (ft-Ib) for the Correlation Monitor HSST Plate 02 Hand-Fit Plots CVGRAPH Plots Capsule Unirradiated irradiated AE Unitradiated Irradiated AE V 123.5 91 -32.5 123.5 91 -32.5 P 123.5 85 -38.5 123.5 85 -38.5 R 123.5 86 -37.5 123.5 86 -37.5 S - - -
123.5 B2.5 -41 l
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-0 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING HYPERBOLIC TANGENT CURVE-FITTING METHOD l
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
. C-1 CVCRAPH 43 Hyperbolic Tangent Curve Printed at 1547f)3 on 10-28-1996 l Page1 ,
Coefficients of Curve 1 l A = 80.09 B = 7/.9 C = 9L49 l TO = 30.93 Equation is CVN : A + B ' [ tanh((T - TO)/C) )
Upper Shelf Energy: 158 Fixed Temp. at 30 ft-lbs -3&9 Temp. at 50 ft-lbs -63 lower Shelf Energy 119 Fixed Material: FORGING SA5083 Heat Number. 21918/38566 Orientation LT Capsule UNIPR Total Fluence 300 cn 250
,C I
a ,
x am I l
o m 1 4
g 150 B_ -
g
'g 0 A z l
> 8 O O So 0
o
- s o
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Setis) Plotted Plant: Pil Cap; UNIRR Material FORGING SA5083 Ori: LT Heat f: 21918/38566 Charpy V-Notch Data Temperature Input CVN Energy Computed CVN Energy Differential
-60 9 20 5 -11W
-60 29 20W &O2
-60 11 5 20W -9.47
-30 32 34.73 - 173
-30 16 3t73 -1&73
-30 28 34 73 -6.73
-10 635 47 3 161
-10 56 47 5 BS
-10 70 47 3 22S
- Data continued on next page "
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-2
~
Page2 Material fDMG SA5083 Ileat Numben 21918/38566 Orientation LT Capsule UNIP.R Total fluence Charpy V-Notch Data (Continued)
Temperature input CVN Energy Computed CVN Energy Differential 40 89 87.79 12 40 84 87.79 -179' 40 86 87.79 -1.79 80 1(Tl 11828 -1128 4
80 115 5 11828 -278 80 111 11828 -728 150 159 14725 11.7 4 3
150 164 14725 16.7 4 210 15 0 1542 -4E 210 153 1542 -12
, 210 153 1542 -L95 i 300 16 0 15756 243 300 155 15756 -256 300 174 15756 16.43 SUM of PEIDUAIS = 15.9 4
s s
1 a
4 A
4 800 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-3 CVGRAPH 41 Hyperbolic Tangent Curve Printed at 155412 on 14-28-1996 Page1 Cbefficients of Curve 1 l A = 4815 8=4715 C = 6531
! TO = -5.62 Equation is: E = A + B * [ tanh((T - TV)/C) i Upper Shelf 2: 953 Temperature at E 3fx -243 lower Shelf II: 1 Fixed Material: IMGING SA5083 Heat Number. 21918/38566 Orientation: LT Capsule UNIRR Total Fluence 200 1
m '
- 150
% l 4 X l N 100 0 o l
~
- U e
y o s.
O s o d
a so -
0 0
a o j
-300 -200 -100 O 100 200 300 400 s00 600 Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap: UNIRR Materiah R)RGING SA5083 Ori: LT Heat l: 21918/38566 Charpy V-Notch Data Temperature Input lateral Erpansion Computed 2 Differential
-60 11
-60 16 -5 29 16 12.99
-60 10
-30 16 -6 27 3133
-30 -433 14 3123
-30 -1733 23 3133
-10 -833 56 44 2 11
-10 47 44.99 2
-10 63 44.99 18 l
" Data continued on next page =
I ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-4 Page2 Materiat MRCING SA5083 Heat Number: 21918/38566 Orientation: LT Capsule UhM Total Muence:
Charpy V-Notch Data (Continued)
Temperature Input lateral Expansion Computed LE Differential 40 76 764 -S 40 71 764 -53 40 74 76S -2S 80 81
! 8 & 91 -7.91 80 88 8 & 91 -31 80 87 8 & 91 -131 150 102 9451 7.48 15 0 101 9451 &48 210 210 99 E7 382 210 98 E7 2S2 300 95 E7 .17 98 9529 17 300 90 9529 -529 300 89 9529 -629 l SUM of PEIDUAIS = -5IE 4
l l
~
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM .
. I C-5 CVGPJLPH 4J Hyperbolic Tangent Curve Printed at 15555 on 10-28-1996 Page1 Coefficients of Curve 1 A = 50 B = 50 C = 93.78 TO = 35.9 Equation is: Shearx = A + B ' I tanh((T - 1D)/C) l Temperature at 50x Shear. 35.9 Material: FDRGING SA5083 Heat Numbec 21918/38566 Orientation: LT Capsule UMRR Total Fluence 100 2 M
g u r cd e
o n
c e
O k 40 0 a
% o h i o
1 2
o
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant P!1 Cap: UNIRR Material: FORGING SA5083 Ori: LT Heat l: 21918/38566 Charpy V-Notch Data Temperature Input Percent Shear Computed Pertent Shear Diffenstial l 1
-60 9 '
11.4 5 -?.45
-60 18 1L45 654
-60 !
12 IL45 54 !
-30 25 19 5 53 .
-30 13 19 5 -6S9 I
-30 18 19 S 9 -1S9
-10 34 3 31 6S8
-10 25 2731 -2 31
-10 30 M31 2S8
" Data continued on next page "
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C
_ -6 Page2 Material: FDRGING SA5083
! Ileat Number: 21918/38566 Orier.tation LT 3 Capsule UNIRR Total Fluence
)
Charpy V-Notch Data (Continued)
Temperature Input Percent Shear Computed Percent Shear
! 40 Differential 63 5118 1061 40 45 5218 -718 s 40 45 5118 i -718 80 79 71S1 71)B
! 80 67 i 71S1 -4S1 !
! 80 61 7131 -1031 150 100 4 9L93 &O6 15 0 100 4 9133 &O6 210 100 3 WS1 238 210 100 WS1 238 210 4
300 10 0 97El 22 100 1
300 9934 3 1 300 100 9964 2 100 99S4 5 SUM of RISIDUAIS = 20S6 i
l 4
4 i
)
1 9
i e
a
'e ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
t
_. C-7 UNIRRADIATED 1 '
CVGRAPH 41 Hyperbolic Tangent Curve Printed at 1638f77 on 10-23-1996 Page1 Coefficients of Curve 1 A = 7259 B = 70.4 C = 9173 TD : 3531 !
l Fquation is CVN : A + B
- 1 tanh((T - E)/C) 1 !
Upper Shelf Energy 143 Fixed Temp at 30 ft-lbs: -313 Temp. at 50 ft-lbs: 3.9 lower Shelf Energy: 239 Fixed Material FORGING SA5083 Heat Number. 21918/38566 Orientation: TL Caprule UNIRR Total Fluence 300 1
l ro 250 Q
., T a
x 2$
- N tw L 150 p h .
m of 100 2; l
- > c !
O n su '
s O u
-300 -200 -100 ' O 100 200 300 400 500 600 Temperature in Degrees F Data Setis) Plotted Plant: P11 Cap.: UNIRR Material: RRGING SA5083 Ori; TL Heat f. 21918/38568 Charpy V-Notch Data Temperature Input CVN Energy Computed CVN Energy Differential
-a) 65 18.96 -1246
-60 12 18.96 -6.96
-60 65 18.96 -1246
. -30 32 30f1 138
-30 53 30El 2138
-30 10 3031 -20S1
-10 41 4126 -26
-10 43 412 6 L73
-10 40 4126 -126
- Data continued on next page "
<se ANALYSIS OF CAPSOLE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-8 UNIRRADIATED Page2 Material: FORCING SA5083 Heat Number. 21918/38566 Orientation: TL Capsule UNIP3 Total Fluence Charpy V-Notch Data (Continued)
Temperature input CVN Energy Computed CVN Energy Differential 10 69 541)6 14 S 3 10 53 54.06 -LO6 l 10 705 541)6 l 16.43 40 76 7537 32 40 66 5 !
75Fl -917 l 80 110 10296 71)3 l 80 905 10296 30
-1246 ST/ 10296 -5.96 210 141 13939 L6 210 1345 13939 -4B9 210 146 13939 64 300 14 9 14243 656 300 148 14243 556 3)0 140 14243 -2 43 SUM of RESIDUAIS =-5.49 4
I i
l l
l l
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM -
l
C-9 UNIRRADIATED -
CVGRAPH 41 Hyperbolic Tangent Curve Printed at 16#33 or.10-23-1996 Page1 itefficients of Curve 1 A = 485 B = 475 C = 79,91 10 = 1031 Equation is E = A + B ' l tanh((T - TD)/C) J Upper Shelf II: 96.01 Temperature at E .Tr -13 lower Shelf LE: 1 Fired Material IDRGING SA5083 Heat Number. 21918/38566 Orientation TL Capsule: UNIPS Total Fluence 200 b 150 6
a X
E4 100 . .,
e r "
4 5
8 a P A 50 0 Pi 0
- a U l j j
-300 -200 -100 0 100 200 300 400 500 600 !
Temperature in Degrees F Data Set (s) Plotted Plant: PIl Cap: UNIP2 Material fDRGING SA5083 Ori: TL Heat l: 21918/38566 Charpy V-Notch Data Temperature input lateral Expansion Computed II Differential
-60 7 14.95 -7.95
-60 13 14.95 -195
-60 7 14.95 -7.95
-30 26 2638 -38
-30 44 2638 17El
-30 13 2638 -1338
-10 35 36E8 -1E8
-10 37 36SB 31
-10 35 36S8 -138
Data continued on nert page
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-10 UNIRRADIATED Page2 Material: }MCING SA5083 Ileat Number: 21918/38566 Orientation TL Capule UhitR Total 71uence Charpy V-Notch Data (Continued)
Temperatun input lateral Expanson Computed I.E Differential 10 10 61 42 12M 45 4832 10 -332 40 59 4G 10M 66 6533 40 El 56 6538 80 82
-9J8 81M .12 80 76 81M -5M 80 79 81M -2R 210 97 9538 161 210 94 9538 -138 210 96 9538 El 300 96 95.95 D4 300 97 9525 1.04 300 98 95S5 204 SUM of RISIDUAIS :-10.45 4
e ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-11 UNIRRADIATED CVCRAPH 41 Hyperbolic Tangent Curve Printed at 162046 on 10-23-1996 Page!
Coefficients of Curve 1 A : 50 B = 50 C = 11453 70:4731 Equation is Shearz : A + B ' l tanh((T - 11))/C) l Temperature at 50x Shear: 473 i Materiah NRCING SA5083 Heat Number 21918/38566 Orientation TL Capsule UMRR Total Fluence 10 0 --
a 8u
/
CC e (
$ su d
4 a a c
e e o A
h# o
~
e au I
o j ,
l
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: PIl Cap; UMRR Materiah MRGING SA5083 Ori: TL Heat h 21918/38566 Charpy V-Notch Data Temperature Input Perant Shear Computed Permat Shear Differential
-60 5 132 -82
-60 9 13 2 -42
-60 7 132 -62
-30 18 20.44 -244
-30 30 20.44 955
-T 17 EH -14
-j0 31 E7 429
-10 27 E7 29
-10 29 E7 229
" Data continued on next page =
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
l
). 6 C-12 UNIRRADIATED i
, Page2 ,
Waterial: NRCING SA5083 lieat Number 21918/38566 Orientation: TL Capsule UNIRR Total Fluence Charpy V-Notch Data (Continued)
Temperature input Percent Shear Computed Percent Shear 10 Differential -
i 43 34D6 8.93 10 37 34D6 293 10 43 34D6 893 40 43 46 5 -359 40 40 fl0 62 4659 -63 63 5 -15 80 55 63.89 '
80
-a69 61 210 63 5 -25 100 210 100 94.43 52 94.43 556 210 100 94.43 536 300 100 9E79 12 300 100 9E79 12 300 100 9E79 12 SUM of RESIDUAIS = 9.75 I 4
. l 1
h
= l I
l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISt.AND Unit 1 REACTOR VESSEL RADIATION SURVEtLLANCE PROGRAM
. 1 l
. C-13 UNIRRADIATED CVCPJLPH 41 Hyperbolic Tangent Curve Printed at 1622:20 on 10-23-1996
- Page!
Coefficients of Curve i A = 4034 B = 38.15 C = 6938 % = -45 Equation is CVN : A + B ' [ tanh((T - %)/C) ]
Upper Shelf Energy: 785 Fixed Temp. at 30 ft-lbs -64.4 Temp. at 50 ft-lbs -269 lower Shelf Energy: 2.19 Fired Materiah WELD Heat Number: 1752 Orientation:
Capsule UNIRR Total Fluenz 300 m 25o
,O .
I a .
% 200 .
' N u
4 150 a>
c .
i:e 100 o a
Z o
- > 5 O
50 Jh --
o a ;
Ml 0
-300 -200 -100 .0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: PIl Cap.: UNIRR Materiah TELD Ori: Heat f.1752 Charpy V-Notch Data Temperature input CVN Energy Computed CVN Energy Differential ;
-100 8 1529 -729
-100 6 1529 -929
-100 255 1529 10 2
-50 51 37E2 1337 1
-50 27 37E2 -1032
-50 32 3732 -5E2
-30 55 4&41 658
-30 64 4&41 1550
-10 41 58 -17
" Data continued on next page "
M ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISt.AND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM -
a .
C-14 UNIRRADIATED j Page2 I
Material TELD Heat Number: 1752 Orientation-Capsule UNIRR Total hence
?
Charpy V-Notch Data (Continued)
Temperature Input CVN Energy Computed CVN Energy Differential
- 10 62 58 199
-10 615 58 149 10 52 654 -114 10 75 65.4 959 10 50 65.4 -15.4 40 77 7234 4E5 40 84 7234 11E5 40 61 7234 -1134 80 85 M42 857 80 95 76.42 1857 80 62 M42 -14.4 2 210 70 78.44 -&44 210 87 M44 855
. 2W 78 M44 .44 SL'M of RISIDUAIS = 153 1 4 .
]
l l
i I
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-15 UNIRRADIATED CVCRAPH 41 Hyperbolic Tangent Curve Pcinted at 162351 on 10-23-1996 Page1 Coefficients of Curve 1 A = 3&7 B = 37.7 C = 7118 ID = -4171 Equation is: E = A + B ' [ tanh((T - 11))/C) l Upper Shelf II: 76.41 Temperature at II 35: -503 laer Shelf LE: 1 Fixed Material WELD Heat Number.1752 Orientation:
Capsule UNIRR Total Fluence:
200 co O 150 a .
4 X 100 O
D o m
's m b
a u c
O d o ll
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap; UNIRR Matenah WELD Ori: Heat l: 1752 Charpy V-Notch Data Temperature input lateral Expansion Computed II -
Differential
-100 7 1433 -72
-100 6 14 2 -&33
-100 23 1433 836
-50 48 3547 1252
-50 26 3147 -9.47
-50 30 3i47 -E47
-30 53 45S9 73
-30 60 45S9 14 3
-10 43 54S4 -IL94
- = Data continued on next page =
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURtEILLANCE PROGRAM
C-16 UNIRRADIATED Page2 Material: WELD Heat Number.1752 Orientation:
Capsule UNIPJt Total Fluence Charpy V-Notch Data (Continued)
Temperature Input lateral Expansion Computed LF. Differential
-10 56 54 S4 1.05
-10 57 54 S4 2.05 10 50 6229 -1229 10 72 6229 9.7 10 52 6229
- -1029 40 71 69.46 153 40 78 69.46 853 40 55 69.46 -14.46 80 82 7333 8.06 80 92 7333 18116 80 65 7333 -8.93 210 69 7634 -734 210 78 7634 If4 210 75 7634 -134 SUM of PEDUAIS = -38 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-17 d
UNIRRADIATED CVGRAPH 4J Hyperbolic Tangent Curve Printed at 162x11 on 10-23-1996
! Pagei Coefficients of Curve i A = 50 B = 50 0 = 949 TO=-2?.03 Equation is Shear /. = A + B ' l tanb((T - 10)/C) l Temperature at 50x Shear: -22 i
Material: WELD Heat Number.1752 Orientation:
Capsule UNIPR Total Fluence 8 f C
a $
d a 0 o d
CO g n M
4 C ce O n b 40 e
20 t{
c)
C)
U j , j
-300 -200 -100 - 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: PI! Cap: UNIPR Materiat NELD Ori: Heat f.1752 Charpy V-Notch Data Tempergure input Percent Shear Computed Perant Shear Differential
-100 9 16 2 -72
-100 5 16 2 -11 2
-100 17 16 2 29
-50 55 35S7 1932
-50 33 35E7 -2.67
-50 29
-30 35E7 -6f7 53 4531 738
-30 61 4531 1538
-10 41 563 -15 3
= Data continued on next page "
as ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
1 ,
C-18 UNIRRADIATED Page2 Material: FELD Heat Number 1752 Orientation-4 Capule LMRR Total Fluence Charpy V-Notch Data (Continued) i Temperature input Pettent Shear Computed Percent Shear
-10 Differential 3 63 563 6S9 a
-10 51 S&3 10 53 57 10 73 66 3 -93 66 3 &73 10 63 66 3 -326 40 77 7&7 -L7 40 82 7&7 329 40 68 78.7 - 10.7 80 93 89 2 143 80 !T1 89M 7.43 80 89 89 2 -2 210 100 99 3 .74 210 100
' 99 3 .74 210 100 99 3 .74 SUM of RESIDUAIS :-1M d
I 4
i j
i 4
i 1
4 1
I.
E i
1 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM -
4
.. C-19 UNIRRADIATED Best-Fit Curve as documented in WCAP-8916 I
(not fit with hyperbclic-tangent function)
Upper Shelf Energy: 211 Fired Temp. at 30 ft-lbs -200 Temp. at 50 ft-lbs -125 lower Shelf Energy-
, Wateriah HEAT AFFD ZONE Heat Number. Orientation:
Capsule UNIRR Total Fluence 300 i w 250 e
i .O 1*
A o
f
,e a
x 4
e 4 150 a f
1 D o i c i N o 4
100 a n 'o a
> 00 0 o _ ,
1 i o i
-300 -200 - 100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: P11 Cap; UNIRR Materiah HEAT AFFD ZONE Ori.: Heatb
,l Charpy V-Notch Data Temperature input CVN Energy Computed CVN Energy Differential
-200 21 45 2 -24 2
-200 18 45 2 -l/52
-200 41 45 2 -452
-17 5 7J 5291 2048
-17 5 22 52S1 -3031
-17 5 105 52S1 5248
- 15 0 48 6 1.11 -1311
-150 92 61.!! 3038
-lS) 73 Sill 1138
" Data continued on next page "
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-20 UNIRRADIATED Page2 Materiat ifEAT AFD ZONE IIeat Numben Orientation-Capsule UNIRR Total Fluence Charpy V-Notch Data (Continued)
Temperature Input O'N Energy Computed O'N Energy Differential
-100 91 79.7 1129
-100 11 4 N7 3429 '
-100 90 79.7 1029
-10 78 117 2 -392
-10 92 117 2 -252
-10 130 117 2 12.76 40 945 13753 -43.03 40 142 1J753 4.46 40 74 13753 -6353 80 154 15218 L81 80 145 15218 -718 80 2115 15218 5921 210 239 18535 53.04 210 189 185.95 104 210 205 185 S 5 19.04 SL'M of PSDUAIS = 4545 4 .
4 ase ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIR!E ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM I
C-21 UNIRRADIATED Best-Fit Curve as documented in WCAP-8916 (not fit with hyperbolic-tangent function) ,
l Upper Shelf E: 92 Temperature at M 3fx -152 !aer SE:lf LE:
Material: IIEAT WD ZONE Heat Number. Orientation:
Capsule UNIRR Total Fluence 200 en
.- 150 a
4 x 100 am
%- P M
.3 !!/
/ B o
o e o A W 60 uu t a C3 D
-300 -200 -100 0 100 200 300 400 500 600 4
Temperature in Degrees F Data Set (s) Plotted
. Plant: PIl Cap: UNIRR Materiah HEAT AFFD ZONE Ori: Heatb Charpy V-Notch Data Temperattae input lateral Erpansion Computed E Differential
-200 12 29W -17W
-200 18 29W -11W
-200 27 29W -2M
, -17 5 48 36fi5 1134
-17 5 18 36S5 -18S 5
'5 67 36S5 3034 4J 36 4171 -7.71
-15 0 56 4171 1228 i
-150 48 4171 428
" Data continued on next page "
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-22 UNIRRADIATED Page2 Materiah HEAT AITD ZONE Heat Number. tene .tation-Capsule LWRR Total Fluence Charpy V-Notch Data (Continued) i Temperature input lateral Erpansion Computed 11 Differential
-10 0 62 57.49 45
-100 67 57.49 95
-100 52 57.49 -149
-10 61 7523 -1423
-10 64 2 7523 -1123
' -10 77 Ta23 1.76 40 73 80.49 -7.49 40 95 -
80.49 14 5 40 62 80.49 -1&49 80 89 8103 5.96 80 93 8103 936 80 91
- 8103 7S6 210 90 8623 176 210 87 8623 .76 210 83
, 8623 -423 4
SUM of RESIDUAIS :-353 d
.l 4
e 4
3 j
e G
e no ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
^
C-23 l UNIRRADIATED Best-Fit Curve as riocumented in WCAP-8916 l l
a (not fit with hyperbolic-tangent function) 0 Temperature at 50x Shear: -110 l 4 Material: HEAT AFFD ZONE Heat Nurnber: Orientation:
Capsule UNIRR Total Fluence
! o 7
- \
a * ' )
e a o i j
4 o
~
l m 60 u 4 4 / o
-4 e a a
2 0 o O ll o D
- o
! u-1 b ((o D i ; j 1
- -300 -200 -100 o 100 200 300 400 500 600 i
Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap: UNIRR Material HEAT AFFD ZONE Ori:
I Heatb Charpy V-Notch Data Temperature Input Percent Shear Computed Perant Shear Differential
-200 17 25m -822 3 -200 17 2532 -832
-200 W 2532 Ll?
- 17 5 W 29.99 7
-175 18 29.99 -1139
-17 5 51 29.99 21
-15 0 W 3451 248
-150 45 3451 10.48
-150 W 3451 248
= Data continued on next page "
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCF PROGRAM -
C-24 UNIRRADIATED 1 Page2 l l' l Material HEAT AITD ZONE Heat Number: Orientation- l Capsule UNIRR Total fluence 1
Charpy V-Notch Data (Continued) l Temperatun input Percent Shear l Computed Perant Shear Differential j
' -100 42 44 2 -238
-100 63 44 2 1851
-100 43 44 2 -138
-10 47 67.74 -15.74 4 -10 47 E74 -1174 i -10 59 6?.74 -3.74 !
40 53 7133 -1&83 i
40 71 7131 -33 40 43 7133 -2833 80 90 7&D4 1L95 80 95 7&04 16.95 6
80 100 7804 21S5 210 100 9126 &73
. 210 130 9126 & 73 210 100 9126 &73 SUM of PElDUAIS : 2113 l
l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM .
C-25 UNIRRADIATED CVCRAPH 41 Hyperbolic Tangent Curve Printed at 165124 on 10-23-1996 Page1 Coefficients of Curve i A = 6254 B = 605 C = 8225 TO = 96D0 Equation is CVN = A + B ' l tanh((T - TO)/C) l Upper Shelf Energy:135 Fixed Temp. at 30 ft-lhe 462 Temp at 50 ft-lbs 33 !aer Shelf Energy: 219 Fixed Material SRM HSSm2 Heat Numben SA53381 Orientation: LT Capsule LMRR Total Fluence 300 rn eso C
I a
x am
. 4 x
e a 150 0
C N V ll o/
Z 8 a 0 W a
u w
-300 -200 -100 0 100 200 300 400 500 600 Temperature in. Degrees F Data Setis) Plotted Plant: Pil Cap: UNIRR Materiah SRM IE!D2 Ori: LT Heat l: SA533B1 Charpy V-Notch Data Temperature input CVN Energy Computed CVN Energy .
Differential
-50
-50 3 . 558 -22 5 52 -2
-50 5 558 -2
-20 6 9 -3
-20 65 9 -25
-20 9 9 0 10 13 5 1551 -201 10 12 1551 -351 10 14 5 1551 -LO!
" Data continued on next page "
9
+
m ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-26 UNIRRADIATED Page2 Material: SRM HS1112 Heat Number. SA533B1 Orientation LT Capsule UNIRR Total Euence Charpy V-Notch Data (Continued)
Temperature input OH Energy Computed GH Energy Differential 40 35 26 5 8.1 40 22 26 5 -45 40 36 26 5 91 85 52 5 4.71 - 2 71 85 585 54.71 3.78 85 415 54.71 -1321 110 635 73 -95 11 0 825 73 9.49 11 0 855 73 1249 160 109 10232 657 i 160 1085 10233 617 !
160 81 102.32 -2132 210 121 11634 465 210 11 7 11634 A5 210 115 11634 -134 300 !?/ 122S5 434 300 12 5 12265 234 l
, . 300 117 5 12265 .
-515
- i SUM of PSUA!S = -615 l
l 9
a=
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-27 UNIRRADIATED l CVGRAPH 4j flyperbolic Tangent Curve Printed at 165252 on 10-23-1996 l -
Pagel Coefficients of Curve 1 A = 43a5 B = 42S5 C = 84D8 IV = 7621 F4uation is E = A + B * [ tanh((T - IV)/C) )
Upper Shelf II: 8S31 Temperature at E 35c S&9 lower Shelf LE:1 Fixed Material: SRM llEV2 lieat Number: SA533B1 Orientation: LT Capsule Uh1RR Total Fluence:
200 l
1 m 1
- 150 a
X 100 i g 9L. ii .
b o s
e
,3 so
~
1 o l
l o
J l -300 -200 -100 - o too 200 aoo 400 soo soo Temperature in Degrees F Data Setts) Plotted Plant: P11 Cap: UNIRR Material:SRM IfEm Ori: LT IIeat f. SA533B1 Charpy V-Notch Data Temperature Input lateral Erpansion Computed E Differential
-50 4 5.03 -103
-50 3 103 -203
-50 5 5.03 -D3
, -20 9 884 J5
-20 6 884 -284
-20 10 834 IJ5 10 14 ISSI -131 10 15 15 S1 -31 10 14 15 S1 -131
- Data continued on next page "
e 1
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 RFACTOR VESSEL RADIATION SURVEILLANCE PROGRAM l _
l
i ,
C-28 I '
I UNIRRADIATED Page2 Material SRM flSSim Heat Number: SA533B1 Orientation LT Capsule UNIRR Total Fluence Charpy V-Notch Data (Continued)
Temperature input lateral Expansion Computal 1.E Differential 40 32 40 2631 52 23 2631 -331 40 32 85 2631 52 85 45 48.06 -32 51 4 &06
' 223 85 42 4&D6 -6.06 110 54 593 -59 110 60 593 M 110 71 593 ILO9 16 0 79 76M 2.92 160 72 76M -4M 16 0 89 *MM -7M 210 87 f291 42 210 84 82S1 12 210 88 8291 52 -
300 84 853 -1S 300 87 859 LO9 4 300 83 853 -29 SUM of RESIDUAIS -103 l
1 G.s ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-29 UNIRRADIATED CVGRAPH 4J Hyperbolic Tangent Curve Printed at 1654f)3 on 10-23-1996 Page1 Coefficients of Curve 1 A = 50 B = 50 C = 10039 E = 8554 Equation is Shearx = A + B '[ tanh((T - E)/C) l Temperature at 50x Shear. 855 Material: SRM HSSm2 Heat Number: SA53381 Orientation: LT Capsule: UNIRR Total Fluence: '
100 c i u ~ V
. C5
.)w a a,
4 g ;
. < 1 c) o h \
5
- a o ),
O
- l C
0 s
u l -300 -200 -100 0 100 200 300 400 500 600 j Temperature in Degrees F Data Set (s) Plotted
- Plant
- Pl! Cap: UNIRR Materiah SRM HSSm2 Ori: LT HeatbSA533B1 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential
-50 9 m 2E2
-50 9 m 2B2 ;
-60
-20 9 m 222 l
13 10.96 2h1
-20 9 10.98 -136
-20 13 10.98 2 01 l 10 23 12 4.72 10 23 18 27 4.72 10 23
)
18 71 4.72 .
- Data continued on next page "
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
1 c 30 UNIRRADIATED Page2 !
Material SRM HS5702 Heat Number. SA533B1 Orientation LT Capsule UNIRR Total Muence Charpy V-Notch Data (Continued) j Temperature input Penent Shear Computed Percent Shear 40 Differential 29 2884 40 35 33 2834 1
40 415 i 29 2884 85 J5 42 49.72 -7.72
, if> 43 49.72
~
85 4 72 41 49.72 4 4 72 110 55 6L88 -638 11 0 58 I 6138 -338 11 0 67 6128 5.11 160 87 8139 5S 16 0 84
' 81J9 2S 16 0 85 8139 3S 210 100 9P.17 732 210 98 4 9P.17 532 210 98 4
300 9P.17 52
' 100 9853 L4 300 100 . 9859 L4
, 300 100 9859 .
L4 i SUM of RESIDUAIS = 'Ji18 i
l t
i -
k
'l I
4
~
ANALYSIS OF CAPSUt.E S FROM THE NORTHERN STATES POWER COMPANY PRAIR!E ISLAND Unit 1 REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM
i C-31 CAPSULE V D' GRAPH 4J Hyperbolic Tangent Curve Printed at 17f)352 on 10-23-1996 Page1 Coefficients of Curve 1 A = 7259 B : 70.4 C = 73 TD = 68E3 Equation is 03 : A + B ' I tanh((T - 11))/C) ]
Upper Shelf Energy:143 Fixed Temp at 30 ft-lbs 17.4 Temp. at 50 ft-lbs 443 Imer Shelf Energy: 219 Fixed Material R)RGING SA5083 Heat Number. 21918/38566 Orientation: LT Capsule V Total Fluenz 300 rn 259
,o.
aI N am X
t2 4 150 a _
- e c 0 cr.a 100 Z 0 U -
o 50 0
O O
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: P11 Cap: V Materiat FORGING SA5083 Ori: LT Heat f. 21918/38566 Charpy V-Notch Data ,
Temperature Input OH Energy Computed O'N Energy Differential
-25 13 1225 .74 O 53 I
2010 3 116 1 0 19 2033 -13A !
25 10 349 -24 3 50 305 55 -24.5 50 785 55 2149 75 78 7&72 .72
- 13) 124 5 115 3 919 150 1265 12931 -231
'" Data continued on next page =
ANALYSIC DF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIR!E ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-32 CAPSULE V Page2 Material FORGING SA5083 lieat Number.21918/30566 Orientation: LT i Cap mle V Total Fluence:
1 Charpy V-Notch Data (Continued) l l Temperature Input O'N Energy Computed O'N Energy Differential !
210 140 14033 -13 300 150 14?.75 724 300 140 14?.75 -2.75 SUM of RIS!DUAIS : 1517 1
1 1
l
~
j i 4
6 d
e l
i H
i l
l j
f l -
, 1 4
8 I O
e
.1 m.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
.. .- _ .-- ~-_
C-33 CAPSULE V CVGRAPH 41 Hyperbolic Tangent Curve Printed at 1722 on 10-23-1996 Page1 Coefficients of Curve 1 A = 4171 B = 4171 C = 7959 E = 6421 Equation is E = A + B ' l tanh((T - TD)/C) l Upper Shelf LE: 8E43 Temperature at E 31 47.7 laer Shelf LE:1 Fired Materiat NRCING SA5083 Heat Number. 21918/38566 Orientation: LT Capsule V Total Fluence:
200 m
= m ,
5 n
M 100 ku o a
- t O
a e o /
A Su
\
o o
o d o 4 ,
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap:V Materiat mRGING SA5083 Ori: LT Heat f. 21918/38566 Charpy V-Notch Data Temperature input lateral Erpansion Computed E Differential
-25 6 92 -32 0 345 1518 1931 0 14 5 1518 -2 25 75 2422 -1&72 50 25 3G16 -1116 50 56 3636 1933 75 40 49.46 -9.46 12 0 745 6955 4.94 150 81 7756 143
" Data continued on next page "
4 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RAD:ATION SURVEILLANCE PROGRAM -
C-34
~
CAPSULE V l Page2 l
Material FORCING SA5083 Ileat Number: 21918/38566 Orientatiotr LT l Capsule V Total Fluence Charpy V-Notch Data (Continued) i Temperature input lateral Expansion Computed 11.
210 Differential 84 8429 300
-29 81 862 l 300
-52 88 862 1.79 SUM of RISDUAIS : 254 l
i l
l l
l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAlRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
G C-35 CAPSULE V CVCRAPH 4J Hyperbolic Tangent Curve Printed at 17ff46 on 10-23-1996 Pagei Coefficients of Curve 1 A = 50 B = 50 C = 95.71 11) = 9032 Equation is Shearx = A + B ' I tanh((T - TO)/C) }
Temperature at 50x Shean 903 j
Material FORGING SA5083 Heat Number: 21918/3566 '
Orientation LT !
Capsule V Total Fluence i 100 l
% OU
/
CO O a
.C 4 4 o a
1 D
c l 40 h 1 1
4 t
- J '
/
0
-300 -200
)
-100 0 100 200 300 400 500 600 Temperature in. Degrees F Plant: PIl Cap:V Material R)RGINGOri:)SA5083 LT Heat f.Data Set (s Plotted 21318/38566 Charpy V-Notch Data Temperature Input Permnt Shear Computed Percent Shear - Differential
-25 5 . 816 - 116 0 10 1103 -103 0 10 1103 -103 25 20 2017 -17 m 2 52 2 50 30 29 2 11 75 53 4131 111 8 120 60 64.78 -4.78 150 70 71.4 9 -7.49
- Data continued on next page "
GBu ANALYSIS OF CAPSULE S FROM THE NOR1HERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-36
- CAPSULE V Page2' Material MRGING SA5083 Heat Number
- 21918/38566 Orientation: LT Capsule V Total Roenee 4
Charpy V-Notch Data (Continued)
Temperature Input Pen:ent Shear Computed Pen:ent Shear t 210 Differential 100 9234 7E5 300 100 9&75 124 300 100 9&75 124 l SUM of RESIDUAIS = -12
't I
2 1
1 4
4 2
E 4
)
l 4 .
1 i
h J
i 1
1 J
1
? ~
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY P.RAIRIE ISLAND Unit 1 j REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 1
- C-37 CAPSULE V l CVGRAPH 43 Hyperbolic Tangent Curve Printed at 1797:20 on 10-23-1996 l
Page1 Coefficients of Curve 1 l
t A = 7859 B = 76.4 C = 7636 1V = 5015 l I
Equation is CVN = A + B ' [ tanh((T - 1D)/C) l Upper Shelf Energy: 155 Fired Temp at 30 ft-lbs -72 Temp. at 50 ft-lbs 201 lower Shelf Energy: 2.19 Fixed Material IDEING SA5083 Heat Number. 21918/38566 Orientation TI, Capsule V Total Fluence 300 ,
l m 250
,C I l a l l g 200 j
- x u _
o o i L 150
, O o d -
m /
100 0
~
(
o, 4
-300 -200 -100 ' O 100 200 300 400 500 600
- Temperature in Degrees F l Data Set (s) Plotted i Plant: PIl Cap: Y Material: FDEING SA5083 Ori: TL Heat f. 21918/38566 Charpy V-Notch Data Temperatum loput CVN Energy Computed CVN Energy Differential
-50 10 !?.53 -253
-25 10 5 20 2 -10.42 0 17 5 3457 -17D7
. 0 30 3457 -457 25 545 543 19 25 85 543 30E9 50 845 7E44 6D5 75 895 102S1 -1311 150 1325 14458 -1248
't
"" Data a>ntinued on acrt page ""
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1
~
REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
- i A
)
c.38 1 s
CAPSULE V Page2 Material FORGING SA5083 Heat Number. 21918/38566 Orientation: TL Capsule: V Total Fluena:
Charpy V-Notch Data (Continued)
Temperature input CVN Energy Computed CVN Energy Differential ,
210 168 15171 1528 !
, 250 166 15418 1131 300 154 154.78 .78 SL1 of RISIDUAIS : 145 i
t n
. < 1 1
l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILIANCE PROGRAM -
,,. C-39 CAPSULE V CVCRAPH 41 Hyperbolic Tangent Curve Printed at 17fE34 on 10-23-1996
. Page1 Coefficients of Curve 1 A = 40.43 B = 39.43 C = 46.1 1D = 2531 Equation is E = A + B ' [ tanh((T - 1D)/C) j Upper Shelf II: 79JI6 Temperature at 11. 3fx 18 9 Imer Shelf LE:1 Fired Material: FDRCING SA5083 Heat Number: 21918/38566 Orientation TI, Capsule V Total Fluence:
20u m
- = 150 l
M e i 1
a x
M ~ 100 o M O {'
<- u
. 3 a o ce !
A Eu !
I o
u g
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Deg.rees F Data Set (s) Plotted Plant: PIl Cap:V Material: FDRGING SA5083 Ori: TL Heat f. 21918/38566 Charpy V-Notch Data Temperature input lateral Erpansion Computed E Differential
-50 55 3fl9 15
-25 4 899 -439 0 95 20.72 - 11 2 0 21 20.72 3 25 355 4036 -4fl6 25 615 40J6 2133 50 555 59.73 -423 75 625 71E7 -937 15 0 6"I 7951 -1251
= Data continued on next page "
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILIANCE PROGRAM
. C-40 CAPSULE V Page2 i
Materfah IURGING SA5083 IIeat Number. 21918/38566 Orientation TL Capsule V Total menz Charpy V-Notch Data (Continued)
Temperature input lateral Expansion Computed II Differential 210 865 7923 6fi6 250 94 5 7936 14E3 300 78 7936 -LB6 SUM of RISIDUAIS :-435 4
l O
e e
Se ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADLATION SURVEILLANCE PROGRAM
_ . _ - _ .- -- . . - = - . _.
J C-41 CAPSULE V' CVCRAPH 41 Hyperbolic Tangent Curve Printed at 17fR47 on 10-23-1996 Page1 l Coefficients of Curve 1 A = 50 B = 50 C = 71.11 1D = 7537 l
Equation is Shearx = A + B ' I tanh((T - 1D)/C) }
Temperature at 50x Shear. 753 Materiah IVRGING SA5083 Heat Number: 21918/38566 Orientation: TL Capsule V Total Fluence 100 2
M a 80 c5 0
C .
m ,
4 s .
c e
O b ** ~
l 4 o !
20 -
U '
-300 -200 -100 o too 200 300 400 soo soo l Temperature in Degrees F Data Set (s) Plotted i Plant: P11 Cap V Materiah IDRGING SA5083 Ori.: TL Heat f. 21918/38566 i
Charpy V-Notch Data Temperature input Permnt Shear Computed Percent Shear Differential
-50 2 2 81 -B1
-25 5 553 -53 0 20 1058 9.41 0 10 1058 -58 25 15 1929 -429 25 30 1929 10.7 50 30 3256 -256 70 40 49 3 -92 15 0 100 8894 ILOS
= Data continued on next page "
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE IStAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-42 CAPSULE V Page2 Material: IURGING SA5083 Heat Number: 21918/38566 Orientation 11 CapsuleY Total Tluence Charpy V-Notch Data (Continued)
Tm ture InputPe at Shear Computed Pereent Shear Differential 9
224 250 100 99 5 ,74 300 100 Wal 3 SUM of PISIDUAIS = 16.17 l
1 e
e a
W ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRA!RIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
l .. C.43 CAPSULE V CVGRAPH 41 Hyperbolic Tangent Curve Printed at 17:11:20 on 10-23-1996 Page1 Coeffkients of Curve 1 A = 4659 B = 44.4 C = 107.48 TO=!?.17 Equation is CVN = A + B * [ tanh((T - 11))/C) l l Upper Shelf Energy: 91 Fixed Temp. at 30 ft-lbs -30 Temp. at 50 ft-lbs 20.4 Irwer Shelf Energy: ?.19 Fired Material: WD Heat Number:1752 Orientation-Capsule V Total Fluena:
300 m aso O
I
)
w aw A
4 u
4 150 e
c kJ n 100 2 [ c b
- (
/
o
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: P11 Cap:V Material: WELD Ori: Heat f: 1752 Charpy V-Notch Data Temperature Input CVN Energy Computed CVN Energy Differential
-50 255 2144 2.05 0 38 4159 -359 32 60 5439 53 75 64 6935 -5S5 11 0 785 78fi2 .12 150 88 8435 334 210 795 8881 -931 210 1025 8881 1168 SLIM of RESIDUAIS = 5.4 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVE!LLANCE PROGRAM -
C-44 d
CAPSULE V CVGRAPH 41 Hyperbolic Tangent Curve Pdnted at 17:1221 on 10-23-1996 Page1 Coefficients of Curve 1 A = 4031 B=3931 0 = 12425 E = 38S ,
I Equation is 12 = A + B
- I tanh((T - M)/C) j !
Upper Shelf II: 7932 Temperature at 12 3E 22 lower Shelf II: 1 Fixed Material: WELD Heat Number.1752 Orientation:
Capsule V Total Fluence 200 m
.= 150 n
4 x M too g O -
$m A * /
[
a s
0
-coo -200 -too - o too 200 soo 400 soo soo Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap:Y Material TELD Ori: Heat f.1752 Charpy V-Notch Data Temperature input lateral Erpansion Computed 12 Differential
-50 13 5 1616 -2B6 0 29 2838 El 2 43 3812 437
. 75 47 51.41 -4.41 110 585 60S3 -213 150 72 6835 364 i 210 67 7491 -791 210 82 7431 7D8 SUM of PJSIDUAIS = -jE 1
1 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-45 CAPSULE V CVGRAPH 43 Hyperbolic Tangent Curve Printed at 17d323 on 14-23-1996 Page1 Coefficients of Curve 1 A = 50 B=50 C = 94D9 M = 40S8 F4 uation is: Shear /. = A + B ' I tanh((T - 2)/C) 1 Temperature at 50/. Shear. 493 Materiah TELD Heat Number.1752 Orientation:
Capsule V Total Fluence:
100 ed a)
OU f
.C m g C a a ca O
y 40 .
Q.4
}
20 J
u
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set {s) Plotted +
Plant: P11 Cap:V Materiah TELD Ori: Heat f.1752 Charpy V-Notch Data Temperature input Percent Shear Computa! Percent Shear Diffemntial
-50 10 1012 -12 0 25 258 -8 32 50 da71 928 75 50 6 113 -1113 11 0 80 78 71 172 150 95 89 3 56 210 100 9a79 32 210 100 9a79 32 SUM of PEDUAIS = 835 en ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER QOMPANY P,8AIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM -
. 1 l
C-46 Best-Fit Curve as documented in WCAP-8916 l
(not fit with hyperbolic-tangent function)
Upper Shelf Energy: <211' Temp. at 30 ft-the <-200 Temp at 50 ft-lhe <-125 lower Shelf Energy:
Material HEAT AITD ZONE Heat Number, uneaution:
Capsule Y Total }1uence 3*
I cn 250
,Q
~
- 'Ia
/
g am
.m k 1s0
/ a CD _
i C <r a N o /
100 a
> a o ,
/
o
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted
, Plant: Pil Cap:Y Materiah HEAT AITD ZONE Ori: Heat f:
Charpy V-Notch Data Temperature Input CVN Energy Computed CVN Energy Differential
-15 0 104 903 1319
-100 73 972 -242
-50 12 8 10352 24.47 75 885 11 & 4 7 -2951 100 119 12127 -72/
150 116 5 126 S 2 -10J2 175 1595 12918 30.31 SUM of PJSIDUAIS : L4
- Upper shelf impact energy not obtainable due to excessive toughness.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM -
C-47 l 1
CAPSULE V
.. l Best-Fit Curve as documented in WCAP-8916 (not fit with hyperbolic-tangent function)
Upper Shelf a 87 Temperatun at 2 Jr -128 , lower Shelf LE:
Material: HEAT AFD ZONE Heat Number. Orientation-Capsule V Total mence:
am i
i
<n
= 150 l
)
a .
. 4 M !
rsa 1, l
e-v U ce f
$ a Y
a su
[ o O
o- ^
-300 -200 -tm o im 200 aoo 400 soo soo 1
Temperature in. Degrees F l Data Setls) Ntted ;
&nt: P11 Cap: Y Material HfAT AFD ZONE Ori: Heatf. i Charpy V-Notch Data Temperatun input lateral Expansion Computai E .
Differential
-150 495 4158 5 91
- 10 0 36 48.03 -1?.03
-50 65 52A5 1214 75 54 6654 -1254 100 735 6957 192 150 68 752 -73 17 5 90 792 10.79 SUM of PJSIDUAIS = 29 O
en ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-48 CAPSULE V
~
Best-Fit Curve as documented in WCAP-8916 (not fit with hyperbolic-tangent function)
Temperature at 50x Shear. -60 Material HFAT AITD 20NE Heat Number: Orientation:
Capsule V Total Fluence 100 D
a $
CO e
ca eu =
=,/
c e
O A
b* f
=
2u u
-300 -200 -100 0 100 200 300 400 500 600 i Temperature in Degrees F Data Setis) Plotted
, Plant: P!1 Cap:Y Materiah HEAT A}TD 20h1 Ori: Heat f:
Charpy V-Notch Data l Temperature input Percent Shear Computed Percent Shear Differential
-150 60 4t01 1&96
-100 30 47 2 -1723
-50 60 54 S4 5.05
' 75 40 7L09 -3L09 100 80 73fl9 61
'. 15 0 85 7E93 BD6 175 100 8L17 1832 SUM of PJ51 DUALS = 199 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM j
C-49 CAPSULE V O'CRAPH 41 Mf perbolic Tangent Curve Printed at 17:1959 on 10-23-1996 Page1 Coefficients of Curve 1 A = 4659 B : 44.4 C = WD9 TO = 1872 F4uation is OW = A + B ' [ tanh((T - 11))/C) ] -
Upper Shelf Energy 91 Fixed Temp, at 30 ft-lbs 149 Temp. at 50 ft-lbs 194S laer Shelf Energy 219 Fixed Material: SRM HSTM2 Heat Number: SA5:DB1 Orientation: LT Capsule V Total Fluence 300 ui asu o
I a
cz., 2$
h
. 4 e
a 150 c) c -
ra 100 u Z 0 W
> 0 0 n so u 0
-300 -200 -too - o too 200 aoo 400 soo soo l
. Temperature in Degrees F l Data Set (s) Plotted Plant: Pil Cap:Y Material: SRM HSm2 Ori: LT Heat f. SA533B1 Charpy V-Notch Data Temperature input GH Energy Computed 03 Energy Differential 75 10 5 10 2 29 150 245 3037 -557 175 465 41D4 5.45
. 210 655 5633 866 250 525 7128 -1938 Z?5 84 7&49 55 300 86 81R 292 350 96.5 87.99 85 SUM of FISIDUAIS : 6D8
~
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C 50 CAPSULE V CVGRAPH 41 Hyperbolic Tangent Curve Printed at 17212 on 10-21-1996 Page1 Coefficients of Curve 1 A = 40E! B = 3932 C = 116.11 1D = 210D5 Equation is: E = A + B ' [ tanh((T - IV)/C) )
Upper Shelf LF.: 8034 Temperature at 2 35: 19? 9 lower Shelf LI:1 Fired Material: SRM 151V2 Heat Numbec SA533B1 Orientation: LT Capsule V Total Fluence 200 rn
- 150 n
. M 100 e '
b 7 a m yo
/
0 4
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set Plant: P11 Cap:V Material:IEf02 SRM(s) Ori:Plotted LT Heat f. SA533B1 Charpy V-Notch Data '
Temperature Input lateral Erpansion Computed E Diffenntial
'S 2 &06 -6DB 15 0 18 2138 -338 17 5 36 29J4 65 210 47 403 6.19 250 46 54 -6 25 59 61 2 -202 300 68 66S8 131 350 76 74D8 191 SUM of RIS!DUAIS = -171 4 15 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILIANCE PROGRAM -
C-51 CAPSULE V CVCRAPH 4J Hyperbolic Tangent Curve Printed at 17:2154 on 10-23-1996 Page1 Coefficients of Curve 1 A = 50 B = 50 C = 10152 % = 217D3 Equation is Shearx = A + B ' [ tanh((T - W)/C) }
Temperature at 50x Shean 217 Material: SPM HSm2 Heat Numben SA533B1 Orientation: LT Capsule V Total Fluena:
100 =
~
w $
. cd
. D
.c l cn ,
4 4
C a a )
c O i b 40 O a a :
a !
j i 0
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F l Data Set (s) Plotted Plant: P11 Cap:Y Material: SPM HSm2 Ori: LT Heat f. SA533B1 Charpy V-Notch Data Temperature loput Per v. 2 v Computed Perant Shear Differential 75 10 E74 425 15 0 20 21M -107 1 70 35 30.4 459 210 50 4634 145 250 50 65SB -15 fib 215 70 70B -5B 300 100 83B 16.32 350 100 932 E79 SUM of RESIDUAIS = 1285 l
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-S2 CAPSULE P CVCRAPH 41 Hyperbolic Tangent Cune Printed at 14:4&49 on 0F20-1996 Page!
Coefficients of Cune 1 A = 7209 B = 69.9 C=8831 % = 45.93 Fquation is CVN = A 4 B * [ tanh((T - %)/C) ]
Upper Shelf Energy:142 Fixed Temp at 30 ft-lbs -153 Temp, at 50 ft-lbs 16 3 Imer Shelf Energy:119 Fired Materiat FORGING SA500 Heat Number: 21918/38566 Orientation: LT Capsule P Total Fluence l 3
l 1
i m aso D~
l l '
a x em ;
u W 150 "
r O ~
o ,
c !
M
> l 100 '
2: :
o l m
,e
-300 -200 - 10 0 o 100 200 300 400 500 soo i Temperature in. Degrees F Data Set (s) Plotted
. Plant: Pil Cap: P Materiah 10RGING SA5083 Ori: LT Heat f. 21918/38566 Charpy V-Notch Data Temperature loput CVN Energy Computed CVN Energy Diffenntial
-50 95 1658 -748
-25 175 2536 -E16
-10 4&5 3103 1146 0 44 3&79 52 25 48 55 3 -738 50 745 753 -3 75 171 9423 2.76 100 106 11033 -413 15 0 129 12931 -31
" Data continued on nert page "
.su ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
'l C-53
~ CAPSULE P Page2 Material FDRGING SA5083 Heat Number. 21918/38566 Orientation: LT CapsuleP Total Fluence Charpy V-Notch Data (Continued)
Temperature Input CVN Energy Computed CYN Energy Differential 200 1475 1373 939 25 1365 1412 350 -4.7 1425 14135 .64 SUM of PEDUAIS = J7 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
-J
l C-54 CAPSULE P CVGPJLPH 4J Hyperbolic Tangent Curve Printed at 145412 on 09-20-1996 Page1 Coefficients of Curve 1 A = 4453 B=4353 C = 8&6 10 = 29.06 Equation is E = A + B ' I tanb((T - IV)/C) l Upper Shelf E 8BD6 Temperature at E 35: 93 lower Shelf G 1 Fixed Material: FORGING SA5083 Heat Number. 21918/38566 Orientation: LT Capsule P Total Fluence am m
.O 150 a
4 X loo
- a n n
()
O a cJ c a su a
~
o U '
-300 -200 -100 ' o 100 200 soo 400 500 soo Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap: P Material JURGING SA5083 Ori: LT Heat f. 21918/38566 Charpy V-Notch Data Temperatun input lateral Expansion Computed 2 Diffenntial
-50 109 1351 -10.41
-25 1539 2034 -514
-10 34 2 2E49 81 0 35.79 30.74 SD5 25 37 4253 -553 50 66S 54E2 1227 75 519 65 21 -1137 10 0 768 7145 134 150 7319 E73 -953
Data continued on next page
ANALYSG OF CAPSOLE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISI.AND Unit 1 REACOR VESSEL RADIATION SURVE!LLANCE PROGRAM
l l l
l* C-SS CAPSULE P Page2 Material: R)RGING SA5083 Heat Numben 21918/38566 Orientation LT Capsule P Total Fluence Charpy V-Notch Data (Continued) l l
Temperature Input lateral Erpansion Computal !.F.
200 Differential 9039 8626 433 275 89.4 87.72 Is l 350 9019 l 88 pig SUM of P2SIDUAIS = -5 I
4 l
l l
1 l
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISt.AND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
. c-se CAPSULE P CVCRAPH 41 Hyperbolic Tangent Curve Printed at 14f6:14 on 09-20-1996 Page1 Coefficients of Curve 1 A = 50 B=50 C = 9145 W = 74J2 Equation is Shearz = A + B ' I tanh((T - 1D)/C) l Temperature at 50x Shean 741 Material PORGING SA5083 Heat Number: 21918/38566 Orientation: LT l Capsule: P Total Fluence:
100 g -
y M a
' cc 0
d c cn so c
e O
b a
l 20 gr I
- o o l l , a c i
l
-coo -200 -too o too 200 aoo 400 soo soo l
Temperature in Degrees F Data Set (s) Plotted Plant: P!1 Cap:P Materiah FDPSLNG SA5083 Ori: LT Ileat f. 21918/38566 Charpy V-Notch Data Temperature input Perant Shear Computed Perant Shear Differential
-50 3 638 -338
-25 7 10.4 8 -148
-10 18 1194 4 115 0 17 16.74 25 25 28 25El 232 50 34 J724 -324 75 53 50.47 252 100 64 63S4 35 15 0 78 83.77 -E77
" Data continued on nert page =
I I
l l
1 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-57
. CAPSULE P Page2 Materiah FORGING SA5083 Heat Number. 21918/38566 Orientation LT Capsule P Total Fluence Charpy V-Notch Data (Continued)
Temperature input Percent Shear Computed Percent Shear Differential 200 10 0 9333 6J6 715 100 9&T2 121 350 100 99.74 25 SUM of RISIDUAIS = 12 4
l l
A r
J l -
e A
d J
e 0
.j J ,,,
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY P,RAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
]
C-58
. CAPSULE P
- CVGRAPH 41 Hyperbolic Tangent Curve Printed at 14511)5 on 09-20-1996 Page1 Coefficients of Curve 1 A = 69D9 B = 66.9 C = 137.41 TO = 94E2 Equation is CVN = A + B * [ tanh((T - 11))/C) ]
l'pper Shelf Energy:136 Fized Temp at 30 ft-lbs 26 Temp at 50 ft-lbs 542 lower Shelf Energy: 219 fired Material ERGING SA500 Heat Number. 2t918/38566 Orientation T1.
Capsule P Total Fluence:
300
,C m a50 I
a '
g am .
4 x t:o b 150 u g n l C U 100 A 1 O '
o/ao 50__
o p i
J 0 0 . .
-a00 -200 -loo o too 200 aoo 400 500 soo
' Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap;P Material NPSING SA5083 Ori TI, Heat f. 21918/38566 Charpy V-Notch Data Temperature input CVN Energy Computed CVN Energy Differentia!
-50 6 IE73 -1&73
-10 18 5 2635 -7SS 0 40 29J5 1034 50 585 481 1039 75 *M 593 1039 100 55 7L71 -l & 71 15 0 91 94S8 -3S8 200 103
- 11225 -925 250 146 12337 2282
" Data continued on next page "
~
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRA!R!E ! GLAND Unit 1 REACTOR VESSEL RAD lATION SURVEtLtXNCE PROGRAM
4 C-59 l -
i .
CAPSULE P Page2 l Material FORGING SA5083 Heat Number 21918/38566 Orientatiom n Capsule P Total Fluence Charpy V-Notch Data (Continued)
Temie ture Input G Energy Computed GW Energy MfenM IE -72 I
@ 137 5 134.44 g l
SUM of REIDUAIS = 1.94 i
4 4
l l
ANALYSIS OF CAPSOLE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1
^
1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
i C-60 :
.. CAPSULE P
- CVGRAPH 41 Hyperbolic Tangent Curve Printed at 1458:38 on 09-20-1996 j l Page1 Coefficients of Curve 1 A = 3L43 B = 30.43 C=74f1 W = 937 Equation is E : A + B ' ( tanh((T - 2)/C) l Upper Shelf II: 6137 Temperature at E 3E 1&1 lower Shelf 12: 1 Fixed Materiah }DRGING SA5083 Heat Number: 21918/38566 Orientation: TL Capsule: P Total Fluence 200 m
O 150 a
. 4 x
M 100 co 4 a a
o o n g Cy #
U a 50 o n
o O
o
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Plant: P11 Cap:P Material: FDRGING Ori: TL )SA5083 Data Set (s Plotted Heat b 21918/38566 Charpy V-Notch Data
, Temperature Input lateral Erpansion Computed 11 Differential
-50 !?19 It29 3
-10 1569 217 -6 0
50 315 M3 2 5509 4654 75 855
.5109 5233 J6 100 472 5694 -9 74 150 5159 605 -89 200 6E69 6151 418 250 7119 6LTI Il42
= 0.t conunued on ont pas. ~
Em ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEll1ANCE PROGRAM '
._ ._. . . _ . . - . . - . --- . - ~ . . . - - _ _ _ _ _ .. - . _ - . .- .
l l' C-61 CAPSULE P Page2 Material: RRGING SA5083 Heat Number: 21918/38566 Orientation: TL Capsule P Total Fluence Charpy V-Notch Data (Continued)
Temperature input lateral Erpansion Computed LE Differential l 350 63.79 6136 133 400 57D9 61El -4.77 SUM of REIDUAIS = .4 l
l l
4 + +
I l
l 4
l l
I l
l l
l l
89 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
. C-62 CAPSULE P CVCPJLPil 41 lipperbolic Tangent Curve Printed at 150045 on 09-20-1996 Page!
Coefficients of Curve 1 A = 50 B = 50 C = 11611 TO = 103J2 Equation is Shearx A + B ' [ tanh((T - 1D)/C) l Temperature at 50/. Shear.103.1 Material MRCING SA5083 Heat Nr.nber. 21918/38566 Orientation TL Cape: P Total Fluence:
100 2 f
o OU
)
Cd o
C 60 4 4 C
0 g' 40
/o 4 a 20 o f
- a 0
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Setts) Plottai Plant: PIl Cap: P Material MRGING SA5083 Ori: TL Pat f. 21918/38566 Charpy V-Notch Data Temperature Input Pertent Shear Computai Percent Shear -
Differential
-50 3 6S7 -367
-10 13 1247 52 0 18 1(47 352 50 31 2859 24 75 45 3&l2 EM i 100 38 48fi5 -10E5
!$0 66 6915 -315 l 200 87 84j3 286 i 250 100 92.62 73 I
- Data contir.ued on next page =
e ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-63 CAPSULE P Page2 Material: fDP.GING SA5083 Heat Number. 21918/38566 Orientation TL Capsule: P Total Fluena:
Charpy V-Notch Data (Continued)
Temperature loput Perant Shear Q>mputed Perant Shear Differential 350 100 9859 L4 400 100 99.4 30 SUM of RESIDUAIS = E09 4
9 4
W ANALYSIS OF CAPSULE S FROM THE NO,RTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
0 C-64
. CAPSULE P
~ O' GRAPH 41 Hyperbolic Tangent Curve Printed at 15M26 on 09-20-1996 Page1 Coefficients of Curve 1 I A = 4259 B = 40.4 C = 1268 TO = 2155 Equation is 03 = A + B * [ tanh((T - 10)/C) J Upper Shelf Energy: M Fixed Temp. at 30 ft-lbs -192 Temp. at 50 ft-lbs 45 lower Shelf Energy: P.19 Fixed Material TELD Heat Number:1752 Orientation:
Capsule P Total Fluence 300 m 250 Q
I a
g em 4 h !
tm !
4 150 1 G) l c -
rza i 100- o o A 0 l
, ) i
-300 -200 -100 ' O 100 200 300 400
. 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: PIl Cap.: P Material: TELD Ori: Heat l: 1752 Charpy V-Notch Data Temperature input 03 Energy ' Computed 07 Energy Differential
-60 205 21 2 -L42
-25 325 2837 412 0 355
- 35.78 -28 50 515 5152 -D2 75 555 5&7 -32 150 71 73S1 -281 250 95 8036 1413 300 835 82D1 L48 SUM of PEIDUAIS = 1217
>=
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPAtN PRA!RIE ISLAND Unit 1 REACTOR VESSEL RAD lATION SURVEILLANCE PROGRAM
C-65 CAPSULE P CVCRAPH (J Hyperbolic Tangent Curve Printed at 15dO56 on 09-20-1990 Page1 Coefficients of Ourve 1 A = 40.96 B = 3936 C = 14134 M = 46FI Equation is E : A + B ' l tanh((T - %)/C) l Upper Shelf 1180.92 Temperature at E 3ir 255 lower Shelf 1.E: 1 Fixed Material: TilD Heat Number: 1752 Orientation-Capsule P Total fluene 200 m
.O 150 a
4 M
100
[
2 7-3ce a 50
~ /
2 U l l I i
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap:P Materiat NI1D Ori: Heat f.1752 Charpy V-Notch Data Temperature input lateral Expansion Computed 11. Differential
-50 1729 1724 DS
-25 2439 2228 211 0 28 2821 -21 1 50 37 4134 -434 75 52 4&78 3 21 i
15 0 665 65.78 .71 i 250 77J9 7651 58 !
300 7759 7&73 -113 l Sl'M of EISIDUAIS = .49 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POVER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION buRVE!LLANCE PROGRAM l
~
l f
e c-se CAPSULE P CVCRAPH 4J Hyperbolic Tangent Curve Printed at 15230 on 09-20-1996
~
Page!
Coefficients of Curve 1 A = 50 B = 50 . C = 127.45 1D = 20.47 L luation is Shearx : A + B
- 1 tanh((T - 1D)/C) l Temperatun at 50x Shear. 20.4 Material WD Heat Number: 1752 Orieutation:
Capsule P Totalfluenz j 100 o [ !
1 i
a $ /
e i e
.c M so l
. < s ,
.c e .
4 O t k 40 e l a
m 7 l U
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F i Data Set (s) Plotted Plant: P11 Cap:P Material: WELD Ori HeatI: 174 Charpy V-Notch Data Temperatun input Percent Shear Computed Permnt Shear Differential
-50 27 242 2.13
-25 35 32 2 211 0 43 4203 .96 50 53 6138 -838 75 71 7017 B2 1 150 95 8&41 658 l 250 100 Irl34 2fi5 300 100 9 & 71 122 l Sl!M of PEIDUAIS : E12 l
l i 1 l
~
l l
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 j REACTOR VE3SEL RADIATION SURVElLLANCE PROGRAM -
)
~.
l l
C-87 CAPSULE P CVCRAPH 41 Hyperbolic Tangent Curve Printed at 121526 on 09-20-1996 Page1
- Coefficients of Curve i A = 7259 B = 70.4 C=9922 1D = -5178 F4 uation is CVN = A + B ' l tanh((T - 10)/C)l Upper Shelf Energy
- 143 Fixed Temp at 30 ft-lbs -1253 Temp at 50 ft-Ibs -883 laer Shelf Energy: 219 Fixed Materiah HEAT AFFD ZONE Heat Number. Orientation:
Capsule P TotalFluenz 30u m 250 4
I a
N N a
4 h ~
e 4
150 g --
c o M ,
Z o o o so , -
o , , ,
l i
-300 -200 -100 0 100 200 300 400 500 600 Temperature in. Degrees F Data Set (s) Plotted Plant: Pil Cap: P Materiah HEAT AffD ZONE Ori: Heat l:
Charpy V-Notch Data Temperatum in;'ut CVN Energy Computed CVN Energy Diffenntial
-150 17 2053 -353
- 10 0 55 4 3.15 1134 0 12 10& 47 -26.47 75 186 13158 5?.41 150 116 5 14031 -2421 250 1275 14?.7 -15 2 SUM of RESIDUAIS = -526 l
i
- l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISt.AND Unit 1 l REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 1
a i
, C-68
, CAPSULE P CVCRAPH 43 Hyperbolic Tangent Curve Printed at 15:3633 on 09-20-1996 Page1 Coefficients of Curve 1 A = 4034 B = 3934 C = 158 TO = -2729 Equation is M = A + B ' I tanh((T - 1D)/C) I j
Upper Shelf LE: 8039 Temperature at E 3E -512 lower Shelf LE: 1 Fired !
Materiah HEAT AITD ZONE Heat Numben Orientation-Capsule P Total Fluena:
2m ,
i J
m
.O 150 a
4 x
100
~
e n b K a So c
/
o U
-300 -200 -100 0 100 200 300 400 500 600 Temperature in D' egrees F Data Set (s) Plotted Plant: Pil Cap: P Materiah HEAT AITD ZONE Ori; Heat i:
Charpy V-Notch Data Temperature input lateral Erpansion Computed M Differential
-150 10 3 15 -4.4
-100 2829 2183 4.47 ;
D 457 473 -21 75 66.09 63fi5 244 150 693 711 -33 200 803 7839 L9 SUM of PEDUAIS :-L49 1
)
ANALYSIS OF CAPSULE S FROM ThE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 !
REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
~ i l
C-69 CAPSULE P CVCRAPH 4J Hyperbolic Tangent Curve Printed at 1540J4 on 09-20-1996 Page1 Coefficients of Curve I A = 50 B = 50 C = 8322 % = -6033 fquation is Shear /. = A + B ' ( tanh((T - W)/C) l Temperature at 5&/. Shear: -60.9 Materiah HEAT AFFD 20hT Heat Number. Orientation-Capsule P Total Fluenz im ; - 2 g
u m >
ce e
A
- 60
. < a ce O -
5 * .
4 o 20
}
s a 0
-300 -200 - 100 ' O 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap:P Materiah HEAT AffD 20h1 Ori: Heat l: l Charpy V-Notch Data Temperaim Input Percent Shear Computed Percent Shear Differential
-15 0 2
-100 10 2 -82 34 2&ll 538 0 *19 8122 -222 75 96 96 2 -2 150 96 99 2 -3J/
250 100 99.94 A5 SL'M of PSDUAIS = -85
. 1 l
I ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM l
. C-70 CAPSULE P CVCRAPH 41 Hyperbolic Tangent Curve Printed at 154232 on 09-20-1996
~
Page!
Coefficients of Curve 1 A = 4359 B = 4L4 C = 4134 11) = 22L71 Equation is CVN = A + B ' I tanh((T - TO)/C) l Upper Shelf Energy: 85 Fixed Temp. at 30 ft-lbs 207S Temp, at 50 ft-lbs 21 lower Shelf Energy: 219 fixed Materiah SPA IEID2 Heat Number: SA533B1 Orientation LT Capsule: P Total Fluence:
300 m 2so
,C I
N 2m .
t:to L 150 c>
c m .
100 t a.
> o o
So y
tP D
ad
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: Pll Cap: P Materiat SPM HSSID2 Ori.: LT Heat f. SA533B1 Charpy V-Notch Data Temperature input CVN Energy Computed CVN Energy Diffenntial 125 9 296 6.03 200 25 23E5 134 210 28 3?.17 -437 225 495 4637 262 l 250 67 6819 l
-L19 300 375 8316 421 350 735 8433 -1123 425 935 8499 85 SUM of PISIDUAIS = 614 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM -
C-71 CAPSULE P CVGP.APH (J Hyperbolic Tangent Curve Printed at 1544:46 on 09-20-1996 Page!
Coefficients of Curve 1 A = 3036 B=29S6 C = 4451 70 = 21134 Equation is 11 : A + B ' I tanh((T - TO)/C) l Upper Shelf LE: 6032 Temperature at II Ex 2173 lower Shelf II: 1 Fired I Material SPM HSSTt2 Heat Number: SA533B1 Orientation LT Capsule P Total Fluena:
20u m
."- 150 E
a 4 x.
100 e
b>
u Og a Su o U, 4 l ,
1 g
-300 -200 - 10 0 0 100 200 300 400 500 600 Temperature in Degrees F DataSet Plant Pil Cap:P MateriahHS57J2 SPM(s) Ori:Plotted LT HeatfSA533B1 Charpy V-Notch Data Temperatum Input lateral Erpansion Computed II Differential 12 5 SD9 2 21 228 200 2U$ 2M8 - 218 210 33.09 30D5 104 225 362 3937 -3S7 250 57S 51S5 5S4 300 472 5933 -12E3 350 683 603 8D9 425 6L4 6032 .47 SUM of PIEUAIS : 195 ANALYSIS OF CAPSULE S FROM THE NCRTHERN STATES POWER COMPANY PRAIRIE ISt.AND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
l
. C-72 CAPSULE P !
CVCP.APH 4J Hyperbolic Tangent Curve Printed at 154&41 on 09-20-1996 Page!
Coefficients of Curve 1 l A=2 B=M C=N3 %=MM Equation is Shear /. = A + B ' I tanh((T - TO)/C) )
Temperature at $b Shear: 2212 Materiah SPM HSSltl2 Heat Number: SA533B1 Orientation: LT l Capsule P Total Fluence 100 l
w $
e o
A cn a> e
< a c
e .
O 1 y su of l
a P I
1 1 au
\
1 l
O
- 1 i 1
-300 -200 - 100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap:P Materiah SPM HSS112 Ori LT Heatf.SA533B1 Charpy V-Notch Data Temperature input Percent Shear Computed Percent Shear' - Differential 12 5 9 4.76 423 l 200 'A 34D4 3.95 1 210 34 4133 -733 225 5B 52.91 l EOS 250 65
' 7038 -5.98 300 100 E06 733 350 100 9821 1.78 425 100 9932 J7
. SUM of PJSIDUAIS = 935 ANALYSIS OF CAf 3ULE S FRCN THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unii i REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
- -=. . . ._ . .- - - - .. .. ..
C-73 CAPSULE R f
CVGRAPH 41 Hyperbolic Tangsct Cceve Printed at 155614 on 09-20-1996 Pagei Coefficients of Utree 1 A = 7359 B = 7L4 C = 10143 11) = 1307/
Equation is CVN : A + B ' l tanh(tT - 11))/C) ]
Upper Shelf Energy:145 Fixed Temp. at 30 ft-lbs 56S Temp. at !D ft-lbs 943 laer Shelf Energy: P.19 Fixed Waterial: It)RGING SA5010 Heat Number: 2i918/38566 Orientation LT Capsule R Tekt fluena:
300 -
m usu
.O I
x em X
4 Q O'
% L50 '
o M A ,
100 1 Z
l l
0 -
n su a Al O o, j
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap: R Material:II)RGING SA5083 Ori: LT Heat f. 21918/38566 Charpy V-Notch Data Temperature Input CVN Energy Computed CVN Energy Differential 0 8 1232 -432 25 17 18 S6 -126 50 10 2711 -1711 50 18 2711 -911 50 44 2711 1638 75 52 3885 1334 150 91 86S8 4D1 200 101 11551 -1451
. 250 135 132.14 235
Data continued on next page
~
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
. - 1 l
, C-74
.. CAPSULE R Page2 j Material: FORCING SA5083 Heat Number. 21918/38566 Orientation LT Capsule: R Total Fluenm Charpy V-Notch Data (Continued) l Temperature input CVN fnerg Computed CYN Energy Differential 149 13932 g17 350 157 14198 Igg
- 139 14422 522 SL'M of RlHDUALS = 731 4
4
. 4 5
4 4
e e
DEW ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE FROGRAM
C-75
.. CAPSULE R
~ CVCRAPH 41 Hyperbolic Tangent Cune Printed at 155821 on 0F20-1996 Page!
Coefficients of Cune 1 A = 4075 B = 3925 C = 8&75 % = 92B1 Equation is: 1.F. : A + B ' [ tanh((T - W)/C) l Upper Shelf 11795 Temperature at E 35: 803 Imer Shelf 111 Fixed Material: FORCING SA5083 Heat Number. 21918/38566 Orientation LT Capsule: R Total Fluenz mo m
.O 150
- 4 a
y N
1 100 e o o w e o 0 o c
a so .
D r
D a
0
-a00 -200 -too o too 200 aoo 400 soo 800 :
Temperature in Degrees F l Material FORGING )SA5083 Data Set (s Plotted Plant: Pll Cap:R Ori: LT Heat f. 21918/38566 Charpy V-Notch Data Temperature input lateral Erpansion Computed E Differential 0 8 932 -1E2 25 14 l
1439 .99 50 13 22E6 -936 50 15 22A6 -7E6 50 34 22a6 1133 75 40 3?.47 752
- 150 63 62.54 .45 200
- 250 66 7
- J -7D6 80 1729 P.7
- nota mounu.i on next p.se =
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY P,RAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILIANCE PROGRAM
. C-76 CAPSULE R Page2 Materiah FORGLNG SA5083 ileat Number. 21918/38566 Orientatiom 1.T Capsuk R Total Fluence Charpy V-Notch Data (Continued)
Temperature loput la Erpanson Compg LE Differential
=
2 W.C
- s37 SUM of PJSIDUAIS -145 i
t
= 4 9
l .
. i l
l <
1 1 .
l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
p C-77 l
CAPSULE R l CVCRAPH (J Hyperbolic Tangent Curve Printed at 16M13 on 09-20-1996
~
Page1 Coemeients of Curve 1 A = 50 B=50 C = 9025 10 = 13125 Equation is Shear /. = A + B ' l tanh((T - 1D)/C) {
Temperature at Sk Shear: 1312 l Materiah IDRGING SA5083 Heat Number. 21918/38566 Orientation: LT Capsule: R Total fluenz
= -
100
)
1 c
a 80 ce e
A cn so a
ce O
b 40 e
a 20 o M) u
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degre.es F Data Set (s) Plottal Plant: P11 Cap:R Materiah FORGING SA5083 Ori: LT Heat f. 21918/38566 Charpy V-Notch Data Temperature input Penent Shear Computed Percent Shear Differential 0 2 5J7 - 117 25 6 BE7 -267 50 11 1417 - 117 50 14 1417 -17 50 18 1417 35! ,
75 30 E33 7116 150 54 6024 -624 l 200 82 82.1 -l j 250 100 9328 6.71 l
! l
" Data conunued on next page "
. i m.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER ~ COMPANY PRAIRIE ISLAND Unit 1 l
~
l REACTOR VESSEL RADIATION SURVE!LIANCE PROGRAM -
1
- C-78 l .
CAPSULE R Page2 Materiah R)RCLNG SA5083 Heat Number: 21918/38566 Orientation 1,7 Capsule R Total D uen Charpy V-Notch Data (Continued) .
Temperature Input Percent Shear Computed Pement Shear Differential 300 100 ME7 232 350 100 9922 .W 400 100 99.74 25 SUM of TEDUAIS : 601 4
f I
i i
i ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PF%lRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-79 CAPSULE R CVGRAPH 41 Hyperbolic Tangent Curve Printed at IEf222 on 09-20-1996 Page!
Coefficients of Curve 1 A = 6559 B = 614 C = 12L63 1D = 130J1 F4 uation is CVN = A + B ' I tanb((T - 70)/Q l Upper Shelf Energy:129 Fixed Temp. at 30 ft-lhe 528 Temp. at 50 fHbc 995 lower Shelf Energy:119 Fixed Material: K%ING SA500 lient Number: 21918/38566 Orientatum: TL Capsule R Total Fluence 300 m 250 eQ I
a z 200 4 x ec L 150--
C) o c U
. [a N ,
z' 0 -
so >
0
- M O 1
Ul l il
, -300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees. F Plant: P!! Cap:R Material: IVRGING Ori:)SA5083 TL Heat f. Data Set (s Plotted 21918/30566 Charpy V-Notch Data Temperature input CVN Energy
] Computed CVN Energy Differential O 11 1555 -455 25 9 21 3 ? -12 3 !
ti0 18 28.98 -10.98 50 36 2898 7D1
, 75 42 3868 331 75 75 3868 3631 100 44 5021 -4 21 150 51 75 51 -2437 200 101 9&48 251
- D t. wnunu.a on next p.e. -
e M
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
. C-80
- .. CAPSULE R l Page2 Lterial
- ERCING SA5083 Beat Number. 21918/38566 Orientathn: TI, Capsule R Total Fluence:
Charpy V-Notch Data (Continued)
Temperature Input O'N Energy g Computed O'N Energy Differential 124 jn4g g 12138 10.5 721 400 2 251 E48 SUM o' RESIDUAIS = n49 4
e I i I
i ~
l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
s C-81 CAPSULE R CVGRAPH 41 Hyperbolic Tangent Curve Printed at 15:26:34 on 10-22-1996
~
Page!
Coefficients of Curve 1 A = 4334 B = 4234 C = 131.15 TO = 11147 Equation is 11 = A + B ' l tanh((T - 1D)/C) l Upper Shelf LE: 85fi8 Temperatun at II 35: 852 laer Shelf LE:1 Fixed Wateriah FORGING SA5083 Heat Number: 21918/38566 Orientation: T!.
Capsule: R Total }1uence:
20u m
= 150 .
n 4 X 100
@ *a .
5 [
t u m e
go
/
D 0
jPQ u, ,
s ; ;
~
-300 -200 -100 0 100 200 300 400 500 600 i
Temperature in Degrees F l Data Set Plant: Pil Cap: R Materiah FORGING SA5083 Ori: (s)
TI, Plotted Heat f: 21918/38566 l
Charpy V-Notch Data Temperature input lateral Expnsion Computed LE Differential :
0 10 1411 -4 11 25 10 18119 -839 50 16 24115 -8B5 50 31 2425 634 75 31 31E7 -A7 75 E5 3151 23fi2 100 36 39A5 -3S5 15 0 46 % 41 -9.41 200 67 6821 -121 l
- "" Data continued on next page ****
e e
em ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
- C-82 i
,. CAPSULE R I Page2
~
Material: IDRCING SA5083
' Heat Number: 21918/38566 Orientation: TL Capsule ?. Total Fluence Charpy V-Notch Data (Continued)
Temperature Input lateral Erpansion Computed LE Differential 250 83 7651 &48 300 79 8U3
@ B4
-PJ3 84 9 .y 1 SUM of RESIDUAIS :-355 l J
. 4 4
l l
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM l
O e
CAPSULE'R CVCRAPli 41 Hyperbolic Tangent Curve Printed at 1&0&43 on 09-20-1996 Page1 Coefficients of Curve 1 A = 50 B=50 C = 10028 1V = 140B3 Equation is Shenrx = A + B
- l tanh((T - IV)/C) l Temperature at 50x Shier. 140 Waterial: F0PSING SA5083 Heat Number: 21918/38566 Orientation: TI, Capsule: R Total Fluence.
im -
e Il y % (
ca e
C D
Su c
e G
$ * /
a 4 /
20
/ u u
-300 --200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap: R Material: FVRCING SA5083 Ori.: TL Heat f. 21918/38566 Charpy V-Notch Data Temperature input Percent Shear Computed Penrnt Shear Differential 0 3 57/ -277 25 6 9J5 - 115 50 11 1423 -323 50 16 1423 L76 75 35 2L46 13.53 75 37 2L46 1553 10 0 19 31R3 -12B3 15 0 40 5435 -1435 200 82 76.7/ 522
Data continued on next page
~
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE istAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM -
, c44 CAPSULE R Page2 Materiah FORGING SA5083 Ileat Number. 21918/38566 Orientation R Capsuk R Total Fluen Charpy V-Notch Data (Continued)
Temperature Input Perant Shear Computed Percent Shear Differential 250 100 89.96 10.03 300 100 9604 3SS 400 100 99.44 55
- SUM of RESIDUAIS = 14.44 s
. 4 J
i J
G S
m ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
- 1 C-85 CAPSULE R CVCRAPH 4J liyperbolic Tangent Curve Printed at 1529f)2 on 10-22-1996 Page1 i Coefficients of Curve 1 I A = 3859 B = 364 C = 13618 11) = 9032 L;uation is CVN = A + B
- l tanh((T - 11))/C) l Upper Shelf Energy: 75 Fixed Temp at 30 ft-lbs 58 Temp. at 50 ft-lhe 1342 lower Shelf Energy: P19 Fixed Materiat TELD IIcat Number.1752 Orientation:
Capsule R Total Fluenm 300 m 25u
,Q I
a x 20u 4 h e
L 150 C.)
c N
100
> u,"
O -
/
50 "
ra s
u, l \
\
-300 -200 -100 0 100 200 300 400 500 600 Temperature in ' Degrees F Data Set (s) Plotted Plant: P11 Cap: R Material: FELD Ori:
Heat f.1752 Charpy V-Notch Data Temperature input CVN Energy Computed CVN Energy Differential 0 17 17.18 -38 25 24 2226 L73 50 24 28 -4 50 23 28 -5 75 47 3439 12 3 150 43 5149 -10.4 9 200 68 62B 519 300 75 7L77 322 SUM of E51DUAIS = 238 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLXNCE PROGRAM
C-86
. CAPSULE R
~
CVGRAPH 4.111yperbolic Tangent Curve Printed at 15:3148 on 10-24-1998 Page1 Coefficients of Cune 1 A = 4155 B = 4055 . C = 2(P.14 TO = 150 Equation is 11 = A + B
- i tanh((T - TO)/C) i Upper Shelf II: 82.1 Temperature at II 3fx 11 7 Imer Shelf II: 1 Fired Materiah WELD Heat Number:132 Orientation: I Capsule R Total Fluence '
200 i
m
.i
.- 150 a
M M 100 5 # l Te a So -
0
/
U
-300 -200 -100 0 100 200 300
. 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap:R Materiah WELD Ori: Heat l: 1752 Charpy V-Notch Data Temperature Input lateral Erpansion Computed II Differential 0 115 15 S 8 -2.48 25 20 1924 .75 50 19
- 2228 -198 50 18 2298 -438 75 40 2738 1283 150 40 4155 -155 200 49 51 2 -238 300 68 67.11 2 SUM of RESIDUAIS -2 tas ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-87 CAPSULE R CVCPAPH 4J Hyperbolic Tangent Curve Printed at 15203 on 10-22-1996 Page1 Coefficients of Curve !
A = 50 B = 50 C : 12P4 10 : 11108 Equation is Shearx : A + B ' l tanh((T - 10)/C) l Temperature at $N Shear: 11 3 4
Materiah TELD Heat Number: 1752 Orientation:
Capsule R Total Fluence 100 g-a 80 a I ce e i t:
- su
< a ;
ce :
O g e =; '
a 0
20 y-u (
)
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: PI! Cap: R Materiah TELD Ori: Heat l: 1752 Charpy V-Notch Data Temperature input Percent Shear Computed Perant Shear Differential 0 17 13 S1 338 25 22 1916 233 50 22 2629 -429 50 20 2629 -629 75 40 34J12 5#1 150 67 6423 236 200 76 8053 -453 300 100 9149 45 SUM of PISIDUAIS = 3.03 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILtANCE PROGRAM -
C-88
. CAPSULE R
~ CVGRAPH 43 Hyperbolic Tangent Curve Printed at 16fE6 on 10-23-1996 Page1 Coefficients of Curve 1 A = 492 B = 47.4 C : 65.08 10 : -2168 F4 uation is CYN : A + B 'I tanh((T - TO)/C) l Upper Shel! Energy: W Fixed Temp. at 30 ft-lbs -503 Temp. at 50 ft-lbs -2L1 lower Shelf Energy: P.19 Fixed Material: HEAT AFFD 205T Heat Number. Orientation:
Capsule R Total Fluence 300 m 2su Q
Ja z em 4 h tw s.4 150 e
c a eel o 100 a -.
o A, so IJ A
U ,
) l l
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap: R Material:IIEAT AFFD ZONE Ori: Heat f:
Charpy V-Notch Data Temperature input CVN Energy Computed CVN Energy Differential
-W 20 10.72 9 71
-50 33 3018 231 0 53 6423 -1133 75 119 E37 2GE2 150 11 4 9651 17.48 150 l 92 9651 200 -451 96 96B9 250 -39 87 96Fl
-9FI SUM of RISIDUAIS = 2&96 4se ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RAD!ATION SURVEILLANCE PROGRAM .
. C-89 CAPSULE R CVCRAPH 41 Hyperbolic Tangent Curve Printed at IE4158 on 10-22-1996 Page!
Coefficients of Curve 1 A = 31E B = 306 C = 71D9 10 = -19.68 F4 uation is E = A + B
- l tanb((T - 70)/C) l Upper Shelf LE: 6221 Temperature at E 3E -IL7 lower Shelf E: 1 Fixed Materiah HEAT AWD ZONE Heat Number. Orientation:
Capsule R Total Fluence sou m i
- 150 a .
. x M 100 a D k
.3 "
m 7 o
(
a so a
A D
-300 -200 - 100 o too 200 aoo a soo soo Temperature in Degrees F Data Set (s) Plotted Plant: Pil . Cap: R Materiah HEAT AWD ZONE Ori: Heatf.
Charpy V-Notch Data Temperature Input lateral Expansion Computed E -
Differential
-W 11 724 175
-50 21 1929 L7 0 34 39El -5 51 75 67 5&22 E77 15 0 77 6L7 1529 150 595 6L7 -22 200 51 6208 -1108 250 54 6218 -al8 SUM of RESIDUAIS = 238 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
o C-90
. CAPSULE R
~ CVCRAPH 41 Hyperbolic Tangent Curve Printed at 16fD24 on 10-22-1996 Page!
Coefficients of Curve 1 A = 50 B=50 C = 954 11) = 2&l2 Equation is Shear /. : A + B ' l tanh((T - 11)}/C) 1 Temperatun at 50/'. Shear. 2E1 Materiah HEAT AITD ZONE Heat Numher: Orientation-Capsule: R Total Fluence:
100 w
80 co 0
.c
- so
< a ce O
5
- a /
0 20 a
J U , .
-300 -200 -100 0 100 200 300 400 500 600 Temperature in ' Degrees F Data ScFs) Plotted Plant: Pil Cap:R Materiah HfAT A}TD ZONE Ori: Heatf.
Charpy V-Notch Data Temperatun input Percent Shear Computed Pemnt Shear Diffenntial
-97 11 6.76 423
-50 23 1627 0 6.72 31 35S7
~5 -4E7 67 72.76 150
-5.76 100 9P.78 150 721 100 9178 721 200 100 7134 2fi5 250 100 99.05 .94 SUM of P251DUAIS = 123
~
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
f C-91 CAPSULE R O'CRAPH 43 Hyperbolic Tangeni Curve Printed at 16M56 on 10-23-1996 Page1 Coefficients of Curve 1 A = 44D9 B=4L9 C = E42 1D = 26&79 Equation is OH = A + B * [ tanh((T - 10)/C) ]
Uppe* Shelf Energy: 80 Fixed Temp. at 30 ft-lbs 2393 Temp. at 50 ft-lhe 280.4 lower Shelf Energy: 219 Fixed Material: SRM IEID2 Heat Numbs. SA533B1 Orientation: LT Capsule R Total Fluence 300 m aso !
Q I
$ 20u I
4 x tw 4 150 D
g -
N 100 Z a D )
su O
W) u ,
t ;
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap: R Material SRM !!SS11E Ori: LT Heatf.SA533B1 Charpy V-Notch Data Temperature input CVN Energy Computed 03 Energy Differential 15 0 9 6S4 235 200 10 15.48 -E48 225 30 2171 628 250 31 34.7 -3.7 300 62 5924 P75 350 72 7E74 -174 375 79 80M -lM 425 91 8414 6a5 SUM of PJSIDUALS = 423 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-92 CAPSULE R CVGRAPH 4j Hyperbolic Tangent Curve Printed at 16M32 on 10-23-1996 Page1 Coefficients of Curve 1 A = 40D3 B = 39D3 C = 144.46 ID = 31828 Equation is 12 = A + B ' [ tanb((T - 1D)/C) l Upper Shelf II: 79M Temperature at 12 35 2995 lower Shelf LE:1 Fixed Materiah SRM HSSID2 Heat Number. SA533B1 Orientation: LT Capsule R Total Fluenz axr en
.- 150
. 6 a
. 4 x
100 l i
2 > l 3ct Y A
/
n lo O
ga D
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: P11 Cap: R Materiah SRM IfSSID2 Ori: LT Heatf.SA533B1 Charpy V-Notch Data Temperature input lateral Expansion Computed 12 Differential 150 6 7.92 6 -L92 200 10 117 -17 225 23 1733 536 250 18 2234 -434 300 41 35J2 537 350 56 4a47 732 375 39 54S2 -15 I 2 425 70 6436 5.43 SUM of PEDUAIS =-P09 s
O ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM -
4 C-93 CAPSULE'R CVGRAPH 4J Hyperbolic Tangent Cune Printed at 1 Ell 48 on 10-23-1996 Page1 Coefficients of Curve 1 A = 50 B = 50 C = 84B8 % = 266.71 Equation is Shearx = A + B
- l tanh((T - M)/C) l l Temperature at 50x Shear. 26E7 Material: SEM HSS1112 Heat Number: SA533B1 Orientation LT Capsule: R Total fluence:
100 --
g
- 80 Cl5 CJ m 80
)m
- J c
c)
O 4
A oc o
as i o
U s 3 ( ) l
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted Plant: P11 Cap:R Materiah SRM IET02 Ori: LT Heat l: SA533B1 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential 15 0 13 6 6.99 200 23 1719 53 225 3? 2722 4.77 250 32 4027 -8 71 300 59 68 5 -95 350 100 87El 1232 375 100 SP.76 723 425 100 97E5 234 SUM of PEDUAIS = 21f23 to ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVElLLANCE PROGRAM
I C-94
- , CAPSULE S CVGPJfH 41 Hyperbolic Tangent Curve Printed at 15
- 1632 on 10-16-1996 Page1 e
Coefficients of Curve 1 A = 7234 B = 7015 C = 9828 TO = 13125 t Equation is CVN = A + B ' l tanb((T - 10)/C) l l Upper Shelf Energy:1425 Fird Temp. at 30 ft-lbs 625 Temp at 50 ft-lbs 988 Imer Shelf Energy: 219 Fixed i l Material: FORGING SA500 Heat Number. 21918/38566 Orientation LT 1
Capsule S
{
Total Fluence:
i
- t i
en 2su
,e
- I
.a x 20u
. . N '
en -
4 150 -
M g
V g o .
100 i Z l 0 ,
1 so a
_13nen I l
D i i i !
-300 -200 -100 0 100 200 000 400 500 800 Temperature in Degrees F Data Set (s) Plottal Plant: Pll Cap S Material FORGING SA5083 Ori: LT Heat f: 21918/38566 Charpy V-Notch Data Temperature input CVN Energy Computal CVN Energy - Differential
-25 9 75 119 25 10 16A8 -638 26 33 16 S 4 16D5 72 35 3453 .46 100 51 50.76 23 125 68 6729 1 150 74 8557 -1157 17 5 106 10126 4 31 250 136
. 131 439
- Data continued on next page "
em ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
. C-95 $
i
~
CAPSULE S Page2 Material: MPSLNG SA5083 IIeat Number. 21918/38566 Orientation: LT Capsule S Total Fluence Charpy V-Notch Data (Continued)
Temperature input GW Energy Computed DH Energy Differential 300 150 13811 1138 350 151 14038 10.11 400 133 14131 -891 SUN of PIS!DUAIS : 2221 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY P.RAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
o C-96 CAPSULE S -
CVCRAPH 4j Hyperbolic Tangent Curve Printed at 1334d5 on 10-16-1996
- 3 Page!
I Coefficients of Curve 1 A = 462 B = 452 C=10538 E = 11531 Equation is II = A + B ' I tanh((T - %)/C) l Upper Shelf II: 92.05 Temperature at II 3fx 88 !aer Shelf II: 1 Fired Materiah MRGING SA5083 Heat Number. 21918/38566 Orientation LT ;
Capsule S Total Fluence: I 20u m
O 150 a
M i
~
100 o o !
bw a J
o i l
e 1 a w
{
D 4
83 i
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted !
Plant: Pil Cap: S Materiah MRGING SA5083 Ori: LT Heat l: 21918/38566 l Charpy V-Notch Data Temperature input lateral Erpansion Computed II Differential
-25 4 69 -23 25 10 1436 -436 26 26 15DB 1031 72 29 28.77 2 100 37 3934 -2S4 12 5 51 50.7 29 150 54 6102 -7D2 13 77 699 7D9 250 93 85.52 7.47
Data continued on next page
AMI.YSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACiC3 VESSEL RADIATION SURVElt1ANCE PROGRAM
-- .~. . _ . - . ._. . . - - ._. -_
la C-97 CAPSULE S Page2 Material: FORGLNG SA5083 Heat Number: 21918/38566 Orientatiom LT Capmle S Total Fluence l Charpy V-Notch Data (Continued)
- Temperature Input lateral Expansion Computed LF. Differential 300 80 89.41 -9.41 350 90 91D1 -101 400 94 91.fi5 234 SUM of RESIDUAIS = .16 i
e ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM -
C-98 CAPSULE S CVGPRH 41 Hyperbolic Tangent Curve Printed at 133704 on 10-16-1996 Page!
Coefficients of Curve 1 A = 50 B = 50 C = 7128 11) = 16125 Equation is Shear /. = A + B ' [ tanh((T - 11))/C) ]
Temperature at 50x Shear.1612 Materiah MRGING SA5083 Heat Number. 21918/38566 Orientation LT Capsule S Total Fluence too 2 -
w a
- ce o
[
C D
su 4 M ce O
4 D
4 a
/
O O O O U
-300 -200 -100 o too 200 300 400 soo soo
. Temperature in Degrees F Data Set Plottal Plant: P!1 Cap:S Materiah MRGING(s)SA5083 Ori: LT Heat f. 21918/38566 Charpy V-Notch Data Temperature Input Perant Shear Computal Perunt Shear Differential
-25 5 53 4.46 25 5 214 25 26 10 2J9 73 72 10 72 2.44 100 15 15 2 -2 125 30 2656 3.43 15 0 35 4?J7 -7J7 175 60 5932 .47 250 100 9234 7ES
Data continued on next page
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEll!.ANCE PROGRAM .
C-99 i
CAPSULE S Page2
- Waterial: F0PCING SA5083 licat Number: 21918/38566 Orientation LT Capsule S Total Fluence Charpy V-Notch Data (Continued)
~
Temperature Input Percent Shear Computed Percent Shear Differential 300 100 98 339 350 100 995 .49 -
400 100 99El .12 SUM of RESIDUAIS = 2437 1 .
b.e l
I i
i i
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAlRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-100
- CAPSULE S
' CVCRAPH 41 Hyperbolic Tangent Curve Printed at 0807M on 10-2h1996 Page1
- Coefficients of Curve 1 A = 6859 B = 66.4 C=1002 % = 109.43 Equation is CVN = A + B ' [ tanh((T - W)/C) !
Upper Shelf Energy:135 Fixed Temp. at 30 ft-lbs 4?.9 Temp. at 50 ft-lbs 80.6 lower Shelf Energy: 219 fixed Material: ERGING SA5083 Heat Number: 21918/38566Orientation: T1, Capsule: S Total Fluence 300 l
m esu A
i I a
x am 4 1 tw i L 150- :
0 co l
c <
M 100
/
u O '
_)
su -
a 4
0 ,
1 1
-300 -200 -100 0 100 200 300 400 s00 600 Temperature in ' Degrees F Data Set Plant: Pil Cap: S Material: ERGING(s) SA500 Ori: TLPlotted Heat b 21918/38566 Charpy V-Notch Data Temperature Input CVN Energy Computed CVN Energy Differential
-25 7 10 S 6 -386 25 22 22 2 .92 50 19
.G22 -1422 60 49 3822 10.71 72 63 44 5 1814 100 59 6235 -335 125 69 7884 -934 17 5 97 106.79 -9.79 225 13 5 12301 IL96
" Data continued on next page "
~
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POV/ER COMPANY PRAIRIE ISLAND (kit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-101 CAPSULE S Page2 Material FORGING SA5083 IIeat Number. 21918/38566 Orientation: TL Capsule S Total Fluence Charpy V-Notch Data (Continued)
Temperature input CVN Energy Computed CVN Energy. Differential 250 138 127.45 1054 300 132 1212 .12 SUM of RESIDUAIS = 952 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLN4CE PROGRAM
C 102 CAPSULE S CVGP.APli 41 Hyperbolic Tangent Cune Printed at 15M11 on 10-16-1996 Page1 Coefficients of Curve 1 A = 4LD5 B = 40.05 C = 85.93 % = 8&l2 F4 uation is E : A + B ' [ tanh((T - N)/C) ]
Upper Shelf 11.: 8L11 Temperature at E 35: 75 lower Shelf 11.: 1 Fixed Material: FORCING SA5083 Heat Numben 21918/38566 Orientation: TL Capsule: S Total Fluence !
20u m
O 150 a
4 >t 100
- a
]to ea es a
M 50 a f" o
a U
s
\
-300 -200 - 100 0 100 200 300 400 500 600 l Temperature in Degrees F Data Set (s) Plotted Plant: PIl Cap:S Material:IDRGING SA5083 Ori: TL Heat f. 21918/38566 Charpy V-Notch Data Temperature input lateral Erpansion Computed LE Differential
-25 2 637 -437 25 14 15.98 -198 50 14 2436 -1036 60 36 2&39 7S 72 45 3162 1137 100 46 4656 -56 12 5 50 5726 -726 17 5 66 7L75 -5.75 225 90 77313 !?.06
= Data continued on next page -
ANALYS!S OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIR!E ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM -
C-103 l - . .
' CAPSULE S Page2 Material: IDRGING SA5083 Orientation: TI, liest Number: 21918/38566 l
Capsule S Total Fluence Charpy V-Notch Data (Continued)
Temperature Input lateral Expansion Computed I.E Differential 250 85 793 5S9 300 70 80M -1054 SUM of RISIDUAIS :-413 4
b anu ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMP /NY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLNJCE PROGRAM
C-104 i
CAPSULE S CVGRAPH 41 Hyperbolic Tangent Curve Printed at 15343! on 10-16-1996 Page1 Coefficients of Curve 1 A = 50 B = 50 C = 8143 TO = 134.47 Equation is Shear /. = A + B ' [ tanh((T - 1D)/C) ]
Temperature at 50x Shear.134.4 ,
Material IDRGING SA5083 Heat Number. 21918/38566 Orientation TL Capsules Total Fluence
== _
100 a au CO e
d m eu .
< a c
e O
4 5
A c ll 2u A
o -
-300 -200 - 10 0 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted
. Plant: Pil Cap: S Materiah IDRGING SA5083 Ori: TL Heat f. 21918/3566 Charpy V-Notch Data Temperature Input Percent Shear Computal Percent Shear Differential
-25 0 214 -214 25 '5 6.76 -1.76 50 10 11A6 -L66 60 20 1436 5f3 72 25 1827 E72 100 30 30.44 .44
% @ MM HM 17 5 65 7254 -754 225 - 100 89.75 1024
" Data continued on next page "
=
ANAt.YSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY. PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
_ . _ . . _ _ _ __- .. - - = - . _ - . -- .. ._ . -
s 1
C-105 CAPSULE S Page 2 Materiah It)RGING SA5083 lleat Number. 21918/38566 Orientation TL Capsule S Total Fluence Charpy V-Notch Data (Continued) y Tempenture Input Percent Shear Computed Percent Shear Differential 250 100 94.09 53 300 100 98.14 1.85 SUM of RISIDUAIS = !?.46 1
4 l
l l
j 1
i ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY.P.RAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
f.
C-106
. CAPSULE S CVGPJLPH 4J Hyperbolic Tangent Curve Printed at 142921 on 11-06-1996 Page1 Coefficients of Curve 1 i
A = 4334 B = 41.15 C = 9194 W = 12827 Equation is CVN = A + B '[ tanh((T - E)/C)l Upper Shelf Energy: 645 Fixed Temp. at 30 ft-lbs 95S Temp. at 50 ft-lbs 1439 Lower Shelf Energy:119 Fixed Materiah WELD Heat Number.1752 Orientation-
) Capsules Total Fluence
- i
! m a
,O
. I a
cr., em 4 x tw L 150 t
. c g .
- N 100 Z o i > F
- O [
u 1 m_ M n 1
-300 -200 -100 0 100 200 300 400 500 600
{ Temperature in Degrees F Data Set (s) Plotted Plant: PIl Cap: S Materiah TElb Ori: Heat f: 1752
, Charpy V-Notch Data Temperature Input CVN Energy Computed CVN Energy Differmtial
-25 8 143 256 25 17 10.76 623 72 24 21S 4 235 100 28 3156 -356 150 45 4
5251 -751 175 67 6L94 1 05
' 225 77 7432 217 300 92 8226 9.73 SUM of PIS!DUAIS = 1734 4
j
~
ANALYSIS OF CAPSULE S FROM THE NORTliERN STATES POWER COMPANY PRA!RIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEll1ANCE PROGRAM
C-107 4
CAPSULE S CVGRAPH 4J Hyperbolic Tangent Curve Printed at 1647:35 on 11-07-1996 Page1 Coefficients of Curve i A = 3828 8=3728 C=10519 1D = 142.03 Equation is 11 : A + B * [ tanh((T - TO)/C) ]
!!pper Shelf II: 7556 Temperature at II 35 1317 lower Shelf II: 1 Fired !
Materiah YllD Heat Numbec 1752 Orientation:
Capsule S Total Fluence !
200 m
O 150 b -
a 4
M 100 eo 4
.S 3 So
[
O T
u , ,
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees. F Data Set (s) Plotted Plant: Pil Cap: S Materiah YELD Ori: Heat f.1752 Charpy V-Notch Data Temperature input lateral Expansion Computed II Differential
-25 3 196 .98 25 12 827 172 72 16 1657 -57 100 24 2413 -J3 15 0 37 4LO9 -4D9 175 53 4959 34 2 64 628 ljg 300 71 72.03 -ID3 SUM of PISIDl'AIS = L49 O
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM -
C-108 ;
CAPSULE S CVCP#H 4J Hyperbolic Tangent Curve Printed at 1&4&34 on 11-07-1996
~
i Page1 l Coefficients of Curve 1 A = 50 B=50 C = 7573 E = 89.06 Equation is Shearx = A + B ' l tanh((T - W)/C) l Temperature at 50x Shean 89 Material YELD Heat Numben 1752 Orientation-Capsule S Total Fluence
} 7 a m -
9 m
E so
, 4 a 7
- c
~
o O
% 4U e
4 1
1 a>
0 0 o ,
J ,
l
-300 -mo -100 o too 200 mo 400 soo 800 l Temperature in Degrees F l Data Set (s) Plotted j Plant PIl Cap: S Materiah YELD Ori: Heat f.1752 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear Differential l
1 i
-25 10 4E8 531 I 25 10 1555 -555 4
72 40 38.92 IM 100 60 5717 222 J
150 80 1033 -333 175 90 90S3 -E3 225 100 9731 2S8 300 10 0 99S2 2 SUM of PJSIDUAIS : 170 I
d ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISl.AND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
C-109 l 4
CAPSULE S CVCRAPH 4J Hyperbolic Tangent Curve Printed at 09362 on 10-18-1996 Page1 Coefficients of Curve 1 A = 69D9 B = 66.9 C = 96.08 TO = L4 Equation is: CVN = A + B ' l tanb((T - TU)/C) l '
Upper Shelf Energy:136 Fixed Temp. at 30 ft-lbs -623 Temp at 50 ft-lbs -263 lower Shelf Energy: 2J9 Fixed Materiah HEAT AFFD ZONE Heat Number: Orientation-Capsule S Total Fluence 30u m 200 Q
T
$ 200
- ' x tw L 150. o 0
C f a M /
2 z n 0 / o Su J
U,
-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plotted
, Plant: PI! Cap.: S Materiah HEAT AFFD ZONE Ori: Heat f:
Charpy V-Notch Data Temperature input CYN Energy Computed CVN Energy - Differential
-100 20 16E5 334
-50 32 0
36 5/ -4M 82 6832 50 1387 58 10031 72 -4231 143 110R 175 32D2 149 13148 300 1651 123 13173 -1?.73 SUM of PJSIDUAIS = 633 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
=_;
s C-110 c
CAPSULE S CVCRAPH 4J Hyperbolic Tangent Curve Printed at 09:3938 on 10-18-1996 Page!
Coefficients of Curve 1 A = 4053 B = 3953 C = 106.41 TO = 75 Equation is E = A + B ' I tanh((T - 1D)/C) l
- Upper Shelf LE
- 80M Temperature at E 3fx -7.4 lower Shelf 2: 1 Fired Materiah HEAT AFFD ZONE Heat Numben - Orientation:
- Capsule S Total Fluence
wo 4
m
.~, 150 i
6 a
M 100 2 4, h o 8
%a so 7,
/
j U 4
-coo -200 -100 o too mo soo 400 soo soo Temperature in ' Degrees F Data Set (s) Plotted Plant: P!1 Cap: S Materiah HEAT AFFD ZONE Ori: Heat l:
Charpy V-Notch Data Temperatun Input lateral Erpansion Computed E Differential
-100 8 1025 -225
-50 16 2LO3 -5.03 0 50 37.75 1224 50 40 5553 -1553 72 71 6133 9.06 17 5 74 7631 -231
,300 m 79.74 225 SUM of PISIDUAIS = -208 ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAlRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEtLLANCE PROGRAM
f 4
C-111 CAPSULE S 1
CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 09.4207 on 10-18-1996
, Page1 l
- Coefficients of Curve 1 1
- A = 50 B = 50 C = 11&D9 !
- 11) = 4Ul5 Equation is Shearx = A + B ' l tanh((T - 11))/C) { i
! Temperature at 50x Shear. 413 Materiah HEAT AITD ZONE Heat Number. Orientatiorr Capsule S Total Iluence
, 10u L 80 i O a 0 C
CO 4
6u ,
i
! 4 4 C l O l i O W g O
j b i
i gg i
U !
, u l
-300 -200 - 10 0' O 100 200 300 400 500 800 i Temperature in Degrees F Data Set (s) Plotted Plant: Pil Cap: S Materiah HEAT AITD ZONE Ori: lleat f:
- Charpy V-Notch Data Temperatun Input Pen:ent Shear Computed Perant Shear Diffematial
! -100 10 829 17
- -50 25 17.42 757
- 0 30 2 98 -296 4
. M M MM -14
! 72 60 6249 -2.49 i 175 100 905- 9.49 300 100 9&75 124 SUM of PISIDUAIS = 11D9 4
- e 1
w I
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1
- REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM 1 _
t C-112 CAPSULE S CVCRAPH 4J Hyperbolic Tangent Curve Printed at 0&l002 on 10-29-1996 Page1 Cbefficients of Curve 1 A = 4234 B = 4015 C = 503 IV = 22828 Equation is CVN = A + B 'l tanh((T - 1D)/C) l Upper Shelf Energy: 825 Fixed Temp. at 30 ft-lbs 2122 Temp. at 50 ft-lbs 2373 lower Shelf Energy: ?19 Fixed Material: SRM ESSIV2 Heat Number: SA533B1 Orientation:
Capsule: S Total Fluenz 300 m 250
,C l
a x em X
tm
'L 15o O
g c .
100 Z g D
so v
/
o J
-soo -200 -100 o too 200 aoo 400 soo soo Temperature in Degrees F Plant: Pil Cap: S Material: SRW)Ori: HS5TEData Heatf.SA533B1 Set (s Pictted Charpy V-Notch Data Temperatun input CVN Energy Computed CVN Energy Differential N 9 E 2 I 200 21 2138 -31 206 19 25S4 -664 !
225 47 39.73 726 250 57 58S8 -168 2 % M 4 350 78 8126 -336 !
400 W 8P.41 458 I SUM of RIEIDUAIS = S3 !
Gen ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEll. LANCE PROGRAM l
4
' C-113 1
i i
This Page Was Intentionally Left Blank.
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l I
i i
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 ,
REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM i
< C-114 CAPSULE'S CVGRAPH U Hyperbolic Tangent Curve Printed at 14248 on 10-18-1996 Page!
Coefficients of Curve 1 A = 3026 B = 2926 C = 48D2 1V = 230 2 Faustion is E = A + B ' I tanh((T - TD)/C) l -
UP per Shelf G 592 Temperature at E 35c 23&4 lower Shelf 21 Fixed Material: SRM HSSIO2 Heat Number: SA533B1 Orientation-Capsules Total Fluence 200 m
O 150 6
a 4 x 100 ce 5a o CY a 50
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-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Data Set (s) Plottal Plant: Pil Cap: S Material: SRM HSSID2 Ori.: Heat l. SA533B1 Charpy V-Notch Data Temperature Input lateral Expansion Computal E Differential 15 0 9 2.96 6D3 200 14 11 71 22 206 13 16.4 4 -144 225 29 2684 P.15 250 41 41.46 .46 2
350 57 sc 2 64 5911 438 400 54 59.46 -546 SUM of RESIDUAIS = 146 8
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ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
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, C-115 CAPSULE S CVGRAPH 41 Hyperbolic Tangent Cune Printed at 15212 on 10-18-1996
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Coefficients of Curre 1 A = 50 B = 50 C = 6853 TO = 24934 Equation is: Shearx = A + B ' [ tanh((T - TO)/C) j Temperature at 50x Shear. 2493 i Material: SRM HSSIV2 Heat Number: SA533B1 Orientation:
Capsule: S Total Fluena:
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-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Plant: PIl Cap:S Material: SRM) Ori: HSSIV2 Heat Data Set (s Plotted
- f. SA533B1 Charpy V-Notch Data Temperature Input Perent Shear Computed Percent Shear . Diffenntial 150 15 514 935 200 20 1833 1D6 206 15 2L76 4 76 225 30 3232 -2S2 250 55 5011 438 300 80 812 -12 350 95 9439 .1 400 100 9&76 123 SUM of PJSIDUAIS = 654 O
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t D-0 APPENDIX D SURVEILLANCE DATA CREDIBILITY EVALUATION 1
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISt.AND Unit 1 REACTOR 'ASSEL RADIATION SURVEILLANCE PROGRAM
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D-1 INTRODUCTION: I Regulatory Guide 1.99, Revision 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently l used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, I Revision 2, describes the methodology for calculating the adjusted reference temperature and l Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data.
The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.
To date, there have been four surveillance capsules removed from the Prairie Island Unit 1 reactor vessel. This capsule data must be shown to be credible. In accordance with the discussion of Regulatory Guide 1.99, Revision 2, there are five requirements that must be met for the surveillance data to be judged credible.
The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Prairie Island Unit 1 reactor vessel surveillance data and determine if the Prairie Island Unit 1 surveillance data is credible.
l Criterion 1: Materials in the capsules should be those Judged most likely to be controlling with regard to radiation embrittlement.
l The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, l ' Fracture Toughness Requirements", December 19,1995 to be:
l
- the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron i radiation damage to be considered in the selection of the most limiting material with regard to radiation damage.' '
The Prairie Island Unit 1 reactor vessel consists of the following beltline region materials:
a) Intermediate shell forging C, heat number 21918/38566 l b) Lower shcell forging D, heat number 21887/38530 c) Circumferential weld wire UM 89, heat number 1752, UM 89 flux, batch number 1230 1
1 Per WCAP-80865, the Prairie Island Unit 1 surveillance program was based on ASTM E185-70, " Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels". Per Section 3.1.2 of ASTM E185-70, 'A minimum test program shall consist of specimens taken from the following locations (1) base metal of one heat, incorporated in the highest flux location of the reactor vessel, that has the highest initial ductile-brittle transition temperature, (2) wald metal, fully representative of fabrication practice used for the wcIds in the highest flux location of the reactor vessel, (weld wire or rod, and flux must come from one of the heats used in the highest flux region of the reactor vessel) and (3) the heat-affected zone of the weldments noted above.'
l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRA!RIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
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t D-2 l Therefore, at the time the Prairie Island Unit 1 surveillance capsule program was developed, j intermediate shell forging C was judged to be most limiting based on the above i recommendations and was utilized in the surveillance program.
l The surveillance program weld for Prairie Island Unit 1 was fabricated using the same heat of 4
weld wire used to fabricate the circumferential weld seam (heat 1752). The results of 4 mechanical property tests performed on the surveillance weld are considered to be j representative of the property changes expected in the reactor vessel beltline seams.
l Therefore, the materials selected for use in the Prairie Island Unit 1 surveillance program were
. those judged to be most likely controlling with regard to radiation embrittlement according to l the accepted methodology at the time the surveillance program was developed. The Prairie
- Island Unit 1 surveillance program meets this criteria.
i b Criterion 2: Scatter in the plots of Charpy energy versus temperature for the 1
Irradiated and unirradiated conditions should be small enough to permit i the determination of the 30 ft-Ib temperature and upper shelf energy, i unambiguously.
l Plots of Charpy energy versus temperature fcr the unirradiated condition are presented in
! WCAP-8086*, "Northem States Power Company Prairie island Unit No.1 Reactor Vessel l Radiation Surveillance Program," dated June 1973. Plots of Charpy energy versus temperature for the irradiated conditions are presented in Appendix C of this report for Capsules V, P, R and S.
Based on engineering judgement, the scatter in the data presented in these plots is small enough to determine the 30 ft-lb temperature and the upper shelf energy of the Prairie Island Unit 1 surveillance materials unambiguously. Therefore, the Prairie Island Unit 1 surveillance program meets this criteria.
Criterion 3: When there are two or mom sets of surveillance data from one reactor, the scatter of ARTuor values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28'F for welds and 17'F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fall this criterion for use in shitt calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82.
The least squares method, as described in Regulatory Position 2.1, will be utilized in determining a best-fit line for this data to determine if this criteria is met.
ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEltlANCE PROGRAM
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Table DI Prairie Island Unit 1 Surveillance Capsule Data Calculation of Best-Fit Line as Described in Position 2.1 of Regulatory Guide 1.99, Revision 2 Material Capsule f'8 FF" ART. FF x ART, FF' 2
(x) (y) (xy) (x )
Intent.ediate Shell V 0.563 0.839 24.07 20.19 0.704 Forging C (Axial) P 1.318 1.077 33.98 36.60 1.160 R 4.478 1.380 84.18 116.17 1.904 S 4.017 1.357 7427 100.78 1.841 V 0.563 0.839 56.36 47.29 0.704 Intermediate Shell P 1.318 1.077 23.11 24.89 1.160 Forghg C (Tangential) R 4.478 1.380 95.85 132.27 1.940 S 4.017 1.357 101.46 137.68 1.841 fw 9.306 49328 615.87 11.21B Weld Metal V 0.563 0.839 34.38 28.84 0.704 P 1.318 1.077 45.15 48.63 1.160 R 4.478 1.380 122.47 169.01 1.904 S 4.017 1.357 160.43 217.70 1.641 fy 4.653 362.0 464.18 ; 5.609 NOTES:
(a) I = Ruence (10" n/ctn', E > 1.0 MeV)
(b) FF = Fluence Factor = f*"'*H ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRA!RIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
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Per the 27B Edition of the CRC Standard Mathematical Tables (page 497), for a straight line fit by the method of
, squares, the values b,and b, are obtained by soMng the normal equatons i
e n b, + b, Ix, = Iy, and
- b,Ix,+ b,4 = Ixy, i These equations can be re-written as follows
, n n i,
E yi - an + bEx, i=1 ist 1 and i
1 n n- n j
[Xyg = 8{ X + b[ XjI 3 l=1 let tal l
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! Intermediate Shell Foraino C:
Based on the data provided in Table D1, these equations become:
1 i 1.) 493.28 = Ba + 9.306b or a = 61.66 - 1.16b
'. and 1
2.) 615.87 = 9.306a + 11.218b
- Thus, by substituting Eq.1 into Eq. 2, b = 107.1. Now, enter b (= 107.1)into Eq. I and a = 62.9. Therefore, the equation of the straight line which provides the best fit in the sense of least squares is
- Y' = 107.1 (X) - 62.9 j The error in predicting a value Y corresponding to a given X va!ue is
- e = Y - Y'.
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l ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
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l Table D2- L cit Evaluation for intermediate Forging Base Material (Orientation) ART. Best Fit ART. Scatter of ART.
FF (30 ft lb)(*F) (*F) ('F) ummummmmmm mmu-unumumummmmmmmmmmmum Intermediate Shell 0.839 24.07 27.0 -2.9 1.077 33.98 52.4 -18.4 1.340 44.18 84.9 -0.7 1.357 74.27 82.4 -8.1 Intermediate Shell 0.839 56.36 27.0 29.4 1.077 23.11 52.4 -29.3 angenti 1.380 95.85 84.9 11.0 1.357 101.46 82.4 19.1 I The scatter of ART, values about a best-fit line drawn, as described in Regulatory Position 2.1, should be less than 17 F for base metal. However, even if the fluence range is large, the scatter should not exceed twice this value (34 F).
As shown above, the error is within 34 F of tha best-fit line. Therefore, this criteria is met for the Prairie Island Unit 1 surveillance forging material.
i Weld Metal: I Based on the data provided in Table D1 the equations become: i 1.) 362.43 = 4a + 4.653b or a : 90.61 1.163b and 2.) 464.18 = 4.653a + 5.609b Thus, by substituting Eq.1 into Eq. 2, b = 216.7. Now, enter b (= 216.7) into Eq.1 and a = 161.5. Therefore, the equation of the straight line which provides the best fit in the sense of least squares is:
Y' = 216.7 (X) - 161.5 The error in predicting a value Y corresponding to a given X value is: e = Y - Y' ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES N!WER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEll'. mF PAGGA AM
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Table D3: Best Fit Evaluation for Weld Metal Base Material ART , Best Fit ART, Scatter of ART, FF (30 ft-lb)(*F) ('F) (*F) mummrsummmmmmmmmsmummmmmmmmmmmmmmemummmmmmmumummmmu Weld Metal 0.839 34.38 20.3 14.1 1.077 45.15 71.9 -26.8 l
. 1 1.380 122.47 137.5 15.0 1
1.357 160.43 132.6 27.8 The scatter of ARTem values about a best-fit line drawn, as described in Regulatory Position 2.1, should be less than 28'F. However, even if the fluence range is large, the scatter should not exceed twice this value (56 F). As shown above, the error is within 56 F of the best-fit line. Therefore, this criteria is met for the Prairie island Unit 1 surveillance weld material.
1 1
s Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding / base metalInterface within +f 25*F.
The Prairie Island Unit 1 capsule specimens are located in the reactor between the thermal shield and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the thermal shield. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions and will not differ by more than 25'F.
Criterion 5: The surveillance data for the corcelation monitor materialin the capsule should fall within the scatter band of the data base for that material.
Correlation monitor material was supplied by the Oak Ridge National Laboratory from plate material used in the AEC-sponsored Heavy Section Steel Technology (HSST) Program. This material, which was obtained from a 12-inch thick A533 Grade B Class 1 plate (HSST Plate 02), was provided to Subcommittee 11 (of ASTM Committee E 10 on Radioisotopes and Radiation Effects) to serve as correlation monitor material in reactor vessel surveillance programs.
The plate was produced by the Lukens Steel Company and heat treated by Combustion Engineering, Inc.
Figure D1 contains a plot of the residual (measured shift minus Regulatory Guide 1.99, Revision 2 shift) versus capsule fluence data. The plot shows the Prairie Island Unit 1 datr. 2s solid points. The data has been shifted such that the mean value is at zero and the two sigma bound at 45 F. All of the Prairie Island Unit 1 correlation monitor material data l falls within the two-sigma scatter band of the A533 Grade B Class 1 data per this criterion.
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Conclusion:
Based on the preceding responses to the criteria of Regulatory Guide 1.99, Revision 2, Section B, and the application of engineering judgement, the Prairie Island Unit 1 surveillance weld metal data is credible.
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ANALYSIS OF CAPSULE S FROM THE NORTHERN STATES POWER COMPANY PRAIRIE ISLAND Unit 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
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