ML20065D806
| ML20065D806 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 02/28/1994 |
| From: | Madeyski A, Terek E, Zawalick S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20065D787 | List: |
| References | |
| WCAP-13949, NUDOCS 9404070239 | |
| Download: ML20065D806 (200) | |
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l-l WESTINGHOUSI! CLASS 3 (Non-Proprietary)
ANALYSIS OF CAPSULE V SPECIMENS AND DOSIMETERS AND ANALYSIS OF CAPSULE Z DOSIMETERS FROM THE DUKE POWER COMPANY MCGUIRE UNIT I REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM l
l l
E. Terek l
! S. L. Zawalick A. Madeyski P. A. Peter l Febniary 1994 Work Perfonned Under Shop Order DXBP-106A Prepared by Westinghouse Electric Corporation for the Duke Power Company Approved by: /
TT A. Mcber, Manaher Structural Reliability and Plant Life Optimization WESTINGIIOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 0 - 1994 Westinghouse Electric Corporation
' PREFACE This report has been technically reviewed and verified.
Reviewer Sections 1 through 5,7,8, Appendix A. J. M. Chicots v Appendix B and Appendix C Section 6 E. P. Lippincott 4,
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l Appendix D B. A. Bishop d uri M O, I
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TABLE OF CONTENTS Section Title Page 1.0
SUMMARY
OF RESULTS 1-1
2.0 INTRODUCTION
2-1
3.0 BACKGROUND
3-1
4.0 DESCRIPTION
OF PROGRAM 4-1 5.0 TESTING OF SPECIMENS FROM CAPSULE V 5-1 5.1 Overview 5-1 5.2 Chagy V-Notch Impact Test Results 5-3 5.3 Tension Test Results 5-5 5.4 Compact Tension Specimen Tests 5-6 5.5 Bend Bar Specimen Tests 5-6 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6-1 6.1 Introduction 6-1 6.2 Discrete Ordinates Analysis 6-2 6.3 Neutron Dosimetry 6-6 6.4 Projections of Pressure Vessel Exposure 6-10 7.0 SURVEILLANCE CAPSULE REMOVAL SCIEDULE 7-1
8.0 REFERENCES
8-1 APPENDIX A LOAD-TIME RECORDS FOR CHARPY SPECBEN TESTS APPEMDlX B IEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION APPENDIX C UPPER SHELF ENERGY EVALUATION APPENDIX D JUSTIFICATION FOR USING DIABLO CANYON UNIT 2 SURVEILLANCE WELD DATA FOR THE PREDICTION OF TIE MCGUIRE UNIT 1 LOWER SIELL LONGITUDINAL WELD SEAM METAL MECHANICAL PROPERTIES il l
LIST OF TABLES Title Page Table 4-1 Chemical Composition (wt%) of the McGuire Unit 1 Reactor Vessel 4-3 Surveillance Materials 4-2 Chemical Composition of Four McGuire Unit 1 Charpy Specimens Removed 4-4 from Surveillance Capsule V 4-3 Heat Treatment of the McGuire Unit 1 Reactor Vessel Surveillance Materials 4-5 4-4 Chemistry Results from the NBS Certified Reference Standards 4-6 4-5 Chemistry Results from the NBS Certified Reference Standards 4-7 5-1 Charpy V-notch Data for the McGuire Unit 1 Intermediate Shell Plate 5-7 2
B5012-1 Irradiated at 550'F to a Fluence of 2.186 x 10" n/cm (E > 1.0 MeV) (Longitudinal Orientation) 5-2 Charpy V-notch Data for the McGuire Unit 1 Intermediate Shell Plate 58 2
B5012-1 Irradiated at 550 F to a Fluence of 2.186 x 10" n/cm (E > 1.0 MeV) (Transverse Orientation) 5-3 Charpy V-notch Data for the McGuire Unit 1 Surveillance Weld Metal 5-9 2
Irradiated at 550*F to a Fluence of 2.186 X 10" n/cm (E > 1.0 MeV) 5-4 Charpy V-notch Data for the McGuire Unit 1 Heat-Affected-Zone (HAZ) Metal 5-10 2
Irradiated at 550 F 1 Fluence of 2.186 X 10" n/cm (E > 1.0 MeV) 5-5 Instrumented Charpy impact Test Results for the McGuire Unit 1 5-11 Intermediate Shell Plate B5012-1 Irradiated at 550 F to a Fluence of 2.186 x 10" n/cm2 (E > 1.0 MeV)(Longitudinal Orientation) lii
LIST OF TABLES (continued)
Table Title. Page 5-6 Instrumented Charpy Impact Test Results for the McGuire Unit 1 5-12 Intermediate Shell Plate B5012-1 Irradiated at 550*F to a Fluence of 2
2.186 x 10" n/cm (E > 1.0 Mev) (Transverse Orientation) 5-7 Instrumented Charpy Impact Test Results for the McGuire Unit 1 5-13 Surveillance Weld Metal Irradiated at 550 F to a Fluence of 2
2.186 x 10" n/cm (E > 1.0 MeV) l 5-8 Instmmented Charpy Impact Test Results for the McGuire Unit 1 5-14 l
Surveillance Heat-Affected-Zone (HAZ) Metal Irradiated at 550 F to a 2
Fluence of 2.186 x 10" v/cm (E > 1.0 MeV) 2 5-9 Effect of 550 F Irradiation to 2.186 x 10" n/cm (E > 1.0 MeV) on the 5-15 Notch Toughness Pmperties of the McGuire Unit 1 Reactor Vessel Surveillance Materials 5-10 Comparison of the McGuire Unit 1 Surveillance Material 30 ft-lb Transition 5-16 l
Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2 Predictions 5-11 Tensile Properties of the McGuire Unit 1 Reactor Vessel Surveillance 5-17 2
Materials Irradiated at 550 F to 2.186 x 10" n/cm (E > 1.0 MeV) 6-1 Calculated Fast Neutron Exposure Rates at the Surveillance Capsule Center 6-14 6-2 Calculated Azimuthal Variation of Fast Neutron Exposure Rates at the 6-15 Pressure Vessel Clad / Base Metal Interface 6-3 Relative Radial Distribution of $(E > 1.0 MeV) within the Pressure Vessel 6-16 Wall iv
i LIST OF TABLES (continued)
Table Title Pace 64 Relative Radial Distribution of $(E > 0.1 MeV) within the Pressure Vessel 6-17 Wall 6-5 Relative Radial Distribution of dpa/sec within the Pressure Vessel Wall 6-18 6-6 Nuclear Parameters used in the Evaluation of Neutron Sensors 6-19 6-7 Monthly Thermal Genention During the First Eight Fuel Cycles of the 6-20 McGuire Unit 1 Reactor 6-8 Measured Sensor Activities and Reaction Rates Surveillance Capsule U 6-21 Saturated Activities and Derived Fast Neutron Flux 6-9 Measured Sensor Activities and Reaction Rates Surveillance Capsule X 6-22 Saturated Activities and Derived Fast Neutron Flux 6-10 Measured Sensor Activities and Reaction Rates Surveillance Capsule V 6-23 Saturated Activities and Derived Fast Neutron Flux 6-11 Measured Sensor Activities and Reaction Rates Surveillance Capsule Z 6-24 Saturated Activities and Derived Fast Neutron Flux 6-12 Summary of Neutron Dosimetry Results Surveillance Capsules U, X V, and Z 6-25 6-13 Comparison of Measured and Ferret Calculated Reaction Rates at the 6-26 Surveillance Capsule Center Surveillance Capsules U, X, V, and Z 6-14 Adjusted Neutron Energy Spectmm at the Center of Surveillance Capsule V 6-27 6-15 Adjusted Neutron Energy Spectrum at the Center of Surveillance Capsule Z 6-28 y
LIST OF TABLES (continued)
Table Title Page s 6-16 Ccmparison of Calculated and Measured Neutron Exposure Levels for McGuire 6-29 Unit 1 Surveillance Capsules U, X, V, and Z 6-17 Neutron Exposure Projections at Key Locations on the Pressure Vessel 6-30 Clad /Dase Metal Interface 6-18 Neutron Exposure Values 6-31 Curves 6-19 Updated Lead Factors for McGuire Unit 1 Surveillance Capsules 6-32 7-1 McGuire Unit ! Reactor Vessel Surveillance Capsule Withdrawal Schedule 7-1 l-f vi
LIST OF ILLUSTRATIONS Figure Title Pace 4-1 Arrangement of Surveillance Capsules in the McGuire Unit 1 Reactor Vessel 4-8 4-2 Diagram Showing the Location of Specimens, Thennal Monitors and Dosimeters 4-9 5-1 Charpy V-Notch Impact Properties for McGuire Unit 1 Reactor Vessel 5-18 Intermediate Shell Plate B5012-1 (Longitudinal Orientation) 5-2 Charpy V-Notch Impact Properties for McGuire Unit 1 Reactor Vessel 5-19 Intermediate Shell Plate B5012-1 (Transverse Orientation) 5-3 Charpy V-Notch Impact Properties for McGuire Unit 1 Reactor Vessel 5-20 Surveillance Weld Metal 5-4 Charpy V-Notch Impact Properties for McGuire Unit 1 Reactor Vessel 5-21 Weld Heat-Affected-Zone Meud 5-5 Charpy Impact Specimen Fracture Surfaces for McGuire Unit 1 Reactor 5-22 Vessel Intermediate Shell Plate B5012-1 (Longitudinal Orientation) l l
5-6 Charpy Impact Specimen Fracture Surfaces for McGuire Unit 1 Reactor 5-23 Vessel Intermediate Shell Plate B5012-1 (Transverse Orientation) 5-7 Charpy impact Specimen Fracture Surfaces for McGuire Unit i Reactor 5-24 Vessel Surveillance Weld Metal 5-8 Charpy impact Specimen Fracture Surfaces for McGuire Unit 1 Reactor 5-25 Vessel Heat-Affected-Zone Metal 5-9 Tensile Pmperties for McGuire Unit 1 Reactor Vessel Intermediate Shell 5-26 Plate B5012-1 (Longitudinal Orientation) vii
i LIST OF ILLUSTRATIONS (continued)
Ficure Title Pace 5-10 Tensile Properties for McGuire Unit 1 Reactor Vessel Intermediate Shell 5-27 Plate B5012-1 (Transverse Orientation) 5-11 Tensile Properties for McGuire Unit 1 Reactor Vessel Surveillance Weld 5-28 Metal 5-12 Fractured Tensile Specimens from McGuire Unit 1 Reactor Vessel 5-29 Intermediate Shell Plate B5012-1 (Longitudinal Orientation) 5-13 Fractured Tensile Specimens from McGuire Unit 1 Reactor Vessel 5-30 Intermediate Shell Plate B5012-1 (Transverse Orientation) 5-14 Fractured Tensile Specimens from McGuire Unit 1 Reactor Vessel 5-31 Surveillance Weld Metal 5-15 Engineering Stress-Strain Curves for Intermediate Shell Plate 35012-1 5-32 Tensile Specimens DL4 and DL5 (Longitudinal Orientation) 5-16 Engineering Stress-Strain Curve for Intermediate Shell Plate B5012-1 5-33 Tensile Specimen DL6 (Longitudinal Orientation) 5-17 Engineering Stress-Strain Curves for Intennediate Shell Plate B5012-1 5-34 Tensile Specimens DT4 and DT5 (Transverse Orientation) 5-18 Engineering Stress-Strain Curve for Intermediate Shell Plate B5012-1 5-35 Tensile Specimen DT6 (Transverse Orientation) 5-19 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens 5-36 DW4 and DW5 5-20 Engineering Stress-Strain Curve for Weld Metal Tensile Specimen DW6 5-37 viii
LIST OF ILLUSTRATIONS (continued)
Figure Title Page l 6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-12 6-2 Axial Distribution of Neutron Fluence (E > 1.0 MeV) Along the 6-13 45 Degree Azimuth ix
SECTION 1.0
SUMMARY
OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule V, the third capsule l
to be removed from the McGuire Unit I reactor pressure vessel and the analysis of the dosimeters contained in capsule Z, the fourth capsule to be removed from the McGuire Unit I reactor vessel, led to the following conclusions:
o Capsule V received an average fast neutron fluence (E > 1.0 MeV) of 2.186 x 10" n/cm2 and carsule Z received an average fast neutron fluence (E > 1.0 MeV) of 2.285 x 10" n/cm2 after 7.24 effective full power years (EFPY) of plant operation.
o Irradiation of the capsule V reactor vessel intermediate shell plate B5012-1 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal orientation), to 2.186 x 10" n/cm2 (E > 1.0 MeV) resulted in a 30 ft-Ib transition temperature increase of 85 F and a 50 ft-lb transition temperature increase of 90*F. This results in an irradiated 30 fr-lb transition temperature of 90*F and an irradiated 50 ft-lb transition temperature of 125 F for the longitudinally oriented specimens.
o Irradiation of the capsule V reactor vessel intermediate shell plate B5012-1 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major rolling direction of the plate (transverse orientation), to 2.186 x 10" n/cm2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 85 F and a 50 ft-Ib transition tempenture increase of 60 F. This results in an irradiated 30 ft-lb transition temperature of 85 F and an irradiated 50 ft-lb transition temperature of 135 F for transversely oriented specimens.
o Irradiation of the capsule V weld metal Charpy specimens to 2.186 x 10" n/cm2 (E > 1.0 MeV) resulted in a 30 h-lb transition temperature increase of 175 F and a 50 ft-lb transition temperature increase of 190 F. This results in an irradiated 30 ft-lb transition temperature of 170 F and an irradiated 50 ft-lb transition temperature of 210#F.
o Irradiation of the capsule V weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 2
2.186 x 10' n/cm (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 135 F and a 50 ft-lb transition temperature increase of 125'F. This results in an irradiated 30 ft-lb transition temperature of 85 F and an irradiated 50 ft-Ib transition temperature of 120 F.
l-1
o The average upper shelf energy of the capsule V intennediate shell plate B5012-1 Charpy specimens (longitudinal orientation) resulted in an average energy decrease of 15 ft-lbs after irradiation to 2.186 x 10" n/cm2(E > 1.0 MeV). This results in an irradiated average upper shelf energy of 125 ft-lbs for the longitudinally oriented specimens.
o The average upper shelf energy of the capsule V intennediate shell plate B5012-1 Charpy specimens (transverse orientation) resulted in an average energy decrease of 7 ft-lbs after irradiation to 2.186 x 10" n/cm2(E > 1.0 MeV). This msults in an irradiated average upper shelf energy of 94 ft-lbs for the transversely oriented specimens.
o The average upper shelf energy of the capsule V weld metal Charpy specimens resulted in 2
an average energy decrease of 40 ft-lbs after irradiation to 2.186 x 10" n/cm (E > 1.0 MeV). This results in an irradiated average upper shelf energy of 72 ft lbs for the weld metal specimens.
l o The average upper shelf energy of the capsule V weld HAZ metal Charpy specimens 2
decreased 35 ft-lb after irradiation to 2.186 x 10" n/cm (E > 1.0 MeV). This results in an irradiated average upper shelf energy of 83 ft-lbs for the weld HAZ metal.
I o The surveillance Capsule V test results indicate that the surveillance material 30 ft-lb transition temperature increases and average upper shelf energy decreases are less than the Regulatory Guide 1.99, Revision 2 predictions.
o The calculated end-of-license (32 EFPY) maximum neutron fluence (E > 1.0 MeV) for the McGuire Unit I reactor vessel based on the capsule V and Z dosimeters is as follows:
Vessel inner mdius' = 2.016 x 10" n/cm 2 2
Vessel 1/4 thickness = 1.101 x 10" n/cm 2
Vessel 3/4 thickness = 2.359 x 10" n/cm
- Clad / base metal interface 1-2
SECTION
2.0 INTRODUCTION
This report presents the results of the examination of the Capsule V test specimens and dosimeters-and the results of the examination of the dosimeters from Capsule Z. These capsules are the third and founh capsules to be removed imm the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Duke Power Company McGuire Unit I reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Duke Power Company McGuire Unit I reactor pressure vessel materials was designed and reconunended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the stactor vessel materials is presented in WCAP-9195, " Duke Power Company William B. McGuire Unit No.1 Reactor Vessel Radiation Surveillance Program"Ul The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-73, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels" Capsules "V" and "Z" were removed from the reactor after 7.24 EFPY of exposure and shipped to the Westinghouse Science and Technology Center Hot Cell Facility, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile specimens from capsule "V" was performed.
This repon summarizes the testing of and the postirradiation data obtained from surveillance capsules "V" and "Z" removed from the Duke Power Company McGuire Unit I reactor vessel and discusses the analysis of the data. For capsule "Z" only the dosimeters were analyzed at this time and the surveillance specimens were placed in storage at the Westinghouse Science and Technology Center Hot Cell Facility.
l l ..
9 d
2-1
SECTION
3.0 BACKGROUND
The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical pre ' ties of low alloy, ferritic pressure vessel steels such as SA533 Grade B Class 1 (base raa'm.u of the McGuire Unit I reactor pressure vessel) are well documented in the literatum. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness during high-energy irradiation.
A method for performing analyses to guard against fast fracture in reactor pressure vessels has been 1
presented in " Protection Against Nonductile Failure," Appendix G to Section III of the ASME Boiler and Pressure Vessel Codem. The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RTsm).
RTsm is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208m) or the temperature 60 F less than the 50 ft-Ib (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major l working direction of the plate. The RTym of a given material is used to index that material to a
( reference stress intensity factor curve (Km curve) which appears in Appendix G to the ASME Codem. The Ks curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the Km curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress
~
intensity factors.
RTym ud, in tum, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material pmperties. The changes in mechanical properties of a given reactor pressure vessel steel due to irradiation can be monitored by a reactor surveillance program, such as the McGuire Unit I reactor vessel radiation surveillance program m , in which a ,
surveillance capsule is periodically removed from the operating nuclear reactor and the 3-1
encapsulated specimens tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTxm) due to irradiation is added to the initial RTsm to adjust the RTsm (ART) for radiation embrittlement. This ART (RTsm initial + ARTsm) is used to index the material to the Ka curve and, in tum, to set operating limits for the riuclear power plant that take into account the effects of irradiation on the reactor vessel materials.
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3-2 L __ __ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ . _
f SECTION
4.0 DESCRIPTION
OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the McGuire Unit I reactor pressure vessel core region (beltline) materials were inserted in the reactor vessel prior to e initial plant start-up. The six capsules were positioned in the reactor vessel between the neutron pads and the vessel wall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core. The capsules contain specimens made from intennediat? shell plate B5012-1 (Heat No. C-4387-2) and weld metal fabricated with 3/16-inch Mil B-4 weld filler wire (tandem wire heat numbers 20291 and 12008) Linde 1092 flux, lot number 3854, which is identical to that used in the actual fabrication of the intermediate shell longitudinal weld seams.
Capsules V and Z were removed after 7.241 effective full power years (EFPY) of plant operation.
These capsules contained Charpy V-notch, tensile, and 1/2T compact specimens (1/2T-CT) made from intermediate shell plate B5012-1 and submerged are weld metal identical to the intennediate shell longitudinal seams and Charpy V-notch specimens from the weld heat affected zone (HAZ) of intermediate shell plate B5012-2.
Test material obtained from the intermediate shell plate (after thennal heat treatment and forming of the plate) was taken at least one plate thickness from the quenched edges of the plate. All test specimens were machined from the 1/4 thickness location of the plate after performing a simulated postweld, stress-relieving treatment on the test material. Specimens from weld metal and HAZ metal were machined from a stress-relieved weldment joining intermediate shell plate B5012-2 and B5012-3. All heat-affected-zone specimens were obtained from the weld heat-affected-zone of intennediate shell plate B5012-2.
l Charpy V-notch impact and tension specimens were machined from intennediate shell plate B5012-1 in both the longitudinal orientation (longitudinal axis of the specimen parallel to the major rolling direction of the plate) and transverse orientation (longitudinal axis of the specimen nonnal to the major rolling direction of the plate). Charpy V-notch and tensile specimens from the weld metal were oriented such that the long dimension of the specimen was normal to the welding direction.
The notch of the weld metal Charpy specimens was machined such that the direction of crack propagation in the specimen is in the welding direction.
I 4-1
~
T .
1 Capsules V and Z, also, contained 1/2T-CT test specimens from intermediate shell plate B5012-1 machined in both the longitudinal and transverse orientations. 1/2T-CT test specimens from the weld metal were machined such that the simulated crack in the specimen would propagate in the direction of welding. All specimens were fatigue pre-cracked according to ASTM E399.
Each capsule contained one bend bar specimen from intermediate shell plate B5012-1. Th'e bend bar specimen was machined with the longitudinal axis of the specimen oriented nomial to the rolling direction of the plate such that the simulated crack would propagate in the rolling direction of the plate. The bend bar specimen was fatigue precracked according to ASTM E399.
)
The chemical composition and heat treaunent of the surveillance material is presented in Tables 4 through 4-3. The chemical analysis reponed in Table 4-1 was obtained from unirradiated material used in the surveillance programufand irradiated material from capsule W 'U . In addition, a I chemical analysis using Inductivelf Coupled Plasma Spectrometry was perfenned on one irradiated Charpy specimen fmm inten:nediate shell plate B5012-1 and three weld metal Chalpy specimens and is report in Table 4-2. The chemistry results from the NBS certified reference standards are reported in Table 4-4 and 4-5. J Capsules V and Z contained dosimeter wires of pure copper, iron, nickel, and aluminum-0.15.
weight percent cobalt wire (cadmium-shielded and unshielded). In addition, cadmium shielded j 23 dosimeters of neptunium (Np25') and uranium (U s) were placed in the capsule to measure the l integrated flux at specific neutron energy levels.
The capsules contained thermal monitors made fmm two low-melting-point eutectic alloys and l l
sealed in Pyrex tubes. These thennal monitors were used to define the maximum temperature attained by the test specimens during irradiation. The composition of the two eutectic 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)
The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in capsules V and Z is showr igure 4-2.
4-2 L
i TABLE 4-1
, Chemical Composition (wt%) of the.McGuire Unit 1 Reactor Vessel Surveillance Materials Intennediate Shell Plate B5012-1 Surveillance Weld Metald Westinghouse CE Analysis CE Analysis- Capsule U M' Element Analysis Analysis C -
0.21 0.10 - -
S -
0.016 0.008 -
N 0.003 -
0.008 -
h Co 0.016 -
0.014 -
Cu 0.087 0.I3N 0.21 0.20 Si - 0.23 0.24 0.23 Mo -
0.57 0.55 0.54 Ni -
0.60 0.88 0.91 Mn -
1.26 1.36 1.19 Cr 0.068 -
0.04 0.05 V 0.003 -
0.04 -
P -
0.010 0.011 0.010 Sn 0.007 -
0.007 --
B <0.003 -
<0.001 - +
Cb <0.001 -
<0.01 - i Ti 0.005 -
<0.01 -
W <0.001 -
<0.01 -
As 0.008 -
0.009 -
c Zr 0.003 -
<0.001 -
.Sb <0.001 -
0.0022 -
Pb 0.001 -
<0.001 - -
-l a) Suiveillance weld . specimens wem made of the same weld wire and flux as the intermediate shell longitudinal weld seams (Tandem weld wire heats 20291 and 12008 -
and Linde 1092 Flux Lot '3854)m b) Ladle analysis (Lukens Steel Co.) '
c) Chemical Analysis by Westinghouse on irradiated Charpy weld specimen DW-15 removed from capsule U.N l
4-3.
1 L . _. - .- - sh
1 TABLE 4-2 Chemical Composition of Four McGuire Unit 1 Charpy Specimens Removed from Suiveillance Capsule V Concentration in Weight Percent Base Metal Weld Metal Specimens Specimen Element DW-22 DW-24 DW-25 DT-25 Fe 95.848 95.880 95.851 96.002-Co 0.015 0.016 0.016 0.015 :
Cr 0.046 0.044 0.044 0.076-Cu 0.195 0.191 0.193 0.117 Mn 1.329 1.331 1.330 1.355 Mo 0.548 0.539 0.550 0.634 Ni 0.870 0.848 0.863 0.643' P 0.013 0.014 0.015 0.011 Ti 0.006 0.006 0.006 0.007 V 0.001 0.001 0.001 <0.001 l Al <0.022 <0.022 <0.022 <0.024 As <0.016 <0.015 <0.016 <0.018 B 0.004 0.004 0.003 0.005 ht 0.013 0.013 0.013 0.013 Sn <0.024 <0.025 <0.025 <0.027 -
W <0.039 <0.040 <0.040 <0.043 Zr <0.010 <0.010 <0.010 <0.011 C 0.096 0.100 0.100 0.219 S 0.0106 0.0096 0.0097 0.0133 Si 0.185- 0.180 0.188 0.216 Analyses Method of Analysis Metals ICPS, Inductively Coupled Plasma Spectrometry Carbon LECO Analyzer Sulfur LECO Combustion-titration Silicon ASTM Gravimetric 4-4 e
'j
)
-l TABLE 4-3 IIeat Treatment of the McGuire Unit 1 Reactor Vessel Sutveillance Materialsl0 Material Temperature ( F) Time (hr) Coolant Surveillance Austenitizing: 4 Water quenched Program 1600 50 Test Plate '
B5012-1 Tempered: 4 Air cooled 1225 25 Stress R.elief: 40 Fumace cooled 1150 25 Weldment Stress Relief: 40 Fumace cooled 1150 1 25 I
t I
4-5
'i k
(
TABLE 4-4 Chemistry Results fmm the NBS Certified Reference Standards ,
Low Alloy Steel: NIST Control Standard Concentration in Weight Percent NIST 361 NIST. 362 I Element Measured Cenified Measured Certified Fe 95.442 95.600 95.370 95.300 Co 0.033 0.032 0.335 0.300 Cr 0.722 0.694 0.321 0.300.
Cu ' O.046 0.042 0.570 0.500 Mn 0.682 0.660 1.104 1.040 Mo 0.208 0.190 0.068 0.068 Ni 2.054 2.000 0.620 0.590 P 0.020 0.014 0.038 0.041 Ti 0.022 0.020 0.030 0.084 V 0.008 0.011 0.041 0.040 Al <0.022 0.021 0.089 0.095 As 0.023 n017 0.104 0.092 B <0.002 0.000 0.004 0.003 Nb 0.023 0.022 0.016 0.290 Sn <0.024 0.010 <0.026 0.016 W <0.020 0.017 0.209 0.200 Zr <0.009 0.009 0.204 0.190 C 0.383 0.383 0.161 0.160 S NA 0.0140 0.0358 0.0360 Si NA 0.222 NA 0.39 NA - Not Analyzed Analyses - Method of Analysis :
Metals ICPS, Inductively Coupled Plasma Spectrometry Carbon LECO Analyzer -;
Sulfur LECO Combustion-titration Silicon ASTM Gravimetric ;
4-6 -
r
t TABLE 4-5 i
Chemistry Results from the NBS Certified Reference Standards Low Alloy Steel: NIST Control Standard Concentration in Weight Percent NIST 363 NIST 364 NIST 365 Element Measured Certified Measured Cenified Measured Cenified Fe 94.065 94.400 96.606 96.700 99.710 99.900 Co 0.052 0.M8 0.175 0.150 0.008 0.007 Cr 1.416 1.310 0.073 0.063 <0.022 0.007 Cu. 0.118 0.100 0.302 0.249 0.007 0.006 Mn 1.598 1.500 0.277 0.255 0.008 0.006 Mo 0.024 0.028 0.551 0.490 <0.006 0.005 Ni 0.339 0.300 0.155 0.144 0.044 0.M1 P 0.033 0.029 <0.012 0.010 <0.011 0.003 Ti 0.N8 0.050 0.257 0.240 0.008 0.001 V 0.340 0.310 0.122 0.105 <0.001 0.001 Al 0.286 0.24 <0.026 OJX)8 <0.025 0.001 As <0.019 0.010 0.052 0.052 <0.019 0.000-B <0.002 0.001 0.015 0.011 <0.002 0.000 Nb 0.050 0.N9 0.042 0.157 0.012 0.000 Sn 0.103 0.lN <0.030 0.008 <0.028 0.000 W <0.045 0.046 <0.048 0.100 <0.M4 0.000 Zr 0.035 0.049 0.069 0.068 <0.004 0.000 C NA 0.620 NA 0.870 NA 0.007-S NA 0.0068 0.0242 0.0250 NA 0.0055 Si NA 0.74 NA 0.065 NA 0.008 NA - Not Analyzed Analyses Method of Analysis Metals ICPS, Inductively Coupled Plasma Spectrometry Carbon LECO Analyzer Sulfur LECO Combustion-titration Silicon ASTM Gravimetric 4-7 i
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a l a Figure 4-2. Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeters .
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.J SECTION 5.0 TESTING OF SPECIMENS FROM CAPSULE V-l
.5.1 Overview i The Charpy V-notch specimens contained in capsule Z were placed in storage and not tested at this time. The specimens contained in capsule V were tested and the results are repotted here.
The post-irradiation mechanical testing.of the Charpy V-notch impact specimens and tensile -
specimens, removed from capsule V, was perfonned in the Remote Metalograhpic Facility at the Westinghouse Science and Technology Center. Testing was performed in accordance with 10CFR50, Appendices G and HW , ASTM Specification E185-82 M and Westinghouse Procedure MHL 8402, Revision 2 as modified be Westinghouse RMF Procedure 8102, Revision 1, and 8103, Revision 1.
Upon receipt of capsules V and Z at the hot cell laboratory, the specimens and spacer blocks were ca.efully removed, inspected for id:ntification number, and checked against the master list in WCAP-9195t u. No discrepancies were found.
Examination of the two low-melting point 579 F (304 C) and 590*F (310 C) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination,'the maximum temperature to which the test specimens were exposed was less than 579 F (304 C).
The Charpy impact tests were performed per ASTM Specification E23-92W and RMF Procedure 8103, Revision 1. on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy impact test machine is instrumented with a GRC 8301 instrumentation system, feeding information -
into an IBM XT Computer. With this system, load-time and energy-time signals can be recorded
~
in addition to the standard measurement of Charpy energy (Eo). From the load-time curve, the load - ,
of general yielding (Pay),'the time to general yielding (tcy), the maximum load (Pu), and the time j to maximum load (tu) can be determined. Under some test conditions, a sharp drop in load ;
indicative of fast fractun: was observed. The load at which fast fracture was initiated is identified as the fast fracture load (Pp), and the load at which fast fracture terminated is identified as the
. arrest load (PJ. 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 r .l energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack 5-1
(E,) is the difference between the total energy to fracture (Ep) and the energy at maximum load (Ed.
The yield stress (cy) was calculated from the three-point bend formula having the following expression:
cy = (Pay
- L) / {B * (W - a)2
- C] (1) where: L = distance between the specimen supports in the impact machine B = the width of the specimen measured parallel to the notch W = height of the specimen, measured perpendicularly to the notch a = notch depth The constant C is dependent on the notch flank angle (Q), notch root radius (p) and the type of loading (ie. pure bending or three-point bending). In three-point bending, for a Charpy specimen in which & = 45 and p = 0.010", Equation 1 is valid with C = 1.21. Therefore, (for L = 4W).
c3 = (Pay
- L) / [B * (W - a):
- 1.21] = (3.3
- Pay
- W) / [B * (W - a)2] (2)
For the Chalpy specimen. B = 0.394" W = 0.394" and a = 0.079" Equation 2 then reduces to:
oy = (33.3
- Pay) (3) whem oy is in units of psi and Por is in units of lbs. The flow stress was calculated from the
=
average of the yield and maxunum loads, also using the three-point bend formula. ;
Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-92M. The lateral expansion was measurcJ using a dial gage rig similar to that shown in the same specification.
Tensile tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specification E8-91" I and E21-79(1988)"", and RMF Pmcedure 8102, Revision 1. All pull rods, grips, and pins were made of Inconel 718. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test 5-2
Extension measurements were made with a linear variable displacement transducer extensometer.
The extensometer knife edges were spring-loaded to the specimen and operated through specimen -
failure. The extensometer gage length was 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-93n21, Elevated test temperatures were obtained with a three-zone electric resistance split-tube fumace with a 9-inch hot zone. All tests were conducted in air. Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperatures. Chromel-alumel thermocouples were positioned at center and each end of the gage section of a dummy specimen in each grip. In the test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range from room temperature to 550 F (288 C). During the '
actual testing, the grip temperatures were used to obtain desired specimen temperatures.
Experiments indicated that this method is accurate to 2 F. :
The yicld 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 criginal cross-sectional area. The final diameter and ~
final gage length were determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.
5.2 Charny V-Notch impact Test Results The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule V, which was irradiated to 2.186 x 10" n/cm2(E > 1.0 MeV), are presented in Tables 5-1 ,
through 5-8 and are compared with unitradiated results ul as shown in Figures %1 through 5-4 The . 1 transition temperature increases and upper shelf energy decreases for the Capsule V materials are summarized in Table 5-9.
Irradiation of the reactor vessel intermediate shell plate B5012-1 Charpy specimens oriented with-the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal orientation) to 2.186 x 10" n/cm2(E > 1.0 MeV) at 550*F (Figure 5-1) resulted in a 30 ft-lb
- transitions temperature increase of 85 F and a 50 ft-lb transition temperature increase of 90 F.-
i 5-3
This resulted in an irradiated 30 ft-lb transition temperature of 90 F and an irradiated 50 ft-lb transition temperature of 125 F (longitudinal orientation).
The average upper shelf energy (USE) of the intermediate shell plate B50121 Charpy specimens (longitudinal orientation) resulted in an energy decrease of 15 ft-lb after irradiation to 2.186 x 10" n/cm2 (E > 1.0 MeV) at 550 F. This results in an irradiated average USE of 125 ft-lb (Figure 5-1).
irradiation of the reactor vessel intermediate shell plate B5012-1 Charpy specimens oriented with the longitudinal axis of the specimen normal to the major rolling direction of the plate (transverse orientation) to 2.186 x 10" n/cm2 (E > 1.0 MeV) at 550 F (Figure 5-2) resulted in a 30 ft-lb transition temperature increase of 85 F and a 50 ft-lb transition temperature increase of 60 F. This results in an irradiated 30 ft-lb transition temperature of 85 F and an irradiated 50 ft-lb transition L temperature of 135*F (transverse orientation).
The average USE of the intermediate shell plate B5012-1 Charpy specimens (transverse orientation) -
2 resulted in an energy decrease of 7 ft-lb after irradiation to 2.186 x 10" n/cm (E > 1.0 MeV) at 550 F. This results in an irradiated average USE of 94 ft-lb (Figure 5-2). ;
~l 2
Irradiation of the surveillance weld metal Charpy specimens to 2.186 x 10" n/cm (E > 1.0 MeV) at 550 F (Figure 5-3) resulted in a 30 ft-lb transition temperature shift of 175*F and a 50 ft-lb transition temperature increase of 190 F. This results in an irradiated 30 ft-lb transition -q
. temperature of 170 F and an irradiated 50 ft-lb transition temperature of 210 F. .I l
The average USE of the surveillance weld metal resulted in an energy decrease of 40 ft-Ib after .
irradiation to 2.186 x 10" n/cm2(E > 1.0 MeV) at 550 F. This resulted in an irradiated average i
USE of 72 ft-lb (Figure 5-3).
Irradiation of the reactor vessel weld HAZ metal Charpy specimens to 2.186 x 10" n/cm2 (E > 1.0 MeV) at 550*F (Figure 5-4) resulted in a 30 ft-lb transition temperature increase of 135 F -
and a 50 ft-lb transition temperature increase of 125*F. This resulted in an irradiated 30 ft-lb transition temperature of 85 F and an irradiated 50 ft-lb transition temperature of 120 F.
The average USE of the weld HAZ metal resulted in an energy decrease of 35 ft-lb after irradiation
. to 2.186 x 10" n/cm2 (E > 1.0 MeV) at 550 F. This resulted in an irradiated average USE of 83 ft-lb (Figure 5-4).
5-4
~- _ _ _ _ __-____ - - _ __________ -
1 l
The fracture appearance of each irradiated Charpy specimen from the various materials is shown in .
Figures 5-5 through 5-8 and show an increasingly ductile or tougher appearance with increasing test temperature.
A comparison of the 30 ft-lb transition temperature increases and upper shelf energy decreases for -
the various McGuire Unit I surveillance materials with predicted values using the methods of NRC ,
Regulatory Guide 1.99, Revision 2"U is presented in Table 5-10 and led to the following !
conclusions:
o The 30 ft-lb transition temperature increases for the capsule V surveillance materials are less than the Regulatory Guide 1.99, Revision 2 prediction.
o The upper shelf energy decreases of the capsule V surveillance materials are less than the Regulatory Guide 1.99, Revision 2 predictions.
5.3 Tension Test Results The results of the tension tests perfonned on the various materials contained in capsule V irradiated i 2
to 2.186 x 10" n/cm (E > 1.0 MeV) are presented in Table 5-11 and are compared with -
unirradiated resultstu as shown in Figures 5-9 through 5-11.
The results of the tension tests performed on the intermediate shell plate B5012-1 (longitudinal ;
2 orientation) indicated that irradiation to 2.186 x 10" n/cm (E > 1.0 MeV) at 550 F caused a 9 to 11 ksi increase in the 0.2 percent offset yield strength and a 7 to 10 ksi increase in the ultimate !
tensile strength when compared to unitradiated data"1 (Figure 5-9).
The results of the tension tests performed on the intennediate shell plate B5012-1 (transverse >
orientation) indicated that irradiation to 2.186 x 10" n/cm* (E > 1.0 MeV) at 550 F caused a 8 to 9 -
ksi increase in the 0.2 percent offset yield strength and a 4 to 8 ksi increase in the ultimate tensile strength 'when compared to unirradiated dataf 4 (Figure 5-10). 'i f
The results of the tension tests performed on the surveillance weld metal indicated that irradiation to 2.186 x 10" n/cm 2(E > 1.0 MeV) at 550 F caused a 20 to 26 ksi increase in the 0.2 percent offset yield strength and a 16 to 20 ksi increase in the ultimate tensile strength when compared to unirradiated data"1 (Figure 5-11).
5-5 l
l
. The fractured tension specimens for the intermediate shell plate B5012-1 material are shown in Figures 5-12 and 5-13, while the fractured tension specimens for the surveillance weld metal are shown in Figure 5-14.
The engineering stress-strain curves for the tension tests are shown in Figures 5-15 through 5-20.
5.4 Compact Tension Specimen Tests Per the surveillance capsule testing contract, the 1/2T compact tension (CT) specimens were not tested. The 1/2T CT specimens are being stored at the Westinghouse Science and Technology -
Center Hot Cell Facility, 5.5_ Bend Bar Specimen Tests Per the surveillance capsule testing contract, the bend bar specimen was not tested. The bend bar specimen is being stored at the Westinghouse Science and Technology Center Hot Cell facility.
5-6
TABLE 51 Charpy V-notch Data for the McGuire Unit 1 Intennediate Shell Plate B5012-1 2
Irradiated at 550 F to a Fluence of 2.186 X 10" n/cm (E > 1.0 MeV)
(Longitudinal Orientation)
Sample Temperature impact Energy Lateral Expansion ' Shear Number (F) (*C) (ft-lb) (J) (rnils) (mm) (%)
DL29 25 -4 9 0.13 7 5 5 DL17 60 16 18 24 18 0.46 10 DL23 75 24 34 46 -27 0.69 25 DL30 100 38 24 33 24 0.61 20 DL22 100 38 40 54 36 0.91 30 DL18 115 46 40 54 38 0.97 45 DL24 125 52 45 61 39 0.99 -45 DLl6 135 57 60 81 50 1.27. 65 DL25 150 66 73 99 55 1.40 70 DL21 175 79 84 114 66 1.68 80 DL20 200 93 90 122 69 1.75 85 DL19 250 121 115 156 86 2.18 100~-
DL28 300 149 122 165 84 '2.13 100 DL26 350 177 135 183 90 2.29 100 DL27 400 204 127 172 77 1.96 100 ym_ -
5-7
- i 9
TABLE 5-2 Charpy V-notch Data for the McGuire Unit 1 Intermediate Shell Plate B5012-1 2
Irradiated at 550 F to a Fluence of 2.186 X 10" n/cm (E > 1.0 MeV)
(Transverse Orientation)'
Sample - Temperature Impact Energy Lateral Expansion : Shear Number (F) ('C) (ft-lb) (J) . (mils) (mm) . (7c)
DT26 25 -4 12 16 13 0.33 10 DT27 50 10 24 33 24 0.61- 15 DT17 65 18 29 39 .25- 0.64 20 DT20 75 24 32 43 28 0.71 25 ,
DT25 100 38 33 45 28 0.71 25 DT23 115 ' 46 47 64 41. 1.04 35 >
DT29 125 52 52 71 ' 44 1.12 50 DT30 150 66 49 66 43 -1.09 55 DT16 175 79 62 84 49 -1.24 65 DT24 200 93 77 104 60 1.52 75 DT18 225 107 96 130 72 1.83 100 DT28 25 0 121 85 115 .75- 1,91 100 DT19 275 135 93 126 79 2.01 100 DT21 300 149 97 132 80 2.03 100 DT22 325 ,
163 99 134 80 2.03 100 5-8
L TABLE 5 Charpy V-notch Data for the McGuire Unit 1 Surveillance Weld Metal 2
Irradiated at 550 F to a Fluence of 2.186 X 10" n/cm (E > 1.0 MeV)
Sample Temperature Impact Energy Lateral Expansion - Shear Number ( F) ( C) (ft-lb) (J) . (mils) (nun) '(%)
DW21 65 18 4 5 6 0.15 10.
DW17 - 100 38 7 9 9 0.23 15 DW27 125 52 23 31 22 0.56 30 DW22 - 150 66 27 37 24 0.61 35 DW26 160 71 32 43 >
0.76 45 DW24 175 79 34 46 33 0.84 45 DW16 200 93 30 41 29 0.74 40 DW25 200 93 38 52 35 .0.89 60 DW19 215 102 60 81 52 1.32 95 DW29 225 107 57 77 49 1.24 90 DW28 250 121 73 99 60' l.52 100 DW18 300 149 64 87 61 1.55' 100 DW30 300 149 70 95 61 1.55 100-DW23 350 177 77 IN 70 1.78 100 DW20 400 2N 76 103. 66 1.68 100 5-9
m TABLE 5-4 Charpy V-notch Data for the McGuire Unit 1 Heat-Affected-Zone (HAZ) Metal 2
Irradiated at 550T to a Fluence of 2.186 X 10" n/cm (E > 1.0 MeV)
Sample Temperature impact Energy Lateral Expansion Shear Number ( 19 ( C) (ft-lb) (J) (mils) - (mm) - (%)
DH21 -25 -32 18 24 17 0.43 20 DH22 -5 -21 18 24 16 0.41 20 DH18 10 -12 7 9 9 0.23 10 DH19 50 10 -15 20 14 0.36 .15 DH27 65 18 29 39 27 0.69 - 35. ,
DH24 85 29 18 24 16 - 0.41 ~25-DH25 90 32 33 45 27 0.69 30 DH26 '100 38 43 58~ 43 1.09 50 DH16 125 52 53 72 46 .l.17 60 DH28 150 66 58 79 47 1.19 75 DH23 -200 93- M 87 53 1.35 85 :
DH30 250 121 70 95 56 1.42 100 DH17 300- 149 94 127 -- 78 1.98. 100 DH2O 350 177 72 98 62 1.57 100' DH29 375 191 94 127- 82 2.08 '100.
f 4
i 1
J 5 10
TABLE 5-5 Instrumented Giarpy impact Test Results for the McGuire Unit 1 Intennediate Stiell Plate B5012-1 Irradiated at 550 F to a Fluence of'2.186 X 10" n/cm2 (E > 1.0 MeV) (Longitudinal Orientation)
Nonnalized Energies 2
(ft-lb/in )
Time Time Fast Otarpy Yield to Max. to Fract. Arrest Yield Flow Test Energy Charpy Max. Prop. Load Yield Load Max. Load Load Stress Stress Sample ' Temp. Eo Er/A Eu/A E/A Pay tcy Pu tu Pp P3 oy (ksi)
No. ('F) (ft-lb) (Ibs) (psec) (Ibs) (psec) (Ibs) (Ibs) (ksi)
DL29 - 25 7 56 23 33 2819 0.12 2919 0.13 2903 212 94 95 DLl7 60 18 145 105 40 3340 0.17 3808 0.31 3795 245 111 119 DL23 75 34 274 195 79 3243 0.16 4095 0.48 4089 951 108 122 DL30" 100 24 193 - - - - - - ' - - - -
DL22 100 40 322 262 60 3216 0.16 4212 0.61 4209 531 107 123 DL18 115 40 322 223 99 3119 0.16 4023 0.55 4016 1548 104 119 DL24 125 45 362 266 97 3139 0.16 4167 0.63 4154 -1460 104 121 DLl6 135 60 483 289 194 3017 0.15 4112 0.68 4029 1381 100 118 DL25 150 73 588 290. 298 2987 0.15 4183 0.68 3746 2010 99 119 DL21 175 84 - 676 285 391 2970 0.16 '4052 0.68 3503 2045 99 117 DL20 200 90 - 725 283 441 2986 0.17 4079 0.69 3539 2125 99 117 DL19" 250 115 926 - - - - - - - - - -
DL28 300- 122 982- 273 709 2829 0.16 3919 0.67 *
- 94 112 DL26 350 135 1087 276 811 2854 0.17 3912 0.69 *
- 95 112 DL27 400 127 .;1023 266 757 2638 0.15- 3798 0.68 *
- 88 107
- Fully ductile fracture.
- Computer data not availabic due to computer malfunction.
- w. __ -.-_.: ._ . .- .- -_ -- _ __ . . . ._
TABLE 5-6 Instmmented Charpy Impact Test Results for the McGuin: Unit 1 Intennediate Shell Plate B5012-1 2
Irradiated at 550 F to a Fluence of 7.186 X 10" n/cm (E > 1.0 MeV) (Transverse Orientation) :
Nonnalized Energics 2
(ft-lb/in )
Time Time Fast Chagiy Yield to Max. to Fract. Arrest Yield . Flow Test Energy Charpy Max. Prop. Load Yield Load Max. Load Load Stress Stress Sampic Temp. En E/A i Es/A E/A Poy toy Pu tu P, P4 cy (ksi)
No. (*F) (ft-lb) (lbs) (psec) (Ibs) (pscc) (lbs) (Ibs) (ksi)
DT26 25 12 97 41 55 3408 0.16 3507 0.17 3498 364 113 115 DT27 50 24 193 130 64 3411 0.17 4020 0.35 4010 596 113 123 DTl7 65 29 234 186 47 3348 0.16 4183 0.46 4170 336 111 125 DT20 75 32 258 174 84 3323 0.17 4027 0.44 4024 878 110 122 DT25 100 33 266 177 89 3242 0.17 4026 0.46 4007 1174 108 121 DT23 115 47 378 287 91 3153 0.16 4141 0.67 4099 813 105 121~
DT29 125 52 419 284 135 3121 0.16 4265 0.65 4249 1768 104 123 DT30 150 49 395 230 175 3057 0.16 4008 0.55 3966 1946 102 117 DT16 175 62 499 277 222 3064 0.17 4020 0.67 3917 1857 102 118 DT24 200 77 620 260 360 3035 0.17 4028 0.63 3652 2341 101 117 r DTIS 225 96 773' 285 488 2899 0.15 4027 0.69 *
- 96 115 DT28 250 85 684 253 432 2896- 0.17 3877 0.63 *
- 96 112 ,
DT19 275 93 749 268' 481 2815 0.15 3992 0.65 *
- 94 113 DT21 300 97 781' 263 518 2745 0.14 3872 0.65 *
- 91 110 DT22 ' 325- 99 797 264 2684 *
- 533 0.15 3883 0.67 89 109
- Fully ductile fracture.
TABLE 5-7 -
Instmmented Charpy Impact Test Results for the McGuire Unit i Surveillance Weld Metal 2
Irradiated at 550 F to a Fluence of 2.186 X 10" n/cm (E > 1.0 MeV)
Nonnalized Energies (ft-lb/in') s Time Time Fast Charpy Yield to Max. to Fract. Arrest Yield Flow Test Energy Charpy Max. Prop. Load ' Yield Load Max. Load Load Stress Stress Sample Temp. En Er/A Eu/A Ei /A Pr o tar Pu tu Pg P3 cy (ksi)
No. (*F) (ft-lb) ; (Ibs) (psec) (lbs) (psec) (lbs) . (Ibs) (ksi) .
DW21 65 4 32 16 16 2275 0.10 2349 0.11 2346 213 76 77 DW17 100 7 56 32 24 3181 0.14 3294 0.16 3268 296 106 108 DW27 125 23 185 130 56 3407 0.17 3911 0.35 3892 835 113 122 DW22 150 27 217 139 78 3200 0.15 3929 0.38 3917 820 106 118
, DW26 160 32 258 178 79 3324 0.17 4024 0.46 4018 1662 110 122-w DW24 . 175 34 274 177 97 3226 0.16 3970 0.46 3967 1663 107 120 DW16 200 30 242 150 91 3433 0.20 3999 0.41 3995 1816 114 123 ,
DW25 '200 38 306 184 122 3218 0.16 3996 0.47 3980 961 107 120 DW19 215 60 483 210 273 -3163 0.16 4064 0.51 3965 -3031- 105 120 DW29 225 57 459 212 247 3309 0.18 4088 0.52 4001 2951-- 110 123 DW28 250 73 588 206 382 3109 0.16 4024 0.51 *
- 103 118 DW18 300 64 515 198 318 3126 0.17 3898 0.50 * -* 104 117 DW30 ~ 300 70 564 200 363 3082 0.16 3890- 0.51 * *
.102 116 d
DW23 350 77 .620 203 418 3001 0.16 3885 0.52 *
- 100 114 DW20 400 76 612 199 413 2838 0.15 3869 0.52 * *- 94 111
- Fully.. ductile fracture. .l f
- . = _ .. _ . . . . . . .-. . _ - - - _ --_-_ _ _ -
TABLE 5-8 Instrumented Charpy impact Test Results for the McGuire Unit i Surveillance Heal-Affected-Zone (llAZ) Metal 2
, Irradiated at 550 F to a Fluence of 2.186 X 10" n/cm (E > 1.0 MeV) 1 Normalized Energies 2
(ft-lb/in )
a Time Time Fast Charpy Yield to Max. to Fract. Arrest Yicid Flow
- Test Energy Charpy Max. Prop. Load Yield Load Max. Load Load Stress Stress Sample Temp. Eo En/A Eu/A Ep /A Por tar Pu tu Pp P3 or (ksi)
No. ( F) (ft-lb) (Ibs) (psec) (Ibs) (psec) (lbs) (Ibs) (ksi)
DIl21 -25 '18 145- 78 67 3834 0.17 3985 0.24 3978 854 127 130 Dil22 -5 18 145 78 67' 3856 0.1/ 3985 0.24 3979 826 128 130 Dill 8 10 7, 56 22 35 2672 0.11 2849 0.13 2829 212 89 92 DH19 50 15 121 59 62 34'82 0.15 3737 0.20 3727 433 116 120.
% DII27 65 29 234 88 146 3528 0.17 3691 0.27 3688 2247 117 120
- Dil24 85 18 145 88 57 3521 0.17 3820 0.27 3817 1144 117 122 Dil25 90 33 266 180 85 3554 0.17 4178 0.44 4175 916 118 128 Dil26 100 43 346 139 207 3322 0.16 3969 0.37 3962 2990 110 121 ,
L DH16 125 53 427 '217 210 3438 0.17 4179 0.52 3882 2162 114 126
, Dil28 150 58- 467 270 197 3284 0.16 4157 0.63 4141 1001 109 124 l-DH23 200 64 515 274 241 3242 0.16 4081 0.64 4065 2461 108- 122 DH30 250 70 564 204 360 3160 . 0.16 3944 0.51 *
- 105 118 I)
DHl7 300 94 757 279 478 2935 0.16 4044 0.67 *
- 97 116 DH20" 350 72'- 580 .- - - - - - - - -- -
DH29 375 94 757 270~ 487 2990 0.16 3880 0.67 *
- 99 114 l '
- Computer data not available due to computer malfunction.
- Fully ductile fracture.
l-E __ _ _. - _
5 TABLE 5-9 2
h Effect of 550 F Irradiation to 2.186 X 10" n/cm (E > 1.0 MeV) on the Notch Toughness Properties of the McGuire Unit 1 Reactor Vessel Surveillance Materials Average 50 ft-lb M Average 30 (ft-Ib) W Average 35 mil Lateral M Average Energy Absorption
- Transition Temperature ( F) Expansion Temperature (*F) Transition Temperatum ( F) at Full Shear (ft-Ib)
Material '
Unitradiated -Irradiated AT Unirradiated hradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AE Plate B5012-1 5 90 85 35 105 70 35 125. 90 140' 125 -'15 (longitudinal)
Plate B5012-1 0 85 85 50 115 65 75 135 60 101 94 . -'7- -
.. (transverse)
Weld Metal -5 170 175 0. 185 185 20 210 190 112 ' 72 '
HAZ Metal -50 - 85 135 - 15 105 120 -5 120 125 118 83 -- 35.
(a) " Average" is defined as the value read fmm the curve fit through the data points of the Charpy tests (see Figures 5-1 through 54) . .
i S
k
._.__m_ m ..m. '- u , .e . .,' . ,m.,
TABLE 5-10 Comparison of the McGuire Unit 1 Surveillance Material 30 ft-Ib Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision ' Predictions 30 ft-Ib Transition Upper Shelf Energy Fluence Temperature Shift Decrease 2-n/cm Material Capsule Predicted ") Measured Predicted 4' Measured (E > 1.0 McV)
( F) ( F) (%) (%)
U 4.712 x 10" 59 45 18 5 Intermediate Shell Plate B-5012-1 X 1.409 x 10" 81 45 22 5 (longitudinal) V 2.186 x 10" 90 85 24 11 U 4.712 x 10" 59 50 18 1 Intermediate Shell Plate B-5012-1 X 1.409 x 10" 81 65 22 0 h (transverse) V 2.186 x 10" 90 85 24 7 U 4.712 x 10"- 168 160 31 33 Weld Metal X 1.409 x 10" 233 165 38 26 V 2.186 x 10" 258 175 42 36 U 4.712 x 10 -
90 -
8 H A Z M etal X 1.409 x 10" -
115 -
19 V 2.186 x 10" -
135 -
30 (a) Based on Regulatory Guide 1.99, Revision 2 methodology using Mean wt. % values of Cu and Ni.-
TABLE 5-11 Tensile Properties of the hicGuire Unit i Reactor Vessel Surveillance hiaterials Irradiated at 550 F to 2.186 X 10" n/cin a (E > 1.0 hieV) 0.2% Yield Ultimate Fracture Fracture Fracture Unifonn Total Reduction Sample Test Temp. Strength Strength Load Stress Strength - Elongation Elongation in Area .
hiatcrial Number ( F) (ksi) (ksi) (kip) (ksi) (ksi) (%) (%) .(%)
Plate B5012-1 DL4 135' 74.9 93.7 2.90 164.1 59.1 12.0 26.6 64 (longitudinalj Plate B5012-1 DL5 300 70.3 88.6 3.00 169.8 61.1 11.5 21.9 64 (longitudinal)
Plate B5012-1 DL6 550 67.5 91.2 3.15 166.9 64.2 10.8 ' 23.3 ~ 62 (Longitudinal Plate B5012-1 DT4, 115 73.8 92.7 3.45 152.8 70.3 12.0 23.3 54 (transverse)
Plate B50121 DT5 250 69.3 86.6 2.95 158.1 60.1 11.1 22.1 62 (transverse)
.a y Plate B5012-1 DT6 550 67.2 88.6 3.10 143.5 63.2 9.9 19.8 56:
(transverse)
Weld Metal DW4. 160 863 97.8 3.45 152.8 70.3 11.5 22.6 54 Weld Metal DW5 .250 83.0 95.7 3.55 190.3 72.3 11 1 21.5 62 -
Weld Metal DW6' 550 78.4 94.7 3.40 157.4 69.3 9.3 19.7 56-3 .-
i t
____..,_____.____._____.____.u.._m__ _ _ _ _ _ _ + u +.r e e- -.e w .-- ,
( C)-
-150 -100 -50 0 50 100: 150 .200 250 l l l I lg(I , I .I I 100 -- - - -
8 80 -
% 60 -
$ 40 I I I I I I 0
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80 C5
@ 60 80 w f= #r 40 -
= o[ -
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-200 -100 0 -100- 200 .300 400 500 TEMPERATURE ('F)
O LMRRMIATO 0
e IRRAMATO GSf 4 FtVDCE 2J86 x 10 n/cn2 E ) to MeV) .
~!
-1
. Figure 5-1 rharpy V-Notch Impact Properties for McGuire Unit 1 Reactor Vessel Intermediate Shell Plate B5012-1 (Longitudinal Orientation) 5-18
, . . . . .,- - - . .. ,. .. - - ..--. l
.l 1
l
( C) -- ,
-150 -100 -50 0 50 100 150' 200 250 i i i .I I l -l l 1 i 100
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=
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-200 -100 0 100 200 300 400 500 <
TEMPERATURE ( F)
O lEdASIATO 2
e IRRAMATO C50'4 FUDCE 2J86 x En/cm E ) 1A NeV)
Figure 5-2 Charpy V-Notch hupact Propenies for McGuire Unit 1 Reactor Vessel Intermediate Shell Plate B5012-1 (Transverse Orientation) 5-19
I
( C)- ,
-150 -100 -50 0 50 - 100 150 200 250-I i i i. I .I I 100
- ^ 3'^1 ^ 2(I o o
@'80
' E 60 -
W o ,
" 40 -
20 -
2 I I I I I I 0
100 2.5 g 8 a-- -
y 80 -
2.0
- -e .
a- 60 -
g o e '
1.5 m.
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1.0 g Wr o o 2
0.5 0
- \ 0 120 160 i
_ p.
100 -
-120 g 80 -
0 , ,
I _ ,
5 60 -
o 80 b
m 19er 5 40 -
c "= In*r -
40 20 -
o
- 2 x o i I
- I i I O O
-200 -100 0 100 200 - 300 400 500 TEMPERATURE ( F) o LMRRADIATE3 4 RRADIATED G50'n rLLEKE 2J86 xa/cn 2 g ) 3,3 g,y) 1Y9 i
Figure 5-3 Charpy V-Notch Impact Properties for McGuire Unit 1 Reactor Vessel ,
Surveillance Weld Metal 5-20
i l
( C) 1
-150 -100 -50 0 50 100 150 200 250 .j l 1 I I l 2 fx l l I I 100
@ 80 o [*
$ 60 -
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- fig I I I I I 't 0 0 160 2L1 140 o o 120 -
160 2100 - 0 o o '
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60 -
o 80 m - m 40 -
og
[= 'r
( .g,. c 40 e 135 20 -
I I > l l I I 0
0
-200 -100 0 100 200 300 400 500 TEMPERATUPI (*D o (MRR/JJATO o IRRADIATO G5(D, FUDCE 2J% x 10 n/m2E ) 1.0 McV)
Figure 5-4 Charpy V-Notch Impact Properties for McGuire Unit 1 Reactor Vessel Weld Heat-Alfected-Zone Metal 5-21
m, _.. .
y...,. . _ . . , . . - - , _ . - -__ . . . _ - . . .
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. ;a.tg DL20 , DL19 DL28 DL26 DL27 l
!=
h l Figure 5-5 Charpy Impact Specimen Fracture Surfaces for McGuire Unit i Reactor ,
Vessel Intermediate Shell Plate B5012-1 (Longitudinal Orientation) . i 5-22 ,[
f
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n;g DT18 DT28 DT19 DT21 DT22 Figure 5-6 Charpy Impact Specimen Fracture Surfaces for McGuire Unit 1 Reactor Vessel Intermediate Shell Plate B5012-1 (Transverse Orientation) 5-23
wuzw 8 L:q P - <
y ~
m w{.w.. - vm y% gC=s ychm . (1
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DW24 DW16 DW25 DW19 DW29
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l DW28 DW18 DW30 DW23 DW20 Figure 5-7 Charpy Impact Specimen Fracture Surfaces for McGuire Unit 1 Reactor vessel Surveillance Weld Metal 5-24
1 i
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Figure 5-9 Tensile Properties for McGuire Unit 1 Reactor Vessel Intermediate Shell Plate B5012-1 (Longitudinal Orientation) 5-26
l J
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Figure 5-10 Tensile Properties for McGuire Unit 1 Reactor Vessel Intermediate Shell Plate B5012-1 (Transverse Orientation) 5-27
(*C) 0 50 100 150 200 .250 303 120
- } l l j 800 l-110 700 100 9 90 -
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l-l' 1
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i Figure 5-12 Fractured Tensile Specimens from McGuire Unit i Reactor Vessel j
intennediate Shell Plate B5012-1 (Longitudinal Orientation) 5-29 1
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I
MCGUl5E UNIT 1 'Y' CAPSULE 100.00 90.00-
/
80.00-70.00-
<n
- 60.00-i
@ 50.00-
=
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0.30 i 0.00 0.10 0.20 i
STRAIN, IN/IN -l f
MCGUIRE UNIT 1 'Y' CAPSULE 100.00 l
90.00-80.00-70.00-
- 60.00-6 ]'
@ 50.00-z '
$ 40.00-30.00-DL5 J
20.00 g 10.00 1
0.00 ! .
0.20 0.00 0.10 STRAIN, IN/IN L I l k l
l Figure 5-15 Engineering Stress-Strain Curves for Intennediate Shell Plate B5012-1 Tensile Specimens DL4 and DL5 (Longitudinal Orientation) 5-32 I
1 m ,
i I
I l.
MCGulRE UNIT 1 Y CAPSULE g 100.00 -!
a 90.00-80.00-70.00- ,
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10.00 0.00 .
0.20 ;
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Figure 5-16 Engineering Stress-Strain Curve for Intermediate Shell Plate B50121 Tensile Specimen DL6 (Longitudinal Orientation) 5-33 ,
.i i
l
t MCGUIRE UNIT 1 "V" CAPSULE 100.00 90.00-80.00-
,- J 70.00-m
- 60.00-tri
@ 50.00-c y 40.00-30.00- DT 4 20.00-115 F 10.00_
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O.20 0.00 0.'10 STRAIN, IN/IN MCGUIRE UNIT 1 "V' CAPSULE 100.00 90.00-80.00-70.00-m 60.00-vi
@ 50.00-c y 40.00-30.00-DT5 20.00-250 F 10.00-0.00 , , ,
0.00 0.10 0.20 STRAIN, IN/IN Figure 5-17 Engineering Stress-Strain Curves for Intermediate Shell Plate B5012 Tensile Specimens DT4 and DT5 (Transverse Orientation) 5-34
MCGUlRE UNIT 1 Y CAPSULE 100.00 90.00-80.00-70.00-rn '
- c:
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vi
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1 Figure 5-18 Engineering Stress-Strain Curve for Intermediate Shell Plate B5012-1 Tensile Specimen DT6 (Transverse Orientation) 5-35
MCGUIRE UNIT 1 "V" CAPSULE 110.00 100.00-90.00-80.00-
[ 70.00-vi 60.00-v) w 50.00-
") 40.00-30.00- DW4 ,
20.00-10.00-
.00 O.10 0.20 STRAIN, IN/IN ,
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- 60.00-VI
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0.00 0.10 0.20 STRAIN, IN/IN Figure 5-19 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens DW4 and DWS 5-36
MCGUIRE UNIT 1 Y CAPSULE 100.00 90.00-80.00-70.00-co
- 60.00-tri
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0.08 0.12 0.16 - 0.20 0.00 0.04 STRAIN, IN/IN l
l l
l Figure 5-20 Engineering Stress-Strain Curve for Weld Metal Tensile Specimen DW6 5-37
SECTION 6.0
. RADIATION ' ANALYSIS AND NEUTRON DOSIMETRY -
6.1 Introduction F
Knowledge of the neutron environment within the reactor pressure vessel and surveilluce capsule .
geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two L
reasons. First,in order to interpret the neutron radiation induced material property changes observed ~
in the test specimens, the neutron environment (enerEy spectrum, flux, fluence) to which the test specimens were exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established .
. between the neutron environment at various positions within the pressure vessel and that experienced by the test specimens. The former requirement is normally met by employing a combination of
. rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The latter information is generally derived solely from -
analysis. ,
h The use of fast neutron fluence (E > 1.0 MeV) to conclate measured material property changes to the ,
neutron exposure of the material has traditionally been accepted for development of damage trend :
curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall.
Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, " Analysis and Interpretation of Light Water Reactor Surveillance Results," recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a data base for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693 "Oiaracterizing Neutron Exposures in Ferritic Steels in Tenns of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the pressure vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, " Radiation Damage to :
Reactor Vessel Materials." l I
6-1 .!
)
+
- ,7
- This section provides the results of the neutmn dosimetry evaluations perfonned in conjunction with the analysis of test speciment contained in surveillance capsules V and Z, withdrawn at the end of the eighth fuel cycle. Also included is an updated evaluation of the dosimetry contained in capsules U and X, withdrawn a* the conclusion of cycles one and five, respectively. This update is based on current state-of-the-art methodology and nuclear data; and, together with the capsule V and Z results, pmvides a consistent up to date data base for use in evaluating the material properties of the McGuire Unit I reactor vesse.l.
In each of the 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 with the projected exposure of the pressure vessel are provided.
6.2 Discrete 3rdinates 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 56*,58.5 ,124 ,236 ,
238.5*, and 304 relative to the core cardinal axis as shown in Figure 4-1. A plan view of a dual surveillance capsule holder attached to the neutron pad is shown in Figure 6-1. The stainless steel specimen containers are 1.182 by 1-inch and approximately 56 inches in height. The containers are i positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 5 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 tne test specimen location, the capsules themselves must be included in the analytical model.
6-2 1
i i
In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, two distinct sets of transport calculations were canied out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters {$(E > 1.0 MeV),$(E > 0.1 MeV), and dpa/sec} thmugh the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the.
surveillance capsules as well as for the determination of exposure parameter ratios; i.e., [dpa/sec]/[$(E
> 1.0 MeV)], within the pressure vessel geometry. The relative radial gradient infonnation was required to permit the projection of measured exposure parameters to locations interior to the pressure vessel wall; i.e., the 1/4T,1/2T, and 3/4T locations.
The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux, . f Q(E > 1.0 MeV), at surveflance capsule positions and at several azimuthal locations on the pressure !
vessel inner radius to neutroa source distributions within the reactor core. The source imponance functions generated from these Mjoint 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 irradiation. They also established the means to perform similar ' ,
E predictions and dosimetry evaluations for all subsequent fuel cycles. It is important to note that the ,
cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core but also accounted for the effects of varying neutron yield per fission and fission spectrum introduced by the build-up of plutonium as the bumup of individual fuel !
assemblies increased.
The absolute cycle specific data from the adjoint evaluations together with the relative neutron energy -
spectra and radial distribution information from the reference forward calculation provided the means to: ,
t I
1- Evaluate neutron dosimetry obtained from surveillance capsules.
2- Extrapolate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel wa!' ,
3- Enable a direct comparison of analytical prediction with measurement. >
4 -- Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves.
6-3
l 2
! 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 DOT two-dimensional discrete ordinates code"4 and the S AILOR cross- -
section library"S. The SAILOR library is a 47 energy group ENDF/B-IV based data set produced specifically for light water reactor applications. In these analyses anisotropic scattering _was treated with a P 3expansion 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 4-loop plants. Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Funhennore, 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 + 20 value for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results.
All adjoint calculations were also carried out using an S, order of angular quadrature and the P3 cmss-section appmximation from the SAILOR library. Adjoint source locations were chosen at several azimuthal locations along the pressure 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:
R(r,0) = [ [ [ 1(r,0,E) S(r,0,E) r dr dB dE r 0 E where: R(r,0) = Q(E > 1.0 MeV) at radius r and azimuthal angle 0.
1(r,0,E)= Adjoint source importance function at radius r, azimuthal angle 0, and neutron source energy E. ,
S(r.0,E)= Neutron source strength at core location r,0 and energy E. ;
1 i
l Although the adjoint importance functions used in this analysis were based on a response function-defined by the threshold neutron flux $(E > 1.0 MeV), prior calculationsus have shown that, while the 6-4
.z--
1 H
hnplementation of low leakage loading pattems significantly impacts'both the magnitude and spatial
~
distribution of the neutron field, changes in the relative neutron energy spectnun are of second order.
Thus, for a given location the ratio of [dpa/sec]/[$(E > 1.0 MeV)] is insensitive to changing core i l
source distributions. In the application of these adjoint importance functions to the McGuire Unit 1 j
- reactor, therefore, the iron atom displacement rates (dpa/sec) and the neutron flux $(E > 0.1 MeV) I were computed on a cycle specific basis by using [dpa/sec]/[$(E > 1.0 MeV)] and [$(E > 0.1 MeV)]/[$(E > 1.0 MeV)} ratios from the forward analysis in conjunction v'ith the cycle specific $(E >
1.0 MeV) solutions from the individual adjobt evaluations.
The reactor core power distributions used in the plant specific adjoint calculations were taken from the fuel cycle design repons for the first eight operating cycles of McGuire Unit 1 U""#*
Selected msults from the neutron transpon analyses are provided in Tables 6-1 through 6-5. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation periods and provide the means to correlate dosimetry results with the corresponding exposure of the pressure vessel wall.
In Table 6-1, the calculated exposure parameters [$(E > 1.0 MeV), Q(E > 0.1 MeV), and dpa/sec] are given at the geometric center of the two surveillance capsule positions for both the reference and the plant specific core power distributions. The plant specific data, based on the adjoint tmnsport 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 plant specific fluence calculations can be compared. Similar data are given in Table 6-2 for the -
pressure vessel inner radius. Again, the three peninent exposure parameters are listed for the reference -
and the cycle one through eight plant specific power distributions. It is important to note that the data for the vessel inner radius were taken at the clad / base metal interface; and, cus, represent the maximum predicted exposu e levels of the vessel wallitself.
Radial gradient information applicable to Q(E > 1.0 MeV), Q(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 through the vessel wall may be obtained by nonnalizing the calculated or projected exposure at the vessel inner radius to the gradient data listed in Tables 6-3 through 6-5.
6-5
r For example, the neutron flux $(E > 1.0 MeV) at the 1/4T depth in the pressure vessel wall along the -
45 azimuth is given by:
4gf45#) = $(220.27, 45') F(225.75, 45 )
i where: $uc(45 ) = Projected neutron flux at the 1/4T position on the 45 azimuth.
$(220.27,45 ) = Projected or calculated neutron flux at the vessel inner radius on the 45*
azimuth.
F(225.75,45') = Ratio of the neu.ron flux at the 1/4T position to the flux at the vessel inner radius for the 45 azimuth. This data is obtained from Table 6-3 Similar expmssions apply for exposure parameters expressed in terms of $(E > 0.1 MeV) and dpa/sec where the attenuation function F is obtained from Tables 6-4 and 6-5, mspectively.
6.3 Neutmn Dosimetry The passive neutron sensors included in the McGuire Unit I surveillance program are listed in Table 6-6. Also given in Table 6-6 are the primary nuclear reactions and associated nuclear comants that were used in the evaluation of the neutron energy spectrum within the surveillance capsules and in the subsequent determination of the various exposure parameters of interest [$(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 imn, nickel, copper, and cobalt-aluminum monitors,in wire fonn, were placed in holes drilled in spacers at several axial levels within the capsules. The cadmium shielded uranimu 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 neutmn 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 irmdiation 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.
6-6
~ . _ _. - _. _. . . - .
0' The operating history of the reactor, The energy response of each monitor.
- The neutron energy spectrum at the monitor location.
The specific activity of each of the neutmn monitors was detennined using established ASTM procedurest2m e m. 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 McGuire Unit i reactor during cycles one through eight was supplied by NUREG-0020.
" Licensed Operating Reactors Status Summary Report," for the applicable period. Th6 irradiation j instory applicable to capsules U, X, V and Z is given in Table 6-7. j I
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 fmm the following equation:
A R=
~ ~
No F Y{ l C j[1-e *'d [c *4]
nf where:
R = Reaction rate averaged over the irradiation period and referenced to operation at a core power level of P ,(rps/ nucleus).
A = Measured specific activity (dps/gm).
No = Number of target element atoms per gram of sensor.
F = Weight fraction of the target isotope in the sensor material.
Y = Number of product atoms produced per reaction.
P; = Avenge core power level during irradiation period j (MW).
P,= Maximum or reference power level of the reactor (MW).
C; = Calculated ratio of $(E > 1.0 MeV) during irradiation period j to the time weighted average $(E > 1.0 MeV) over the entire irradiatian period.
A = Decay constant of the product isotope (1/sec).
t, = Length of irradiation period j (sec).
t, = Decay time following irradiation period j (sec).
and the summation is carried out over the total number of monthly intervals comprising the irradiation period.
6-7
In the equation describing the reaction rate calculation, the ratio [P]/[P,,,]
3 accounts for month by month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio C,, which can be calculated for each fuel cycle using the adjoint transport technology discussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in flux j level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single j l
cycle irradiation C; is normally taken to be 1.0. However, for multiple cycle irradiations, particularly l 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 U, X, V, and Z, the flux level term in the reaction rate calculations was developed from the plant specific analysis provided in Table 6-1. Measured and saturated reaction product specific activities as well as the derived full power reaction rates are listed in Tables 6-8 through 611 for capsules U, X, V, and Z, respectively.
Values of key fast neutron exposure parameters were derived from the measured reaction rates using l
the FERRET least squares adjustment code"1 The FERRET approach used the measured reaction rate data, sensor reaction cross-sections, and a calculated trial spectmm as input and proceeded to adjust the group fluxes from the trial spectrum to produce a best fit (in a least squares sense) to the measured '
reaction rate data. The " measured" exposure parameters along with the associated uncertainties were then obtained from the adjusted spectrum.
In the FERRET evaluations, a log-normal least squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertainties and correlations. In general, the measured values f are linearly related to the flux Q by some response matrix A:
g8 ") = { A [ 4 ")
8 where i indexes the measured value; belonging to a single data set s, g designates the energy group, and ct delineates spectra that may be simultaneously adjusted. For example, R, = [ og 4, e-6-8
relates a set of measured reaction rates R, to a single spectrum $, by the multigroup reaction cross-section oi,. The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with large assigned uncertainties.
In the least squares adjustment, the continuous quantities (i.e., neutron spectra and cross-sections) were approximated in a multi-gmup format consisting of 53 energy groups. The trial input spectrum was l
converted to the FERRET 53 group structure using the SAND-11 code". This procedure was carried out by first expanding the 47 group calculated spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure in regions where group boundaries do not coincide. The 620 point spectrum was then re-collapsed into the group stmeture used in FERRET.
The sensor set reaction cross-sections, obtained from the ENDF/B-V dosimetry file, were also collapsed into the 53 energy group structure using the SAND-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 cmss-section uncertainties in the form of a 53 x 53 covariance matrix for each sensor reaction were also constructed from the infonnation contained on the ENDF/B-V data files. These matrices included energy group to energy group uncenainty 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 adiustment 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-V data files, the covariance matrix for the input trial spectrum was constructed fmm the following relation:
M,1 =
Rl + R, R,i P,i .
l where R, specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the set I of values. The fractional uncertainties R, specify additional random uncertainties for group g that are correlated with a correlation matrix given by:
I P,, = [1 -0] 6,, + 0 e -" !
where: l 6-9 ;
H = f8-8 2 2y The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes short range correlations over a group range y(0 specifies the strength of the latter tenn). The value of S is I when g = g' and 0 otherwise. For the trial spectrum used in the current evaluations, a short range correlation of y = 6 groups was used. This choice implies that neighboring groups are strongly correlated when 0 is close to 1. Stmng long range correlations (or anti-correlations) were justified based on information presented by R. E. Maerkerau Maerker's results are closely duplicated when y = 6.
The uncenainties associated with the measured reaction rates included both statistical (counting) and systematic components. The systematic component of the overall uncertainty accounts for counter efficiency, counter calibrations, irradiation history corrections, and corrections for competing reactions in the individual sensors.
I Results of the FERRET evaluations of the capsules U, X, V, and Z dosimetry are given in Tables 6-12 and 6-13. The data summarized in these tables include fast neutmn exposure evaluations in terms of O(E > 1.0 MeV), @(E > 0.1 MeV), and dpa. In general good results were achieved in the fits of the -
adjusted spectra to the individual measured reaction rates. The adjusted spectra from the least squares evaluations are given in Tables 6-14 and 6-15 in the FERRET 53 energy gmup structure. The results for capsules V and Z are consistent with results obtained from similar evaluations of dosimetry from other Westinghouse reactors.
6.4 Projections of Pressure Vessel Exposure Neutron exposure projections at key locations on the pressure vessel inner radius are given in Table 6-
- 17. Along wid the current _(7.241 EFPY) exposvre, projections are also provided for exposure periods of 16 EFPY and 32 EFPY. In computing these vessel exposures, the calculated values fmm Table ( .2 were scaled by the avera3e measurement / cal.:ulation ratios (M/C) obseived from the evaluations of dosimetry from capsules U, X, V, and Z for each f 4t nr atmn exposure parameter. This procedure resulted in bias factors of 1.23,1.23, and 1.18 be og ap,, lied to the calculated values of C(E > 1.3 -
MeV), $(E > 0.1 MeV), and dpa. respectively. Pa.,.:ctions for future operation were based on the 6-10
assumption that the average exposure rates characteristic of the cycle one through eight irradiation would continue to be applicable throughout plant life.
For McGuire Unit 1, the uncertainty in each individual capsule derived fluence is estimated to consist of a 9% random component and a 5 % systematic component, and the extrapolation uncenainty is estimated to be SE A statistical combination of these uncertainties for the two capsules produces an overall uncertainty estimate in the exposure of the pressure vessel wall in the beltline region of 8%
(10) for fluence above 1 MeV.
In the calculation of exposure gradients for use in the development of heatup and cooldown curves for the McGuire Unit I reactor coolant system, exposure projections to 16 EFPY and 32 EFPY were also employed. Data based on both a $(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 RTsm vs fluence curves, dpa equivalent fast neutron fluence levels for the 1/4T and 3/4T positions were defined by the relations:
4(1/4T) = 4(07) dpa(1/4T) dpa(07)
I and l 4(3/4T) = 4(0T) @(3/4T) dpa(07)
Using this approach results in the dpa equivalent fluence values listed in Table 6-18. In Table 6-19 updated lead factors are listed for each of the McGuire Unit I surveillance capsules. Lead factor 'data based on the accumulated fluence through cycle eight are provided for each remaining capsule.
6-11
FIGURE 6-1 PLAN VIEW OF A DUAL REACTOR VESSEL SURVEILLANCE CAPSULE ;
1 (TYPICAL) .;
Co 58.5' ~ 81.08 - ,
- 6 7 _.8s m 8 ,~:
I ,
I!
[ _cNEUTRON PAQ N
a i
f s
j 6-12
i FIGURE 6-2 AXIAL DISTRIBUTION OF NEUTRON FLUENCE (E > 1.0 MEV)
ALONG THE 45 DEGREE AZIMUTH
- 1. 0 E + 2 0 t e
i i l
m . .
! _ ; _ _,_ _ _ _ _ _ _ _ _ _ _ _ ,I E
O 1.0 E + 19. . , , - - -
s -
C ' !
\
W p l
- i \
i , 1 f
8 ' f X N g / ! ! \
a I
- u. ,
I i
C ,
o u
I
.i i
5 1.0 E + 18 ! ,
e . . .
i z '
, , 1 i
! l i
ye '
Mi7.24 J EEPYL-- 15 EFPY ::
e
$," 32? EFPY.
l
" h i 1.0 E+ 17 0 1 2 3 4 5 6 7 8 9 101112 Distance From Core Bottom (ft) 6-13 l 1
+
TABLE 6-1 CALCULATED FAST NEUTRON EXPOSURE RATES AT THE SURVEILLANCE CAPSULE CENTER 2
CALCULATED FLUX $(E >1.0 MeV) [n/cm -sec] AT THE SURVEILLANCE CAPSULES-CAPSULE LOCATION 31.5* 34.0 -
CYCLE 1 8.5777E+10 9.8957E+10 CYCLE 2 1.0586E+11 1.2074E+11 CYCLE 3 7.6650E+10 8.6146E+10 CYCLE 4 7.0165E+10 7.9365E+10 .
CYCLE 5 6.6807E+10 7.4018E+10 CYCLE 6 7.1250E+10 7.9446E+10 CYCLE 7 6.9907E+10 7.8699E+10 CYCLE 8 6.9324E+ 10 - 7.7147E+10 -'i 2
CALCULATED FLUX Q(E >0.1 MeV) [n/cm -sec] AT THE SURVEILLANCE CAPSULES CAPSULE LOCATION ,
31.5 34.0*
CYCLE 1 3.8N2E+11 4.4699E+ 11.
CYCLE 2 4.6949E+11 5.4538E+11 CYCLE 3 3.3994E+11 3.8912E+11 -
CYCLE 4 3.1118E+11 3.5849E+11 CYCLE 5 2.9629E+11 3.3434E+11 CYCLE 6 3.1599E+11 3.5886E+11 CYCLE 7 3.10ME+11 3.5548E+11 CYCLE 8 3.0745E+ 11 3.4847E+11 -
CALCULATED 1ron Displacement Rate [dpa/sec] AT TIIE SURVEILLANCE CAPSULES CAPSULE LOCATION 31.5 34.0 CYCLE 1 1.7052E-10 1.9861E-10 CYCLE 2 2.lM5E-10 2.4233E-10 CYCLE 3 1.5238E-10 1.7290E-10 CYCLE 4 1.3949E-10 1.5929E-10 CYCLE 5 1.3281E-10 1.4855E-10 CYCLE 6 1.4165E-10 1.5945E-10. I CYCLE 7 1.3898E-10 1.5795E-10 CYCLE 8 1.3782E-10 1.5483E-10 6-14 l
TABLE 6-2 CALCULATED AZIMUTilAL VARIATION OF FAST NEUTRON EXPOSURE RATES' AT THE PRESSURE VESSEL CLAD /B ASE METAL INTERFACE 6m >1.0 MeV) ln/cm 2-seci 0 DEG 15 DEG 30 DEG 45 DEG CYCLE 1 1.0932E+10 1.6394E+10 1.3024E+10 1.8845E+10 CYCLE 2 1.4170E+10 2.1316E+10 1.6230E+10 2.2679E+10 CYCLE 3 1.0672E+10 1.5620E+10 1.1687E+10 1.6069E+10 CYCLE 4 1.0200E+10 1.5141E+10 1.0913E+10 1.5155E+10 CYCLE 5 1.0791E+10 - 1.5222E+10 1.0304E+10 1.3717E+10 '
CYCLE 6 1.0161E+10 1.5184E+10 1.0971E+10 1.4801E+10 CYCLE 7 9.4351E+09 1.3867E+10 1.0624E+10 1.4737E+10 CYCLE 8 1.0125E+10 1.5096E+10 1.0738E+10 - 1.4410E+10 2
&G >0.I MeV) In/cm -secl ODEG 15 DEG 30DEG 45 DEG CYCLE 1 2.2769E+10 3.4456E+10 3.3513E+10 4.7190E+10 CYCLE 2 2.9513E+10 4.4801E+10 4.1763E+10 5.6790E+10 CYCLE 3 2.2227E+10 3.2830E+10 3.0073E+10 4.0238E+10 CYCLE 4 2.1244E+10 3.1823E+10 2.8081E+10 3.7950E+10 CYCLE 5 2.2475E+10 3.1993E+10 2.6514E+10 3.4349E+10 CYCLE 6 2.1163E+10 3.1913E+10 2.8230E+10 3.7063E+10 CYCLE 7 1.9651E+10 2.9145E+10 2.7337E+10 3.6903E+10 CYCLE 8 2.10S8E+10 3.1728E+10 2.7631E+10 3.6084E+10 Iron Displacement Rate [dpa/seci g 0DEG 15 DEG 30 DEG 45 DEG CYCLE 1 1.6963E-11' 2.5244E-1I 2.1204E-11 2.9967E-11 CYCLE 2 2.1988E-11 3.2823E-11 2.6423E-11 3.6063E CYCLE 3 1.6560E-11 2.4052E-11 1.9027E-11 2.5552E-11 '
CYCLE 4 1.5828E-Il 2.3314E-Il 1.7767E-Il 2.4099E-11 CYCLE 5 1.6745E-11 2.3439E-11 1.6775E-Il 2.1812E-ll ,
CYCLE 6 1.5767E-11 2.3381E-11 1.786lE-11 2.3536E-11 CYCLE 7 1.4641E-11 2.1353E-11 1.7296E-ll' 2.3434E '
CYCLE 8 1.5711E-11 2.3245E Ii 1.7482E-11; 2.2914E >!
- j I
6-15
TABLE 6-3 RELATIVE RADIAL DISTRIBUTION OF $(E > 1.0 MeV)
WITHIN THE PRESSURE VESSEL WALL Radius (cm) O. 0 15.0 25.0* 35.0* 45.0 220.27* 1.00 1.00 1.00- 1.00 1.00 ' .
220.64 0.976 0.979 0.980 0.977 0.979 221.66 0.888 0.891 0.893 0.891 0.889 ~
222.99 0.768 0.770 0.772 0.770 0.766 224.31 0.653 0.653 0.657 0.655 0.648 225.63 0.551 0.550 0.554 0.552 0.543-226.95 0.462 . 0.460 0.465 0.463 OA52 228.28 0.386 0.384 0.388 0.386 0.375 229.60 0.321 0.319 0.324 0.321 0.311
.230.92 0.267 0.263 0.275 0.267 0.257 232.25 0.221 0.219 0.225 0.221 0.211 233.57 0.183 0.181 0.185 0.183 0.174 234.89 0.151 0.149 0.153 0.151 0.142 236.22 0.124 0.122 0.126 0.124 0.116 237.54 0.102 0.100 0.104 0.102 0.0945 238.86 0.0828 0.0817 0.0846 .0835 0.0762 240.19 0.0671 0.0660 0.0689 .0679 0.0608 241.51 0.0538- 0.0522 0.0550- 0.0545 0.0471 242.175 0.0506 0.0488 0.0518 0.0521 0.0438 l l
l NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius 6-16
p TABLE 64 RELATIVE RADIAL DISTRIBUTION OF Q(E > 0.1 MeV)
WITHIN THE PRESSURE VESSEL WALL Radius -
(cm) 0. 0" 15.0 25.0 35.0 45.0*
220.27m' l.00- 1.00 1.00 1.00 1.00 220.64 1.00 1.00 1.00 1.00 1.00 j 221.66 1.00 1.00 1.00 0.999 0.995 222.99 0.974 0.969 0.974 0.959 0.956 224.31 0.927 0.920 0.927 0.907 0.901 225.63 0.874 0.865 0.874 0.850 '0.842 226.95 0.818 0.808 0.818 0.792 0.782 228.28 C.761 0.750 0.716 0.734 0.721 229.60 0.705 0.693 0.704 0.677 0.662 230.92 0.649 0.637 0.649 0.621 0.605 232.25 0.594 0.582 0.594 0.567 0.549 233.57 0.540 0.529 0.542 0.515 0.495 ,
234.89 0.487. 0.478 0.490 0.465 0.443-236.22 0.436 0.428 0.440 0.416 - 0.392 237.54 0.386 0.380 0.392 0.369 0.343 238.86 0.337 0.333 0.344 0.324 0.295 ,
240.19 0.289 0.287 0.298 0.279 0.248 241.51 0.244 0.238 0.249 0.233 0.201-242.175 0.233 0.226 .237 0.223 0.188 P
NOTES: 1) Base Metal Inner Radius
- 2) Base Metal Outer Radius i
l 6-17 l
4
TABLE 6-5 IELATIVE RADIAL DISTRIBUTION OF dpa/sec WITHIN THE PRESSURE VESSEL WALL Radius (cm) 0. 0 15.0 25.0* 35.0* 45.0 220.27W l.00 1.00 1.00 1.00 1.00-220.64 - 0.984 0.981 0.984 0.983 0.984 221.66 0.912 .0.909 0.917 0.921 0.915 222.99 0.815 0.812 0.826 0.833 0.821 224.31 0.722 0.719 0.737 - 0.747 0.730 -
225.63 0.638 0.634 0.656 0.668 0.647 226.95 0.563 0.559 0.584 0.597 0.572 228.28 0.497 0.493 0.519 0.533 0.506 229.60 0.439 0.435 0.462 0.475 0.447 230.92 0.387 0.383 0.410 0.423 0.394 232.25 0.341 0.338 0.364 0.376 0.347 233.57 0.300 0.297 0.322 0.334 0.305 ,
234.89 0.263 0.261 0.285 0.295 0.266 236.22 0.230 0.228 0.250 0.260 0.231
~
'37.54 0.199 0.198. 0.218 0.227 0.199 l
l 238.86 0.171 0.170 0.189 0.196 0.169 240.19 0.145 0.144 0.161 0.167 0.140 241.51 0.121 0.119 0.135 0.139 0.113 242.17* 0.116 0.113 0.128 0.134 0.106 -
NOTES: 1) Base Metal Inner Radius y 2) Base Metal Outer Radius i
6-18 .
, j l
TABLE 6-6 NUCLEAR PARAMETERS USED IN THE EVALUATION OF NEUTRON SENSORS j
- Reaction Target Fission Monitor of Weight Response Product Yield Material Interest Fraction Rance Half-Life (%)
Copper- Cu"(n,cx)Co* 0.6917 E > 4.7 MeV 5.271 yrs Iron Fe"(n,p)Mn" 0.0580 E > 1.0 MeV 312.5 days Nickel Ni (n,p)Co" S8 0.6827 E > 1.0 MeV ' 70.78 days Uranium-238* U238(n,0Cs "
1 1.0 E > 0.4 MeV 30.17 yrs - 6.00 Neptunium-237* Np2 (n,0Cs " 1.0 E > 0.08 MeV 30.17 yrs 6.27 Cobalt-Aluminum
- Co"(n,y)Co* 0.0015 0.4ev>E> 0.015 MeV 5.271 yrs Cobalt-Aluminum Co"(n,y)Co* 0.0015 E > 0415 MeV 5.271 yrs
- Denotes that monitor is cadmium shielded.
s
^
6 j
1 6-19
- i
TABLE 6-7 l a
MONTHLY THERMAL GENERATION DURING THE FIRST EIGHT FUEL CYCLES OF THE McGUIRE UNIT 1 REACTOR ;
Thermal Generation Thermal Generation Year Month (MW-h6 Year Month (MW-h6
'7 2,534,622 1981 10 222,129 8 1,847,703 11 791,796 9- 221,936 12 84,955 10 .
0 1982 1 1,273,713 11 1,101,430 2 1,054,504 12 2,251,637 530,357 19B8 3 2,379,899-3
- 4 1,192,627 . 2,314,778'
'5 1,484,279 3 2,380,942 6 1,266,005 4 2,314,604-7 561,572 5 2,528,964 8 1,467,549 6 2,281,133 9 1,454,192 7 2,485,619 10 1,338,949 8 - 2,522,007' 527,736 9 2,437,962 11 12 1,265,980 10 898,152 1993 1 857,172 1 0 2 e }12 9,881 g 1989 1 2,245.259 3 ~
4 0 2 2,286,701 5 187,441 3 569,136 6 1,953,836 4 0 7 1,975,352 5 1,664,673 8 1,129,545 6 2,412,733 9 2,260,850 7 2,368,C74 10 1,935,163 8 2,300,333 1,623,905 9 2,425,699 11 2,011,387 0 2,518,C39 12 1 1984 1 2,222,932 1 2,456,C58 2,530,503 2
3 1'889,758 O
1990 '.
1 608,302 4 0 4 0 5 + 1,994,665 3 0 2,266,395 4 0 6
7 2,319,518 5 6
,.,156,153 521, m 8 2,370,549 9 2,480,194 ]e 2,494,695 2,124,989
- 0 2,042,345 1,553,C94 9 2,200,793 11 ,
226,629 10 1,063,174 12 1985 1 2,356,943 11 1,146,029 2 2,088,791 12 2,518,560 3 1,694,129 1991 1 2,519,888 4 596,719 2 1 818 761 8
c 0 3 2,521'.731, 5 117 845 4 993 7 2,344$677 5' l'274,'480 1, 443 8 2,533,613 j 2,446,944 2,516,832 9 2,329,794 '
10 '2,533,965 0 2,534,664 1,540,353 9 1,557,266 11 2,425,449 10 0 12 1986 1 2,471,699 11 0 2,022,499 12 1,566,100-2 2,089,263 1992 1 1,277,016-1 4 2,204,598- 2 752,634 5 1,001,197 3 2,438,606 ,
6 0 4 2,390,345 5 305,612 6 680,274 7 0 7 2,285,861 8 0 8 2,514,974 9 944,223 9 2,450,774 10 2.293,046 -
10- 2,514,483 11 44,556 11 2,451,002 12 2'516,554 12 2,526,799 1987 1 2,534,776 1993 1 2,543,'424 1,861,001 2- 2.288,198 2
3 2,537,575 3 904'384 4 2,287,291 5 2,495,567 6 2,441,831 6-20
t TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES -
SURVEILLANCE CAPSULE U SATURATED ACTIVITIES AND DERIVED FAST NEUTRON FLUX MEASURED SATURATED REACTION MONITOR AND ACTIVITY ACTIVITY RATE AXIAL LOCATION (dis /sec-cm) (dis /sec-em) (rps/ nucleus)
Co-63 (nfr) Co-60 84-2731 TOP - 4.510E+04 3.873E+05 .
84-2737 MID 4.740E+N 4.070E+05 84-2743 BOT 4.530E+M 3.890E+05 AVERAGES 4.593E+04 3.945E+05 6.018E-17 Fe-54 (nm) Mn-54 84-2730 TOP -1.100E+06 3.935E+06 84-2736 MID 1.130E+06 4.M2E+06 84 2742 BOT 1.150E+06 4.114E+06 AVERAGES 1.127E+06 4.030E+06 6.444E-15 Ni-58 (n.p) Co-58 84 2734 TOP 4.950E+06 6.149E+07
- 84-2740 MID 5.160E+06 6.409E+07 84-2746 BOT 5.340E+06 6.633E+07 AVERAGES 5.150E+06 6.397E+07 9.134E-15 Co-59 (n3) Co-60 84 2732 TOP 1.310E+07 1.125E+08 :
84-2738 MID 1.170E+07 1.005E+08 -
84-2744 BOT 9.270E+06 7.961E+07 84-2735 TOP 1.250E+07 1.073E+08 q 84-2741 MID 9.780E+06 8.399E+07 84-2747 BOT 1.210E+07 1.039E+08 AVERAGES 1.141E+07 9.797E+07 ' .392E-l'2 6
Co-59 (n3) Co-60 84-2733 TOP 6.440E+06 5.530E+07 ,
' 84-2739 MID - 6.360E+06 5.462E+07 84-2745 BOT 6.470E+06 5.556E+07 AVERAGES 6.423E+06 5.516E+07 3.599E-12 U-238 (n.O Cs-137 84-2729 MID 1.800E+05 7.464E+06 ' 4.919E No-237 (n.O Cs-137 84-2728 MID 1.650E+06 6.842E+07 4.295E-13 ,
6-21 ;
TABLE 6-9 hiEASURED SENSOR ACTIVITIES AND REACTION RATES-SURVEILLANCE CAPSULE X SATURATED ACTIVITIES AND DERIVED FAST NEUTRON FLUX MEASURED SATURATED REACTION MONITOR AND ACTIVITY ACTIVITY RATE AXIAL LOCATION (dis /sec-Rm) (dit/sec-em) (rps/ nucleus)
Cu-63 (n.a) Co-60 89-336 TOP 1.260E+05 3.479E+05 89-342 MID- 1.250E+05 3.451E+05 89-348 BOT 1.290E+05 3.561E+05 AVERAGES 1.267E+05 3.497E+05 . 5.335E-17 Fe-54 (n,p) Mn-54 89-337 TOP 1.650E+06 3.292E+06 89-343 MID 1.640E+06 3.272E+%
89-349 BOT 1.730E+06 3.451E+06 AVERAGES - 1.673E+06 3.338E+06 5.338E-15 Ni-58 (n.p) Co-58 89-338 TOP 1.100E+07 5.284E+07 89-344 MID 1.070E+07 5.140E+07 89-350 BOT 1.120E+07 5.380E+07 AVERAGES 1.097E+07 5.268E+07 7.522E-15 Co-59 (n.y) Co-60 89-333 TOP 2.760E+07 7.620E+07 89 339 MID 2.950E+07 8.144E+07 89-345 BOT 2.490E+07 6.874E+07 89-334 TOP 3.140E+07 8.669E+07 89 340 MID 3350E+07 9.249E+07 89-346 BOT 3.000E+07 8.282E+07 AVERAGES 2.948E+07 8.140E+07 5.311E-12 Co-59 (nty) Co-60 89-335 TOP 1.680E+07 4.638E+07 89-341 MID 1.680E+07 4.638E+07 89 347 BOT 1.550E+07 4.279E+07 AVERAGES 1.637E+07 4.519E+07 2.948E U-238 (n.0 Cs-137 89-331 MID 6.270E+05 6.889E+06 4.540E-14 No-237 (n.O Cs-137 89 332 MID 3.910E+% 4.296E+07 2.697E-13 6-22 i
I
)
TABLE 610 l
MEASURED SENS'OR ACITVITIES AND REACTION RATES' 4 SURVEILLANCE CAPSULE V l R
SATURATED ACTIVITIES AND DERIVED FAST NEUTRON FLUX MEASURED SATURATED REACTION MONITOR AND ACTIVITY ACTIVITY RATE AXTAL LOCATION (dis /sec-cm) (dis /sec.cm) (ms/ nucleus) 4 Cu-63 (n.a) Co-60 .
93-4158 TOP 1.470E+05 3.169E+05 934163 MID 1.610E+05 3.470E+05 934169 BOT 1.450E+05 3.126E+05 AVERAGES 1.510E+05 3.255E+05 4.965 0 17 Fe-54 (nm) Mn-54 93-4160 TOP 1.280E+06 2.938E+06 93-4165 MID 1.280E+06 2.938E+06 L 93-4171 BOT 1.200E+06 2.755E+06 AVERAGES 1.253E+06 2.877E+06 4.600E-15 Ni 58 (nm) Co-58 93-4159 TOP 4.730E+06 4.487E+07 93-4164 MID 4.810E+06 4.563E+07 93-4170 BOT 4.740E+06 4.497E+07 <
AVERAGES 4.760E+% 4.516E+07 6.448E-15 Co-59 (n.y) Co-60 93-4155 TOP 2.740E+07 5.906E+07 93-4161 MID 2.940E+07 6337E+07
' 93-4166 BOT 2.600E+07 5.604E+07 93-4156 TOP 3.220E+07 6.941E+07 93 4162 MID 2.430E+07 5.238E+07 934167 BOT 2.970E+07 6.402E+07 AVERAGES 6.071E+07 6.071E+07 3.961E-12 Co-59 (n.y) Cm60 93-4157 TOP 1.690E+07 3.643E+07 4162 MID 1.620E+07 3.492E+07 .
934168 BOT 1.630E+07 3.514E+07 - ,
AVERAGES 1.647E+07 3.549E+07 2.316E 12 U-238 (n,0 Cs-137 93-4153 MID 8380E+05 5.785E+06 3.812E-14 ,
Nr>.237 (n.O Cs-137 93-4154 MID 3.110E+06 2.147E+07 1348E-13 6-23
,_______J___ _____1.._______________ ._ . ._._u -- ._ .-.._r ,
' TABLE 6-11 MEASURED SENSOR ACTIVITIES AND REACTION RATES ,
SURVEILLANCE CAPSULE Z SATURATED ACTIVITIES AND DERIVED FAST NEUTRON FLUX MEASURED SATURATED REACTION MONITOR AND ACTIVITY ACTIVITY RATE AX1AL LOCATION - (dis /sec-em) (dis /sec-am) : (rps/ nucleus), .
Cu-63 (n,cr) Co-60 93-4176 TOP l530E+05 3.311E+05 93-4183 MID 1.610E+05 3.484E+05
- 93-4189 BOT 1.570E+05 3.398E+05 AVERAGES 1.570E+05 3.398E+05 5.184E-17 Fe-54 (nm) Mn-54 93-4178 TOP 1.290E+06 2.994E+06 93-4185 MID 1.350E+06 3.133E+06 93-4191 BOT 1.310E+06 3.041E+06 AVERAGES 1317E+06 3.056E+06 4.886E-15 Ni-58 (nm) Co-58 93-4177 TOP 5.090E+06 4.896E+07 93-4184 MID 5.220E+06 5.021E+07 93 4190 BOT 5.110E+06 4.915E+07 AVERAGES 5.140E+06 4.944E+07 7.059E-15 Co-59 (n y) Co-60 93-4179 TOP 2.800E+07 6.060E+07 93-4180 MID 3.010E+07 6.514E+07 93-4186 BOT 3.030E+07 6.558E+07 93-4174 TOP 3.430E+07 7,423E+07 93-4181 MID 3.350E+07 7.250E+07 93-4187 BOT 3.540E+07 7.661E+07 ,
AVERAGES 6.911E+07 6.911E+07 4.509E-12 Co-59 (n.y) Co-60 93-4175 TOP 1.820E+07 3.939E+07 '
93-4182 MID 1.780E+07 3.852E+07 93-4188 BOT 1.890E+07 4.090E+07 AVERAGES 1.830E+07 3.961E+07 2.584E-12 U-238 (n.O Cs-137 93-4172 MID 9.570E+05 6.611E+06 '4357E 14-No-237 (n.O Cs-137 93-4173 MID 6 80E+06 4.546E+07 ' 2.854E-13 i
l.
- l. 6-24 l' ,
1 r
i-- y v( .ww-y ym--+ , g- - r--- ty---w-- ,yv ng- w r e -r
+.
TABLE 6-12
SUMMARY
OF NEUTRON DOSIMETRY RESULTS SURVEILLANCE CAPSULES U, X, V, AND Z l
-1
)
Calculation of Meuured Fluence for Capsule U Flux Time Fluence Uncertainty Men Fluence < 0.414 ev = (Meas Hux < 0.414) * (EFPS) 1.146E+11 3.437E+07 ' 3.939E+18 - 22 Men Fluence > 0.1 Mev = (Meu Hux > .1) * (EFPS) 6.348E+11 3.437E+07 2.182E+19 18 Meas Fluence > 1.0 Mev = (Meu Hux > I) * (EFPS) 1.371E+11 3.437E+07 4.712E+18 9 .
dpa 2.678E-10 3.437E+07 - 9.204E.03 13 Calatladon of Measured Fluence for Capsule X Flux Time Fluence Uncenainty
- Men Fluence < 0.414 ev = (Meas Flux < 0.414; * (EFPS) 9.608E+10 1.358E+08 1.304E+19 22 Meas Fluence > 0.1.Mev = (Meas Hux > .1) * (EFPS) ' 4.308E+11 1.358E+08 5.848E+19 15 Meas Fluence > 1.0 Mev = (Meas Hux > 1) * (EFPS) . 1.038E+11 1.358E+08 .1.409E+19 8 dpa 1.908E-10 1.358E+08 2.590E42 11 Calculation of Measured Huence for Capsule V Hux Time Fluence Uncenainty Meas Fluence < 0.414 ev = (Meas Flux < 0.414) * (EFPS) 6.830E+10 2.285E+08 1.561E+19 23 ,
Meas Fluence > 0.1 Mev = (Meas Hux > .1) * (EFPS) 4.480E+11 2.285E+08 1.024E+20 21 --
Meas Fluence > 1.0 Mev = (Meu Hux > I) * (EFPS) 9.565E+10 2.285E+08 2.186E+19 - 10 dpa' l.885E.10 2.285E+08 4.308E42 16 . ;
Calculation of Measured Fluence for Capsule Z Flux Time Fluence Uncenainty Men Fluence < 0.414 ev = (Meas Flux < 0.414) * (EFPS) 7.958E+10 - 2.285E+08 1.819E+19 - 23 Meas Fluence > 0.1 Mev = (Meas Hux > .1) * (EFPS) 4.397E+11 2.225E+08 .1.005E+20 15 Meas Fluence > 1.0 Mev = (Meas Flux > 1) * (EFPS) 1.000E+11 2.285E+08 2.285E+19 - 8 dpa 1.899E.10 2.285E+08 4.340E.02 1I Y
t s
6-25 ;
1 o .:
.l
-w
TABLE 6-13 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCE CAPSULE CENTER SURVEILLANCE CAPSULES U, X, V, and Z ADJUSTED REACTION MEASURED CALCULATION C/M CAPSULE U WITHDRAWN EOC 1 Cu63 (n,a) Co60 6.02E-17 - 6.12E-17 1.02 Fe54 (n.p) Mn54 6.44E-15 6.48E 15 1.01 NiS8 (n p) CoS8 9.13E-15 9.07E-15 0.99-CoS9 (n,y) Co60 6.39E-12 6.34E-12 0.99 CoS9 (n,y) Co60 (Cd) 3.60E-12 3.62E-12 1,00 U238 (n,0 Csl37 (Cd) 4.28E-14 4.05E-14 0.94 Np237 (n,f) Csl37 (Cd) 4.29E-13 4.46E-13 1.04 CAPSULE X WITHDRAWN EOC 5 Cu63 (n.a) Co60 5.33E 17 5.40E-17 1.01 Fe54 (n,p) Mn54 5.34E-15 5.41E-15 1.01 NiS8 (n p) CoS8 7.52E-15 7.50E-15 1.00 CoS9 (n,y) Co60 5.31E-12 5.26E-12 0.99-CoS9 (n,y) Co60 (Cd) 2.95E-12 2.96E-12 1,00 U238 (n,f) Csl37 (Cd) 3.80E-14 3.25E-14 0.86 Np237 (n,f) Cs137. (Cd) . 2.70E-13 2.97E-13 1.10 CAPSULE V WITHDRAWN EOC 8 Cu63 (n,a) Co60 4.96E-17 4.98E-17 1.00 i Fe54 (n,p) Mn54 4.60E-15 4.68E-15 1.02 NiS8 (n.p) CoS8 6.45E-15 6.46E-15 1.00 '
CoS9 (n,y) Co60 3.96E-12 3.93E-12 0.99 CoS9 (n,y) Co60 (Cd) 2.32E-12 2.33E-12 1.01 U238 (n,f) Csl37 (Cd) . 3.12E-14 2.86E-14 0.92 CAPSULE Z WITHDRAWN EOC 8 Cu63 (n,a) Co60 5.18E-17 5.21E-17 1.00 Fe54 (n.p) Mn54 4.89E-15 5.01E-15 1.03 NiS8 (n,p) CoS8 .7.06E-15 7.02E-15 1.00 Co59 (n,y) Co60 4.5 IE-12 4.47E-12 0.99 CoS9 (n,y) Co60 (Cd) 2.58E-12 2.60E-12 -1.00 U238 (n.f) Csl37 (Cd) 3.53E-14 3.07E-14 0.87 Np237 (n,f) Cs137. (Cd) . 2.85E-13 3.03E-13 1.06 1
6-26
- _ . . -___.________-._._________._i____._______________b_______.- ___.__mm r
TABLE 6-14 i
ADJUSTED NEUTRON ENERGY SPECTRUM AT THE.
CENTER OF SURVEILLANCE CAPSULE V ENERGY ADJUSTED FLUX ENERGY. ADJUSTED FLUX 2
GROUP (MeV) (n/cm2 4ec) GROUP (McV) (n/cm -sec) 1 1.733E+01 7.385E+06 27 1.503E-02 1.533E+10 2 1.492E+01 1.634E+07 28 9.119E-03 1.876E+10 3 1350E+01 6.144E+07 29 5.531E-03 2.196E+10 4 1.162E+01 1.370E+08 30 3355E-03 6.909E+09 ~
5 1.000E+01 2.980E+08 31 2.839E-03 6.653E+09 '
6 8.607E+00 5.023E+08 32 2.404E-03 6.499E+09 7 7.408E+00 1.144E+09 33 2.035E-03 1.916E+10 8 6.065E+00 1.633E+09 34 1.234E-03 1.840E+10 ,
9 4.966E+00 3.430E4)9 35 7.485E-04 1.628E+10 10 3.679E+00 4.629E+09 36 4.540E44 1.400E+10 11 2.865E+00 9.975E+09 37 2.754E44 1.628E+10 12 2.231E+00 1.427E+10 38 1.670E41 1.731E+10 13 1.738E+00 2.058E+10 39 1.013E-04 1.771E+10 14 1353E+00 2.358E+10 40 6.144E-05 1.776E+10 l
15 1.108E+00 4.505E+10 41 3.727E-05 1.745E+10 16 8.208E-01 5.104E+10 42 2.260E-05 1.692E+10 17 6393E-01 5.556E+10 43 1371E-05 1.629E+10 18 4.979E-01 3.837E+10 44 8315E-06 1.546E+10 19 3.877E-01 . 5.484E+10 '45 5.043E-06 1.442E+10 20 3.020E-01 5.929E+10 46 3.059E-06 1337E+10 21 1.832E-01 5.487E+10 47 1.855E-06 1.201E+10 22 1.111E-01 4.125E+10 48 1.125E-06 9.178E+09
-23 6.738E-02 3.196E+10 49 6.826E-07 1.104E+10 24 4.087E-02 1.566E+10 50 4.140E 1331E+10 .
I 25 2.554E-02 2.160E+10 51 2.511E-07 1.264E+10 26 1.989E-02 1.036E+10 52 1.523E-07 1.149E+10 53 9.237E-08 3.085E+10
- l. .
!* l Note: Tabulated energy levels represent the upper energy in each group. ,
6 !
I l
TABLE 6-15 1 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCE CAPSULE Z ENEROY ADJUSTED FLUX- ENERGY ADIUSTED FLUX 2
GROUP (MeV) 2 (n/cm -sec) GROUP (MeV) (n/cm -sec) ,
1 1.733E+01 7.283E+06 27 1.503E-02 1.582E+10 '
2 1.492E+01 1.626E+07 28 9.119E 1.965E+10 3 1350E+01 6.198E+07 29 5.531E-03 2318E+10 4 1.162E+01 1.405E+08 30 3355E-03 7.367E+09 -
5 1.000E+01 3.091E+08 31 2.839E-03 7.146E+09 6 8.607E+00 5.288E+08 32 2.404E-03 7.020E+09 7 7.408E+00 1.216E+09 33 2.035E-03 2.071E+10 8 6.065E+00 1.748E+09 34 1.234E-03 1.991E+10 9 4.966E+00 3.727E+09 35 7.485E-M 1.773E+10 10 3.679E+00 5.068E+09 36 4.540E-04 1.536E+10 11 2.865E+00 1.083E+10 37 2.754E-04 1.769E+10 12 2.231E+00 1.531E+10 38 1.670E-04 1.939E+10 13 1.738E+00 2.168E+10 39 1.013E-N 1.946E+10 14 1353E+00 2.405E+10 40 6.144E-05 1.930E+10 15 1.108E+00 4.488E+10 41 3.727E-05 1.899E+10 16 8.208E-01 4.994E+10 42 2.260E-05 1.844E+10 17 6393E-01 5368E+10 43 1371E-05 1.778E+10 18 4.979E-01 3.675E+10 44 8315E-06 1.691E+10 19 3.877E-01 5.229E+10 45 5.043E-06 1.577E+10 20 3.020E-01 5.648E+10 46 3.059E-06 1.467E+10 21 1.832E-01 5.259E+10 47 1.855E-06 1324E+10 22 1.111E-01 3.996E+10 48 1.125E-06 1.013E+10 23 6.738E 3.135E+10 49 6.826E-07 1.231E+10 24 4.087E-02 1.565E+10 50 4.140E-07 1.501F+10 -
25 2.554E-02 2.186E+10 - 51 2.511E-07 1.444E+10 26 1.989E-02 1.053E+10 52 1.523E-07 1.327E+10 53 9.237E-08 3.687E+10 _ ;
4-Note: Tabulated energy levels represent the upper energy in each group.
~
6-28 4
i i
TABLE 6-16 COMPARISON OF CALCULATED AND MEASURED NEUTRON EXPOSURE LEVELS FOR McGUIRE UNIT 1 SURVEILLANCE CAPSULES U, X, V, AND Z Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE Rate for Capsule U Calculated Measured C/M 1/ C/M Fluence (E > 1.0 Mev) [n/cm2-sec] 3.4012E+18 4.7119E+18 0.722 1385 Fluence (E > 0.1 Mev) [n/cm2-sec] 1.5363E+19 2.1817E+19 0.704 1.420 dpa 6.8262E-03 9.2038E-03 0.742 1348 Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE Rate for Capsule X Calculated Measured C/M 1/ C/M Fluence (E > 1.0 Mev) [n/cm2-sec] 1.2430E+19 1.4091E+19 0.882 1.134 Fluence (E > 0.1 Mev) [n/cm2-sec) 5.6146E+19 5.8483E+19 0.960 1.(M2 dpa 2.4947E 2.5902E-02 0.963 1.038 Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE Rate for Cepsule V Calculated Measured C/M 1/ C/M Fluence (E > 1.0 Mev) [n/cm2-sec) 1.7461E+19 2.1858E+19 0.799 1.252 Fluence (E > 0.1 Mev) [n/cm2-sec] 7.7440E+19 - 1.0238E+2 ' O.756 1.322 dpa- 3.4712E-02 4.3076E-02 0.806 1.241 Comparison of Calculated and Measured INTEGRATED Neutron EXPOSURE Rate for Capsule Z Calculated Measured C/M 1/ C/M Fluence (E > 1.0 Mev) [n/cm2-sec] 1.9701E+19 2 ?852E+19 0.862 1.160 Fluence (E > 0.1 Mev) [n/cm2-sec) 8.8989E+19 1.0048E+20 0.886 1.129 dpa 3.9540E-02 4.3396E-02 0.911 1.098 l
l I:
i 6-29 l:
- _ -__ ___ _ ___ ___ _ ______ = _ -- -
l l
. TABLE 6-17 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS i ON THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE ,
BEST ESTIMATE EXPOSURE (7.241 EFPY) AT THE PRESSURE VESSEL INNER RADIUS 0DEG 15 DEG 30 DEG 45 DEG E > 1.0 3.015E+18 4.457E+18 3.301E+18 4.562E+18 E > 0.1. 6.257E+18 9.333E+18 8.464E+18 1.138E+19 dpa 4.483E-03 6.576E-03 5.151E-03 6.951E-03 BEST ESTIMATE EXTRAPOLATION FLUX AT THE PRESSURE VESSEL INNER RADIUS 0DEG 15 DEG 30 DEG 45 DEG E > 1.0 1319E+10 1.950E+10 1.445E+10 1.996E+10 ,
E > 0.1 2.738E+10 4.084E+10 ' 3.7NE+10 4.980E+10 dpa 1.962E-11 2.878E-11 2.254E-11 3.N2E-11 ~
'(
ESTIMATE EXPOSURE (16.0 EFPY) AT THE PRESSURE VESSEL INNT.R RADIUS - .
ODEG 15 DEG 30 DEG 45 DEG E > 1.0 6.662E+18 9.847E+18 7.295E+18 1.008E+19 -
E > 0.1 1.382E+19 2.062E+19 1.870E+19 2.515E+19 dpa 9.906E-03 1.453E-02 1.138E4)2 1.536E-02 ESTIMATE EXPOSURE (32.0 EFPY) AT 'IEE PRESSURE VESSEL INNER RADIUS 0DEG 15 DEG 30 DEG 45 DEG E > 1.0 1332E+19 1.969E+19 1.459E+19 2.016E+19 E > 0.1 2.765E+19 4.124E+19 3.740E+19 5.029E+19 ,
dpa 1.981E-02 2.906E-02 2.276E-02 3.072E-02
+
t f
6-30 I
1 I
' TABLE 6-18 NEITIRON EXPOSURE VALUES .,
l FLUENCE BASED ON E > 1.0 MeV SLOPE .
0DEG 15 DEG 30 DEG 45 DEG !
1 16 EFPY FLUENCE SURFACE 6.662E+18 9.847E+18 7.295E+18 1.008E+19 .
1/4T 3.617E+18 5.327E+18 3.983E+18 5,503E+18 .
3/4T 7.727E+17 1.123E+18 8.608E+17 1.179E+18 32 EFPY FLUENG
- SURFAG 1.332E+19 1.969E+19 1.459E+19 2.016E+19 -
1/4T 7.234E+18 1.065E+19 7.966E+18 1.101E+19 3/4T 1.545E+18 . 2.245E+18 1.722E+18 2.359E+18 FLUENCE BASED ON dpa SLOPE 4 0DEG 15 DEG 30 DEG 45 DEG 16 EFPY FLUENCE SURFAG 6.662E+18 9.847E+18 7.295E+18 1.008E+19 1/4T 4.203E+18 6.164E+18 4.741E+18 6.693E+18 -
3/4T 1.459E+18 2.137E+18 1.743E+18 2.510E+18 l
32 EFPY FLUENCE j SURFAG 1.332E+19 1.969E+19 1.459E+19 2.016E+19 1/4T . 8.407E+18 1.233E+19 9.483E+18 1.339E+19 ;
3/4T 2.918E+18 4.274E+18 3.487E+18 5.019E+18 r
t
.T I
P 6-31 -}
i
,l TABLE 6-19 UPDATED LEAD FACTORS FOR BYRON UNIT 1 SURVEILLANCE CAPSULES CAPSULE LEAD FACTOR- ].
U 5.25 V 4,72*
X 531 _i W 532 l Y 4.72 l Z 532* '
-i
- WITHDRAWN EOC 8, BASIS FOR THIS ANALYSIS l
b S
Y f
i 6-32 l
- c ..~
SECTION 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following surveillance capsule removal schedule meets the requirements of ASTM E185-82 and is recommended for future capsules to be removed from the McGuire Unit I reactor vessel:
l TABLE 7-1 McGuire Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule . !
Removal Time Fluence 2
Capsule Location Lead Facter (EFPY)W (n/cm )*
U 56* 5.25 1.06 4.712 x 10" M X 236 5.31 4.33 1.409 x 10" "
V 58.5 4.72 7.24 2.186 x 10" N Z 3(M 5.32 7.24M 2.285 x 10" N 1 Y 238.5 4.72 10 2.97 x 10" W 124 5.32 Stand-By --
(a) Effective Full Power Years (EFPY) fmm plant startup.
(b) Actual measured neutmn fluence (c) Capsule Z was removed and disassembled. The specimens were placed in storage and the dosimeters analyzed.
(d) E > 1.0 MeV 7 - . - . . _ - ___ _ _ . _ . . _ _ _ .
p ..p .,44 gA.im e .m .m. 4 .. .n ,-.m a m aw % s 2.+. .+,- .m.a -a wae s.4*= a ,e a war w.ver -a.=-,w w w mee m > m = 4ms a m- -
--.:..a e- s e-- a,+ es nsa -- --
-ma me. smaw r. s a a= m..w is. e ./. A m. . a. A . 4 1
I T
i g ,-
+
i i
b.
2 1
5 7
B y'
e f
n 9
B -
t i ' i I
if
. ~
f i
4
+
7 3
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?
a k
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L t
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t 0,
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- k i.
I.
y 1
1-s G.
h^
)
re NTw w w-t lut h ie we em w- w v'* ewwNm er aw- 956M n* w An-T '*e "
W" 'V-*=F"'***W'MP-- ty-=='sW"'-- fM' f *'wP VFt- 9
SECTION
8.0 REFERENCES
- 1. Davidson, J.A. and Yanichko, S.E., Duke Power Company William B. McGuire Unit No.1 Reactor Vessel Radiation Surveillance Program, WCAP-9195, November 1977,
- 2.Section III of the ASME Boiler and Pressure Vessel Code, Appendix 0, Protection Against Nonductile Failure.
- 3. ASTM E208, Standard Test Methodfor Conducting Drop 4Veight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.
- 4. Yanichko, S.E., Congedo, T.V., Kaiser W T., Analysis of Capsule Ufrom the Duke Power l
Company McGuire Unit 1 Reactor Vessel Radiation Surveillance Program, WCAP-10786, February 1985.
- 5. Yanichko, S.E., Anderson, S.L., Albertin, L., Ray, N.K., Analysis of Capsule Xfrom the Duke Power Company McGuire Unit i Reactor Vessel Radiation Surveillance Program, WCAP-12354, August 1989.
- 6. Code of Federal Regulations,10CFR50, Appendix G, Fracture Toughness Requirements, and Appendix H. Reactor Vessel Material Surveillance Program Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C.
- 7. ASTM E185-82, Standard Practicefor Conducting Surveillance Testsfor Light 4 Vater Cooled Nuclear Power Reactor Vessels, E706 (IF), in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1993.
- 8. ASTM E23-92, Standard Test Methods for Notched Bar impact Testing of Metallic Materials, in +
ASTM Standards, Section 3, Amencan Society for Testing and Materials, Philadelphia, PA, ,
1992.'
f
- 9. ASTM A370-92, Standard Test Methods and Definitions for Mechanical Testing of Steel Products,in ASTM Standards, Section 3,' American Society for Testing and Materials, Philadelphia, PA,1992.
8-1
l lo, ASTM E8-91, Standard Test Methods of Tension Testing of Metallic Materials. in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphir., PA,1992, 1I. ASTM E21~79(l988), Standard Practicefor Elevated Temperature Tension Tests of Metallic Materials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA,1991.
I 2. ASTM E83-93, Standard Practicefor Verification and Classification of Extensometers, in .
ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.
- 13. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.
Nuclear Regulatory Commission, May,1988.
- 14. R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L. Ziegler, Nuclear Rocket Shielding . Methods, Modification, Updating and input Data Preparation. Vol. 5--Two-Dimensional Discrete Ordinates Transport Technique, WANL-PR(LL)-034, Vol. 5, August 1970,
- 15. ORNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded,47 Neutron,20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors.
I 6. R. E. Maerker, et al. Accountingfor Changing Source Distributions in Light Water Reactor Surveillance Dosimetry Analysis. Nuclear Science and Engineering, Volume 94, Pages 291308, 1986.
- 17. A. Saced. D. R. Gibson, M. A. Kotun, "The Nuclear Design and Core Physics Characteristics of the W. B. McGuire Unit 1 Nuclear Power Plant, Cycle 1," WCAP-9323-R1, August, 11,-1978.
- 18. ' J. R. Lesko, C. R. Savage, R.W. Miller, L. R. Rios, R. M. Turcovski, "The Nuclear Design of the W. B. McGuire Unit 1 Nuclear Power Plant, Cycle 2," WCAP-16463, January, l984.
- 19. J. R. Lesk.0, R.W. Miller, L. R. Rios Et. Al., "The Nuclear Design of the W. B. McGuire Unit i Nuclear Power Plant, Cycle 3," WCAP-10782, February,1985.
l 8-2 l
-l L
s
- 20. J. R. Lesko, L. R. Rios, D. A. Johnson, Et. Al., "The Nucleat Design of the W. B. AfcGuire Unit 1 Nuclear Power Plant, Cycle 4,".WCAP 11141, May,1986.
21.. J. R. Lesko, P. D. Banning, M. A Kotun, Et. Al., "The Nuclear Design of the W. B. AlcGuire Unit i Nuclear Power Plant, Cycle 5 " WCAP-1IS89,0ctober,1987.
4
- 22. J. R. Lesko, R.W. Miller, M. A Kotun, "The Nuclear Design of the W, B. McGuire Unit I-Nuclear Power Plant, Cycle 6," WCAP-ll2014, November,1988.
- 23. J. R. Lesko, M. A Kotun, "The Nuclear Desten of the W. B. AlcGuire Unit 1 Nuclear Power Plant, Cycle 7." WCAP-12544, April,1990.
- 24. Tonya M. IIall, "McGuire 1 Cycle 8 Final Fuel Cycle Design, Revision 3," Design engineering, Duke Power Company, October.1991.
- 25. ASTM Designation E482-89, Standard Guidefor Application of Neutron Transport Methodsfor Reactor Vessel Surveillance, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 26. ASTM Designation ES60-84, Standard Reconunended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, in ASTM Standards, Section'12, American Society for Testing and Materials, Philadelphia, PA,1993.
- 27. ASTM Designation E693-79, Standard Practicefor Characterizing Netaron 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.
1
- 28. ASTM Designation E706-87, Standard ofaster Afatrixfor Light-Water Reactor Pressure Vessel Surveillance Standard,in ASTM Standards, Section 12, American Society for Testing and .
Materials. Philadelphia, PA,1993.
- 29. ASTM Designation E853-87, Standard Practicefor Analysis and interpretation of Light-Water Reactor Surveillance Results, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1993.
8-3
) '
]
-I
' 30.- ASTM Designation E261-90, Standard niethodfor Determining _ Neutron Flux, Fluence, and Spectra by Radioactivation Techniques, in ASTM Standards, Section 12, American Society for --
Testing and Materials, Philadelphia, PA,1993.
- 31. ASTM Designation E262-86, Standard Afethodfor Afeasuring Thermal Neutron Flux by Radioactivation Techniques, in AUTM Standards, Section 12, American Society for Testing and :
Materials, Philadelphia, PA,1993.
t
- 32. ASTM Designation E263-88, Standard biethodfor Determining Fast-Neutron Flux Density by Radioactivation of fron, in ASTM Standards,- Section 12, American Society for Testing and i Materials, Philadelphia, PA,1993.
f
- 33. ASTM Designation E264-92, Standard biethodfor Determining Fast-Neutron Flux Density by Radioactivation of Nickel, in ASTM Standards, Section 12, American Society for Testing end Materials, Philadelphia, PA,19'J3.
34, ASTM Designation E481-92, Standard Afethodfor hieasuring Neutron-Flux Density by .
Radioactivation of Cobalt and Silver, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadciphia, PA,1993.
- 35. ASTM Designation E523-87, Standard Afethodfor Determining Fast-Neutron Flux Density by Radioactivation of Copper, in ASTM Standards, Section 12, American Society for Testing and {
Materials, Philadelphia, PA 1993.
- 36. ASTM Designation E705-90, Standard biethodfor hieasuring Reaction Rates by Radioactivation of Uranium.238, in ASTM Standards, Section 12. American Society for Testing and Materials, ;
Philadelphia, PA,1993.
-i 1
- 37. ASTM Designation E705-90, Standard biethodfor hieasuring Fast-Neutron Flux Density by; Radioactivation ofNeptunium-237,in ASTM Standards, Section 12. American Society for ,
Testing and Materials, Philadelphia, PA,1993.
4 38, ASTM Designation E1005-84, Standard Afethodfor Application and Analysis of Radiometric .
Afonitorsfor Reactor Vessel Surveillance, in ASTM Standards, Section 12. American Society for-Testing and Materials, Philadelphia, PA,1993.
8-4
.I
1' -;
. 39.' F. A. Schmittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford Engineering q Development Laboratory, Richland, WA, September 1979.
- 40. 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. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.
- 41. R. E. Macrker, et al., Development and Demonstration of an Advanced Methodologyfor LWR .
Dosimetry Applications. EPRI-NP-2188,1981.
I i
l~
i 8-5
-=_-_:-__-___-___-______-_-___--_______________________-___-______-______-_-__-_-__---__ .. .. .
r ,
1 l
APPENDIX A l l l.
Load-Time Records for Charpy Specimen Tests-i s
1 l
L-t__ __ _ _ _ _ . _ . . _ _ _ _ _ _ _ __ _ __ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ - -
o 2
O D A A O O L L E T R S .
U E T R C R A A R
- _
F g .
=, P P .
D -
A O - / l l i I i i I L
M d U r o _
M I
c e
X r A
e E
M m .
M I
it-s T d g a o
P l d
l I I i O I I I ) I 1 1 I I i I e
i z
l a
e d_
I 1
A -
t, e r .
u ig F
=
a m
t L
A R
E D 4
t A
E O -
G L
= O y L 1 1 l j l i , I E y gi g =.
P Y f t --
O40,
,0
n.
l i
i l
NCCU1RC et V* DL29 . a
., i 4
i j
l
? ._
E4 one h
99- n w
v_
f,1
.a
- f. -
. 4
^^ , i 6.0 o .-
3.6 4.8 ,
.0 1.2 2.4 ;
T!?tC ( MSCC )
MCGUIRE #1 "V" DL29 :
, necu1RC et v- our s
, i M
o e a
n S 9- n v
v-a i
"f.
--- i a i o i 3.6- 4.8 6.0
.D 1.2 2.4 TIMC ( MSCC >
MCGUIRE #1 "V" Batch:DL17 Figure A-2. Load-time records for Specimens DL29 and DL17 A-2 l
- -. - 1 J
MCCU1Rf #1 'V' OL23
, s s s
s 4
M 2*
m a
$ at-w
- t 3.6 4.8 6.0
.D 1.2 2.4 TIPC ( MSEC )
MCGUIRE #1 "V" D7.23 :
The load-time record for Specimen DL30 is not available because of computer system malfunction l
1 Figure A-3. Load-time records for Specimens DL23 and DL30
-I I
A-3 l I
4
'2 MCCUIRE 81 "V" OL22
, a 1 4
i j
7 e_
3- & 4-S n*-
- w v, -
u
- 9. -
o- , '^^T ,
4.8 6.O 1.2 2.4 3.6
.0 ( MSEC )
TIMC MCGUIRE #1 "V" DL22 :
MCCUIRC 61 'Y' DL18
, s u a s 3
?_
a
- t. 3 t- n w
l 9-e i i o i 3.6 4.8 6.0
.D 1.2 2.4 TINC ( ftSEC )
i MCGUIRE #1 "V" DL18 :
Figure A-4. Load-time records for Specimens DL22 and DL18 A-4
- MCcUIRE 41 'V* d ' ' ,
~
n a w a
3 *M-w N
af -
I 3.6' 4.8 6. 0 g 1. 2 ' 2.4' rinc < mcc >
MCGUIRE #1 "V" DL2 (y' :
4 necuzRc et "v' a.16 9
e
, . i
~
a ._
h
$ e*-
v i _
f _ - - - .,
- 4.s 6. 8 g 3. a ' a.4' 3.6 Tinc < mcc >
MCGUIRE #1 "V"-
DL16 :
Figure A-5. Load-time records for Specimens DL24 and DL16 A-5
MCCU1RC et 'V' DL25
, e i i i
g a4 li 3: n3_
v e
.a y_
f _
s i
. 6. 0 o .
2.4 3.6 4.8 I .0 1.2 Til'E < MSEC )
l l
MCGUIRE #1 "V" DL25 :
MCCUIRE S1 "V" OL21
, , i a i
7m
$ 3_ n v
a q
f -,___ ,
o n
- 3. 6 4.9 6. 0
.D 1.2 2.4 TIME < MSEC )
MCGUIRE #1 "V" DL21 :
Figure A-6. Load-time records for Specimens DL25 and DL21 A4
OL20
,_ _MCCU1RC et *va i i i g 4 h
39-ci w
q-3.6 4.8 6.0
.D 1.2 2.4 TIME ( fCEC )
MCGUIRE #1 "V" DL20 :
The load-time record for Specimen DL19 is not available because of computer system malfunction Figure A-7. Load-time records for Specimens DL20 and DL19 R
A-7
'l J
MCCU1RE et "V" DL28
, s 4
i i j
7m m
ci w
.J 9.
I f ,
e .
- 3. 6 i
4.8 6.0
.D 1.2 2.4 TIME < PtSEC )
MCGUIRE #1 "V" DL28 :
4 MCCUIRC #1 *v" Out6 a
i s 4
j 7_
e S *e-w
.a
- n. -
f .
o .
3.6 4.8 6.O
.D 1.2 2. 4 -
TIME ( PTSEC ) >
MCGUIRE #1 "V" DL26 -:
Figure A-8.' Load time records for Specimens DL28 and DL26 A4
l l
I MCCUIRE #1 "Y* CL27
, a i a
g a a4 2
a 51- n w
u 3.6 4.8 6. 0
.D 1.2 2.4 TIMC ( MSEC )
MCGUIRE #1 "V" DL27 :
MCCUIRC #1 "V" DT26
, i i i j i l
?._
EE 2
4 ue
$ 9- "
v f
w a
- f - i i
e i 6.0 0 1.2 2. 4 3.6 4.8 TIME ( MSEC )
MCGUIRE #1 "V" DT2r :
i Figure A-9. Load-time records for Specimens DL27 and DT26 j 1
1 A9
E PCCUIRC #1 "V" - 0T27
, o . .
g 7 ._
h S *e-v q_
, i.? ....
( fGEC )
... ... .0 TIMC MCGUIRE #1 "V" DT27 :
l MCCU1RE #1 "v' DT17
- i 8
. i #
e h
" ci w
a 3.
J 4 g
2.4 3.6 4.8 6. 0
., 1.2 TIME C FCEC )
MCGUIRE #1 "V" DT17 :
Figure A-10. Load-time records for Specimens DT27 and DT17 A-10
.l I 1
i MCCUIRC #1 ava DT20 i
- i 3
-j
. i a
- 1 1
1
- )
7 m-SI E
a g .- . -
a w
w
.a u-
, ,__ _7 6.0
.o 't.2 2.4 3.6 4.8 T!!E C MSTC )
MCGUIRE #1 "V" DT20 :
, MCCUIRC 01 "Y' 0725 a i a j i 7 m-E4 '
7
~
$ *n-v N
q-
, 7-____ , ,
6.0
.0 1.2 2.4 3.6 4.8 TIME C MSEC )
--MCGUIRE #1 "V" DT25 :
Figure A-11. Load-time records for Specimens DT20 and DT25 -
A-11 l
2
]
, NCCUIRE 01 "V* DT23
, a i g i 7 e_
3 *n_
v y-i o i i 2.4 3.6 4.8 6. 0
.D 1.2 T1?E ( MSEC >
MCGUIRE fl "V" DT23 :
McCUIRE e1 *va OT29
, s a i
j i 9 e_
m.,_
e v
y_
7 - _.
1.2 2.4 3.6 4.8 6.0
.0 Tite ( PCEC )
MCGUIRE #1 "V" DT29 :
Figure A-12. Load. time records for Specimens DT23 and DT29 A 12
'l 1
i 1
MCCUtRC et av* 0730 i i
,. i a
&4
~
n -
n l
y P
~
s i
J n_
. i l
)
o 4.8 6.O i
2.4 3.6 '
.D 1. 2 .
T!f1C ( MSEC ) ;
MCGUIRE #1 "V" DT30 :
I
)
I Mccurnt et ava oTis a
, i i -l
. a
? e_ ,
.1
&,4 .- !
a A
i
$ *n.- 1 i
w v.
n-
, 4.8 6. 0 2.4 3. 6
.0 1.2 Titt ( ttSEC >
MCGUIRE #1 "V" DT16 :
Figure A 13. Load time records for Specimens DT30 and DT16 A 13
MCCUIRE e1 "V"' DT24
, i i i 4 j
2 a_
a4 7
a S n9-v as q_
n a
. 6. 0 o s 2.4 3.6 4.8
.D 1.2 TIPC ( MSEC )
MCGUIRE #1 "V" DT24 :
MCCUIRE 01 *V* DT19
, 3 4 I i 7 e_
a4 7
s S n9-v l -
! J 9-o ,
4.8 6. 0
.D 1.2 2.4 3.6 t
' TIME - ( MSEC )
MCGUIRE #1 "V" DT18 :
l :
Figure A.14. Load-time records for Specimens DT24 and DT18 A 14
Mccu!RC #1 "V" 0T29
, i i i
. 4 7 m_
9 *a-v s
N q_
, i o . ,
3.6 4.8 6. 0
.0 1.2 2.4 TINC ( MSCC >
MCGUIRE #1 "V" DT28 :
MccutRc et v- orar
, i i i j i h
$ *n-w
+-
S_
o .
3.6 4.8 6. 0
.D 1.2 2.4 TINC ( MSEC )
MCGUIRE #1 "V" DT19 .:
Figure A-15. Load-time records for Specimens DT28 and DT19 A 15
NCCU1RC 41 ava DT21 e
, i s i
g 2 e-a4 39- e v
y-
- ~
. . i o .
3.6 4.8 6. 0
.0 1.2 2.4 TIME < ftSEC )
MCGUIRE #1 "V" DT21 :
nCouinc et v- 0722 4
7 e-a4 e
S 9- n w
.,a ru -
N o .
6.0
.D 1.2 2.4 3.6 4.8 TIME < ft EC )
MCGUIRE #1 "V" DT22 :
Figure A-16. Load-time records for Specimens DT21 and DT22
[
l A-16 L.
tcCU1RE et *va cual
, i i s e j
- e a4 m
S a9-v v
a
^! -
, e .
o ,
- 6. 0
.D 1.2 2.4 3.6 4.8 TIMC ( ttSEC >
MCGUIRE #1 "V" DW21 :
l, MCCUIRE 01 'V' DW17
, i
, . i
- e-f-
39- "
v r i
.a aJ e .
i 2.4 3. 6 4.8 6. 0
.9 1.2 TIME ( PtSEC )
MCGUIRE #1 "V" DW17 :
Figure A-17, Load-time records for Specimens DW21 and DW17 A.17 I
, MCCU1RE el "V" DW27 1 i a g a i
- e i 4 2
3 ., _
es v
e, N
N_
I o I i i
- 6. 0
.D 1. 2 - 2.4 3.6 4.8 TIME < MSEC )
MCGUIRE #1 "V" DW27 :
l I
MCCU1RE 01 "V" DU22
! i i 3 4 e t.
a4 7
n S*- m w
uno N b
- i. -
N_
l i
i i e i 4.8 6.0 1.2 2.4 3. 6
.D TIME < PCEC )
MCGUIRE #1 "V" DW22 : ,
Figure A.18. Load-time records for Specimens DW27 and DW22 A-18 i-
i l
i l
4 NCCUIRE #1 'v* Ou26 i
9 e
i i i
7 m-AJ li a
m 4
" d v
e l
8 9-e a
o i i
4.8 6. O 2.4 3.6
.9 1.2 TIME < tCCC )
MCGUIRE #1 "V" DW26 :
l
, MCCUIRE #1 *V" 3 424 i
s a g a i
)
Om E4
~
a w
a q-C i .
o i 4.8 6. O
.D 1.2 2.4 3.6 TIMC ( MSEC ) l l
MCGUIRE #1 "V" DW24 :
i Figure A-19. Load-time records for Specimens DW26 and DW24 ,
A 19
, MCCU1RC 61 v' DWI6 i a i e g
7.-
h a
39- M w
w-
.s q-o . .
3.6 4.8 6. 0
.D 1.2 2.4 TIME ( MSCC )
MCGUIRE #1 "V" Batch:DW16 Mcculac en v- Duas
, i i i i h
n S M9-w a
y-
?
I . i o 4 3.6 4.8 6.0
.D 1.2 2.4 TIME ( MSEC )
MCGUIRE #1 "V" DW25 :
Figure A-20. Load-time records for Specimens DW16 and DW25 A-20
?-
MCCUIRE #1 'V" DW19
, a i a g i O=-
a4 2
a 39- n v
y-o .
2.4 3.6 4.8 6.0
.D ' 1. 2 TIME ( M:EC )
MCGUIRE #1 "V" DW19 :
MCCUIRE 81 "V" Du29 ,
i i j
~
e *-
a e
39- n w
a q-o .
2.4 3.6 4.8 6.0
.D 1.2 TIME ( MSEC )
MCGUIRE #1 "V" DW29 :
Figure A-21. Load time records for Specimens DW19 and DW29 A-21
MCCU!RE #1 "V"- Ou?8
, s a 4
i w
~
.m. *-
o e a
a
' s e-n w
_ as 3-q-
- 2 i . .
3.6 4.8 6. 0
.0 1.2 2.4 l.
' T!MC ( ICEC )
MCGUIRE #1 "V" DW28 :
t I
i MCCUIRE 81 "V" Dule
, i e i j j i l
7 m-a4
~
m a
= *-
m v
A I -
i 3-l i
o .
2.4 i
3.6 4,8 6.0
.D 1.2 TIME ( ftSEC >
MCGUIRE #1 "V" DW18 :
Figure A 22. Load-time records for Specimens DW28 and DW18 i-I f
' A-22 l
i s
NCGU!RE 01 "V' Du30 i a i j
84 a
3 e, _
M _j w
v, a
u-
- i o .
1.2 2.4 3.6 .4.8 6.0
.D TIME ( MSEC )
MCGUIRE #1 "V" DW30 :
l l
, MCcutRE #1 "Y" Du23 i . . ,
e 7 m- -
.E m
4 A
1' =e-is v
I e-i, J eu -
1
.D 1.2 2.4 3.6 4.8 6.0 i
TIME < M::EC >
MCGUIRE #1 "V" DW23 :
Figure A-23. Load-time records for Specimens DW30 and DW23 A 23
a> - -
i l
1 MCCUIRC e1 *V' Ou20 ,
, i a -s i
j i l
7-._
a4 2
4 5 *n-v 3.6 4.8 6. 0
,3 1.2 2.4 TIMC ( MSCC )
MCGUIRE #1 "V" DW20 :
- l. :
, MCCU1RC 01 "V" DH21 i i i g i 2 ._
a4 2
^
9 *M-fi -
v s,
q_
o . i 3.6, 4.8 6.0
.D- 1.2 2.4' TIMC ( MSCC ) -
MCGUIRE /1 "V" DH21 :
Figure A-24. Load-time records for Specimens DW20 and DH21 A-24
- (p r' 2
MCCUIRE 81 "Y' DH22
. I I I 4 4
7 m-84
$ n *-
w w-a n-
~ ^ -
o 4.8 6. 0
.D 1.2 2.4 3. 6 TINC ( tcEC )
MCGUIRE /1 "V" Dil22 :
necurst et av* oHis
, i e
. i i e
7e S4 e
e
.s.
34- M v
a 9"
4.8 6. 0
.0 1.2 2.4 3.6 rinc ( n:Ec >
MCGUIRE #1 "V" DlI18 :
Figure A.25. Load. time records for Specimens D1122 and Dil18 I
l A 25
MCCUIRC 41 "V" OH19
, s :
i e j
7 ._
a4 iii a
S M9- .
w a
"f -
I w -_ _. __
o i i 3.6 4.8 6. 0
.D 1.2 2. 4 ftrc < Msce >
MCGUIRE #1 "V" DH19 :
NC!1J1RC #1 "Y" OH27
, a i i i g
? ..
h t S 9- n v
v- t
~
- n. -
f ---
. . i 4.8
.D 1. 2 - 2.4 3.6 6.0
-TIMc ( MSEC >
MCGUIRE #1 "V" DH27 :
e l
Figure A-26. Load-time records for Specimens DH19 and DH27 l l
A 26
NCCutRc #1 Y' CH24
, i i e
g i 7 m_
&4 1ll n
S 9,-
c ,
w n-h o 6.0 1.2 2. 4 3.6 4.8
.V TITE < PtSEC )
MCGUIRE #1 "V" DH24 :
m;curac et v- aes i i i j .
t.
y**._
~
I l
m
) m I-s 2- es v
e-
.a q- -
o ,
1.2 2.4 3.6 4.8 6. 0
.D TIMC ( PtSEC )
MCGUIRE #1 "V" DH25 :
J Figure A.27. Load-time records for Specimens DH24 and DH25 l
A-27 1
r--
tCCU1RE 91 'v' OH26 j
? ._
S4
~
A 3 *a -
w a
ru _
4.8 6.0 1.2 2.4 3.6
! .D TIME ( ttSEC )
l MCGUIRE #1 "V" DH26 :
l MCCUIRC 41 'V* W16
, i i g
- e-a4 7
n 59- n w
a su .
o , s
- 3. 6 4.8 6. 0
.D 1.2 2.4 TIME ( tGEC )
MCGUIRE #1 "V" DH16 :
Figure A-28. Load-time records for Specimens DH26 and DH16 A 28 l-
MCCUIRC et "V' ' DH28 i
9 i e i
?-
84 3*- n w
y-o , i i
6.0
.O 1.2 2.4 3.6 4.8 T!!T ( MSEC )
MCGUIRE #1 "V" DH28 :
, MCoutRc et v- DH23 i i i g i 7 m-h
$ n*-
v
.a y-I a G &
.D 1.2 s
2.4 .
3.6 4.8 6.0 TIME ( PCEC )
MCGUIRE #1 "V" DH23 :
Figure A-29. Load-time records for Specimens DH28 and DH23 A 29
)
.a 3
MCCUIRC et av* D430
, a i i
j 7=_
h S n9- .
w w_
q.
3.6 4.8 6. 0
.D 1.2 2.4 TIME < MSEC )
i MCGUIRE #1 "V" DH30 :
Mccurat si v- ossy
. i . . e j
0=_ _
h s *.n -
w k
$~
a N.-
~l l
o i 6.0 i
1
.D 1.2 2.4 3.6 4.8
( MSEC ) )
TIMt MCGUIRE #1 "V" DH17 :
Figure A-30. Load-time reccrds for Specimens DH30 and DH17 A 30
The load-time record for Specimen DH2O is not available because of computer system malfunction
, Mccutar et v- otas i i
- e i 7_
m.,_
M w
[ h
.a N-
, i o . .
4.8 6. 0
[' .D 1.2 2.4 3.6 TIME ( MSEC )
MCGUIRE #1 "V" DB29 :
pH :
Figure A-31. Load time records for Specimens DH2O and DH29 l
- i i
A-31 .
, .)
'i l
I i
i 1
5
+
l t
APPENDIX B :
Heatup and Cooldown Limit Curves for Normal Operation - ,
t l?
i i
k f
f e
1 L
4 e
', E_
. . .- . . .. =. _. --. ._... . .
l TABLE OF CONTENTS ,
1 Section Title Pane LIST OF ILLUSTRATIONS B-2 l LIST OF TABLES B2 l
l B-1 INTRODUCTION B-3 l B-2 FRACTURE TOUGHNESS PROPERTIES B-3 i
B-3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE ' B-4 !
l RELATIONSHIPS j B-4 IIEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT B-7 CURVES ,
1 B-5 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE B-9 j B-6 REI~ERENCES B-21 ATTACllMENT B1 DATA POINTS FOR HEATUP AND COOLDOWN B-24 ;
i
'l CURVES (WITHOUT MARGINS FOR INSTRUMENTATION ERRORS)
I ATTACHMENT B2 HEATUP AND COOLDOWN CURVES AND DATA POINTS B OVITH MARGINS OF 12*F AND 30 PSIG FOR INSTRUMENTATION ERRORS) i
'd B1 Id
- -m-
, . . . ~ . . . . . . _ _. . .. . ._
LIST OF ILLUSTRATIONS Figure Title Page B1 McGuire Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate B-19 of 60*F/ar) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors)
B-2 McGuire Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown B-20 l
Rates up to 100*F/hr) Applicable for the First 16 EFPY (Without Margins ,
for Instnimentation Errors) f B-3 McGuire Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate - ' B-27 of 60*F/hr) Applicable for the First 16 EFPY (With Margins of 12*F and 30 psig for Instrumentation Errors)
B-4 McGuire Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown B 28 Rates up to 100*F/hr) Applicable for the First 16 EFPY (With Margins of 12*F and 30 psig for Instnzmentation Errors) i.
, LIST OF TABLES Table Title Page-B-1 Calculation of Average Cu and Ni Weight Percent for McGuire Unit 1 B-Il B-2 McGuire Unit 1 Reactor Vessel Toughness Table (Unirradiated) B 15 B-3 Calculation of Chemistry Factors Using Surveillance Capsule Data B-16 B-4 Summary of Adjusted Reference Temperatures (ART's) at 1/4-t and 3/4-t B-17 ;
Locations for 16 EFPY B-5 Calculation of Adjusted Reference Temperatures at 16 EFPY for the B-18 Limiting McGuire Unit 1 Reactor Vessel Materials - lower Shell Longitudinal Welds 3-442A & C and Lower Shell Plate B5013-2 B-2 i
. ._ _. - - . . ~ ,
. . _ - _ . ._ m .
B-1. INTRODUCTION l Heatup and cooldown limit curves are calculated using the most limiting value of RTym (reference
]
~ nil-ductility temperature) corresponding to the limiting beltline region material for the reactor vessel.
The most limiting RTxm of the material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties and estimating the ,
radiation-induced ARTsm. The unirradiated RT 3m is designated as the higher of either the drop weight nil-ductility transition temperature (NDIT) or the temperature at which the material exhibits at least 50 fr-lb of impact energy and 35-millateral expansion (normal to the major working direction) minus 60'F.
RTum increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTym t aany time period in the reactor's life, ARTym due to the radiation exposure associated with that time period must be added to the original unirradiated RT 3 m. The extent of the shift in RT ym is enhanced by certam chemical elements (such as copper and nickel) present in reactor vessel
! steels. The Puclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2 (Radiation Embrittlement of Reactor Vessel Materials) inn. Regulatory Guide 1.99, Revision 2 is used for the calculation of Adjusted Reference Temperature values (irradiated RTym with margins for uncertainties) at 1/4-t and 3/4-t locations. "t" is the thickness of the vessel at the beltline region measured from the clad / base metal interface. ,
l The pressure-temperature limit curves in Figures B-1 and B-2 of this report do not include margins for . <
instrumentation errors or for pressure differences between the wide-range pressure transmitter and the' limiting reactor vessel beltline region. The pressure-temperature limit curves in Attachment B2 of this report only include instrumentation error margins of 12 F and 30 psig; no margins are included for .
pressure differences between the wide-range pressure transmitter and the limiting reactor vessel beltline i
region.
l' B-2. FRACTURE TOUGHNESS PROPERTIES i The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are 4 determined in accordance with the NRC Regulatory Standard Review Flantau. The pre irradiation fracture-toughness properties of the McGuire Unit I reactor vessel are presented in Table B-l. The -
post-irradiation fracture toughness properties of the reactor vessel beltline material were obtained
}
B-3 e
q
'directly from the McGuire Unit I and Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Propams, Credible surveillance data is available for three capsules (Capsules U, X and V) for McGuire Unit I and two capsules (Capsules U and X) for Diablo Canyon Unit 2.
B 3. CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Kr, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, Ku, for the metal temperature at that time. K uis obtained from the reference fracture toughness curve, defined in Appendix G of the ASME Code!"'I. The Ku curve is given by the following equation:
l Ku = 26.78 + 1.223
- e tous o am . imi
_ (i) l where. ,
Ku = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTym Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Codd*'l as follows:
C
- Km + Krr s Ku (2) where.
Km = stress intensity factor caused by membrane (pressure) stress Krr = stress intensity factor caused by the thermal gradient < ;
Ku = reference stress intensity factor as a function of temperature relative to the RTm of the -
material C = 2.0 for Level A and Level B service limits B-4
C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical At any time during the heatup or cooldown transient, Ku is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTm, and the reference fracture toughness curve.
The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kr r, for the reference flaw are computed.
From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall.
During cooldown, the controlling location of the flaw is always at the inside of the wall because the i thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates.
Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constmeted for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4-t vessel location is at a higher temperature than the fluid adjacent to the vessel irmer diameter, This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of Ku at the 1/4-t location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in Ku exceeds Krr, the calculated allowable pressure during cooldown will be greater than the steady-state value.
The above procedures are needed because there is no direct control on temperature at the 1/4-t location .
and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
4 B-5
i l
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4-t defect !
at the inside of the wall, The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the .
coolant temperature; therefore, the K3 for the 1/4-t crack during heatup is lower than the Ku for the 1/4-t crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower Ku's do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the -
1/4-t flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady state and finite.
heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4-t deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These ,
thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along j the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for both the steady state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under .
consideration. The use of the composite curve is necessary to set conservative heatup limitations 4
. because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.
Finally, the 1980 Amendment to 10CFR50" has a rule which addresses the metal temperature of the i closure he6 flange and vessel flange regions. This mie states that the metal temperature of the i
closure flange regions must exceed the material unirradiated RTm by at least 120*F for normal
- B-6
)
i 1:
operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for McGuire Unit 1).
Table B 2 indicates that the limiting unirradiated RT 3m of 40*F occurs in the closure head flange of the McGuire Unit I reactor vessel, so the minimum allowable temperature of this region is 160'F at pressures greater than 621 psig ' Itis limit is shown in Figures B-1 and B-2 whenever applicable.
Figures B-3 and B-4 in Attachment B2 contain margins of 12*F and 30 psig for instntmentation errors; for these curves the minimum allowable temperature is 172*F at pressures greater than 591 psig.
B-4. HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Fressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using t
the methods discussed in Section B-3 a3:1 of this repon.
The time and position dependent temperature solution utilized in both the heatup and cooldown analyses was based on the one-dimensional transient heat conduction equation:
pC H = K [ 32T + 1 BT ]
dt dr2 r 3r With the following boundary conditions applied at the inner and outer radii of the pressure vessel:
At r = ri, -K BT = h (T - T,)
Br At r = r , ,3T = 0 dr where h= heat transfer coefficient between the coolant and the vessel wall, a conservative value of 7000 (Btu /hr-ft' *F) was used p = density (@70*F = 490.9 lbm/ff, @ 550 F = 484,7 lbm/ff) ,
C = specific heat (@70 F = 0.l(M Btu /lbm *F, @ 550 F = 0.130 Btu /lbm- F)
K = conductivity (@70 F = 26.42 Btu /hr ft- F, @ 550 F = 23.90 Btu /hr-ft *F) '
T = wall temperature variable r = radius variable t = time variable l
j l
B7 l
The above is solved numerically to obtain the results of the heat transfer analysis as input; position and time dependent temperature distributions of hoop thermal stress were calculated for each heatup and cooldown rateA Since indication of reactor vessel beltline pressure is not available on the plant, the pressure difference between the wide range pressure transmitter and the limiting beltline region must be accounted for when using pressure-temperature limit curves presented in Figures B-1 through B-4. The limit curves presented in Figures B-3 and B-4 include only instrumentation error margins.
Figure B 1 presents the heatup curve without margins for instrumentation errors and pressure differences using a heatup rate of 60*F/hr applicable for the first 16 EFPY. Figure B-2 presents the cooldown curves without margins using cooldown rates up to 100*F/hr applicable for the first 16 -
EFPY. Margins of 12*F and 30 psig for possible instrumentation errors are included in the development of heatup and cooldown cun'es found in Attachment B2. Allowable combinations of -
temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures B 1 through B-4. This is in addition to other criteria which must be met before the reactor is made critical.
The reactor must not be made critical until pressure-temperature combinations are to the right of the -
criticality limit line shown in Figures B-1 and B-3. The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10CFR Part 50. The governing equation for the hydrostatic test is defined in Appendix G to Section XI of the ASME Code as follows:
1.5 Km .5 Ku, where, ;
Km is the stress intensity factor covered by membrane (pressure) stress, K 3= 26.78 + 1.233S e " * *"' ",
T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature.
The criticality limit curves shown in Figures B 1 and B-3 specify pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference.B4.
The pressure-temperature limits for core operation (except for low power physics tests) are that the B-8
l reactor vessel must be at a temperature equal to or higher than the minimum temperature required for j the inservice hydrostatic test, and at least 40*F higher than the minimum permissible temperature in
~t he corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section B-3. The minimum temperature for the inservice hydrostatic leak test (without margins for instrumentation errors) for the McGuire Unit I reactor vessel at 16 EFPY is 282*F. A vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40*F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.
Additionally, the minimum temperature for the inservice hydrostatic leak test (with instrumentation error margins of 12*F and 30 psig) for the McGuire Unit I reactor vessel at 16 EFPY is 295'F. This . ,
is shown in Figure B-3. ,
Figures B-1 through B-4 define limits for ensuring prevention of nonductile failure for the McGuire .
r Unit I reactor vessel. The data points used to develop the heatup and cooldown pressure-temperature limit curves shown in Figures B-1 through B-4 are presented in Attachments B1 and B2.'
B-5. CALCULATION OF ADJUSTED REFEIULNCE TEMPERATURE i
From Regulatory Guide 1.99, Revision 2rnu he t adjusted reference temperature (ART) for each material in the beltline is given by the following expression:
ART = Initial RTsm + ARTym + Margin (3)
Initial RTym is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code. If measured values of initial RTym for the material in' question are nuc available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.
ARTsm is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:
- f a2s .m i. o (4)
To calculate ARTym at any depth (e.g., at 1/4-t or 3/4-t), the following formula must first be used to attenuate the fluence at the specific depth.
I B-9
]
l 4
- )
f <e.g. x) = f.ma.
- e (5)
I where x (in inches) is the depth into the vessel wall measured from the vessel clad / base metal ;
i interface. The resultant fluence is then put into equation (4) to calculate ART, at the specific depth.
The calculated surface fluence for McGuire Unit 1 intermediate and lower shells and circumferential weld at 16 EFPY is 1.008 x 10" n/cm2 ts"1, The calculated surface fluences for McGuire Unit I longitudinal welds at 16 EFPY at the 0 and 30 azimuthal angles is 6.662 x idf n/cm2 and 7.295 x 10" n/cm2, respectively!""1 P
Material property values were obtained from material test certifications from the original fabrication as well as the additional material chemistry tests performed as part of the surveillance capsule testing programtssi. In addition, pertinent chemistry test results were also obtained from other plants with similar weld materials. Specifically, Diablo Canyon Unit 2 surveillance data was used to calculated a chemistry factor for the lower shell longitudinal welds. Justification for the use of this data can be found in Appendix D of this report. The average copper and nickel values were calculated for each of the beltline region materials using all of the available material chemistry information as shown in Table B-1.
CF (*F) is the chemistry factor, obtained from Tables in Reference B1, using the average values of the copper and nickel content calculated in Table B-1 and reported in Table B-2.- The chemistry factors were also calculated using the surveillance capsule data in Table B-3. All materials in the beltline region of McGuire Unit I reactor vessel were considered in determining the limiting material. The results of the ART's at 1/4-t and 3/4-t are summarized in Table B-4. From Table B-4, it can be seen that the limiting materials are the Lower Shell Longitudinal Welds 3-442A & C and Lower Shell Plate B5013-2 for heatup and cooldown curves applicable up to 16 EFPY. Sample calculations to determine the ART values for the lower shell longitudinal welds and lower shell plate at 16 EFPY are shown in Table B-5.
)
i s
B-10
Table B-1 Calculation of Average Cu and Ni Weight Percent for McGuire Unit 1 Intermediate Shell Longitudinal Weld Seams,2-442A, B & C (Ht. 20291 & 12008, Linde 1092, Flux Lot No. 3854) -
Reference wt. % wt. % 1 Cu Ni B5 0.21 0.88 B6 0.20 0.91 B31 0.195 0.87 B31 0.191 0.848 B31 0.193 0.863 Average 0.20 0.87 Intermediate to Lower Shell Circumferential Weld, 9-442 (Ht. 83640, Linde 0091, Flux Lot No. 3490)
Reference wt. % wt. %
Cu Ni B7 0.050 ---
B8 0.050 0.120 Average 0.050 0.120 B-11
if <
l ..
Table B-1 (Continued)
IAwer Shell Longitudinal Weld Seams. 3-442A. B & C - l (Ht. 21935 & 12008, Linde 1092, Flux Lot No. 3889) l l
l Reference wt. % ' wt. %
Cu Ni B9 0.220 --
B10 0.200 ---
B11 0.22 0.83 B12 0.23 - 0.90 0.21 0.76 0.22 0.90 B13 0.219 0.86 0.212 0.88-0.213 0.90 B14 0.225 0.875 0.213 0.856 0.225 0.877 ,
Average 0.22 0.86 Lower Shell IAngitudinal Weld Seams. 3-442A (Root Weld)* -
(Ht. 305424, Linde 1092, Flux lot No. 3889) ,
Reference wt. % wt. %
Cu Ni BIS 0.300 0.640 B16 0.260 0.620 t
B34 0.230 0.637 Average - 0.263 0.632
- The lower shelllongitudinal welds also contained a different weld wire heat in the double U root area of the weld. Since the root weld chemistry is not more limiting than the above weld data, it was not utilized in the evaluations.
B-12
-e l
')
1 Table B-1 (Continued) l Lower Shell' Longitudinal Weld Seams. 3-442B & C (Root Weld)* l (Ht. 21935, Linde 1092, Flux Lot No. ";889) !
Reference wt. % wt. %
Cu Ni B17 0.200 ---
B12 0.21 0.68 0.71 l Average 0.21 0.70 Intermediate Shell Plate. B5012-1 (Ht. C4387-2)
Reference wt. % wt. % -
Cu Ni B 18, B19 0.13 0.60 B5 0.087 ---
B20 ---
0.58 B31 0.117 0.643 Average 0.11 0.61 Intermediate Shell Plate, B5012-2 (Ht. C4417-3)
Reference wt. % wt. %
Cu- Ni B18, B21 0.14 0.62
[
B22 --
0.60 - ,
Average 0.14 0.61
- The lower shell longitudinal welds also contained a different weld wire heat in the double U root area of the weld. Since the root weld chemistry is not more limiting than the above weld data, it was not utilized in the evaluations.
B-13
Table B-1 (Continued)
Intermediate Shell Plate, B5012-3 Git.' C4377-2)
Reference wt. % wt. %
Cu Ni B18, B23 0.11 0.66 B20 --
0.65 Average 0.11 0.66 lower Shell Plate, B5013-1 git. C4315-1)
Reference wt. % wt. %
Cu Ni B18. B24 0.14 0.56 B25 ---
0.59 Average 0.14 0.58 Lower Shell Plate B5013-2 Git. C4374-2)
Reference wt. % wt. %
Cu Ni -
B18, B26 0,10 0.52 B27 ---
0.50 Average 0.10 0.51 Lower Shell Plate, B5013-3 Git. C4371-2)
Reference wt. % wt. %
Cu Ni B18, B28 0.10 0.55 B29 --
0.54 Average 0.10 0.55-B-14
Table B 2 McGuire Unit 1 Reactor Vessel Toughness Table (Unirradiated)
Material Description _ Cu (%)
- Ni(%)* I (*F) tam m Intermediate Shell Plate, B5012-1 0.11 0.61- 34 ,
Intermediate Shell Plate, B5012-2 0.14 0.61 - 0 l Intermediate Shell Plate, B5012-3 0.11 0.66 -13 Imwer Shell Plate, B5013-1 0.14 0.58 0. 1 lower Shell Plate, B5013-2 0.10 0.51 30 J Lower Shell Plate, B5013-3 0.10 0.55 15 i Inter. Shell longitudinal Welds, 2-442A, B & C 0.20 0.87 -50 Lower Shell IAngitudinal Welds,3-442A, B & C 0.22 0.86 - -50" Circumferential Weld, 9-442 0.05 0.12 ' 70 (a) Initial RTm values were estimated per U.S. NRC Standard Review Plan. The initial RTm values for the plates and welds are measured values, except for the lower shell longitudinal welds 'i (which are generic values).
- Diablo Canyon Unit 2 initial RTm value. Justification can be found in Appendix D of this report.
h
?
B-15 l
l j
l Table B-3 Calculation of Chemistry Factors Using Surveillance Capsule Data Material Capsule Fluence FF ARTym FF*ARTym FF2 Inter. Shell Plate, U 4.719 x 10" O.790 45 35.550 0.624 B5012-1 X 1.4091 x 10! ' l.095 45 49.275 1.199 (Longitudinal)
V 2.1858 x 10t ' 1.212 85 103.020 1.469 28 Inter. Shell Plate, U 4.719 x 10 0.790 50 39.500 0.624 B5012-1 X 1.4091 x 10 ' 2 1.095 65 71.175 1.199 (Transverse)
V 2.1858 x 10t ' l.212 85 103.020 1.469 Sum: 401.54 6.584 Chemistry Factor = 401.54 + 6.584 = 61.0 = 61 Intermediate Shell U 4.719 x 10 28 0.790 160 126.400 0.624 Iengitudinal Welds, 2-442A, B & C X 1.4091x 10 '2 l.095 165 180.675 1.199 V 2.1858 x 10 l.212 175 212.100 1.469 Sum: 519.175 3.292 Chemistry Factor = 519.175 + 3.292 = 157.7 = 158 Using Diablo Canyon Unit 2 Surveillance Data Material Capsule Fluence FF ARTym FF*ARTym FF2 Lower Shell U 3.51 x 10 O.711 174 123.7 0.506 Longitudinal Welds 3-442A, B, C X 8.87 x 10'8 0.966 204.2 197.3 0.933 Sum: 321.0 1.439 Chemistry Factor = 321.0 + 1.439 = 223.1 ,
Justification for the use of this data can be found in Appendix D of this report.
B-16
Table B-4 Summary of Adjusted Reference Temperatures (ART's) at 1/4-t and 3/4-t Locations for 16 EFPY -
Component 16 EFPY RTm 1/4-t (*F) 3/4-t (*F)
Intermediate Shell Plate, B5012-1 131.61 '111.41 Using Stuveillance Capsule Data" 103.30 86.69 Intermediate Shell Plate, B5012-2 119.95 92.65 Intermediate Shell Plate, B5012-3 85.21 M.82 Lower Shell Plate, B50131 118.96 91.98 Lower Shell Plate, B5013-2 119.73 102.03*
Lower Shell Plate, B5013-3 104.73 87.03 Inter. Shell Longitudinal Weld, 2-442A 157.93 105.89 Using Surveillance Capsule Data" 95.56 55.29
- Inter. Shell Longitudinal Welds, 2-442B & C 162.92 110.02 Using Surveillance Capsule Data" 99.42 58.48 l
I l: Lower Shell Longitudinal Weld, 3-442B 165.45 112.03 Using Surveillance Capsule Data"* 143.99 87.13--
Lower Shell Longitudinal Welds 3-442A & C 170.57 116.27 Using Surveillance Capsule Data"* 149.45* 91.65 .
Circumferential Weld, 9-442 -1.76 -23.43 e
- These ART numbers were used to generate heatup and cooldown curves.
- Numbers were calculated using a chemistry factor (CF) based on McGuire Unit 1 surveillance capsule data.
- Numbers were calculated using a chemistry factor (CF) based on Diablo Canyon Unit 2 surveillance capsule data. Justification for the use of this data can be found in Appendix D of this report.
1 g
l'
- 1. -
f B-17
-i L --____._ _.
Table B-5 Calculation of Adjusted Reference Temperamres at 16 EFPY for the Limiting McGuire Unit 1 '
Reactor Vessel Materials - Lower Shell Longitudinal Welds 3-442A & C and Lower Shell Plate B5013 2 Parameter Values Operating Time 16 EFPY .
Material Lower Shell Lower Shell Plate longitudinal Welds, B5013-2 3-442A&C Location 1/4-t 3/4-t Chemistry Factor, CF (*F) - 223,1 65.0 Fluence, f (10t ' n/cm )
- 2 0.4348 0.2134 Fluence Factor, ff 0.768 0.585 ARTxm = CF x ff (*F) 171.448 38.028
, Initial RTym, I (*F) -50 30 l
Margin, M (*F)
- 28 34 Adjusted Reference Temperature, (*F) 149.45 102.03 per Regulatory Guide 1.99, Revision 2 (a) Fluence, f, is based upon fa (10# n/cm2 . F>l MeV) = 0.7295 at 16 EFPY for the lower shell longitudinal welds. Fluence, f, is based upon fa (10 ' n/cm2 , E>l MeV) = 1.008 at 16 EFPY for the lower shell plate. The McGuire Unit I reactor vessel wall thickness is 8.625 inches at the beltline region.
(b) Margin is calculated as, M = 2 Yoi: # y,2, o, = standard deviation for the initial RTm: o, = 0 when Iis a measured value' s ai = 17 when I is a generic value For plates and forgings: o, = 17 when stuveillance capsule data is not used o, = 8.5 when surveillance capsule data is used For welds: o, = 28 when surveillance capsule data is not used o, = 14 when surveillance capsule data is used o 3not to exceed 0.5*.iRTw pes Regulatory Guide 199, Revision 2.
B-18
MATERIAL PROPFRTY BASIS LIMITING MATERIALS: LOWER SHELL LONGITUDINAL WELDS 3-442A & C AND LOWER SHELL PLATE B5013-2 LIMITING ART AT 16 EFPY: 1/4-t, 149.5'F 3/4-t, 102.0'F 2,500 iiiiiiiiiiiii ,
LEAK TEST LIMIT --_ ,1 ,
2,250 l l }
i I I I ( l R I O 2.,000 '
/ /
W h 1,750 UNACCEPTABLE OPERATlON
/ /
/ f Q ! .
L 1,500 / / -
g / / ACCEPTABLE g / / OPERATION l((((
e 1.,250 HEATUP RATE N [
Q- /
1.,000 /
y Q /
+J '
d 750 <
O '
[ 500 -
p 7- CRITICALITY LIMIT BASED ON /
250 - -. INSERVICE HYDROSTATIC TEST Z; TEMPERATUAE (282 'F) FOR THE
- SEAVICE PEAIOD UP TO 16 EFPY j 0 '''''''''''''''''' '
O 50 100 150 200 250 300 350 400 450 500 I nd i cated Temperature (Deg . F)
I Figure B-1 McGuire Unit 1 Reactor Coolant System lleatup Limitations (Heatup Rate of l 60*F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation l Errors)
B-19 e
MATERIAL PROPERTY BASIS LIMITING MATERIALS: LOWER SHELL LONGITUDINAL WELDS 3-442A & C AND LOWER SHELL PLATE B5013-2 LIMITING ART AT 16 EFPY: 1/4-t, 149.5'F 3/4-t, 102.0*F 2.,500 f
2.,250 )
n Ti.'
O) 2.,000 t
.._ f W '
O.
y 1.,750 UNACCEPTABLE ,/
OPERATlON j
@ /
L U
1.,500
/
/
W /
W '
e 1.,250 ,
ACCEPTABLE y j OPERATlON 1.,000 /
_O
@ n y a cd 750 cagggg,y , 7 0 RATES 'F/ HR . ,,, [
-- - -m V 500 III: *W '
C 20 Ao
- g /
/
l
~~~-
50 _, -
250 IZZ 'UU l
l 0
0 50 100 150 200 250 300 350 400 450 500 1ndicated Temperature CDeg. F) l l
Figure B-2 McGuire Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100*F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors)
B-20
B-6. REFERENCES
[Bl] Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials" U.S. Nuclear Regulatory Commission, May,1988.
[B2] " Fracture Toughness Requirements", Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Pour Plants, LWR Edition. NUREG-0800,1981.
[B3] ASME Boiler and Pressure Vessel Code,Section XI, Division 1 - Appendixes, " Rules for Inservice Instruction of Nuclear Power Plant Components, Appendix G. Fracture Toughness Criteria for Protection Against Failure", pp. 401-411,1989 Edition, American Society of Mechanical Engineers, New York,1989.
[B4] Code of Federal Regulations,10CFR50 Appendix G, " Fracture Toughness Requirements", U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Vol. 48 No.104, May 27,1983.
[B5] WCAP-9195, " Duke Power Cotupany William B. McGuire Unit No.1 Reactor Vessel Radiation Surveillance Program", J. A. Davidson and S. E. Yanichko, November 1977.
[B6] WCAP-10786, " Analysis of Capsule U from the Duke Power Company McGuire Unit i Reactor Vessel Radiation Surveillance Program", S. E. Yanichko, et al., February 1985.
[B7] Combustion Engineering, Inc., Metallurgical Research & Development, " Chemical Analysis of Wire-Flux Test Coupon", Job Number D32255,8-15-72.
[B8] WCAP-8819, " Central Nuclear de Almaraz, Almaraz Unit No.1 Reactor Vessel Radiation Surveillance Program", R. A. Smith, et al., December 1976.
[B9] Combustion Engineering, Inc., Welding Material Certification and Release for Section III, " Chemical Analysis of Test Weld Sample", August 12,1969.
[B10] Combustion Engineering, Inc., Metallurgical Research & Development, " Chemical Analysis of Wire-Flux Test Coupon", Job Number X-32255,10-14-69.
[Bil] WCAP-8783, " Pacific Gas and Electric Company Diablo Canyon Unit No. 2 Reactor Vessel Radiation Surveillance Program", J. A. Davidson and S. E. Yanichko, December 1976.
[B12] WCAP-10472, " Evaluation of Diablo Canyon Units 1 and 2 Reactor Vessel Beltline Weld Chemistry", S. E. Yanichko and M. K. Kunka, September 1983. (Proprietary)
[B13] WCAP ll851, " Analysis of Capsule U from the Pacific Gas and Electric Company Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program", S. E. Yanichko, et al., May 1988.
B-21
y i
J
.- [B14] WCAP-12811, " Analysis of Capsule X from the Pacific Gas and Electric Company.
1 Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program", E.'Terek, et al.,
December 1990. !
[B15] Combustion Engineering, Inc., Metallurgical Research & Development, " Chemical
. Analysis of Wire-Flux Test Coupon",2-10-70.
[B16] WCAP-8457, "Duquesne Light Company Beaver Valley Unit No.1 Reactor Vessel Radiation Surveillance Program", J. A. Davidson, et al., October 1974.
[B17] Combstion Engineering,Inc., Welding Material Cenification and Release for Section III, " Chemical Analysis of Test Weld Sample", Sample No. D-7279, August 12,1969.
[B18] Lukens Steel Company Letter from John A. Soltesz to S. B. Yanichko dated December 6,1973, Ladle Copper Analysis for Lukens Vessel Materials.
[B19] Combustion Engineering, Inc., Metallurgical Research & Development Department,
" Materials Certification Report", Job No. V-70333-001, March 5,1%9.
[B20] Lukens Steel Company, " Test Cenificate", Mill Order No. 20241-2, corrected copy dated 7-23-68.
[B21] Combustion Engineering, Inc., Metallurgical Research & Development Department,
" Materials Certification Report", Job No. V-70333-004, corrected ropy dated 5/4n0.
[B22] Lukens Steel Company, " Test Certificate", Mill Order No. 20241-2, dated 7-24-68.
[B23] Combustion Engineering, Inc., Metallurgical Research & Development Department,
" Materials Cenification Report", Job No. V-70333-007, conected copy dated 3/31n0.
[B24] Combustion Engineering, Inc., Metallurgical Research & Development Department,
" Materials Cenification Repon", Job No. V-70334-001, September 4,1%9.
[B25] Lukens Steel Company, " Test Cenificate", Mill Order No. 20241-3, dated 5-25-68.
[B26] Combustion Engineering, Inc., Metallurgical Research & Development Department,
" Materials Certification Report", Job No. V-70334-005, September 4,1%9.
[B27] Lukens Steel Company, " Test Cenificate" Mill Order No. 20241-3, dated 710-68.
[B28] Combustion Engineering. Inc., Metallurgical Research & Development Depanment,
" Materials Certification Report". Job No. V-70334-009, September 4,1%9.
[B29] Lukens Steel Company, " Test Cenificate", Mill Order No. 20241-3, dated 6-24-68.
[B30] WCAP-12354, " Analysis of Capsule X from the Duke Power Company McGuire Unit 1 Reactor Vessel Radiation Surveillance Programs", S.E. Yanichko, et al., August 1989.
B
[B31] We-stinghouse Electric Corporation Nuclear Service Division CMT - Analytical Laboratory, Waltz Mill Site, Analytical Request #15211, Alloy Analysis - Steel, Duke Power Company McGuire Nuclear Plant Unit 1, Lawrence Kardos, November 15,1993.
B-22 i
1
1 I
[B32] ' WCAls-7924-A,
- Basis for Heatup and Cooldown Limit Curves". W. S. Hazelton,' et al.', . '
April 1975. l
[B33] Fluence data given in Section 6.0 of this report.
[B34] WCAP-10867, " Analysis of Capsule U from the Duquesne Light Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program", R. S. Boggs, et al., ,
September 1985.
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B-23 ;
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t ATTACHMENT B1 DATA POINTS FOR HEATUP AND COOLDOWN CURVFS (WITHOUT MARGINS FOR INSTRUMENTATION ERRORS) i o
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B-24 er m +- .
McGuire Unit 1 IIeatup and Cooldown Data Without Margins at 16 EFPY E
Cooldown Curves lleatup Curve leak Test Data Steady State 20 Dill CD 40 Di'G CD 60 Di'G CD 100 DEG CD 60 DEG IlU Criticality Ilmit T P T P T P T P T P T P i P T P 75 529.22 75 486.15 75 442.18 75 397.27 75 3G539 75 517.92 282 0.00 261 2000 80 534.63 80 491.75 80 447.94 80 40336 80 311.12 80 517.92 282 525.46 282 2435 85 540.46 85 497.81 85 4 M.26 85 409.98 85 318.48 85 - 517.92 282 520.40 90 546.6I 90 504.23 90 461.06 90 417.06 90 326 42 90 517.92 T't2 517.92 95 553.35 95 511.27 95 468.44 95 424.83 95 335.13 95 517.% 282 ~. 517.%
100 560.58 100 518.83 100 47638 100 433.20 100 344.55 100 519.92 282 519.92 105 56837 105 527.00 105 4&4.99 105 442.29 105 354.83 105 523.85 282 523.85 110 576.73 110 53538 110 4 % 25 110 452.02 110 365.93 110 529.37 282 529.37 115 585.73 IIS 545.16 115 504.17 115 462.64 It5 377.% 115 536.56 282 536.56 120 595.27 120 55535 120 514.95 120 474.09 120 391.01 120 545.06 282 545.06 125 605.67 125 56634 125 526.61 125 486.49 125 405.19 125 555.17 282 555.17 -
130 616.85 130 578.16 130 53915 130 499.76 130 420.44 130 566.65 282 566.65 135 621.00 135 5%B9 135 552.60 135 514.21 135 437.05 135 579.58 282 579.58 140 621.00 140 6G4.48 140 567.18 140 529.78 140 454.90 140 593.78 282 593.78 145 621.00 145 619.25 145 582.92 145 '546.50 145 47432 145 609.59 282 609.59 150 621.00 150 621.00 150 59934 150 56463 150 495 28 150 621.00 282 626.88 -
155 621.00 155 621.00 155 618.02 155 584.22 155 517.88 155 621.00 282 645.61 160 621.00 160 621.00 160 621.00 160 605.18 160 54236 160 621.00 282 621.00 160 703.60 100 670 48 160 637.67 165 627.% 165 568.72 160 666.10 282 666.10 165 722.15 165 690.27 165 658.75 170 65233 170 597.14 165 68830 282- 68830 Y 170 741.90 170 711.43 170 681.56 175 678.79 '175 627.99 170 712.19 282 712.19 to
" 175 76335 175 73434 175 706.02 180 707.11 180 661.10 175 738.12 282 738.12 180 78634 180 758.85 180 732.46 185 737.79 185 6%B6 180 765.99 282 765.99 185 810.% . 185 '78535 185 7to.83 190 770.73 190 735.54 185 795.97 282 795.97 IW) 83735 100 813.75 190 79131 195 806.18 195 777.17 100 828.44 282 828.44 195 865.95 145 844.24 105 824 34 200 84433 200 821.98 195 863.26 282 863.26 200 8 % 48 200 877.21 200 859.71. 205 885.44 205 87033 200 8 % 48 282 8M48 205 929.45 205 912.55 205 . 89736 210 ' 929.85 210 92233 205 929.45 282 929.45 210 764.76 210 950.51 210 938.66 2 15 977.52 215 978.40 210 964.76 282 964.76 215 1002.71 215 99132 215 982.71 220 1028.57 220 1038.64 215 100231 282 1002.71 220 1043.42 220 1035.18 220 1030D7 225 108338 220 1043.42 282 1043.42 225 1087.17 225. 103236 225 1081.05 225 1087.17 282 1087.17 230 !!34.18 230 1133.06 230 1134.18 282 !!34.18 235 1184.70 235 II84.70 282 I184.70 240 1238.98 240 1238.98 282 1238.98 245 1297.13 ' 245 1287.52 285 1287.52 250 1359.59 250 1337.80 290 1337.80 255 1426.60 255 1391.53 295 1391.53 260 1498.41 260 1449.24 300 1449.24 265 1575.51 265 -1510.82 305 1? ' ' 82 [
270 1658.07 270 1577.19 310 - 15 9 275 1746.46 275 1647.73 315 164 1 280 1841.27 280 1723.83 320 1723 285 1932.98 285 1H05.03 325 1805.
- 29) 2051.60 290 1891.36 330 1891.86 295 2167.68 295 1984.71 335 1984.71 300 2291.83 300 2084.23 340 - 2084.23 305 2424 34 305 2190.26 345 2190.26 -
310 2303.62 350 2303.62-315 2424.65 355 . 2424.65
-i
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ATTACIIMENT B2
'IIEATUP AND COOLDOWN CURVES AND DATA POINTS 1
S (WITil MARGINS OF 12*F AND 30 PSIG FOR INSTRUMENTATION ERRORS)
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MATERIAL PROPERTY BASIS LIMITING MATERIALS: LOWER SHELL LONGITUDINAL WELDS 3-442A & C AND LOWER SHELL PLATE B5013-2 LIMITING ART AT 16 EFPY: 1/4-t, 149.5'F 3/4-t, 102.0*F 2,500 ,iiiiiiiiii, ,
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LEAK TEST LIMIT x [ f f 2,250 ] l l f I I
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INSERVICE HrDROSTATIC TEST TEMPERATURE C295 'F) FOR THE i SERVICE PEAIOD UP TO 16 EFPY l O I O 50 100 150 200 250 300 350 400 450 500 I nd i cated Temperature C Deg . FJ ;
Figure B-3 McGuire Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 60*F/hr) Applicable for the First 16 EFPY (With Margias of 12*F and 30 psig for Instrumentation Errors)
B-27
i MATERIAL PROPERTY BASIS LIMITING MATERIALS LOWER SHELL LONGITUDINAL WELDS 3-442A & C AND LOWER SHELL PLATE B5013-2 LIMITING ART AT 16 EFPY: 1/4-t, 149.5'F 3/4-t, 102.0*F 2,500 I
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O 50 100 150 200 250 300 350 400 450 500 I nd i cated Temperature (Deg . F) l Figure B-4 McGuire Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100'F/hr) Applicable for the First 16 EFPY (With Margins of 12*F and 30 psig for Instrumentation Errors) j B-28 1
McGuire Unit i Heatup and Cooldown Data With Margins at 16 EFPY Cooldown Curves llcatup Curve trakTest Data Steady State 20 DEG CD 40 DI:.0 CD 60 DliG CD 100 DFG CD 60 DEG IIU Criticality llmit T P T P T P T P T P T P T P T P 77 489.44 77 446.12 77 401.80 77 356.47 77 262.54 77 487.92 295 0 00 274 2010 82 494.18 82 450.93 82 406.77 82 361.64 82 268.19 82 487.92 295 4%I8 295 2485 87 499.22 87 456.I5 87 412.I8 87 367.27 87 27439 87 487.92 295 499.22 92 SM63 92 461.75 92 417.94 92 37336 92 281.12 92 487.92 295 495.46 97 510.46 97 467.81 97 424.26 97 - 379.98 97 288.48 97 487.92 295 490.40 102 516 61 102 474.23 102 431.06 102 387.06 102 296.42 102 487.92 295 487.92 107 52335 107 481.27 107 438.44 107 394.83 107 305.13 107 487. % 205 487.96 I12 530 58 112 488.83 112 44638 112 '403.20 112 314.55 - 112 489.92 295 489.92 117 53837 117 4W.00 117 454.99 117 412.29 117 324.83 !!7 493.85 205 - 493.85 122 546.73 122 505.78 122 461.25 122 422.02 122 335.93 122 49937 295 49937 I27 555.73 127 515.I6 127 474.17 127 432.64 127 347.96 127 - 506.56 295 506.56 132 565.27 132 52535 132 484.95 132 444.09 132 - 361.01 I32 515.06 295 5I5.06 137 575.67 137 53634 137 4 %.61 137 456.49 137 375 19 137 525.17 2 05 525.17 142 586 85 142 548.16 142 509.15 142 469.76 142 390.44 142 536.65 295 536 65 147 591.00 147 560.89 147 522.60 147 484.21 147 407.05 147 549.58 295 549.58 152 591.00 152 574.48 152 537.18 152 499.78 152 424.90 152 563.78 295 563.78 157 591.00 157 589.25 '157 552.92 157 516.50 157 44432 157 579.59 295 579.59 162 591 00 162 591.00 162 569.74 162 534.63 162 465.28 162 591.00 295 596 88 167 591.00 167 591.00 167 SR8.02 167 554.22 167 487.88 167 591.00 295 615.61 172 591.00 172 591.00 172 591.00 -172 575.18 172 51236 172 591.00 295 591.00 172 673.60 172 640.48 172 607.67 I77 597.96 177 538.72 172 636.10 295 636.10 g 177 . 692.15 177 660.27 177 628.75 182 62233 182 567.14 1 77 65830 295 65830 e 182 711.90 182 681.43 182 651.56- 187- 618.79 187 597.94 182 682.19 295 682.19
,y 187 73335 187 7G134 187 676.02 192 677.11 192 631.10 187 708.12 295 708.12 192 75634 142 728.85 192 702.46 197. 707.79 197 666.86 192 735.99 295' 73599 197 780.96 197 75535 197 730 83 202 740.73 202 705.54 IW 765.97 295 765.97
-202 80735 202 783.75 202 76131 207 776 18 207 747.17 202 798.44 295 798.44 207 835.95 207 814.24 207 79434 212 81433 212 791.98 207 833.26 295 833.26 212 866.48 212 847.21 212 829.71 217 855.44 - 217 84033 212 866.48 295 866.48 217 809.45 - 217 88355 217 867.76 222 899.85 222 89233 217 899.45 - 295 899.45-222 934.76 222 92d51 222 908.66 227 947.52 227 948.40 222 934.76 295 934.76 -
227 972.71 227 % I.32 227 952.71 232 998.57 232 1008.64 227 972.71 295 972.71 232 1013.42 232 1005.18 ' 232 1000.07 237 1053.78 -232 1013.42 295 1013 42 237 1057.17 237 105236 -237 1051.05 -237 1057.17 295 1057.17 242 I104.18 242 1103 06 242 1104.18 295 t 1RI8 247 I I54.70 247 II54.70 295 I154.70 252 1208.98 252 1208.98 295 1208.98 257~ 1267,13 257 1257.52 297 1257.52 262 1329.59 262 1307.80 302 1307.80 267. I3 % .60 267 1361.53 307 1361.53 272 1468.41 272 1419.24 312 1419.24 277 1545.51 277 1480.82 317 1480.82 282 1628.07 2 82 1547.19 322 1547.19 287 1716.46 287 1617.73 327 1617.73 292 1811.27 292 1693.83 332 1693.83 297 1912.98 2M 1775 93 337 1775.03 302 2021.60 302 1861.86 342 1861.86 307 2137.63 307 1954.71' 347 -1954.71 312 2261.83 312 2054.23 352 2054.23 317 239434 317 2160.26 357- 2160.26 322 2273.62 362 2273.62 327 2394.65 367 2394.65 -
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o APPENDIX C T
4 Upper Shelf Energy Evaluation i
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Background:
The low upper shelf Charpy energy concem is associated with the determination of acceptable reactor vessel toughness when the vessel is operating at normal temperatures of 525'F or higher (the " upper shelf" of measured fracture toughness).
In 1973, in an effort to improve the quality of reactor pressure vessel integrity and to base the assessment of vessel integrity on a theoretical rather than an empirical basis, the concept of fracture mechanics techniques was implemented through ASME activities and NRC regulation. These requirements are included in 10CFR50, Appendix G, " Fracture Toughness Requirements". 10CFR50 Appendix G requires utilities to submit an analysis at least 3 years prior to the time that the upper shelf energy of any of the reactor vessel material is predicted to drop below 50 ft-lb as measured by Charpy V-notch specimen testing. One source of the irradiated materials that can be used to make this prediction on a plant specific basis is the reactor vessel surveillance capsule program. l The surveillance -
capsules contain material identical to or representative of the critical reactor vessel materials that would be irradiated during plant operation. The surveillance capsules are attached inside the reactor vessel at locations designed to provide a higher irradiation rate than the reactor vessel itself, thus providing an irradiation " lead" factor that allows for prediction of future vessel irradiation damage.
There are titree methods that can be applied to estimate the upper shelf energy:
o Use of Revision 2 of Regulatory Criteria 1.99; o Use of plant specific surveillance capsule data; and o Use of altemative industry trend curves. ,
in the event that the 50 ft-lb requirement cannot be satisfied as stated in 10CFR50, Appendix G, or by attemative procedures acceptable to the NRC, a reactor may continue to operate provided all the following requirements of 10CFR50, Appendix G, paragraph V.C are satisfied:
- 1. A volumetric examination of 100 percent of the beltline materials that do not satisfy the ,
requirement of Section V.B of Appendix G is made, and any flaws are characterized ;
according to Section XI of the ASME Code and as otherwise specified by the Director, Office of Nuclear Reactor Regulation.
C-1
i:
. 2. Additional evidence of the fracture toughness of the beltline materials after exposure to
- neutron irradiation must be obtained from results of supplemental fracture toughness tests.
- 3. ' An analysis is performed that conservatively demonstrates, making appropriate allowances for all uncertainties, the existence of equivalent margins of safety for continued operation.
The issue of low upper shelf toughness has been a subject of active concem with the NRC for several years. In 1981, NUREG 0774 was issued. This document suggested an analysis pmcedure for determining the margins of safety present in reactor vessels. However, no definite criteria for acceptance margins were given. In 1982 a fonnal request was issued to the ASME Code Section XI committee to develop more specific criteria. In 1989 the ASME completed development of alternative criteria to define acceptable reactor vessel integrity when the Charpy upper shelf energy drops below 50 ft-lb. These criteria were transmitted to the NRC for their evaluation and incorporation into the regulations goveming reactor vessel integrity.
. Methodology:
The methodology presented in Regulatory Guide 1.99, Revision 2, was used to predict the changes in upper shelf energy of the McGuire Unit i reactor vessel beltline materials. For the materials contained in the reactor vessel surveillance program, the decrease in upper shelf energy was obtained by plotting the reduced plant surveillance data on Figure 2 of Regulatory Guide 1.99, Revision 2, and fitting the data with a line parallel to the existing lines as the upper bound of all the data. This line was used in preference to the existing graph to determine upper shelf energy values reported for the surveillance materials. For the beltline materials that were not in the surveillance program, the Charpy upper shelf energy was assumed to decrease as a function of fluence and copper content as indicated in Figure 2 j of Regulatory Guide 1.99. Revision 2.
Results:
i
']
The results of the upper shelf energy calculations are given in Table C-1.
i
-l 1
C-2
1 s
Conclusions:
i Upper shelf energy values for each of the beltline region materials in the McGuire Unit l' reactor .
vessel were calculated for .32 and 48 effective full power years (EFPY) of operation and the results -
presented in Table C-1. The results indicate that all of the beltline region materials will remain above the 50 ft-Ib screening criteria for both 32 and 48 EFPY.'
C-3
.l__
TABLE C-l' McGuire Unit 1 Beltline Material Upper Shelf Energy Values Initial Upper Shelf Upper Shelf Energy - Upper Shelf Energy-Energy at 32 EFPY at 48 EFPY Material Description (ft-lb) (ft-lb) (ft-lb)
Intermediate Shell B5012-1 101 91 90 Intermediate Shell B5012-2 105 80 78 Intennediate Shell B5012-3 112 88 86 Lower Shell B5013-1 95 72 70 Lower Shell B5013-2 115 92 90 Lower Shell B5013-3 103 82 80 Girth Weld >l26 >101 >98 Intermediate Shell 112 69 . 65-Longitudinal Weld Seams l
2-442B & 2442C Intermediate Shell 112 71 66 Longitudinal Weld Seam 2-442A Lower Shell Longitudinal >90 >59 >57 Weld Scams 3-442A & 3442C Lower Shell Longitudinal >90 >58 >$6 Weld Seam 3442B C-4
, ., . .- . . . , . ,. .~ .. .- . - - .
=!
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il; APPENDIX D Justification for sing Diablo Canyon Unit 2 Surveillance Weld !
Data for the Prediction of the McGuire Unit i Lower Shell Longitudinal Weld Seam Metal Mechanical Properties f
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Purpose:
~
The purpose of this appendix is to document an evaluation of the Diablo Canyon _ Unit 2 - 3 surveillance weld metal and the McGuire Unit I reactor vessel lower shell longitudinal weld seam metal to determine the feasibility of using the Diablo Canyon Unit 2 surveillance weld metal as a ,
credible data source for the calculation of the adjusted RTym of the limiting McGuire Unit I lower shell longitudinal weld seam.
Itackground:
Based on the calculational methods of Regulatory Guide 1.99, Revision 2. the limiting beltline -
material (in tenns of RTsm) in the McGuire Unit I reactor vessel is the intennediate shell longitudinal weld seam. Thus, this material was chosen for the surveillance program weld metal. -
However, when surveillance data from the McGuire Unit I program is used to predict the adjusted RTsm of the intermediate shell longitudinal weld seams, it is no longer limiting. Applying the results of the McGuire Unit I surveillance program to calculate the adjusted RTsm of the intermediate shell longitudinal weld seams changes the limiting material to the lower shell -
longitudinal weld seam metal. j During the review of the draft irport on the surveillance capsule V analysis, the Duke Power.
Company noticed that the limiting material for the generation of the heatup/cooldown curves had changed from the intennediate shell longitudinal weld seam to the lower shell longitudinal weld i
seam. This prompted a request to use Diablo Canyon Unit 2 surveillance weld data in the calculation of the RTym for the limiting lower shell longitudinal weld seam and the subsequent request for the development of this appendix and the generation of new heatup/cooldown curves for McGuire Unit 1.
Later Westinghouse received a call from the Duke Power Company asking about the feasibility of using the Diablo Canyon Unit 2 surveillance weld data to predict the RTm of the lower shell longitudinal weld seams of McGuire Unit 1. Since the lower shell longitudinal weld seam is the'-
limiting material for the generation of the McGuire Unit I heatup and cooldown curves, a benefit in RTa would give McGuire Unit I more operating margin when heating up and cooling down ,
the plant.
J D-1 ,
I i
I
Methodology:
The evaluation described in this appendix was perfonned utilizing the following methodology:
The evaluation of the Diablo Canyon Unit 2 surveillance weld metal and McGuire Unit i lower shell longitudinal weld seam data was based on the following:
. What weld wire heat number was used to fabricate the welds,
- What flux and flux lot number were used to fabricate the welds,
- What vender fabricated the welds and in what time frame.
- What heat treatment did each weld receive.
- Is the chemistry of both welds similar,
- Is the initial RTsor of both welds the same or relatively close,
- Is the initial upper shelf energy of both welds the same or relatively close,
- Is the geometry of both plants similar,
- Is the type of fuel in both plants the same,
- Are the fuel loading patterns in both plants similar (ie, low leakage, etc.).
- What is the projected 32 effective full power year surface fluence of each plant,
- What vessel inlet temperatures do the plants operate at,
- What are the differences in the capsule lead factors of both plants, and
- Can the criteria for credibility in Regulatory Guide 1.99, Revision 2, be met for McGuire .
Unit I when applied to the Diablo Canyon Unit 2 weld data ?
Development and documentation of the justification to use the Diablo Canyon Unit 2 suiveillance -
weld data for the McGuire Unit i lower shell longitudinal weld seam metal mechanical property >
predictions was based on the answers to the above questions.
Evaluation:
The comparison of the Diab!o Canyon Unit 2 surveillance weld metal relative to the McGuire Unit I lower shell longitudinal weld seam metal is summarized as' follows:
What weld wire heat numbers, flux, and flux lot numbers were used to fabricate the welds 7.
The Diablo Canyon Unit 2 surveillance weld is a Tandem weld fabricated with weld wire heat numbers 12008 and 21935 using Linde 1092 Flux Lot No. 3869.
D-2
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i The McGuire Unit I lower shell longitudinal weld seams are Tandem weld fabricated with )
weld wire heat numbers 12008 and 21935 using Linde 1092 Flux Lot No. 3889. !
i What vender fabricated the welds and in what time frame ?
^
The Diablo Canyon Unit 2 surveillance weld was fabricated by Combustion Engineering, Inc. in the late 1960's and early 1970's. a The McGuire Unit I welds were also fabricated by Combustion Engineering, Inc. in the-lau ~0's and early 1970's.
What heat treatment did each weld receive ?
The Diablo Canyon Unit 2 surveillance weld metal post weld heat treatment was at 1150*F ll1 125'F for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> and fumace cooled. .
The McGuire Unit I lower shell longitudinal weld seams also received a post weld heat ' -
treatment of i150 F 1 52 F for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> and fumace cooled.
Is the chemistry of both welds similar ?
l The unitradiated chemistry of the Diablo Canyon Unit 2 surveillance weld and the McGuire Unit I Lower Shell Longitudinal Weld Seam metal is provided in Table D-1. .
s ,
A review of the data presented in Table D-1 reveals that the chemistry of the Diablo -
Canyon Unit 2 surveillance weld metal is very similar to the chemistry of the McGuire
- . Unit I lower shell longitudinal weld metal.
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Is the initial RTsm of both welds the same or relatively close ?
t' The initial RT Nm. of the Diablo Canyon Unit 2 surveillance weld metal is -50 F.
The initial RTsm of the McGuire Unit I lower shell longitudinal weld seams is not known so a generic value of-56*F is used.
D-3
l-TABLE D-1 Unitradiated Chemistry of the Diablo Canyon Unit 2 Surveillance Weld and the Mcguire Unit 1 Lower Shell Longitudinal Weld Seams Diablo Canyon Unit 2 CEm Westinghousem Element McGuire Unit IW Analysis Analysis 0.11 i C 0.16 0.13 S 0.010 0.010 0.011 N - 0.008 -
Co - 0.012 -
Cu 0.22 0.22 0.20 Si -
0.22 0.15 l
Mo - 0.47 0.55 Ni - 0.83 -
Mn - 1.32 1.38 Cr - 0.031 -
f.
V -
0.001 -
P 0.015 0.017 0.0 15 Sn -
0.010 -
Al - 0.009 -
D-4 l-
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- Is the initial upper shelf energy of both welds the same or relatively close ?
j The unirradiated Diablo Canyon Unit 2 surveillance weld metal initial upper shelf energy is ;
124 ft-lb.
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No documentation could be located to detennine the unirradiated upper shelf energy of the McGuire Unit I lower shell longitudinal weld seams, is the geometry of both plants similar ?
Both Diablo Canyon Unit 2 and McGuim Unit I have a reactor vessel inner radius of 173 inches, a reactor vessel beltline thickness of 8.625 inches and a power rating of 3565 MWT, both are Westinghouse 4 loop NSSS plants, have neutron pads and the surveillance; capsules are located at the same azimuthal angles, a
Is the type of fuel in both plants the same ?
t Both Diablo Canyon Unit 2 and McGuire Unit I use l7X17 rod array fuel assemblies.
Are the fuel loading patterns in both plants similar (ie. Iow leakage) ?
Diablo Canyon Unit 2 utilizes a low leakage fuel management scheme.
McGuire Unit I currently utilizes a low leakage fuel management scheme.
Is the projected surface fluence at 32 effective full power year (EFPY) relatively close ?
The Diablo Canyon Unit 2 projected clad / base metal interface fluence (n/cm2 . E >l1.0 MeV) at 32 EFPY and various azimuthal angles is:
0* 15" 30* 45 9.83 x 10" 1.46 x 10"- I.19 x 10" 1.70 x 10" '
D-5 m
2 The McGuin: Unit l' projected clad / base metal interface fluence (n/cm , E > 1.0 MeV) at 32 EFPY and various azimuthal angles is: ;
O 15* 30 45* ,
1.332 x 10 l.969 x 10 l.459 x 10 ' 2.016 x 10
The difference in these fluence values is believed to be due to Diablo Canyon Unit 2 utilizing a _
~
low leakage loading pattern from the start of operation while a low leakage loading pattern was not -
initiated in McGuire Unit I until after several operating cycles.
What vessel inlet temperature do the plants operate at ? ,
Diablo Canyon Unit 2 operates with at vessel inlet temperature of approximately 545.1 F.
McGuire Unit 1 operates with at vessel inlet temperature of approximately 557.9 F.
What are the capsule lead factors for both plants ? >
The surveillance capsule lead factors for Diablo Canyon Unit 2 and McGuire Unit I s surveillance capsules is presented in Table D-2. ,
Based on the infonnation provided in Table D-2, the lead factors of the surveillance capsules in both plants are essentially equivalent. However, when the lower average flux rate on the Diablo Canyon Unit 2 capsules is used with the higher average flux rate on the .;
McGuire Unit I reactor vessel, the result is that the Diablo Canyon Unit 2 capsules actually ,
have a slightly lower lead factor than the McGuire Unit I capsules.
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.i D-6 f
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TABLE D-2 Surveillance Capsule Lead Factors for Diablo Canyon Unit 2 and McGuire Unit 1 Diablo Canyon Unit 2nu McGuire Unit 1"'I Capsule Location Lead Factor Capsule ~ Location . Lead Factor U 56* 5.28 U 56 5.25 V 58.5 4.62 V 58.5 4.72 W 124 5.28 W 124 5.32 X 236 5.28 X 236 5.31 Y 238.5* 4.62 Y 238 5 4.72 Z 304 5.28 Z 304 5.32 Can the credibility criteria of Regulatory Guide 1.99, Revision 2, be met for McGuire Unit I when applied to the Diablo Canyon Unit 2 surveillance weld metal ?
I) " Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement according to the recommendations of this guide."
When the results of the tested surveillance capsules fmm McGuire Unit I are applied in the calculation of the 32 EFPY adjusted RTa's of the McGuire Unit I beltline materials, the lower shell longitudiral weld metal becomes limiting. Hence, the weld metal in the Diablo Canyon Unit 2 surveillance program is the li.miting material in the McGuire Unit I rextor -
vessel. Thus, this criteria is met.
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l D-7 P
d J
'2) " Scatter in the plots of Charpy energy versus temperature for the irradiated and unitradiated conditions should be small enough to permit the detennination of the 30-foot-pound; temperature and the upper-shelf energy unambiguously."
A review of the Charpy plots presented in the analyses of the two Diablo Canyon Unit 2 l surveillance capsule tested to date was performed. The scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions was judged to be small enough to permit the detennination of the 30-foot-pound temperature and the upper- ;
shelf energy unambiguously. Thus, this criteria is met. .
- 3) "When there are two or more sets of surveillance data from one reactor, the scatter of ;
ARTsn 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 fail this criterion for use in shift 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 185-82". ,
There are two sets of test data from the Diablo Canyon Unit 2 surveillance program. A -
review of the measured and predicted ARTmn's in these analyses for the weld metal showed that the scatter in the ARTun's of the weld metal is less than 28'F. Thus, this criteria is met.
- 4) "The irradiation temperature of the Charpy specimens in the capsule should match vessel wall temperature at the clad / base metal interface withini 25'F."
Both the Diablo Canyon Unit 2 surveillance program and the McGuire Unit I surveillance program are based on ASTM E185-73. Per ASTM E185-73, " Specimens shall be inradiated j at a location in the reactor that duplicates as closely as possible the neutron-flux spectrum.
temperature history, and maximum accumulated neutron fluence experienced by the reactor:
vessel." The Diablo Canyon Unit 2 and McGuire Unit I surveillance capsules were installed between the neutron pad and the reactor vessel wall in order to fulfill the quirements of ASTM E185-73. Based on the location of the capsules in the vessel, the :
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- m. , . -. _ -- . _. . . . .. .
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temperature should be the same as the vessel inlet water temperature and the temperature of-the reactor vessel wall in the beltline region will also be the smne as the vessel inlet water. j The irradiation temperatures of the surveillance capsules is judged to be within= 25 F of the _ vessel wall temperature at the clad / base metal interface.1 Diablo Canyon Unit 2 operates ;
at vessel inlet temperature of 545.1"F and McGuire Unit I operates at a vessel inlet temperature of 557.9 F. However. NRC currently believes that the irradiation damage of-'
the material increases as the irradiation temperature of the material decreases. Therefore, ;
' the Diablo Canyon Unit 2 test data should be slightly more conservative than if the material -
were irradiated in the McGuire Unit I reactor vessel. Thus, based on this rational, this criteria is also met.
- 5) "The surveillance data for the correlation monitor material in the capsule'should fall within the scatter band of the data base for this material."
Neither the McGuire Unit i or the Diablo Canyon Unit 2 surveillance programs contain- l correlation monitor material. Therefore, this critena is not applicable. ,
Results: :
The evaluation of the Diablo Canyon Unit 2 surveillance weld metal and McGuire Unit I' lower t shell longitudinal weld seam metal results in the following:
Both the Diablo Canyon Unit 2 surveillance weld metal and the McGuire Unit I lower-shell longitudinal weld seams are Tarydem welds fabricated with weld wire heat numbers 21935 and 12008 using Linde 1092 Flux- .
Both the Diablo Canyon Unit 2 surveillance weld metal and the McGuire Unit I lower shell welds were fabricated by Combus'.lon Engineering, Inc. in the late 1960's / early 1970's.
Both the Diablo Canyon Unit 2 surv'eillance weld metal and the McGuire Unit 1. lower shell longitudinal weld searns had a post weld heat treatment of 1150 F15*F 2 for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> and were fumace cooled.
r D-9 i d
A review of the available data for both the Diablo Canyon Unit 2 surveillance weld metal' and the McGuire Unit I lower shell welds indicates that the weld metal chemistry is L _
similar.
The initial RTsm of the Diablo Canyon Unit 2 surveillance weld metal is measured and is -
y 50 F. The initial RTym of the McGuire Unit I lower shell weld seams is a generic value ,
m of-56 F. A higher initial RTsm value is conservative, thus, the use of -50 F for the ' W initial RTunt of the McGuire Unit I lower shell longitudinal weld seams is conservative and acceptable.
The initial upper shelf energy of the Diablo Canyon Unit 2 surveillance material was :
measured to be 124 ft-lb.' No data for initial upper. shelf energy of the McGuire Unit I lower shell longitudinal weld seams could be located. Therefore, this criterion was not used for this evaluation.
-q Both plants' have vessels with a beltline thickness of 8.625 inches and an inner radius of! j 173 inches, both have four loops, both have 17X17 rod array fuel assemblics, both plants ,
J have neutron pads and the surveillance capsules in both plants are located at the same -
azimuthal angles and at the same distance from the center of the core. Thus; Diablo -
Canyon Unit 2 and McGuire Unit I have similar geometry.
Both Diablo Canyon Unit 2 and McGuire Unit I have a power rating of 3565 MWt and .,
both plants are currently using a low leakage core loading pattem.
l 2
The projected 32 full effective power year (EFPY) fluence (n/cm , E > 1.0 MeV) of the
~
McGuire Unit i reactor vessel beltline material is higher than that for Diablo Canyon Unit ~
- 2. However, since the neutron flux rate of the Diablo Canyon Unit 2 surveillance capsules is less than the neutron flux rate of the McGuire Unit I surveillance capsules, the data ,
obtained from the Diablo Canyon Unit 2 surveillance program should give a better prediction of the McGuire Unit i vessel beltline materials than the McGuire Unit'l surveillance program because the integrated average lead factors are slightly lower. t i
D-10 ,
The . vessel coolant inlet temperature of the Diablo Canyon Unit 2 reactor vessel is 545.1 Fl
- and the vessel coolant inlet temperature of the McGuire Unit I reactor vessel is 557.9 EL The current belief is that a lower irradiation temperature causes greater damage to the
~
material. Thus, the lower operating temperature of Diablo Canyon Unit 2 makes its.
surveillance results slightly conservative when applied to the McGuire Unit i reactor vessel-beltline materials.
The applicable credibility criteria of Regulatory Guide 1.99. Revision 2, forjudging the credibility of surveillance data are met for McGuire Unit I when applied to the Diablo Canyon Unit 2 surveillance weld metal.
Conclusions:
Based on the above evaluation, using the Diablo Canyon Unit 2 surveillance weld data to predict the mechanical propenies of the McGuire Unit I lower shell longitudinal weld seam metal is -
justified and is recommended for calculations of the irradiated mechanical propenies of the McGuire Unit I lower shell longitudinal weld seam material.
l D-11 .
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