ML17334B466
| ML17334B466 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 02/28/1993 |
| From: | Shaun Anderson, Chicots J, Meyer T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML17334B467 | List: |
| References | |
| WCAP-13515, NUDOCS 9303180156 | |
| Download: ML17334B466 (19) | |
Text
WCAP-13515 WESTINGHOUSE CLASS 3 9303i8015b 9303i2 PDR ADOCK 050003ib-P PDR ANALYSIS OF CAPSULE U
FROM THE INDIANA MICHIGAN POWER COMPANY D.
C.
COOK UNIT 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM AMERICAN ELECTRIC POW'ER SERVICE CORPORATION APPROVED IN GENERAL 0
APPiiOVED EXCEPT AS NOTED NOT APPROVED OR REFERENCE ONLY e6ucad& i@hvuo~re J.
M. Chicots S.
L. Anderson A. Madeyski February 1993 Work Performed Under Shop Order AFFP-106 Prepared by Westinghouse Electric Corporation for the Indiana Michigan Power Company Approved by:
T. A. Meyer, Ma ager Structural Reliability and Plant Life Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O.
Box 355 Pittsburgh, Pennsylvania 15230-0355 1993 Westinghouse Electric Corp.
All Rights Reserved
II
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PREFACE This report has been technically reviewed and verified.
Reviewer:
Sections 1 through 5, 7, 8 and Appendices A and B
E. Terek Section 6
E.
P. Lippincott Appendix C
H. J.
Malone
TABLE OF CONTENTS Section Title
~Pa e
1.0
SUMMARY
OF RESULTS
2.0 INTRODUCTION
2-1
3.0 BACKGROUND
3-1
4.0 DESCRIPTION
OF PROGRAM 4-1 5.0 TESTING OF SPECIMENS FROM CAPSULE U
5.1 Overview 5.2 Charpy V-Notch Impact Test Results 5.3 Tension Test Results 5.4 Wedge Opening Loading Specimens 5-1 5-1 5-4 5-6 5-7 6.0 RADIATION ANALYSIS AND NEUTRON DOSIMETRY
- 6. 1 Introduction 6.2 Discrete Ordinates Analysis 6.3 Neutron Dosimetry 6-2 6-7 7.0 SURVEILLANCE CAPSULE REMOVAL SCHEDULE 7-1
8.0 REFERENCES
8-1 APPENDIX A LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS APPENDIX B PHOTOGRAPHS OF CHARPY, TENSILE AND WOL SPECIMENS PRIOR TO TESTING APPENDIX C
HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION
LIST OF TABLES Table Title
~Pa e
4-1 Chemical Composition and Heat Treatment of the D.
C.
Cook Unit 2 Reactor Vessel Surveillance Materials 4-3 5-1 Charpy V-Notch Impact Data for the D. C.
Cook Unit 2 Intermediate Shell Plate C5521-2 Irradiated at 550'F, Fluence 1.58 x 10 n/cm (E > 1.0 MeV) 5-8 5-2 Charpy V-Notch Impact Data for the D. C.
Cook Unit 2 Reactor Vessel Weld Metal and HAZ Metal Irradiated at 550'F, Fluence 1.58 x 10 n/cm (E > 1.0 MeV) 5-9 5-3 Instrumented Charpy Impact Test Results for the D.
C.
Cook 5-10 Unit 2 Intermediate Shell Plate C5521-2 Irradiated 'at 550'F, Fluence 1.58 x 10 n/cm (E > 1.0 MeV)
Instrumented Charpy Impact Test Results for the D.
C.
Cook Unit 2 Weld Metal and Heat-Affected-Zone (HAZ) Metal, Irradiated at 550'F, Fluence 1.58 x 10 n/cm (E > 1.0 MeV) 5-11 5-5 Effect of 550'F Irradiation to 1.58 x 10 n/cm (E > 1.0 MeV) on the Notch Toughness Properties of the D.
C.
Cook Unit 2 Reactor Vessel Surveillance Materials 5-12 111
LIST OF TABLES (Continued)
Table Title
~Pa e
5-6 Comparison of the D. C.
Cook Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99 Revision 2
Predictions 5-13 5-7 Projected End of License (32 EFPY)
RTNDT and Upper Shelf Energy Values for D.
C.
Cook Unit 2 Beltline Region Materials per Regulatory Guide 1.99, Revision 2
5-14 5-8 Tensile Properties for the D. C.
Cook Unit 2 Reactor Vessel 5-15 Surveillance Materials Irradiated at 550'F to 1.58 x 10 n/cm (E > 1.0 HeV) 6-1 Calculated Fast Neutron Exposure Parameters at the Surveillance Capsule Center 6-14 6-2 Calculated Fast Neutron Exposure Rates at the Pressure Vessel Clad/Base Metal Interface 6-15 6-3 Relative Radial Distributions of Neutron Flux (E > 1.0 HeV) 6-17 within the Pressure Vessel Wall 6-4 Relative Radial Distributions of Neutron Flux (E > 0. 1 HeV) 6-18 within the Pressure Vessel Wall 6-5 Relative Radial Distributions of Iron Displacement Rate (dpa) within the Pressure Vessel Wall 6-19
LIST OF TABLES (Continued)
Table Title
~Pa e
6-6 Nuclear Parameters for Neutron Flux Monitors 6-20 6-7 Monthly Thermal Generation During the First Eight Fuel Cycles of the D.
C.
Cook Unit 2 Reactor 6-21 6-8 Measured Sensor Activities and Reactions Rates 6-22 6-9 Summary of Neutron Dosimetry Results 6-24 6-10 Comparison of Measured and FERRET Calculated Reaction Rates at the Surveillance Capsule Center 6-25 6-11 Adjusted Neutron Energy Spectrum at the Surveillance Capsule Center 6-26 6-12 Comparison of Calculated and Measured Exposure Levels for Capsule U
6-27 6-13 Neutron Exposure Projections at Key Locations on the Pressure Vessel Clad/Base Metal Interface'-28 6-14 Neutron Exposure Values for Use in the Generation of Heatup/Cooldown Curves 6-29 6-15 Updated Lead Factors for D.
C.
Cook Unit 2 Surveillance Capsules 6-30
LIST OF ILLUSTRATIONS Ficiure Title
~Pa e
4-1 Arrangement of Surveillance Capsules in the D. C.
Cook Unit 2 Reactor Vessel 4-4 Capsule U Diagram Showing Location of Specimens, Thermal Monitors and Dosimeters 4-5 5-1 Charpy V-Notch Impact Properties for D.
C.
Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Longitudinal Orientation) 5-16 5-2 5-3 Charpy V-Notch Impact Properties for D. C.
Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation)
Charpy V-Notch Impact Properties for D. C.
Cook Unit 2 Reactor Vessel Surveillance Weld Metal 5-17 5-18 ~
5-4 Charpy V-Notch Impact Properties for D. C.
Cook Unit 2 Reactor Vessel Weld Heat-Affected-Zone Metal 5-19 5-5 Charpy Impact Specimen Fracture Surfaces for D.
C.
Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Longitudinal Orientation) 5-20 5-6 Charpy Impact Specimen Fracture Surfaces for D. C.
Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation) 5-21 5-7 Charpy Impact Specimen Fracture Surfaces for D.
C.
Cook Unit 2 Reactor Vessel Surveillance Weld Metal 5-22 vi
LIST OF ILLUSTRATIONS (Continued)
~Fi ure Title
~pa e
5-8 Charpy Impact Specimen Fracture Surfaces for D.
C.
Cook Unit 2 Reactor Vessel Weld Heat-Affected-Zone Metal 5-23 5-9 Tensile Properties for D.
C.
Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation) 5-24 5-10 Tensile Properties for D. C.
Cook Unit 2 Reactor Vessel Surveillance Weld Metal 5-25 5-11 Fractured Tensile Specimens from D.
C.
Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation) 5-26 5-12 Fractured Tensile Specimens from D. C.
Cook Unit 2 Reactor 5-27 Vessel Surveillance Weld Metal 5-13 Engineering Stress-Strain Curves for Intermediate Shell Plate C5521-2 Tensile Specimens MTll and MT12 (Transverse Orientation) 5-28 5-14 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens MW11 and MW12 5-29 6-1 Plan View of a Dual Reactor Vessel Surveillance Capsule 6-13
SECTION 1.0
SUMMARY
OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule U, the fourth capsule removed from the Indiana Michigan Power Company D.
C.
Cook Unit 2 reactor pressure
- vessel, led to the following conclusions:
o The capsule received an average fast neutron fluence (E > 1.0 MeV) of 1.58 x 10 n/cm after 8.65 EFPY of plant operation.
o Irradiation of the reactor vessel intermediate shell plate C5521-2 Char py specimens, oriented with the longitudinal axis of the specimen parallel to the maj or rolling direction (longitudinal orientation), to 1.58 x 10 n/cm (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 95'F and a 50 ft-lb transition temperature increase of 110'F.
This results in a 30 ft-lb transition temperature of 120'F and a 50 ft-lb transition temperature of 165'F for longitudinally oriented specimens.
o Irradiation of the reactor vessel intermediate shell plate C5521-2 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major rolling direction (transverse orientation), to 1.58 x 10 n/cm (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 130'F and a 50 ft-lb transition temperature increase of 135'F.
This results in a 30 ft-lb transition temperature of 160'F and a 50 ft-lb transition temperature of 205'F for transversely oriented specimens.
o The weld metal Charpy specimens irradiated to 1.58 x 101 n/cm (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 75'F and a 50 ft-lb transition temperature increase of 40'F.
This results in a 30 ft-lb transition temperature of 85'F and a 50 ft-lb transition temperature of 110'F for the weld metal.
1-1
o Irradiation of the reactor vessel weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 1.58 x 10 n/cm (E > 1.0 HeV) resulted in a 30 ft-lb transition temperature increase of 105'F and a 50 ft-lb transition temperature increase of 110'F.
This results in a 30 ft-lb transition temperature of 45'F and a 50 ft-lb transition temperature of 80'F for the weld HAZ metal.
o Irradiation of intermediate shell plate C5521-2 (longitudinal orientation) to 1.58 x 10 n/cm (E > 1.0 HeV) resulted in an average upper shelf energy decrease of 16 ft-lbs, resulting in an upper shelf energy of ill ft-lbs.
o Irradiation of -intermediate shell plate C5521-2 (transverse orientation) to 1.58 x 10 n/cm (E > 1.0 NeV) resulted in an average upper shelf energy decrease of 14 ft-lb, resulting in an upper shelf energy of 72 ft-lbs.
o The average upper shelf energy of the weld metal decreased 6 ft-lb after irradiation to 1.58 x 10 n/cm (E > 1.0 NeV).
This results in an upper shelf energy of 71 ft-lb for the weld metal.
o The average upper shelf energy of the weld HAZ metal decreased 33 ft-lb after irradiation to 1.58 x 10 n/cm (E > 1.0 HeV).
This results in an upper shelf energy of 82 ft-lb for the weld HAZ metal.
o The surveillance capsule U test results indicate that the intermediate shell plate C5521-2 (longitudinal) and the surveillance weld metal 30 ft-lb transition temperature shift is in good agreement with the Regulatory Guide 1.99 Revision 2 predictions.
- However, comparison of the 30 ft-lb.transition temperature increase for the intermediate shell plate C5521-2 (transverse) is 33'F greater than the Regulatory Guide 1.99 Revision 2 predictions.
Regulatory Guide 1.99 Revision 2 requires a
2 sigma allowance of 34'F for base metal be added to the predicted reference transition temperature to obtain a conservative upper bound value.
Thus, the reference 1-2
transition temperature increase for the intermediate shell plate C5521-2 (transverse) is bounded by the 2 sigma'llowance for shift prediction.
h o
The surveillance capsule materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are expected to maintain an upper shelf energy of greater than 50 ft-lb throughout the life (32 EFPY) of the vessel as required by 10CFR50, Appendix G.
o The calculated end-of-life (32 EFPY) maximum neutron fluence (E > 1.0 MeV) for the D.
C.
Cook Unit 2 reactor vessel is as follows:
Vessel inner radius
- 1.71 x 10 n/cm Vessel 1/4 thickness 9.02 x 10 n/cm Vessel 3/4 thickness 1.80 x 10 n/cm
- Clad/base metal interface 1-3
SECTION
2.0 INTRODUCTION
This report presents the results of the examination of capsule U, the fourth capsule to be removed from the reactor in the continuing surveillance progr am which monitors the effects of neutron irradiation on the Indiana Michigan Power Company D.
C.
Cook Unit 2 reactor pressure vessel materials under actual operating conditions.
The surveillance program for the D.
C.
Cook Unit 2 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation.
A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials is presented in WCAP-8512, entitled "American Electric Power Company Donald C.
Cook Unit No.
2 Reactor Vessel Radiation Surveillance Program" by J.
A. Davidson, et al~ ~.
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".
Westinghouse Power Systems personnel were contr acted to aid in the preparation of procedures for removing capsule U from the reactor and its shipment to the Westinghouse Science and Technology Center Hot Cell Facility, where, the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.
This report summarizes the testing of and the postirradiation data obtained from surveillance capsule U removed from the Indiana Michigan Power Company D.
C.
Cook Unit 2 reactor vessel and discusses the analysis of the data.
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 properties of low alloy, ferritic pressure vessel steels such as SA 533 Grade B Class I (base material of the D.
C.
Cook Unit 2 reactor pressure vessel shell plate) are well documented in the literature.
Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.
A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in "Protection Against Nonductile Failure,"
Appendix G to Section III of the ASNE Boiler and Pressure Vessel Code~4~.
The method uses fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT).
RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTH E-208)~
~ or the temperature 60'F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the plate.
The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KIR curve) which appears in Appendix G to the ASHE Code.
The KIR 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 KIR 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.
3-1
RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material proper ties.
The radiation embrittlement changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance
- program, such as the D.
C.
Cook Unit 2 Reactor Vessel Radiation Surveillance ProgramI~~,
in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested.
The increase in the average Charpy V-notch 30 ft-lb temperature (hRTNDT) due to irradiation is added to the original RTNpT to adjust the RTNDT for radiation embrittlement.
This adjusted RTNpT (RTNDT initial +
hRTNpT) is used to index the material to the K~R curve and, in turn, to set operating limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.
3-2
SECTION
4.0 DESCRIPTION
OF PROGRAM In accordance with WCAP-8512 [I], eight surveillance capsules for monitoring the effects of neutron exposure on the D.
C.
Cook Unit 2 reactor pressure vessel core region material were inserted in the reactor vessel prior to initial plant start-up.
The eight capsules were positioned in the reactor vessel between the thermal shield and the vessel wall as shown in Figure 4-1.
The vertical center of the capsules is opposite the vertical center of the core.
Capsule U was removed after 8.65 Effective Full Power Years (EFPY) of plant operation.
This capsule contained Charpy V-notch and tensile specimens (Figure 4-2) from the intermediate shell plate C5521-2 and charpy V-notch tensile and wedge load opening (WOL) specimens
.from submerged arc weld metal representative of that used in the original fabrication.
Capsule U also contained Charpy V-notch specimens from weld Heat-Affected-Zone (HAZ) material.
All heat-affected zone specimens were obtained from within the HAZ of the intermediate shell plate C5521-2.
All test specimens were machined from the I/4 thickness location of the plate after a simulated postweld stress-relieving treatment on the test material was performed.
The test specimens represent material taken at least one plate thickness from the quenched ends.
Base metal Charpy V-notch impact specimens were machined in both the longitudinal orientation (longitudinal axis of the specimen parallel to the major working direction) and transverse orientation (longitudinal axis of the specimen perpendicular to the major working direction).
Charpy V-notch and tensile specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the weld direction.
The WOL specimens were machined from the weldment such that the long dimension was parallrl to the weld direction.
The notch was machined such that the direction of crack propagation in the specimen was in the weld direction.
The chemical composition and heat treatment of the surveillance material is presented in Table 4-1.
Capsule U contained dosimeter wires of pure copper, iron, nickel, and aluminum-0. 15 weight percent cobalt wire (cadmium-shielded and unshielded).
In
)
and uranium (U
) were placed in the capsule to measure the integrated flux at specific neutron energy levels.
Thermal monitors made from two low-melting eutectic alloys and sealed in Pyrex tubes were included in the capsule.
These thermal monitors were used to define the maximum temperature attained by the test specimens during irradiation.
The composition of the two alloys and their melting points are as follows:
2.5X Ag, 97.5X Pb Helting Point:
579'F (304'C) 1.75X Ag, 0.75X Sn, 97.5X Pb Melting Point:
590'F (310'C) 4-2
TABLE 4-1 CHEMICAL COMPOSITION AND HEAT TREATMENT OF THE D.
C.
COOK UNIT 2 REACTOR VESSEL SURVEILLANCE MATERIALS (1)
Chemical Com osition wtX Intermediate Shell Plate C5521-2 Weld Metal Element C
S N
Co Cu Si Mo Ni Mn Cr V
P Sn Westinghouse Anal sis 0.220 0.014 0.014 0.016 0.110 0.270 0.550 0.580 1.280 0.072 0.001 0.017 0.013 Lukens Steel Anal sis 0.21 0.015 0.14 0.16 0.50 0.58 1.29
'.013 Westinghouse Anal sis 0.110 0.012 0.006 0.032 0.055 0.440 0.540 0.970 1.330 0.068 0.001 0.022 0.006 Chicago Bridge
& Iron Anal sis 0.08 0.016 0.05 0.36 0.96 1.42 0.07 0.019 Material Heat Treatment Histor Tem erature
'F
~Time Hr Coolant Intermediate Shell
- Plate, C5521-2 1650-1750 1550-1650 1200-1300 1150+
25 4 1/2 5
4 1/2 51 1/2 Water quenched Water quenched Air cooled Furnace cooled Weld Metal 1140+
25 Furnace cooled Notes:
(1)
WCAP-8512, Table A-2
270' (220')
Y (320)
W084)
Z (356')
180 "
0 V (f76')
S (4')
u()40')
T (40')
REACTOR VESSEL THERMALSHIELD CORE BARREL Figure 4-1.
Arrangement of Surveillance Capsules in the D.
C.
Cook Unit 2 Reactor Vessel 4-4