ML022940497

WCAP-15916, Rev. 0, Analysis of Capsule X from the Florida Power and Light Co Turkey Point Unit 3 Reactor Vessel Radiation Surveillance Program, Table of Contents - Appendix B, Page B-17

ML022940497
Person / Time
Site:	Turkey Point
Issue date:	09/30/2002
From:	Conermann J, Ledger J, Roberts G Westinghouse
To:	Office of Nuclear Reactor Regulation
References
FOIA/PA-2005-0108 WCAP-15916, Rev 0
Download: ML022940497 (127)

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Westinghouse Non-Proprietary Class 3 September 2002 WCAP-15916 Revision 0 Analysis of Capsule X from the Florida Power and Light Company Turkey Point Unit 3 Reactor Vessel Radiation Surveillance Program

, Westing*house

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15916, Revision 0 Analysis of Capsule X from the Florida Power and Light Company Turkey Point Unit 3 Reactor Vessel Radiation Surveillance Program J. H. Ledger G. K. Roberts J. Conermann September 2002 Approved'.

J. A. Gresham, ýManager Engineering & Materials Technology Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355

©2002Westinghouse Electric Company LLC All Rights Reserved Turkey Point Unit 3 Capsule X

iii TABLE OF CONTENTS LIST OF TA B LES ........................................................................................................................................ iv LIST OF FIGU RE S ...................................................................................................................................... vi PR EFAC E ................................................................................................................................................. viii EXECUTIVE

SUMMARY

(OR) ABSTRACT ........................................................................................... ix 1

SUMMARY

OF RESULTS .......................................................................................................... 1-1 2 INTRODU CTION ........................................................................................................................ 2-1 3 BA CKGR OUN D ......................................................................................................................... 3-1 4 DESCRIPTION OF PROGRAM ................................................................................................. 4-1 5 TESTING OF SPECIMENS FROM CAPSULE X ..................................................................... 5-1 5.1 O VERVIEW .................................................................................................................... 5-1 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS .......................................................... 5-3 5.3 TENSILE TEST RESULTS ............................................................................................ 5-5 5.4 COMPACT TENSION TESTS ....................................................................................... 5-5 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY ................................................... 6-1 6.1 INTRODU CTION ........................................................................................................ 6-1 6.2 DISCRETE ORDINATES ANALYSIS ........................................................................... 6-2 6.3 NEUTRON DOSIMETRY .............................................................................................. 6-5 6.4 CALCULATIONAL UNCERTAINTIES ........................................................................ 6-5 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE ............................................................ 7-1 8 REFEREN CES ............................................................................................................................. 8-1 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS ................................................ A-1 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS ......................... B-1 APPENDIX C CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING HYPERBOLIC TANGENT CURVE-FITTING METHOD .......................................................... C-1

iv LIST OF TABLES Table 4-1 Chemical Composition of the Turkey Point Unit 3 Reactor Vessel Surveillance M aterials .......................... :.......................................................................... 4-2 Table 4-2 Heat Treatment of the Turkey Point Unit 3 Reactor Vessel Surveillance Materials ....... 4-3 Table 5-1 Charpy V-Notch Data for the Turkey Point Unit 3 Intermediate Reactor Vessel Lower Shell Forging 123S266VA1 Irradiated to a Fluence of 2.90 x 1019 n/cm2 (E > 1.0 MeV) (Tangential Orientation) .......................... 5-6 Table 5-2 Charpy V-notch Data for the Turkey Point Unit 3 Weld Data Irradiated to a Fluence of 2.90 x 1019 n/cm 2 (E > 1.0 M eV) .................................................................. 5-7 Table 5-3 Charpy V-notch Data for the Turkey Point Unit 3 HAZ Metal Irradiated to a Fluence of 2.90 x 10 9 n/cm2 (E > 1.0 M eV) .................................................................. 5-8 Table 5-4 Charpy V-notch Data for the Turkey Point Unit 3 ASTM Correlation Monitor Material Irradiated to a Fluence of 2.90 x 1019 n/cm 2 (E> 1.0 MeV) ............................. 5-9 Table 5-5 Instrumented Charpy Impact Test Results for the Turkey Point Unit 3 Lower Shell Forging 123S266VA1 Reactor Vessel Irradiated to a Fluence of 2.90 x 10'9 n/cm2 (E>1.0 M eV) (Tangential Orientation) ......................................................................... 5-10 Table 5-6 Instrumented Charpy Impact Test Results for the Turkey Point Unit 3 Reactor Vessel Weld Metal Irradiated to a Fluence of 2.90 x 10") n/cm2 (E>1.0 MeV) ...... 5-11 Table 5-7 Instrumented Charpy Impact Test Results for the Turkey Point Unit 3 Reactor Vessel HAZ Metal Irradiated to a Fluence of 2.90 x 1019 n/cm2 (E> 1.0 MeV) ............ 5-12 Table 5-8 Instrumented Charpy Impact Test Results for the Turkey Point Unit 3 ASTM Correlation Monitor Material Irradiated to a Fluence of 2.90 x 10'9 n/cm 2 (E>! .0 M eV) .................................................................. 5-13 Table 5-9 The Effect of 550'F Irradiation at 2.90 x 10'9 (E>I .0 MeV) on the Notch Toughness Properties of the Turkey Point Unit 3 Reactor Vessel Surveillance Capsule Materials 5-14 Table 5-10 Comparison of the Turkey Point Unit 3 Reactor Vessel Surveillance Capsule Charpy Impact Test Results with Regulatory Guide 1.99 Revision 2 Predictions .................... 5-15 Table 5-11 Tensile Properties for Turkey Point Unit 3 Reactor Vessel Material Irradiated to a Fluence of 2.90 x 10'9 n/cm 2 (E> 1.0M eV) .................................................................. 5-16 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures Rates at the Surveillance Capsule Center ............................................................................................ 6-9

v LIST OF TABLES (Cont.)

Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface ........................................ 6-17 Table 6-3 Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Intermediate Shell Course to Lower Shell Course Girth Weld C lad/Base M etal Interfaces ........................................................................................... 6-21 Table 6-4 Relative Radial Distribution of Neutron Fluence (E > 1.0 MeV) within the R eactor Vessel Wall ...................................................................................................... 6-25 Table 6-5 Relative Radial Distribution of Iron Atom Displacements (dpa) within the Reactor Vessel Wall .................................................................................................................... 6-25 Table 6-6 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Turkey Point U nit 3 .................................................................................................................. 6-26 Table 6-7 Calculated Surveillance Capsule Lead Factors ............................................................. 6-26 Table 7-1 Turkey Point Unit 3 and Unit 4 Reactor Vessel Surveillance Capsule Withdraw al Schedule ........- ............................................................................................ 7-1 Table A-I Nuclear Parameters Used In The Evaluation Of Neutron Sensors .......................... A-10 Table A-2 Monthly Thermal Generation During The First Eighteen Fuel Cycles Of The Turkey Point Unit 3 Reactor (Reactor Power of 2200 MWt through October 11, 1996 and 2300 M Wt thereafter) ........................................................................................ A-I I Table A-3 Calculated C, Factors at the Surveillance Capsule Center Core Midplane Elevation ..A- 15 Table A-4 Measured Sensor Activities And Reaction Rates Surveillance C apsule T ................................................................................. A -17 Surveillance Capsule S .......................................................................... A-18 Surveillance Capsule V ......................................... A-19 Surveillance Capsule X ................................................................................. A -20 Table A-5 Comparison of Measured, Calculated and Best Estimate Reaction Rates At The Surveillance Capsule Center ............................................. A-21 Table A-6 Comparison of Calculated and Best Estimate Exposure Rates At The Surveillance Capsule Center ............................................................................................................ A -23 Table A-7 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions ................................................................................ A-24 Table A-8 Comparison of Best Estimate/Calculated (BE/C) Exposure Rate Ratios ..................... A-24

vi LIST OF FIGURES Figure 4-1 Original Arrangement of Surveillance Capsules in the Turkey Point Unit 3 Reactor V essel ................................................................................................................. 4-4 Figure 4-2 Capsule X Diagram Showing the Location of Specimens, Thermal Monitors, and D osim eters ...................................................................................................................... 4-5 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Turkey Point Unit 3 Reactor Vessel Shell Forging 123S266VA1 (Tangential Orientation) ....................................... 5-17 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Turkey Point Unit 3 Reactor Vessel Shell Forging 123S266VA1 (Tangential Orientation) ...................................... 5-18 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Turkey Point Unit 3 Reactor Vessel Shell Forging 123S266VAI (Tangential Orientation) ....................................... 5-19 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Turkey Point Unit 3 Reactor Vessel Weld M etal ........................................................................................... 5-20 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Turkey Point Unit 3 Reactor Vessel Weld M etal ........................................................................................... 5-21 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Turkey Point Unit 3 Reactor Vessel Weld M etal ........................................................................................... 5-22 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for Turkey Point Unit 3 Reactor Vessel Weld Heat-Affected-Zone Material ...................................................... 5-23 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for Turkey Point Unit 3 Reactor Vessel Weld Heat-Affected-Zone Material ...................................................... 5-24 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for Turkey Point Unit 3 Reactor Vessel Weld Heat-Affected-Zone Material ...................................................... 5-25 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for Turkey Point Unit 3 ASTM Correlation M onitor M aterial ............................................................................ 5-26 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for Turkey Point Unit 3 ASTM Correlation Monitor Material ............................................................................ 5-27 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for Turkey Point Unit 3 ASTM Correlation M onitor M atenal ............................................................................ 5-28 Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Turkey Point Unit 3 Reactor Vessel Forging 123S266VA I (Tangential Orientation) ............................................................ 5-29

vii LIST OF FIGURES (Cont.)

Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Turkey Point Unit 3 Reactor Vessel W eld M etal ........................................................................................... 5-30 Figure 5-15 Charpy Impact Specimen Fracture Surfaces for Turkey Point Unit 3 Reactor Vessel W eld HAZ M etal .................................................................................. 5-31 Figure 5-16 Charpy Impact Specimen Fracture Surfaces for Turkey Point Unit 3 ASTM Correlation Monitor Material ............................................................................ 5-32 Figure 5-17 Tensile Properties for Turkey Point Umt3 Reactor Vessel Lower Shell Forging 123S266VA 1 (Tangential Orientation) ......................................................................... 5-33 Figure 5-18 Tensile Properties for Turkey Point Unit3 Reactor Vessel Weld Metal ........................ 5-34 Figure 5-19 Fractured Tensile Specimens for Turkey Point Unit 3 Reactor Vessel Shell Forging 123S266VA 1 (Tangential Orientation) ............................................................ 5-35 Figure 5-20 Fractured Tensile Specimens for Turkey Point Unit 3 Reactor Vessel Weld Metal ...... 5-36 Figure 5-21 Engineering Stress-Strain Curves for Turkey Point Unit 3 Reactor Vessel Forging 123S266VA 1, Tensile Specimens S13 and S 14 ............................................................ 5-37 Figure 5-22 Engineering Stress-Strain Curves for Turkey Point Unit 3 Reactor Weld Metal, Tensile Specimens W 3 and W 4 ..................................................................................... 5-38 Figure 6-I Turkey Point Unit 3 r, 0 Reactor Geometry at the Core Midplane ................................. 6-7 Figure 6-2 Turkey Point Unit 3 r,z Reactor Geometry ..................................................................... 6-8

viii PREFACE This report has been technically reviewed and verified by:

Reviewer:

Sections 1 through 5, 7, 8, Appendices B, C and D T. J. Laubham Section 6 and Appendix A S. L. Anderson

ix EXECUTIVE

SUMMARY

The purpose of this report is to document the results of the testing of surveillance Capsule X specimens and dosimeters from the Turkey Point Unit 3 reactor vessel. Capsule X was removed at 19.85 EFPY and post irradiation mechanical testing of the Capsule X Charpy V-notch and tensile specimens was 2

performed along with a fluence evaluation. The surveillance Capsule X fluence was 2.90 x 1019 n/cm after 19.85 EFPY of plant operation. A brief summary of the Charpy V-notch testing results can be found in Section 1 and the updated capsule removal schedule can be found in Section 7. The results of the capsule analysis are within the expected range for both materials tested and dosimetry calculations.

1-1 1

SUMMARY

OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule X, the fourth capsule to be removed from the Turkey Point Unit 3 reactor pressure vessel, resulted in the following conclusions:

0 Capsule X received an average fast neutron calculated fluence (E > 1.0 MeV) of 2.90 x 1029 n/cm2 after 19.85 effective full power years of operation. This capsule was relocated from the "X" position to the "T" position in 1990 in order to accelerate neutron accumulation and better suit the program intent.

0 Irradiation of the reactor vessel lower shell forging 123S266VA1 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction of the plate (tangential orientation), to 2.90 x 10'9 n/cm2 (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 72.44°F and a 50 ft-lb transition temperature increase of 63.4F. This results in an irradiated 30 ft-lb transition temperature of 9.14'F and an irradiated 50 ft-lb transition temperature of 27.95°F for the longitudinally oriented specimens.

Irradiation of the weld metal Charpy'specimens to 2.90 x 10'9 n/cm2 (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 191.06'F. This results in an irradiated 30 ft-lb transition temperature of 190.97°F.

2 Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 2.90 x 10 i9 n/cm (E> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 26 0 F and a 50 ft-lb transition temperature increase of 18.72°F. This results in an irradiated 30 ft-lb transition temperature of-45.77°F and an irradiated 50 ft-lb transition temperature of-30.55°F.

Irradiation of the Correlation Monitor Material Charpy specimens to 2.90 x 1 019 n/cm 2 (E> 1.0 MeV) resulted in a 30 fl-lb transition temperature increase of 126.86°F and a 50 ft-lb transition temperature increase of 127.45°F. This results in an irradiated 30 ft-lb transition temperature of 156.96°F and an irradiated 50 ft-lb transition temperature of 202.030 F.

• The average upper shelf energy of the lower shell forging 123S266VA1 (tangential orientation) resulted in no energy decrease after irradiation to 2.90 x 10'9 n/cm 2 (E> 1.0 MeV). This results in an irradiated average upper shelf energy of 148 ft-lb for the tangentially oriented specimens.

0 The average upper shelf energy of the weld metal Charpy specimens resulted an average energy decrease of 19.7 ft-lb after irradiation to 2.90 x 10'9 n/cm 2 (E> 1.0 MeV). Hence, this results in an irradiated average tipper shelf energy of 45 fl-lb for the weld metal specimens. As expected, this result is below the IOCFR50 Appendix G requirement. The required analysis is documented in reference 19.

The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an average energy decrease of 19 ft-lb after irradiation to 2.90 x 10'9 n/cm 2 (E> 1.0MeV). This results in an irradiated average upper shelf energy of 158 ft-lb for the weld HAZ metal.

Turkey Point Unit 3 Capsule X

1-2

• The average upper shelf energy of the Correlation Monitor Material Charpy specimens resulted in an average energy decrease of 0.5 ft-lb after irradiation to 2.90 x 10'9 n/cm 2 (E> 1.0MeV).

This results in an irradiated average upper shelf energy of 67 ft-lb for the Correlation Monitor Material.

• A comparison of the Turkey Point Unit 3 Reactor Vessel beltline materials test results with the Regulatory Guide 1.99 Revision 2 predictions led to the following conclusions:

- The measured 30 ft-lb shift in transition temperature values of the lower shell forging contained in Capsule X (Tangential) is greater than Regulatory Guide 1.99 Revision 2 predictions. However, the shift value is less than two sigma allowance by Regulatory Guide 1.99 Revision 2.

- The measured 30 ft-lb shift in transition temperature values of the weld metal contained in Capsule X is less than the Regulatory Guide 1.99 Revision 2 predictions.

- The measured percent decrease in upper shelf energy of the Capsule X surveillance materials is less than the Regulatory Guide 1.99 Revision 2 predictions.

• The peak calculated end-of-license (32 EFPY) and end-of-license renewal (48 EFPY) neutron fluence (E> 1.0 MeV) at the core midplane for the Turkey Point Unit 3 reactor vessel using the Regulatory Guide 1.99, Revision 2 attenuation formula (ie. Equation # 3 in the guide; f(dpth) =

fsuace

• e (-o.24x)) is as follows:

Calculated (32 EFPY): Vessel inner radius* = 4.03 x 1019 n/cm2 Vessel 1/4 thickness = 2.53 x 1019n/cm 2 Vessel 3/4 thickness = 9.98 x 1018 n/cm 2 Calculated (48 EFPY): Vessel inner radius* = 5.91 x 1019 n/cm2 Vessel 1/4 thickness = 3.71 x 10'9 n/cm2 Vessel 3/4 thickness = 1.46 x 10'9 n/cm2

• Clad/base metal interface Note: These fluence levels are calculated without the use of part length burnable absorber (Hf) assemblies in the core.

Turkey Point Unit 3 Capsule X

2-1 2 INTRODUCTION This report presents the results of the examination of Capsule X, the fourth capsule removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Florida Power and Light Company Turkey Point Unit 3 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the Turkey Point Unit 3 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-7656, entitled "Florida Power and Light Co. Turkey Point Unit No. 3 Reactor Vessel Radiation Surveillance Program" by S. E. Yanichko, tIl. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-66, "Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors"' 31 .

Subsequently, the Unit 3 program was integrated with the Unit 4 program. This is documented in reference 20 and approved by the NRC in reference 21. Program integration was facilitated by the fact that both units have the identical limiting material, weld SA-1 101 in the lower to intermediate girth welds. Both surveillance programs contain weldments made of the same weld wire, Page wire heat 71249. Therefore the results of this capsule represent results for both units 3 and 4.

Capsule X was removed from the reactor after 19.85 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 surveillance specimens was performed.

This report summarizes the analysis of the post-irradiation data obtained from surveillance Capsule X removed from the Florida Power and Light Company Turkey Point Unit 3 reactor vessel and discusses the analysis of the data.

Turkey Point Unit 3 Capsule X

3-1 3 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 SA508 Class 2 (base material of the Turkey Point Unit 3 reactor pressure vessel forging) 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 during high energy irradiation.

A method for ensuring the integrity of reactor pressure vessels has been presented in "Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code141. The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RTNDT).

RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-20815 1) or the temperature 60'F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal to the major working direction of the material.

The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KI, curve) which appears in Appendix G to the ASME Code. The KI, curve is a lower bound of static fracture toughness results obtained from several heats 6f pressure vessel steel. When a given material is indexed to the KI, 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.

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 properties. The radiation embrittlement changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program, t1 such as the Turkey Point Unit 3 reactor vessel radiation surveillance program , 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 (ARTNDT) due to irradiation is added to the initial RTNDT. along with a margin term(M) to cover uncertainties, to adjust the RTNDT for radiation embrittlement. This RTNDT (RTNDT initial + M + ARTNDT) is used to index the material to the K1, curve and, in turn, to set operating limits for the nuclear power plant that take into account the effects of irradiation on the reactor vessel materials.

Turkey Point Unit 3 Capsule X

4-1 4 DESCRIPTION OF PROGRAM Eight surveillance capsules for monitoring the effects of neutron exposure on the Turkey Point Unit 3 reactor pressure-vessel core region material were inserted in the reactor vessel prior to initial plant startup. The capsules were positioned in the reactor vessel between the thermal shield and the vessel wall at locations shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.

Capsule X (Figure 4-2) was removed after 19.85 Effective Full Power Years (EFPY) of plant operation.

This capsule was relocated from the "X" position to the "T" position in 1990 in order to accelerate neutron accumulation and better suit the program intent. This capsule contained Charpy V-notch, tensile specimens and WOL fracture mechanics specimens from the reactor vessel lower shell forging 123S266VA 1, weld metal representative of the beltline region weld seams, Charpy V-notch specimens from weld heat-affected zone (HAZ) material and Charpy V-notch specimens from ASTM correlation monitor material (A302 Grade B).

The chemistry and heat treatment of the surveillance material are presented in Table 4-1 and Table 4-2, respectively. The chemical analyses reported in table 4-1 were obtained from unirradiated material used in the surveillance program.

All test specimens were machined from the 1A thickness location. Test specimens represent material taken at least one forging thickness from the quenched end of the forging. All base inaterial Charpy V notch impact and tensile specimens were oriented with the longitudinal axis of the specimen parallel to (tangential orientation) the principal working direction of the forging. Charpy V-notch specimens from the weld metal were oriented with the longitudinal axis of the specimens transverse to the weld direction.

Tensile specimens were oriented with the longitudinal axis of the specimens parallel to the weld. The WOL test specimens were machined with the simulated crack of the specimen to the surfaces and hoop direction of the forging.

Capsule X contained dosimeters of Copper, Nickel and Aluminum-Cobalt wire (cadmium-shielded and unshielded), and Neptunium (Np 237 ) and Uranium (U238) which measure the integrated flux at specific neutron energy levels.

Thermal monitors were made from two low-melting eutectic alloys and sealed in Pyrex tubes that were included in the capsule and were located as shown in Figure 4-2. The two eutectic alloys and their melting points are:

2.5% Ag, 97.5% Pb Melting Point 579F (304'C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point 590'F (310'F)

The arrangement of the various mechanical test specimens, dosimeters and thermal monitors contained in Capsule X are shown in Figure 4-2.

Turkey Point Unit 3 Capsule X

4-2 Table 4-1*** Chemical Composition of the Turkey Point Unit 3 Reactor Vessel Surveillance Materials Element Intermediate Shell Forging Lower Shell Forging Weld Metal 123P461VA1 123S266VA1 C 0.20 0.19/021 0076 Mn 0.64/0.64 061 /0.62 1.26 P 0.010 0010 0.011 S 0.010 0.008 0.018 Si 0.26 020/0.19 0.66 Ni 0.70 0.68 /0.66 0.57 Cr 0.40/0.39 038 0.14 V 0 02 0.02 0.002 Mo 0 62 0.58 / 0.59 0.42 Co 0.011/0010 0015/0.016 0.001 Cu 0 058 (0.07)** 0.079 (0.07)** 0.31 Sn 0010 0008 0.004 Zn 0.001 0001 0003 Al 0005 0005 0.015 N, 0003 0.003 0.012 Ti 0.001

• 0001* 0.001 Sb 0.001* 0 001* 0001 A, 0.005* 0 005* 0.005 B 0 003* 0.003* 0.003*

Zr 0 001* 0.001* 0001

• Not Deected The number indicates the minimum limit of detection

• • Copper Content Reported by Bethlehem Steel Co.

• • • Table taken directly from WCAP-76561i Turkey Point Unit 3 Capsule X

4-3 Table 4-2* Heat Treatment of the Turkey Point Unit 3 Reactor Vessel Surveillance Materials Material Temperature ('F) Time (hr) Coolant 1550 13 Water quenched Intermediate and Lower 1210 18 Air Cooled Shell Forgings 1125 10.5 Furnace Cooled Weldment 1125 10.25 Furnace Cooled 1

• Taken directly from WCAP-7656 !

Turkey Point Unit 3 Capsule X

4-4 900 z

1800 Figure 4-1 Original Arrangement of Surveillance Capsules in the Turkey Point Unit 3 Reactor Vessel Turkey Point Unit 3 Capsule X

4-5 SPECIMEN NUMBERING CODE S- Forging 123S266-VAI W- Weld Metal H- Heat-Affected-Zone SURVEILLANCE CAPSULE X R- ASTM Correlation Monitor Material 2 37 Np DOSIMETER CHARPYS CHAR PYS S64 IR54 TENSILE WOL CHARPYS S62I R52 WOL S6s~olol CHARPYS BLOCK CHARPYS WOL CHAR PYS TENSILE CHARPYS CHARPYS S63 R53 A

Cu Co ji Co(Cd}

II II I II LJ LJ ri 579°FF I 579°F ITOR MONI MONITOR LIL CENTER REGION OF VESSEL TO TOP OF VESSEL TO BOTTOM OF VESSEL Figure 4-2 Capsule X Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeters Turkey Point Urut 3 Capsule X

5-1 5 TESTING OF SPECIMENS FROM CAPSULE X 5.1 OVERVIEW The post-irradiation mechanical testing of the Charpy V-notch and tensile specimens was performed at the Westinghouse Science and Technology Center Laboratory with consultation by Westinghouse Energy 21 Systems personnel. Testing was performed in accordance with I OCFR50, Appendices G and H' , ASTM Specification El 85-82[t6, and Westinghouse Remote Metallographic Facility (RMF) Procedure RMF 8402, Revision 1 and 8103, Revision 1.

Upon receipt of the capsule at the hot cell laboratory, the capsule was visually examinated and photographed for identification purposes. The specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-7656"'. No discrepancies were found.

Examination of the two low-melting point 304'C (579°F) and 310°C (590'F) 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 304'C (579°F).

8 The Charpy impact tests were performed per ASTM Specification E23-98' j 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 930-1 instrumentation system feeding information into an IBM compatible computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED). From the load-time curve (Appendix A), the load of general yielding (P0 v), the time to general yielding (tGy), the maximum load (PM). and the time to maximum load (t51) can be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (Pr), and the load at which fast fracture terminated is identified as the arrest load (PA).

The energy at maximum load (ENI) was determined by comparing the energy-time record and the load time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (Ep) is the difference between the total energy to fracture (ED) and the energy at maximum load (EMI).

The yield stress (cry) was calculated from the three-point bend formula having the following expression:

a=(PGY*L)I[B *(W-a)2 *C] ()

where: L = distance between the specimen supports in the impact machine B = the width of the specimen measured parallel to the notch W = height of the specimen, measured perpendicularly to the notch a = notch depth The constant C is dependent on the notch flank angle (4), notch root radius (p) and the type of loading (i.e., pure bending or three-point bending). In three-point bending, for a Charpy specimen in which *=

450 and p = 0.010 inch, Equation I is valid with C = 1.21. Therefore, (for L = 4W),

Turkey Point Unit 3 Capsule X

5-2 0;=(PGy*L)/[B*(W-a)2 *l.21]=(3.33*Pc *W) /[B*(W -a)2] (2)

For the Charpy specimen, B = 0.394 inch, W = 0.394 inch and a = 0.079 inch. Equation 2 then reduces to:

o=33.3 *Per (3) where a, is in units of psi and PGy is in units of lbs. The flow stress'was calculated from the average of the yield and maximum loads, also using the three-point bend formula.

The symbol A in columns 4, 5, and 6 of Tables 5-5 through 5-8 is the cross-section area under the notch of the Charpy specimens:

A = B * (W - a) = 0.1241 sq.in. (4)

Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM Specification E23-98tsJ and A370-97a'91. The lateral expansion was measured 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 E8-99" 01 and E21-92(1998)"', and RMF Procedure 8102, Revision 1. All pull rods, grips, and pins were made of Inconel 718. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.

Extension measurements were made with a linear variable displacement transducer (LVDT) 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 E83-93" 21 .

Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9 inch hot zone. All tests were conducted in air.

The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.

Turkey Point Unit 3 Capsule X

5-3 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule X irradiated to approximately 2.90 x 1019 n/cm2 in 19.85EFPY are presented in Tables 5-1 through 5-8, and are compared with unirradiated results as shown' in Figures 5-1 through 5-12. The transition temperature increases and upper shelf energy decreases for the Capsule X material are shown in Table 5-10.

• Capsule X received an average fast neutron calculated fluence (E > 1.0 MeV) of 2.90 x 1019 n/cm2 after 19.85 effective full power yearý of operation.

0 Irradiation of the reactor vessel lower shell forging 123S266VA1 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction of the plate (tangential orientation), to 2.90 x 1019 n/cm 2 (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 72.44°F and a 50 ft-lb transition temperature increase of 63.41F. This results in an irradiated 30 ft-lb transition temperature of 9.14'F and an irradiated 50 ft-lb transition temperature of 27.95°F for the longitudinally oriented specimens.

• Irradiation of the weld metal Charpy specimens to 2.90 x 1019 n/cm 2 (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 191.06°F. This results in an irradiated 30 ft-lb transition temperature of 190.97°F.

Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 2.90 x 1019 n/cm 2 (E> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 26°F and a 50 ft-lb transition temperature increase of 18.72°F. This results in an irradiated 30 ft-lb transition temperature of-45.77'F and an irradiated 50 ft-lb transition temperature of-30.55°F.

Irradiation of the Correlation Monitor Material Charpy specimens to 2.90 x 1019 n/cm2 (E> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 126.86'F and a 50 ft-lb transition temperature increase of 127.45°F. This results in an irradiated 30 ft-lb transition temperature of 156.96°F and an irradiated 50 ft-lb transition temperature of 202.03'F.

The average upper shelf energy of the lower shell forging 123S266VA1 (tangential orientation) resulted in no energy decrease after irradiation to 2.90 x 1019 n/cm 2 (E> 1.0 MeV). This results in an irradiated average upper shelf energy of 148 ft-lb for the tangentially oriented specimens.

The average upper shelf energy of the weld metal Charpy specimens resulted an average energy decrease of 19.7 ft-lb after irradiation to 2.90 x 10' 9 n/cm2 (E> 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 45 ft-lb for the weld metal specimens.

The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an average energy decrease of 19 ft-lb after irradiation to 2.90 x 1019 n/cm 2 (E> 1.0MeV). This results in an irradiated average upper shelf energy of 158 ft-lb for the weld HAZ metal.

The average upper shelf energy of the Correlation Monitor Material Charpy specimens resulted in an average energy decrease of 0.5 ft-lb after irradiation to 2.90 x 1019 n/cm 2 (E> 1.0MeV).

This results in an irradiated average upper shelf energy of 67 ft-lb for the Correlation Monitor Material.

Turkey Point Unit 3 Capsule X

5-4 The Fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-13 through 5-16 shows an increasingly ductile or tougher appearance with increasing test temperature.

The load time records for individual instrumented Charpy specimens tests are shown in Appendix B.

A comparison of the Turkey Point Unit 3 Reactor Vessel beltline materials test results with the Regulatory Guide 1.99 Revision 2 predictions led to the following conclusions:

The measured 30 ft-lb shift in transition temperature values of the lower shell forging contained in Capsule X (Tangential) is greater than Regulatory Guide 1.99 Revision 2 predictions. However, the shift value is less than two sigma allowance by Regulatory Guide 1.99 Revision 2.

The measured 30 ft-lb shift in transition temperature values of the weld metal contained in Capsule X is less than the Regulatory Guide 1.99 Revision 2 predictions.

- The measured percent decrease in upper shelf energy of the Capsule X surveillance materials is less than the Regulatory Guide 1.99 Revision 2 predictions.

The Charpy V-Notch data presented in this report is based on a re-plot of all capsule data using CVGRAPH, Version 4.1, which is a hyperbolic tangent curve fitting program. Appendix C presents the CVGRAPH, Version 4. 1, Charpy V-Notch plots and the program input data.

Turkey Point Unit 3 Capsule X

5-5 5.3 TENSILE TEST RESULTS The results of the tensile tests performed on forging 123S266VA I (tangential orientation) and weld metal irradiated to 2.90 x 1019 n/cm2 are shown in Table 5-11 and are compared to the unirradiated results as shown in Figures 5-17 and 5-18.

The results of the tensile tests performed on the Lower Shell Forging 123S266VA 1 indicated that irradiation to 2.90 x 10'9 n/cm 2 (E>l1.0 MeV) caused an approximate increase of 10 to 15 ksi in the 0.2 percent offset yield strength and approximately 8 to 12 ksi increase in the ultimate tensile strength when "comparedto the unirradiated data"' (Figure 5-17).

The results of the tensile tests performed on the Weld material indicated that irradiation to 2.90 x 10'9 n/cm 2 (E>i .0 MeV) caused an approximate increase of 18 to 20 ksi in the 0.2 percent offset yield strength and approximately 16 to 18 ksi increase in the ultimate tensile strength when compared to the unirradiated data"'l (Figure 5-18).

'Fractured tension specimens for each of the materials are shown in Figures 5-19 and 5-20. Typical stress strain curves for the tension specimens are shown in Figures 5-21 and 5-22.

5.4 COMPACT TENSION TESTS Per the surveillance capsule testing contract with Florida Power and Light Company, the WOL Fracture Mechanics specimens will not be tested and will be stored at the Hot Cell at the Westinghouse Science and Technology Center.

Turkey Point Unit 3 Capsule X

5-6 Table 5-1 Charpy V-notch Impact Data for the Turkey Point Unit 3 Reactor Vessel Lower Shell Forging 123S266 Irradiated to a Fluence of 2.90 x 1019 n/cm 2 (E> 1.0 MeV)

(Tangential Orientation)

Sample Temperature Impact Energy Lateral Expansion Shear Number F C ft-lbs Joules mils mm  %

S62 0 -18 6 8 0 0.00 2 S60 25 -4 46 62 30 0.76 10 S63 50 10 97 132 59 1.50 20 S65 90 32 122 165 67 1.70 35 566 130 54 144 195 82 2.08 80 S64 150 66 150 203 86 2.18 100 S61 160 71 164 222 80 2.03 100 S59 180 82 173 235 82 2.08 100 Turkey Point Unit 3 Capsule X

5-7 Table 5-2 Charpy V-notch Impact Data for the Turkey Point Unit 3 Reactor Vessel Weld Data Irradiated to a Fluence of 2.90 x 1019 nlcm 2 (E> 1.0 MeV)

Sample Temperature Impact Energy Lateral Expansion Shear Number F C ft-lbs Joules mils mm  %

W13 20 -7 5 7 0 0.00 5 W12 80 27 10 14 4 0.10 '15 WlO 125 52 10 14 4 0.10 25 W14 150 66 18 24 7 0.18 20 Wi1 175 79 27 37 18 0.46 35 W16 225 107 36 49 20 0.51 65 W15 250 121 42 57 29 0.74 95 W9 325 163 48 65 33 0.84 100 Turkey Point Unit 3 Capsule X

5-8 Table 5-3 Charpy V-notch Impact Data for the Turkey Point Unit 3 Reactor Vessel HAZ Metal Irradiated to a Fluence of 2.90 x 1019 n/cm 2 (E> 1.0 MeV)

Sample Temperature Impact Energy Lateral Expansion Shear Number F C ft-lbs Joules mils mm  %

H9 -30 -34 22 30 8 0.20 5 H12 -30 -34 130 176 72 1.83 90 HII -10 -23 12 16 7 0.18 20 HI0 20 -7 187 254 86 2.18 100 H14 25 -4 134 182 72 1.83 60 H16 75 24 142 193 79 2.01 70 H15 100 38 148 201 87 2.21 100 H13 130 54 167 226 86 2.18 100 Turkey Point Unit 3 Capsule X

5-9 Table 5-4 Charpy V-notch Impact Data for the Turkey Point Unit 3 ASTM Correlation Monitor Material Irradiated to a Fluence of 2.90 x 109 n/cm2 (E> 1.0 MeV)

Sample Temperature Impact Energy Lateral Expansion Shear Number F C ft-lbs Joules mils mm  %

R51 60 16 8 11 1 0.03 5 R50 100 38 14 19 9 0.23 25 R49 150 66 27 37 18 0.46 30 R55 200 93 38 52 28 0.71 60 R54 225 107 67 91 44 1.12 90 R56 250 121 65 88 47 1.19 100 R52 275 135 68 92 45 1.14 100 R53 325 163 69 94 55 1.40 100 Turkey Point Unit 3 Capsule X

5-10 Table 5-5 Instrumented Charpy Impact Test Results for the Turkey Point Unit 3 Lower Shell Forging 123S266VA1 Reactor Vessel Irradiated to a Fluence of 2.90 x 1019 n/cmnz (E> 1.0 MeV) (Tangential Orientation)

Test Charpy Normalized Energies Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy (ftlb/ini) Load to Load Maximum Load Load Stress Stress Number Yield (CF) (ft.lb) Charpy Maximum Prop (kips) (kips) (Wsec) (kips) (kips) (ksi) (ksi)

Ed/A Em/A Ep/A Iec)

S62 0 6 48 25 23 9.72 0.1681 2592.4 0.1598 2589.97 0 0 43 S60 25 46 371 327 44 3485.11 0.1549 4533.08 0.6856 4508.7 0 116 134 S63 50 97 782 320 462 3212.39 0.1427 4386.76 0.6942 3572.04 0 107 127 S65 90 122 983 307 676 3122.08 0.144 4272.83 0.6881 3102.66 405.43 104 123 S66 130 144 1160 298 862 2992.8 0.1464 4208.94 0.6844 1740.12 675.09 100 120 S64 150 150 1209 311 898 3080.7 0.144 4369.39 0.6881 n/a n/a 103 124 S61 160 164 1321 292 1029 2838.65 0.1427 4122.97 0.6905 n/a n/a 95 116 S59 180 173 1394 290 1104 2794.61 0.1415 4070.62 0.6942 n/a n/a 93 114 Turkey Point Unit 3 Carpul X

5-11 Table 5-6 Instrumented Charpy Impact Test Results for the Turkey Point Unit 3 Reactor Vessel Weld Metal Irradiated to a Fluence of 2.90 x 1019 n/cm 2 (E> 1.0 MeV)

Test Charpy Normalized Energies Yield Time Maximum Time to Fracture Arrest Yield Flow Sample Temp Energy (ft-lb/in2 ) Load to Load Maximum Load Load Stress Stress Number Yield (NmF) (ft-Il)) Charpy Maximum Prop (kips) (kips) (pIsec) (kips) (kips) (ksi) (ksi)

Em/A Ep/A (psec)

Ed/A W13 20 5 40 17 24 2150.21 0.1171 2172.08 0.1232 2172.08 0 72 72 W12 80 10 81 44 37 3833.72 0.1537 4063.26 0.1732 4053.49 0 128 131 WIO 125 10 81 44 37 3662.22 0.1476 4061.3 0.1757 4061.3 0 122 129 W14 150 18 145 75 70 3553.97 0.144 4596.24 0.2269 4523.01 0 118 136 WII 175 27 218 91 127 4031.42 0.1915 4636.86 0.2696 4605.25 325.82 134 144 W16 225 36 290 182 108 3545.57 0.1598 4567.26 0.4209 4484.75 1050.81 118 135 W15 250 42 338 163 176 3504.34 0.144 4542.94 0.3758 4051.75 1263.42 117 134 W9 325 48 387 173 214 3360.16 0.1464 4417.05 0.405 n/a n/a 112 129 Turkey Point Unit 3 Capsule X

5-12 Table 5-7 Instrumented Charpy Impact Test Results for the Turkey Point Unit 3 Reactor Vessel HAZ Metal Irradiated to a Fluence of 2.90 x 1019 n/cm 2 (E> 1.0 MeV)

Normalized Energies Time Test Charpy Notmline) Yield To Maximum Time to Fracture Arrest Yield Flow Sambe Temp Energy Load to Load Maximum Load Load Stress Stress

("F) (ft-lb) Charpy Maximum Prop (kips) Yield (kips) (jIsec) (kips) (kips) (ksi) (ksi)

Ed/A Em/A Ep/A (_sec)

H9 -30 22 177 79 98 3996.64 0.1562 4789.94 0.2306 4668.56 0 133 146 H12 -30 130 1047 381 666 3977.92 0.1549 5174.22 0.7027 3238.75 911.81 132 152 HII -10 12 97 40 56 6873.73 0.1537 3939.64 0.1659 3932.31 261.18 129 130 HIO 20 187 1507 382 1125 3907.16 0.1513 5156.97 0.71 n/a n/a 130 151 H14 25 134 1080 364 716 3766.83 0.1525 5010.25 0.5942 3015.91 87.77 125 146 H16 75 142 1144 443 701 3602.79 0.1525 4881.83 0.8528 2628.86 656.61 120 141 HI5 100 148 1192 350 842 3536.88 0.1537 4881.83 0.6966 n/a n/a 118 140 H13 130 167 1346 351 994 3595.14 0.1525 4867.3 0.6978 n/a n/a 120 141 Turkey Point Unit 3 Capsule X

5-13 Table 5-8 Instrumented Charpy Impact Test Results for Turkey Point Unit 3 ASTM Correlation Monitor Material Irradiated to a Fluence or 2.90 x 1019 n/cm 2 (E> 1.0 MeV)

Normalized Energies Time Test Charpy (otmline) Yield To Maximum Time to Fracture Arrest Yield Flow Number Temp Energy Load Yield Load Maximum Load Load Stress Stress (7F) (ft-lb) Charpy Maximum Prop (kips) (kips) (qsec) (kips) (kips) (ksi) (ksi)

Ed/A Em/A Ep/A (Psec)

R51 60 8 64 32 33 3311.69 0.144 3353.03 0.1513 3353.03 0 110 H11 R50 100 14 113 46 67 3555.47 0.1525 3954.85 0.1842 3903.71 440.78 118 125 R49 150 27 218 116 101 3234.15 0.1427 4221.95 0.3123 4221.95 543.9 108 124 R55 200 38 306 151 155 3101.39 0.1525 4202.43 0.3916 4139.24 1489.93 103 122 R54 225 67 540 224 316 3137.86 0.1464 4417.77 0.5148 4085.04 2280.53 104 126 R56 250 65 524 195 328 3106.28 0.1476 4257.48 0.4734 n/a n/a 103 123 R52 275 68 548 206 342 3237.98 0.1537 4569.82 0.4746 n/a n/a 108 130 R53 325 69 556 193 363 3130.56 0.1574 4270.49 0.4734 n/a n/a 104 123 Turkey Point Unit 3 Capqule X

5-14 Table 5-9 The Effect of 550 0F Irradiation at 2.90 x 10i"' n/cm 2 (E>I.0 MeV) on the Notch Toughness Properties of the Turkey Point Unit 3 Reactor Vessel Surveillance Capsule Materials Average 30 (ft-tb) Average 35 mil Lateral Average 50 ft-lb Average Energy Absorption Material Transition Temperature ('F) Expansion Temperature ('F) Transition Temperature ('F) at Full Shear (ft-lb)

Umirradiatcd Irradiated AT Unirradijted Irradiated AT Unirradiated Irradiated AT Unirradiatcd Irradiated AT Lower Forgig -633 9.14 72.44 -58.25 31.57 89.82 -35.44 27.95 63.4 148 162 14 123S266VAI Weld Metal -08 190.97 191.06 2628 n/a n/a 59.7 n/a n/a 64.7 45 -19.7 HAZ Metal -71.77 -45.77 26 -49.78 -20.61 29.17 -4928 -30.55 18.72 177 158 -19 Correlation 30.1 156.96 126.86 48.11 224.5 17638 74.58 202.03 127.45 67.5 67 -0.5 Material Turkey Point Unit 3 Capsule X

5-15 Table 5-10 Comparison of the Turkey Point Unit 3 Reactor Vessel Surveillance Capsule Charpy Impact Test Results with Regulatory Guide 1.99 Revision 2 Predictions 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Capsule Fluence Predicted Measured Predicted Measured (x 1019 nlcm2) (OF) ( 0F) (%) (%)

Intermediate Shell T 0.739 33.85 11.48 17 0 Forging Heat# 123P461VA1 1.72 42.55 2.83 20 12 S 1.72 58.65 48.55 20 18 Lower Shell Forging V 1.53 57.12 42.68 20 0 Heat # 123S266VA1 x 2.90 65.43 72.44 24 0 T 0.739 153.31 163.87 34 10 Weld Metal V 1.53 187.66 180.77 39 26 x 2.90 214.97 191.06 48 30 T 0.739 - - 14.72 .. 6 HAZ Metal V 1.53 -- -2.13 .. 8 x 2.90 - - 26 . 11 S 1.72 -- 106.6 __ 11 T 0.739 - - 86.66 .. 0 Correlation Material V 1.53 - - 100.32 .. 3 X 2.90 - - 126.86 .. 1 Turkey Point Unit 3 Capsule X

5-16 Table 5-11 Tensile Properties for Turkey Point Unit 3 Reactor Vessel Material Irradiated to a fluence of 2.90 x 1019 n/tcm 2 (E > 1.0 MeV)

Material Sample Test 0.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Number Temp. Strength Strength Load Stress (ksi) Strength Elongation Elongation in Area (OF) (ksi) (ksi) (kip) (ksi) (%) (%) (%)

Plate S13 200 61.6 82.5 2.56 178.9 52.1 9.8 22.4 71 Plate S14 550 63.2 88.3 3.15 152.8 64.2 10.2 19.9 58 Weld W3 300 92.2 104.9 4.00 177.1 81.5 10.1 19.4 54 Weld W4 550 87.1 104.2 4.00 172.8 81.5 9.3 18.3 53 Turkey Point Unit 3 Capsule X

5-17 i I LOWER SHELL-FORGING 123S266VA-1 (TANG)

CVGRAPH 41 Hyperbohc Tangent Curve Printed at 1134:l6 on 06-17-2002 Results Curve Fluence ISE d-LSE USE d-ISE T o 30 d-T o 30 T o 50 d-T o 50 1 0 2.19 0 140 0 -63.3 0 -3544 0 2 0 aI9 0 122 -26 -14.74 4855 42 3965 3 0 219 0 166.5 185 -2062 42.68 339 38M4 4 0 Z19 0 162 14 914 72,44 2795 634

-4 IA z0

-300 -200 -100 0 100 200 300 '400 500 600 Temperature in Degrees F Curve Legend 10- 20 ---------- 30 - 4 Data Set(s) Plotted Curve Plant Capsule Material OrL Heat#

l TP3 UNIRR FORGING SA5082 LT 12SE66VA-1 2 TP3 S FORGING SA5082 LT 12M6VA-I 3 TP3 V FORGING SA50M2 LT 123I26VA-1 4 TP3 x FORGING SA5082 LT 123S266VA-1 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for Turkey Point Unit 3 Reactor Vessel Lower Forging 123S266VA1 (Tangential Orientation)

Turkey Point Urut 3 Capsule X

5-18 LOWER SHELL FORGING 123S266VA-1 (TANG)

CYGRAPH 41 Hyperbolic Tangent Curve Pnnted at 12.0310 on 06-17-2002 PReults Curve Fluence USE d-USE T o LE35 d-T o LE35 1 0 7967 0 -5825 0 2 0 9?04 12-7 -156 56.68 3 0 8098 131 257 60.3 4 0 79.81 14 3157 8912

--4 (D

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend 10- 20 ---------- 30 - 4 Data Set(s) Plotted Curve Plant Cansule Material Ori. Heatff 1 TP3 UNIRR FORGING SA5082 LT 123I66VA-I 2 TP3 S FORGING SAM32 LT 123S204&-1 3 TP3 V FORGING SA5032 LT 123266 VA-I 4 TI3 X FORGING SA5082 LT 123S266VA-I Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Turkey Point Unit 3 Reactor Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for Turkey Point Unit 3 Reactor Vessel Lower Forging 123S266VAI (Tangential Orientation)

Turkey Point UOit 3 Capsule X

5-19 LOWER SHELL FORGING 123S266VA-1 (TANG)

CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 121743 on 06-17-2U(2 Results Ctrv fliinre T o 5*s~rhear d-T o 50/. Shear Curv o 0r/ hea TFlunce 1 0 1312 0 0 17.34 421 2

3 0 63.75 50.62 0 9656 8343 4

4-)

r)

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend I-- 20 0.------- 30 Data Set(s) Plotted r'urre Plant Causule Material Or lleati Plant CaDsule Cure 1 TP3 UNIRR FORGING SA5082 Li' 123S266VA-I TP3 S FORGING SA5082 Li' 12366VA-1 2

3 TP3 V FORGING SA5082 LT i23=26VA-1 4 TP3 x FORGING SAMO&2 LT 123S26VA-I Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for Turkey Point Unit 3 Reactor Vessel Lower Forging 123S266VA1 (Tangential Orientation)

Turkey Point Urut 3 Capsule X

5-20 SURVEILLANCE PROGRAM WELD METAL CVGRAPIt 41 Hyperbolic Tangent Curve Printed at 112309 on 06-19-2002 Results Curve Fluence LSE d-1SE USE d-USE T o 30 d-T o 30 T o 50 d-T o 50 1 0 219 0 6469 0 -08 0 5-20 597 0 2 0 22 o 515 -619 163.78 163137 244.99 18528 3 0 22 0 47.79 -16.89 180.6 18077 4 0 22 0 45 -1969 190.97 191.6 5 0 22 0 42 -22.69 21618 21627 U)

T bfl 0) z cJ)

-300 -200 -100 0 100 200 300 400 50 6w0 Temperature in Degrees F Curve legend I0O- 20----------- 30 4 - 5-Data Set(s) Plotted Curve Plant Capsule Material OrL Heat#

1 TP3 UNIRR WELD LINDE 80 71249 2 TP3 T WELE) LINDE 80 71249 3 TP3 V WELD LINDE 80 71249 4 TP3 X WELD LINDE 80 71249 5 TP3 A5 IELD LINDE 80 71249 FLUX LOT 8445 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for Turkey Point Unit 3 Reactor Vessel Weld Metal Turkey Point Unit 3 Capsule X

5-21

- SURVEILLANCE PROGRAM WELD METAL CVGRAPH 41 Hlyperbolic Tangent Curve Printed at IM1359 on 06-25-2002 Results Curve Fluence USE d-USE T e LE35 d-T o LE35 1 0 7526 0 2628 0 5-21 2 0 5727 -17M 243.44 21715 3 0 45 -3026 22583 199.54 4 0 3413 -4U2 5 0 45BI -2954 26M9 2366 Cl)

Xt.

COr 04

-300 -200 -100 0 100 2W0- 3w 400 500 600 Temperature in Degrees F Curve Legend 1 D 20 ---------- 30 4 - 5 -

Data Set(s) Plotted Curve Plant Capsule Material On. Heat#

1 TP3 UNIRR WELD LINDE 60 71249 2 TP3 T WELD LINDE 80 71249 3 TP3 V WELD LINDE 80 71249 4 TP3 X WELD LINDE 80 71249 5 TP3 A5 WELD LINDE 80 71249 FLUX LOT 8445 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for Turkey Point Unit 3 Reactor Vessel Weld Metal Turkey Point Urut 3 Capsule X

5-22 SURVEILLANCE PROGRAM WELD METAL CVGRAPH 41 Hyperbolic Tangent Curve Printed at 1014310 on 06-25-2002 Results Curve Fluence T a 50q/. Shear d-T o 50"/Sheair T o 50% Shear d-T 0 W/ Shedr 1 0 2906 0 2 0 21703 17.96 3 0 21EL43 189.37 4 0 19031 1615 5 0 16083 13177 cf) 4.)

C-)

0.,

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend 10O 20G.-.----- 30- 4 - 5-Data Set(s) Plotted Curve Plant Canule Material rin 11eatI/

Curve Plant Capsule Material Ori Ileatl I TP3 UNIRR WELD LINDE 80 71249 2 TP3 T WELD LINDE 80 71249 3 TP3 V WELD LINDE 80 71249 4 TP3 X WELD LINDE 80 .71249 5 TP3 A5 WELD LINDE 80 71249 FLUX LOT 8445 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for Turkey Point Unit 3 Reactor Vessel Weld Metal Turkey Point Unit 3 Capsule X

5-23 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for Turkey Point Unit 3 Reactor Vessel Weld Heat-Affected-Zone Metal Turkey Point Unit 3 Capsule X

5-24 HEAT AFFECTED ZONE CYGRAPH 41 Hyperbolic Tangent Curve Pnnted at 095'28 on 06-18-2002 Re"ults Curve Fluence USE d-USE T o LE35 d-T

• LE35 1 0 91.54 0 -4978 0 2 0 70.72 -20.81 -5449 -4.71 3 0 9477 323 -384 1124 4 0 87.61 -3.92 -20.61 29J7 U) 4)

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend io- 20---------- 30 4 Data Set(s) Plotted Curve Plant CaDsule Material Ori Heat5 1 TP3 UNIRR HEAT AFFD ZONE Or Heat 2 TP3 T HEAT AFFD ZONE 3 TP3 V HEAT AFFD ZONE 4 TP3 X HEAT AFFD ZONE Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for Turkey Point Unit 3 Reactor Vessel Weld Heat-Affected-Zone Metal Turkey Point Unit 3 Capsule X

5-25 HEAT AFFECTED ZONE CVGRAPH 41 lHyperbohc Tangent Curve Printed at 1003:50 on 06-15-2002 Results Curve Fluence T c 5a/ Shear d-T c 50/. Shear 0 -3L%5 0 2 0 --5296 -20.41 3 0 656 3911 4 0 -1171 20.3 C)

C-)

0-4

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend iD- 20C------ 4 `-

Data Set(s) Plotted Curve Plant Capsule Material On Heat#

1 T'3 UNIRR HEAT AFFD ZONE 2 TP3 T HEAT AFFD ZONE 3 TP3 V HEAT AFFD ZONE 4 TFP1 x BEAT AFFD ZONE Figure 5-9 Charpy V-Notch Percent Shear vs. Tempeirature for Turkey Point Unit 3 Reactor Vessel Weld Heat-Affected-Zone Metal Turkey Point Unit 3 Capsule X

5-26 STANDARD REFERENCE MATERIAL CVGRAPH 41 Hyperbolic Tangent Curve Printed at 1023.35 on 06-18-2002 Results Curve Fluence ISE d-LSE USE d-USE T o 30 d-T o 30 T o 50 d-T o 50 1 0 219 0 67.5 0 301 0 7458 0 2 0 22 0 60 -75 136.7 106.6 23376 15917 3 0 219 0 72 45 11676 6666 18459 11001 4 0 219 0 6559 -1.9 13042 10032 19164 11706 5 0 2.19 0 67 -5 156%6 126.86 202.03 127 45

/)

-0 C-)

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend ta-- 20 ----------. 30- 4" 5v Data Set(s) Plotted Curse Plant Capsule Material Ori Heat#

I TP3 UNIRR SRI SA302B 2 TP3 S SRM SA302B 3 TP3 T S*1MSA302B 4 TP3 V SRM SA302B 5 TP3 X SPI SA30213 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for Turkey Point Unit 3 ASTM Correlation Monitor Material Turkey Point Unit 3 Capsule X

5-27 STANDARD REFERENCE MATERIAL CVGRAPH 41 Hyperbohc Tangent Curve Printed at 10-3608 on 06-18-2002 Results Curve Fluence USE d-USE T o LE35 d-T o LE35 1 0 117.73 0 48.11 0 5-27 2 0 47.48 -7025 16811 120 3 0 5749 -6024 17959 13147 4 0 6357 -5416 1633 11519 5 0 18237 6463 2245 17638

(/2 0)

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend I0- 20 ---------- 3 0- 4 - 5-Data Set(s) Plotted Curve Plant CaDsule Material OrL Heat#

1 TP3 UNIRR SRM SA302B 2 TP3 S SA302B 3 TP3 T SEM SA302B 4 TP3 V SRM SA302B 5 TP3 X SRM SA302B Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for Turkey Point Unit 3 ASTM Correlation Monitor Material Turkey Pount Unit 3 Capsule X

5-28 STANDARD REFERENCE MATERIAL CVGRAPH 41 H)perbohic Tangent Cune Printed at 104913 on 06-18-2002 Results r~i,,qr F*11 tIipnr'*

T o .51l/. Shear d-T

• 50r/ Shear r,,-- Fluence T c 50/ Shear 1 0 4312 0 2 0 17138 12826 3 0 17812 135 4 0 163.47 12035 5 0 170.5 12738 Q-4 C/)

0.)

C.)

0-)

-300 -200 -100 0 100 200 300 400 500 600 Temperature in Degrees F Curve Legend 1 o- 2----- --- 30 4 - 5-Data Set(s) Plotted Cnrve Plant Cansule Material Ori Heat#

Cuv Plant... . amile Material....

I TP3 UNIRR SRM SA302B 2 TP3 S SUM SA302B 3 TP3 T SRM SA302B 4 TP3 V SRM SA302B 5 TPM x SRM S,302B Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for Turkey Point Unit 3 ASTM Correlation Monitor Material Turkey Point Urnt 3 Capsule X

5-29 S62, 0F S60, 25°F S63, 50 0 F S65, 90-F S66, 130°F S64, 150TF S61, 160°F S59,180°F Figure 5-13 Charpy Impact Specimen Fracture Surfaces for Turkey Point Unit 3 Reactor Vessel Forging 123S266VA1 (Tangential Orientation)

Turkey Point Unit 3 Capsule X

5-30 I

W13, 20F W12, 80WF W10,125-F W14, 150F Wll, 175-F W16, 225F W15, 250° W9,325° Figure 5-14 Charpy Impact Specimen Fracture Surfaces for Turkey Point Unit 3 Reactor Vessel Weld Metal Turkey Point Umt 3 Capsule X

5-31 H12, -30°F H9, -30°F H1l, -100 F H1O, 20°F H14, 25F I

H16,75°F H15, 100-F H13, 130° Figure 5-15 Charpy Impact Specimen Fracture Surfaces for Turkey Point Unit 3 Reactor Vessel Weld HAZ Metal Turkey Point Unit 3 Capsule X

5-32 I

R51, 60°F R50, 100°F R49,150°F R55, 200°F R54, 225-F R56, 250-F R52, 275-F R53, 325°F Figure 5-16 Charpy Impact Specimen Fracture Surfaces for ASTM Correlation Material Turkey Point Unit 3 Capsule X

5-33 (0°0 0 50 100 150 200 250 300 350 120 I I I I I I I _

800 110 100 ULTIMATE TENSILE STRENGTH 700 90 600

-- A*'-.-.A A o.-

S80 co70 500 0 2% YIELD STRENGTH

, 60 0 0 400 020 50 o 8 00 300 40 LEGEND:

A 0 UNIRRADIATED 19 2 0 A

• IRRADIATED TO A FLUENCE OF 2.90 X 10 nlcm (E>1.OMeV) AT 550 F 80 REDUCTION IN AREA 2 70 _A

- 60

-50 S40 TOTAL ELONGATION S30 6 6 0 8 20 - 0e A - A 10 -- A I 1I UNIFORM ELONGATION I I 0

0 100 200 300 400 500 600 700 EMPERATURE (TF)

Figure 5-17 Tensile Properties for Turkey Point Unit 3 Reactor Vessel Shell Forging 123S266VA1 (Tangential Orientation)

Turkey Point Unit 3 Capsule X

5-34 (0 C) 0 50 100 150 200 250 300 350 120 i 800 110 - ULTIMATE TENSILE STRENGTH A -A 100 - 700 90- "*A _ ---- 600 d0

""80 - 8"

"'& - 0 - 500 0.2% YIELD STRENGTH 60 - -400 50 40 7300 LEGEND:

A 0 UNIRRADIATED 19 2 0 A. *IRRADIATED TO A FLUENCE OF 2.90 X 10 n/cm (E>1.0MeV) AT 550 F 80 70 REDUCTION INAREA

. 60

>-,50 "A A

-4

~40 L_)

S30 TOTAL ELONGATION 20 8 ____o_

10 UNIFORM ELONGATION 0 1 1 1 1 1 1 0 100 200 300 400 500" 600 700 TEMPERATURE (OF)

Figure 5-18 Tensile Properties for Turkey Point Unit 3 Reactor Vessel Weld Metal Turkey Point Unit 3 Capsule X

5-35 Specimen S 13 Tested at 200°F IUl IlI II JUw LIJ.

l JUII*

Specimen S 14 Tested at 550WF Figure 5-19 Fractured Tensile Specimens for Turkey Point Unit 3 Reactor Vessel Shell Forging 123S266VA1 (Tangential Orientation)

Turkey Point Unit 3 Capsule X

5-36 Specimen W3 Tested at 300'F Specimen W4 Tested at 550'F Figure 5-20 Fractured Tensile Specimens for Turkey Point Unit 3 Reactor Vessel Surveillance Weld Metal Turkey Point Unit 3 Capsule X

5-37 STRESS-STRAIN CURVE TURKEY POINT UNIT 3 X" CAPSULE 100 90 80 70 60 C0) w 50 cr IU-40 30 S13 200 F 20 10 0

0 005 01 015 02 025 03 STRAIN, IN/IN STRESS-STRAIN CURVE TURKEY POINT UNIT 3 "X" CAPSULE 100 90 80 70 60 U) 50 U) 40 S14 30 550 F 20 10 0

0 0-05 01 0.15 02 025 03 STRAIN, IN/IN Figure 5-21 Engineering Stress-Strain Curves for Turkey Point Unit 3 Reactor Vessel Forging 123S266VA1, Tensile Specimens S13 and S14 Turkey Point Urut 3 Capsule X

5-38 STRESS-STRAIN CURVE TURKEY POINT UNIT 3 "X" CAPSULE 100 80 60 U)

LU I-40 W3 300 F 20 0

0 005 01 0.15 02 0 25 03 STRAIN, IN/IN STRESS-STRAIN CURVE "TURKEY POINT UNIT 3 "X" CAPSULE 100 80 60 w

Lit U) 40 W4 550 F 20 0

0 005 0.1 0.15 02 025 03 STRAIN, IN/IN Figure 5-22 Engineering Stress-Strain Curves for Turkey Point Unit 3 Reactor Weld Metal, Tensile Specimens W3 and W4 Turkey Point Umt 3 Capsule X

6-1 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6.1 INTRODUCTION

This section describes a discrete ordinates S,, transport analysis performed for the Turkey Point Unit 3 reactor to determine the neutron radiation environment within the reactor pressure vessel and surveillance capsules. In this analysis, fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements (dpa) were established on a plant and fuel cycle specific basis. An evaluation of the most recent dosimetry sensor set from Capsule X, withdrawn at the end of the eighteenth plant operating cycle, is provided. In addition, to provide an up-to-date data base applicable to the Turkey Point Unit 3 reactor, sensor sets from previously withdrawn capsules (T, S, and V) were re analyzed using the current dosimetry evaluation methodology. These dosimetry updates are presented in Appendix A of this report. Comparisons of the results from these dosimetry evaluations with the analytical predictions served to validate the plant specific neutron transport calculations. These validated calculations subsequently formed the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 54 Effective Full Power Years (EFPY). These projections also account for a plant uprating, from 2200 MWt to 2300 MWt, which began during the fifteenth operating cycle.

The use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for the development of damage trend curves as well as for the implementation of trend curve data to assess the condition of the vessel. 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 and improved accuracy in the evaluation of damage gradients through the reactor vessel wall.

Because of thi,, 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 database for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, "Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials."

All of the calculations and dosimetry evaluations described in this section and in Appendix A were based on the latest available nuclear cross-section data derived from ENDF/B-VI and made use of the latest available calculational tools. Furthermore, the neutron transport and dosimetry evaluation methodologies follow the guidance and meet the requirements of Regulatory Guide 1.190, "Calculational and Dosimetry 4

Methods for Determining Pressure Vessel Neutron Fluence.""l Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC approved methodology described in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 1996.1'51 The specific calculational Turkey Point Unit 3 Capsule X

6-2 methods applied are also consistent with those described in WCAP-15557, "Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology."t 16]

6.2 Discrete Ordinates Analysis A plan view of the Turkey Point Unit 3 reactor geometry at the core midplane is shown in Figure 4-1.

Eight irradiation capsules attached to the thermal shield are included in the reactor design that constitutes the reactor vessel surveillance program. The capsules are located at azimuthal angles of 2700 (0' from the core cardinal axes), 2800 (100 from the core cardinal axes), 2900 (200 from the core cardinal axes),

300 and 1500 (300 from the core cardinal axes), and 40', 500, and 2300 (400 from the core cardinal axes) as shown in Figure 4-1. The stainless steel specimen containers are 1-inch square by 56 inches in height.

The containers are 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 the test specimen location, the capsules themselves must be included in the analytical model.

In performing the fast neutron exposure evaluations for the Turkey Point Unit 3 reactor vessel and surveillance capsules, a series of fuel cycle specific forward transport calculations were carried out using the following three-dimensional flux synthesis technique:

0(r.O,z) =Vr,0) . (r, z) 0(r) where 4(rO,z) is the synthesized three-dimensional neutron flux distribution, ý(rO) is the transport solution in rO geometry, ý(rz) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and 4(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the rO two-dimensional calculation. This synthesis procedure was carried out for each operating cycle at Turkey Point Unit 3.

For selected Turkey Point Unit 3 fuel cycles that utilized part-length absorber rods in fuel assemblies located on the core flats to suppress the vessel fluence, the multi-channel analysis form of the three dimensional synthesis equation was used as promulgated in Regulatory Guide 1.1901141. Specifically, the transport analyses for Cycles 9 through 19 of Turkey Point Unit 3 were based on the following equation:

0(r,O,z) = 0.(r,O) * ,(r,z) + *(r,Oj* 4(rz) 0,(r) 4(r) where the first term, denoted as channel "A", represents all assemblies in the core except for those containing part-length absorber rods, whereas the second term, referred to as channel "B" only represents the assemblies on the core flats that contain the part-length absorber rods.

For the Turkey Point Unit 3 transport calculations, the r,0 model depicted in Figure 6-1 was utilized since the reactor is octant symmetric. This rO model includes the core, the reactor internals, the thermal shield Turkey Point Unit 3 Capsule X

6-3

-- including explicit representati6ns of the surveillance capsules at 00,. 100, 200, 300, and 40', the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and -the primary biological shield wall. This model formed the basis for the calculated results and enabled making comparisons to the surveillance capsule dosimetry evaluations. In addition, maximum neutron exposures'at the pressure vessel wall were derived based on a variant of this model in which the material composition of the surveillance capsules were redefined as downcomer water such that fast flux multipliers were determined and applied at selected azimuths along the vessel inner radius relative to the calculated results that were derived with the capsules present. In developing this analytical model, nominal design dimensions were employed for the various structural components. Likewise, water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full power operating conditions. The coolant densities were treated on a fuel cycle specific basis. The reactor core itself was treated as a homogeneous mixture of-fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, et cetera. The geometric mesh description of the rO reactor model consisted of 161 radial by 107 azimuthal intervals. Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the rO calculations was set at a value of 0.001.

The rz model used for the Turkey Point Unit 3 calculations that is shown in Figure 6-2 extended radially from the centerline of the reactor core out to a location interior to the primary biological shield and over an axial span from an elevation 1-foot below the active fuel to approximately I-foot above the active fuel. As in the case of the rO model, nominal design dimensions and full power coolant densities were employed in the calculations. In this case, the homogenous core region was treated as an equivalent cylinder with a volume equal to that of the active core zone. The stainless steel former plates located between the core baffle and core barrel regions were also explicitly included in the model. The r,z geometric mesh description of the reactor model consisted of 158 radial by 106 axial intervals. As in the case of the rO calculations, mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the r,z calculations was also set at a value of 0.001.

The one-dimensional radial model used in the synthesis procedure consisted of the same 158 radial mesh intervals included in the r,z model. Thus, radial synthesis factors could be determined on a meshwise basis throughout the entire geometry.

The core power distributions used in the plant specific transport analysis were taken from the appropriate Turkey Point Unit 3 fuel cycle design reports as well as supplemental material provided by the utility.

The data extracted from the design reports represented cycle dependent fuel assembly enrichments; burnups, and axial power distributions. This information was used to develop spatial and energy dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle averaged neutron flux, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of individual fuel assemblies. From these assembly dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

Turkey Point Unit 3 Capsule X

6-4 All of the transport calculations supporting this analysis were carried out using the DORT discrete ordinates code Version 3 .11171 and the BUGLE-96 cross-section library.tiS] The BUGLE-96 library provides a 67 group coupled neutron-gamma ray cross-section data set produced specifically for light water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P5 legendre expansion and angular discretization was modeled with an S1 6 order of angular quadrature.

Energy and space dependent core power distributions, as well as system operating temperatures, were treated on a fuel cycle specific basis.

Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-7. In Table 6-I, the calculated exposure rates and integrated exposures, expressed in terms of both neutron fluence (E > 1.0 MeV) and dpa, are given at the radial and azimuthal center of the five azimuthally symmetric surveillance capsule positions (0', 100, 200, 300, and 400). Also note that Table 6-1 presents calculated exposure rates and integrated exposures for Capsule X, which was irradiated at a 40' location during Cycles 1 through 11, and subsequently moved to a 00 location until it was removed from service at the end of Cycle 18. These results, representative of the axial midplane of the active core. etablish the calculated exposure of the surveillance capsules withdrawn to date as well as projected into the future.

Similar information is provided in Tables 6-2 and 6-3 for the reactor vessel inner radius. The vessel data given in Table 6-2 are representative of the axial location of the maximum neutron exposure at each of the four azimuthal locations, whereas comparable results that are summarized in Table 6-3 are maximum values taken at the intermediate shell course to lower shell course girth (circumferential) weld located approximately 22.8 inches below the core midplane. It is also important to note that the data for the vessel inner radius were taken at the clad/base metal interface, and thus, represent the naximum calculated exposure levels of the vessel forgings and welds.

Both calculated fluence (E > 1.0 MeV) and dpa data are provided in Table 6-1 through Table 6-3. These data tabulations include both plant and fuel cycle specific calculated neutron exposures at the end of the eighteenth operating fuel cycle as well as projections for the current operating fuel cycle. i.e.. Cycle 19, and future projections to 32, 48, and 54 effective full power years (EFPY). The projections were based on the assumption that the radial power distribution from fuel cycle 7 (using the core octant most compatible with current core designs) was representative of future plant operation since the use of part length absorbers in assemblies located on the core flats was conservatively assumed to be discontinued.

All remaining core parameters were obtained from the current operating cycle 19 design. The future projections are also based on the current reactor power level of 2300 MWt.

Radial gradient information applicable to fast (E > 1.0 MeV) neutron fluence and dpa are given in Tables 6-4 and 6-5, respectively. The data, based on the cumulative integrated exposures from Cycles I through 19, are presented on a relative basis for each exposure parameter at several azimuthal locations.

Exposure distributions through the vessel wall may be obtained by multiplying the calculated exposure at the vessel inner radius by the gradient data listed in Tables 6-4 and 6-5.

The calculated fast neutron exposures for the four surveillance capsules withdrawn from the Turkey Point Unit 3 reactor are provided in Table 6-6. These assigned neutron exposure levels are based on the plant and fuel cycle specific neutron transport calculations performed for the Turkey Point Unit 3 reactor.

Updated lead factors for the Turkey Point Unit 3 surveillance capsules are provided in Table 6-7. The capsule lead factor is defined as the ratio of the calculated fluence (E > 1.0 MeV) at the geometric center of the surveillance capsule to the corresponding maximum calculated fluence at the pressure vessel Turkey Point Unit 3 Capsule X

6-5 clad/base metal interface. In Table 6:7, the lead factors for capsules tlihi hhve been withdrawn from the reactor (T, S, V, and X) were based on the calculated fluence values for the irradiation period corresponding to the time of withdrawal for the individual capsules. For the capsules remaining in the reactor (U, Y, W, and Z), the lead factors correspond to the calculated fluence values at the end of Cycle 19, the current operating fuel cycle for Turkey Point Unit 3.

6.3 Neutron Dosimetry The validity of the calculated neutron exposures previously reported in Section 6.2 is demonstrated by a direct comparison against the measured sensor reaction rates and via a least squares evaluation performed for each of the capsule dosimetry sets. However, since the neutron dosimetry measurement data merely serves to validate the calculated results, only the direct comparison of measured-to-calculated results for the most recent surveillance capsule removed from service is provided in this section of the report. For completeness, the assessment of all measured dosimetry removed to date, based on both direct and least squares evaluation comparisons, is documented in Appendix A.

The direct comparison of measured versus calculated fast neutron threshold reaction rates for the se'nsors from Capsule X, that was withdrawn from Turkey Point Unit 3 at the end of the eighteenth fuel cycle, is summarized below.

Reaction Rates (ms/atom) M/C Reaction Measured Calculated Ratio 63Cu(n,ox)"Co 3.36E-17 3.19E-17 1.05 54Fe(np) 54 Mn 3.1613-15 3.411E-15 0.93 238U(n,p) 137Cs (Cd) 1.36E-14 1.64E- 14 237 0.83 Np(n,f)137Cs (Cd) 1.23E-13 1.23E-13 1.00 Average: 0.95

% Standard Deviation: 1 10.2 The mcasured-to-calculated (M/C) reaction rate ratios for the Capsule X threshold reactions range from 0.83 to 1.05, and the average M/C ratio is 0.95 +/- 10.2% (]a). This direct comparison falls well within the +/- 20% criterion specified in Regulatory Guide 1.190; furthermore, it is consistent with the full set of comparisons given in Appendix A for all measured dosimetry removed to date from the Turkey Point Unit 3 reactor. As a result, these comparisons validate the-current analytical results described in Section 6.2 and are deemed applicable for Turkey Point Unit 3.

6.4 Calculational Uncertainties The uncertainty associated with the calculated neutron exposure of the Turkey Point Unit 3 surveillance capsule and reactor pressure vessel is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology was carried out in the following four stages:

I- Comparison of calculations with benchmark measurements from the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL).

2- Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H. B. Robinson power reactor benchmark experiment.

Turkey Point Unit 3 Capsule X

6-6 3 - An analytical sensitivity study addressing the uncertainty components resulting important input parameters applicable to the plant specific transport calculations used in the neutron exposure assessments.

4- Comparisons of the plant specific calculations with all available dosimetry results from the Turkey Point Unit 3 surveillance program.

The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operational or geometric variables that impact power reactor calculations. The second phase of the qualification (H. B. Robinson comparisons) addressed uncertainties in these additional areas that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations. The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant specific input parameters. The overall calculational uncertainty applicable to the Turkey Point Unit 3 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Turkey Point Unit 3 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule and pressure vessel neutron exposures previously described in Section 6.2. As such, the validation of the Turkey Point Unit 3 analytical model based on the measured plant dosimetry is completely described in Appendix A.

The following summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference 3.

Capsule Vessel IR PCA Comparisons 3% 3%

H. B. Robinson Comparisons 3% 3%

Analytical Sensitivity Studies 10% 11%

Additional Uncertainty for Factors not Explicitly Evaluated 5% 5%

Net Calculational Uncertainty 12% 13%

The net calculational uncertainty was determined by combining the individual components in quadrature.

Therefore, the resultant uncertainty was random and no systematic bias was applied to the analytical results.

The plant specific measurement comparisons described in Appendix A support these uncertainty assessments for Turkey Point Unit 3.

Turkey Point Unit 3 Capsule X

6-7

%" Figure 6-1 Turkey Point Unit 3 rO Reactor Geometry at the Core Midplane E

R Axis (cm)

Note- For core reload designs that have part length absorber rods installed. Zone A represents the fuel assemblies that do not contain these rods and Zone B represents the fuel assemblies where these absorber rods are used Turkey Point Unit 3 Capsule X

6-8 Figure 6-2 Turkey Point Unit 3 rz Reactor Geometry I-cx2 Cd, oo Zone A N0 (0

I.)

W)

(0D 1

I** J.....L..............IL..L'__________________

_______ -k ____

0 50 100 150 260 250 300 R Axis (cm)

Note: For core reload designs that have part length absorber rods installed. Zone A represents the fuel assemblies that do not contain these rods and Zone B represents the fuel assemblies %%herethese absorber rods are used.

Turkey Point Unit 3 Capsule X

6-9 Table 6-1 Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Neutrons (E > 1.0 MeV)

Cumulative Cumulative Neutron Flux (E > 1.0 MeV)

Cycle Irradiation Irradiation [n/cm 2-s]

Length Time Time Cycle [EFPSI [EFPS] [EFPY], 00 100 - 20° I 3.62E+07 3.62E+07 1.15 1.66E+I1 1.16E+I 1 5.14E+10 2 2.45E+07 6.06E+07 1.92 1.61E+ I1 1.13E+11 4.88E+10 3 2.40E+07 8.46E+07 2.68 1.54E+11 1.19E+ 11 5.73E+10 4 2.46E+07 1.09E+08 3.46 1.56E+11 1.17E+ 11 4.911E+10 5 2.45E+07 1.34E+08 4.24 2.OOE+l1 1.41E+11 5.20E+ 10 6 1.59E+07 1.50E+08 4.74 1.85E+11 1.31E+I 1 4.81E+10 7 2.90E+07 1.79E+08 5.66 1.40E+11 9.88E+10 4.55E+10 8 4.30E+07 2.22E+08 7.03 1.47E+ I I 9.89E+ 10 4.56E+10 9 3.26E+07 2.54E+08 8.06 7.15E+ 10 5.72E+10 3.90E+ I0 10 3.95E+07 2.94E+08 9.31 6.84E+ 10 5.61 E+ 10 3.80E+10 11 4.13E+07 3.35E+08 10.62: 7.33E+10 5.76E+10 3.75E+10 12 3.91E+07 3.74E+08 11.86 7.16E+10 5.72E+10 3.77E+10 13 3.95E+07 4.14E+08 13.11 7.02E+ 10 5.47E+ 10 3.53E+10 14 3.93E+07 4.53E+08 14.36 7.46E+10 5.91E+10 3.90E+10 15 4.18E+07 4.95E+08 15.68 7.55E+ 10 5.89E+ 10 3.75E+ 10 16 4.38E+07 5.39E+08 17.07 7.71E+10 6.25E+10 4.53E+10 17 4.12E+07 5.80E+08 18.37 7.62E+ 10 5.811E+10 3.47E+310 18 4.65E+07 6.26E+08 19.85: 6.64E+ 10 5.411E+10 4.04E+ I0 19 (Pjt.) 4.15E+07 6.68E+08 21.16 7.64E+ 10 5.92E+ I0 3.65E+ 10 Future 3.42E+08 1.01E+09 32.00 1.OOE+1 I 8.30E+10 4.81E+10 Future 5.05E+08 1.51E+09 48.00 1.00E+ I1 8.30E+ 10 4.81 E+ I0 Future 1.89E+08 1.70E+09 54.00 1.00E+ I1 8.30E+10 4.81E+10 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Turkey Point Unit 3 Capsule X

6-10 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Neutrons (E > 1.0 MeV)

Cumulative Cumulative Neutron Flux (E > 1.0 MeV)

Cycle Irradiation Irradiation [n/cm 2-s]

Length Time Time 400 - 0' Cycle IEFPS] [EFPS] [EFPY] 300 400 (Cap. X*)

I 3.62E+07 3.62E+07 1.15 3.90E+10 2.59E+10 2.59E+10 2 2.45E+07 6.06E+07 1.92 3.66E+ 10 2.37E+ 10 2.37E+I0 3 2.40E+07 8.46E+07 2.68 4.48E+10 3.03E+10 3.03E+10 4 2.46E+07 1.09E+08 3.46 3.49E+10 2.64E+ 10 2.64E+ 10 5 2.45E+07 1.34E+08 4.24 3.23E+ 10 2.07E+ 10 2.07E+ 10 6 1.59E+07 1.50E+08 4.74 2.96E+ 10 1.99E+ 10 1.99E+ I0 7 2.90E+07 1.79E+08 5.66 3.34E+I0 2.27E+10 2.27E+10 8 4.30E+07 2.22E+08 7.03 3.29E+ 10 2.20E+10 2.20E+ 10 9 3.26E+07 2.54E+08 8.06 3.23E+10 2.35E+ 10 2.35E+10 10 3.95E+07 2.94E+08 9.31 2.99E+10 2.05E+ 10 2.05 E+ 10 II 4.13E+07 3.35E+08 10.62 2.96E+10 1.98E+10 1.98E+10 12 3.91E+07 3.74E+08 11.86 3.OOE+ 10 2.04E+ 10 7.16E+ 10 13 3.95E+07 4.14E+08 13.11 2.81E+10 2.05E+10 7.02E+10 14 3.93E+07 4.53E+08 14.36 3.14E+10 2.28E+10 7.46E+ 10 15 4.18E+07 4.95E+08 15.68 2.97E+ 10 2.07E+ 10 7.55E+10 16 4.38E+07 5.39E+08 17.07 3.57E+10 2.15E+10 7.71E+10 17 4.12E+07 5.80E+08 18.37 2.82E+ 10 1.92E+ 10 7.62E+ 10 18 4.65E+07 6.26E+08 19.85 3.25E+ 10 2.16E+ 10 6 64E+10 19 (Pjt.) 4.15E+07 6.68E+08 21.16 2.94E+10 2.02E+10 7.64E+10 Future 3.42E+08 1.01E+09 32.00 3.73E+10 2.52E+10 1.OOE+I I Future 5.05E+08 1.51E+09 48.00 3.73E+10 2.52E+10 1 OOE+I1 Future I 89E+08 1.70E+09 54.00 3.73E+10 2.52E+10 i.00E+ I1 Note: Neutron exposure .alues reported for the surveillance capsules are centered at the core midplane.

• Capsule X was irradiated at a 400 location during Cycles I through II followed by a 00 location dunrng Cycles 12 through 18 Ahen it was subsequently removed from ser,,ice Turkey Point Unit 3 Capsule X

6-11 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Neutrons (E > 1.0 MeV)

Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation [n/cm 2]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 100 200 I 3.62E+07 3.62E+07 1.15 5.99E+ 18 4.21E+18 1.86E+18 2 2.45E+07 6.06E+07 1.92 9.92E+ 18 6.97E+18 3.05E+ 18 3 2.40E+07 8.46E+07 2.68 1.36E+ 19 9.84E+ 18 4.43E+18 4 2.46E+07 1.09E+08 3.46 1.75E+19 1.27E+ 19 5.64E+18 5 2.45E+07 1.34E+08 4.24 2.24E+ 19 1.62E+ 19 6.911E+18 6 1.59E+07 1.50E+08 4.74 2.53E+ 19 1.83E+ 19 7.68E+ 18 7 2.90E+07 1.79E+08 5.66 2.94E+19 2.1 ]E+19 9.OOE+ 18 8 4.30E+07 2.22E+08 7.03 3.57E+ 19 2.54E+ 19 I.101E+19 9 3.26E+07 2.54E+08 8.06 3.80E+19 2.73E+19 1.22E+19 10 3.95E+07 2.94E+08 9.31 4.07E+ 19 2.95E+ 19 1.37E+ 19 11 4.13E+07 3.35E+08 10.62 4.38E+19 3.19E+ 19 1.53E+ 19 12, 3.91E+07 3.74E+08 11.86' 4.66E+19 3.41E+19 1.68E+19 13- 3.95E+07 4.14E+08 13.11 4.93E+ 19 3.63E+19 1.82E+19 14 3.93E+07 4.53E+08 14.36 5.23E+19 3.86E+19 1.97E+19 15 4.18E+07 4.95E+08 15.68 5.54E+19 4.10E+19 2.13E+19 16 4.38E+07 5.39E+08 17.07 5.88E+19 4.38E+19 2.32E+19 17 4.12E+07 5.80E+08 18.37 6.19E+ 19 4.62E+ 19 2.47E+ 19 18 4.65E+07 6.26E+08 19.85 6.50E+19 4.87E+19 2.65E+19 19 (Pjt.) 4.15E+07 6.68E+08 21.16 6.82E+-19 5.1IE+19 2.81E+19 Future 3.42E+08 1.OIE+09 32.00 1.03E+20 7.95E+19 4.45E+ 19 Future 5.05E+08 1.51 E+09 48.00, 1.53E+20 1.2 1E+20 6.88E+19 Future 1.89E+08 1.70E+09 54.00 1.72E+20 1.37E+20 7.79E+ 19 Note. Neutron exposure %aluesreported for the sur eillance capsules are centered at the core midplane Turkey Point Unit 3 Capsule X

6-12 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Neutrons (E > 1.0 MeV)

Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation [n/cm2]

Length Time Time 400 --* 0' Cycle [EFPS] [EFPS] [EFPY] 300 400 (Cap. X*)

I 3.62E+07 3.62E+07 1.15 1.41E+18 9.36E+ 17 9.36E+ 17 2 2.45E+07 6.06E+07 1.92 2.31E+18 1.52E+ 18 1.52E+18 3 2.40E+07 8.46E+07 2.68 3.38E+ 18 2.24E+ 18 2.24E+ 18 4 2.46E+07 1.09E+08 3.46 4.24E+ 18 2.89E+ 18 2.89E+ 18 5 2.45E+07 1.34E+08 4.24 5.03E+ 18 3.40E+ 18 3.40E+ 18 6 1.59E+07 1.50E+08 4.74 5.50E+ 18 3.72E+ 18 3.72E+ 18 7 2.90E+07 1.79E+08 5.66 6.47E+18 4.38E+18 4.38E+18 8 4.30E+07 2.22E+08 7.03 7.89E+18 5.32E+18 5.32E+18 9 3.26E+07 2.54E+08 8.06 8.94E+ 18 6.09E+ 18 6.09E+ 18 10 3.95E+07 2.94E+08 9.31 1.01E+19 6.90E+18 6.90E+18 11 4.13E+07 3.35E+08 10.62 l. 14E+ 19 7.72E+ 18 7.72E+ 18 12 3.91E+07 3.74E+08 11.86 1.25E+19 8.51E+18 1.05E+19 13 3.95E+07 4.14E+08 13.11 1.36E+19 9.33E+18 1.33E+19 14 3.93E+07 4.53E+08 14.36 1.49E+ 19 1.02E+ 19 1.62E+ 19 15 4.18E+07 4.95E+08 15.68 i.61E+19 1.11E+19 1.94E+19 16 4.38E+07 5.39E+08 17.07 1.77E+19 1.20E+ 19 2.27E+ 19 17 4.12E+07 5.80E+08 18.37 1.88E+ 19 1.28E+ 19 2.59E+ 19 18 4.65E+07 6.26E+08 19.85 2.03E+ 19 1.38E+19 2.90E+ 19 19 (Pjt.) 4.15E+07 6.68E+08 21.16 2.16E+ 19 1.47E+ 19 3.21 E+ 19 Future 3.42E+08 1.01E+09 32.00 3.43E+19 2.33E+19 6.65E+19 Future 5.05E+08 1.51E+09 48.00 5.31 E+ 19 3.60E+ 19 1.17E+20 Future 1.89E+08 1.70E+09 54.00 6.02E+19 4.08E+ 19 1.36E+20 Note: Neutron exposure values reported for the surseillance capsules are centered at the core nudplane.

• Capsule X was irradiated at a 40' location during Cycles I through 11 ollloued by a 0' location during Cycles 12 through 18 when it was subsequently remox ed from sen ice.

Turkey Point Unit 3 Capsule X

6-13 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center IRON ATOM DISPLACEMENTS Cumulative Cumulative Displacement Rate Cycle Irradiation Irradiation [dpa/s]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 100 200 1 3.62E+07 3.62E+07 1.15 2.80E- 10 1.99E-10 8.41E-I 1 2 2.45E+07 6.06E+07 1.92 2.72E- 10 1.93E-10 8.00E- 11 3 2.40E+07 8.46E+07 2.68 2.61E-10 2.03E-1 0 9.38E-! 1 4 2.46E+07 1.09E+08 3.46 2.64E-l 0 2.00E- 10 8.04E- I1 5 2.45E+07 1.34E+08 4.24 3.38E-10 2.41E-10 8.53E- Il 6 1.59E+07 1.50E+08 4.74 - 3.14E-10 2.24E-10 7.88E-1 i 7 2.90E+07 1.79E+08 5.66 2.37E- 10 1.68E- 10 7.44E- 11 8 4.30E+07 2.22E+08 7.03 2.47E- 10 1.68E- 10 7.46E- 11 9 3.26E0-07 2.54E+08 8.06 1.20E- 10 9.68E- I I 6.35E- I I 10 3.95E+07 2.94E+08 9.31 1.15E- 10 9.49E- I1 6.18E- I I I1 4.13E+07 3.35E+08 10.62 1.23E-10 9.75E-11 6.11E-I1

-12 3.9 1E+07 3.74E+08 11.86 1.20E-10 9.67E-11 6.14E- 11 13 3.95E+07 4.14E+08 13.11 1.18E-10 9.25E-I I 5.74E-I1 14 3.93E+07 4.53E+08 14.36 1.25E-10 9.99E-Il 6.35E-I 1 15 4.18E+07 4.95E+08 15.68 1.27E-10 9.96E- 11 6.11E-11 16 4.38E+07 5.39E+08 17.07 1.30E- 10 1.06E- 10 7.37E- I 17 4.1213+07 5.80E+08 18.37 1.28E-10 9.8313-11 5.65E-11 18 4.65E+07 6.26E+08 19.85 1.12E-10 9.14E-11 6.577E-11 19 (Pjt.) 4.15E+07 6.68E+08 21.16 1.28E-I0 1.00E-10 5.94E-11 Future 3.42E+08 1.01E+09 32.00 1.69E-10 1.411E-10 7.84E-1 I Future 5.05E+08 1.511E+09 48.00 1.69E-10 1.41E-10 7.84E-11 Future 1.89E+08 1.70E+09 54.00 1.69E-10 1.41E-10 7.84E1-1 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Turkey Point Unit 3 Capsule X

6-14 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center IRON ATOM DISPLACEMENTS Cumulative Cumulative Displacement Rate Cycle Irradiation Irradiation [dpa/s]

Length Time Time 400 - 00 Cycle [EFPS] [EFPS] [EFPY] 300 400 (Cap. X*)

I 3.62E+07 3.62E+07 1.15 6.40E-11 4.20E-11 4.20E-11 2 2.45E+07 6.06E+07 1.92 6.00E-11 3.84E- I I 3.84E- 11 3 2.40E+07 8.46E+07 2.68 7.35E- I1 4.90E-1 1 4.90E- 11 4 2.46E+07 1.09E+08 3.46 5.70E-11 4.27E- I1 4.27E- 11 5 2.45E+07 1.34E+08 4.24 5.29E- I1 3.35E-1 1 3.35E- 11 6 1.59E+07 1.50E+08 4.74 4.84E-I 1 3.22E-1 I 3.22E- 11 7 2.90E+07 1.79E+08 5.66 5.47E-1 I 3.67E-1 1 3.67E-1 I 8 4.30E+07 2.22E+08 7.03 5.39E-I I 3.57E- 1I 3.57E- I1 9 3.26E+07 2.54E+08 8.06 5.29E- 1 3.80E-1 3.80E-! I 10 3.95E+07 2.94E+08 9.31 4.88E-11 3.31E- 11 3.3 1E-I I 11 4.13E+07 3.35E+08 10.62 4.84E-1 1 3.20E- II 3.20E- 11 12 3.91E+07 3.74E+08 11.86 4.91E-I 1 3.29E- 11 1.20E-10 13 3.95E+07 4.14E+08 13.11 4.60E-11 3.32E-11 1.18E-10 14 3.93E+07 4.53E+08 14.36 5.13E-1 I 3.68E-11 1.25E-10 15 4.18E+07 4.95E+08 15.68 4.86E-1I 3.35E-I I 1.27E-10 16 4.38E+07 5.39E+08 17.07 5.84E-I I 3.48E-11 1.30E-10 17 4.12E+07 5.80E+08 18.37 4.60E-11 3.10E-11 I1.28E-10 18 4.65E+07 6.26E+08 19.85 5.31E-11 3.49E-11 1.12E-10 19 (Pjt.) 4.15E+07 6.68E+08 21.16 4.8013-11 3.27E-11 1.28E-10 Future 3.42E+08 1.01E+09 32.00 6.10E-1 I 4.08E-11 1.69E- 10 Future 5.05E+08 1.511E+09 48.00 6.10E-1 1 4.08E- 11 1.69E-10 Future 1.89E+08 1.70E+09 54.00 6.10E-1 1 4.08E-11 1.69E-10 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

• Capsule X was irradiated at a 400 location during Cycles I through I I followed by a 0' location during Cycles 12 through 18 "hen it was subsequently removed from ser%ice Turkey Point Unit 3 Capsule X

6-15 "Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center IRON ATOM DISPLACEMENTS Cumulative Cumulative Displacements Cycle Irradiation Irradiation [dpa]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 100 200 1 3.62E+07 3.62E+07 1.15 1.01 E-02 7.188E-03 3.04E-03 2 2.45E+07 6.06E+07 1.92 1.68E-02 1.1913-02 5.00E-03 3 2.40E+07 8.46E+07 2.68 2.31E-02 1.68E-02 7.25E-03 4 2.46E+07 1.09E+08 3.46 2.96E-02, 2.17E-02 9.23E-03 5 2.45E+07 1.34E+08 4.24 3.79E-02 2.76E-02 1.1 3E-02 6 1.59E+07 1.50E+08 4.74 4.2813-02 3.12E-02 1.26E-02 7 2.90E+07 1.79E+08 5.66 4.97E-02 3.61 E-02 1.4713-02 8 4.30E+07 2.22E+08 7.03 6.04E-02 4.33E-02 1.79E-02 9 3.26E+07 2.54E+08 8.06 6.43E-02 4.65E-02 2.00E-02 10 3.95E+07 2.94E+08 9.31 6.88E-02 5.02E-02 2.2513-02 I1 4.13E+07 3.35E+08 10.62 7.39E-02 5.42E-02 2.50E-02 12 3.911E+07 3.74E+08 11.86 7.86E-02 5.8013-02 2.74E-02 13 3.95E+07 4.14E+08 13.11 8.3313-02 6.1713-02 2.97E-02 14 3.93E+07 4.53E+08 14.36 8.82E-02 6.56E-02 3.211E-02 15 4.18E+07 4.95E+08 15.68 9.35E-02 6.97E-02 3.47E-02 16 4.38E+07 5.39E+08 17.07 9.92E-02 7.44E-02 3.79E-02 17 4.12E+07 5.80E+08 18.37 1.0513-01 7.84E-02 4.0313-02 18 4.65E+07 6.26E+08 19.85 1.1013-01 8.27E-02 4.33E-02 19 (Pjt.) 4.15E+07 6.68E+08 21.16 1.1513-01 8.68E-02 4.5813-02 Future 3.42E+08 1.01E+09 32.00 1.73E-01 1.35E-01 7.2613-02 Future 5.05E+08 1.511E+09 48.00 2.58E-01 2.06E-01 1.122E-01 Future 1.89E+08 1.70E+09 54.00 2.90E-01 2.33E-01 1.27E-01 Note. Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Turkey Point Unit 3 Capsule X

6-16 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center IRON ATOM DISPLACEMENTS Cumulative Cumulative Displacements Cycle Irradiation Irradiation [dpa]

Length Time Time 40' -+ 0' Cycle [EFPS] [EFPS] [EFPY] 300 400° (Cap. X*)

1 3.62E+07 3.62E+07 1.15 2.31E-03 1.52E-03 1.52E-03 2 2.45E+07 6.06E+07 1.92 3.78E-03 2.46E-03 2.46E-03 3 2.40E+07 8.46E+07 2.68 5.55E-03 3.63E-03 3.63E-03 4 2.46E+07 1.09E+08 3.46 6.95E-03 4.69E-03 4.69E-03 5 2.45E+07 1.34E+08 4.24 8.25E-03 5.51 E-03 5 51 E-03 6 1.59E+07 1.50E+08 4.74 9.02E-03 6.02E-03 6.02E-03 7 2.90E+07 1.79E+08 5.66 i .06E-02 7.08E-03 7.08E-03 8 4.30E+07 2.22E+08 7.03 1.29E-02 8.62E-03 8.62E-03 9 3.26E+07 2.54E+08 8.06 1.47E-02 9.86E-03 9.86E-03' 10 3.95E+07 2.94E+08 9.31 1.66E-02 1.12E-02 1.12E-02 II 4.13E+07 3.35E+08 10.62 1.86E-02 1.25E-02 1.25E-02 12 3.91 E+07 3.74E+08 11.86 2.05E-02 1.38E-02 1.72E-02 13 3.95E+07 4.14E+08 13.11 2.23E-02 1.51E-02 2.19E-02 14 3.93E+07 4.53E+08 14.36 2.43E-02 1.65E-02 2.68E-02 15 4.18E+07 4.95E+08 15.68 2.64E-02 1.79E-02 3.2 1E-02 16 4.38E+07 5.39E+08 17.07 2.89E-02 1.95E-02 3.78E-02 17 4.12E+07 5.80E+08 18.37 3.08E-02 2.07E-02 4.30E-02 18 4.65E+07 6.26E+08 19.85 3.33E-02 2.24E-02 4.82E-02 19 (Pjt.) 4.15E+07 6.68E+08 21.16 3.53E-02 2.37E-02 5.36E-02 Future 3.42E+08 1.01E+09 32.00 5.61E-02 3.77E-02 1.11 E-01 Future 5.05E+08 1.5 1E+09 48.00 8.69E-02 5.82E-02 1.97E-01 Future 1.89E+08 1.70E+09 54.00 9.85E-02 6.60E-02 2.29E-01 Note: Neutron exposure values reported for the sureillance capsules are centered at the core midplane.

Capsule X was irradiated at a 40' location during Cycles I through II followed by a 0' location during Cycles 12 through 18 v,hen it Aas subsequently removed from service.

Turkey Point Unit 3 Capsule X

6-17 Table 6-2 Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Inter'face Cumulative Cumulative Neutron Flux (E > 1.0 MeV)

Cycle Irradiation Irradiation [n/cm 2-s]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 150 300 450 1 3.62E+07 3.62E+07 1.15 6.07E+10 2.78E+10 1.55E+10 1.01E+10 2 2.45E+07 6.06E+07 1.92 5.93E+ 10 2.70E+ 10 1.46E+ 10 9.29E+09 3 2.40E+07 8.46E+07 2.68 5.66E+ 10 2.94E+ 10 1.75E+ 10 1.1713+10 4 2.46E+07 1.09E+08 3.46 5.68E+10 2.76E+10 1.38E+10 1.03E+10 5 2.45E+07 1.34E+08 4.24 7.19E+ 10 3.14E+ 10 1.28E+10 7.98E+09 6 1.59E+07 1.50E+08 4.74 6.66E+ 10 2.91 E+ 10 1.17E+10 7.73E+09 7 2.90E+07 1.79E+08 5.66 5.09E+ 10 2.41 E+ 10 1.32E+ 10 8.87E+09 S8 4.30E+07. 2.22E+08 7.03 5.27E+10 2.411E+10 1.30E+1 0 8.55E+09 9 3.26E+07 2.54E+08 8.06 3.23E+ 10 1.88E+10 1.27E+ 10 9.20E+09

,10 3.95E+07 2.94E+08 9.31 3.1OE+ 10 1.87E+10 1.1913+10 8.03E+09 11 4.13E+07 3.35E+08 10.62 3.31E+10 1.86E+10 1.21E+10 7.97E+09 12 3.911E+07 3.74E+08 11.86 3.24E+10 1.8613+10 1.18E+10 7.911E+09 13 3.95E+07 4.1413+08 13.11 3.17E+10 1.78E+10 1.13E+10 8.1313+09 14-, 3.93E+07 4.53E+08 14.36 3.35E+ 10 1.92E+ 10 1.24E+ 10 8.9313+09 15 4.18E+07 4.95E+08 15.68 3.411E+10 1.89E+ 10' 1.18E+10 8.09E+09 16 4.38E+07 5.39E+08 17.07 3.49E+10 2.13E+10 1.411E+10 8.22E+09 17 4.12E+07 5.80E+08 18.37 3.42E+ 10 1.80E+10 1.11 E+ 10 7.44E+09 18 4.65E+07 6.26E+08 19.85 2.99E+ 10 1.86E+10 1.28E+ 10 8.35E+09 19 (Pjt.) 4.15E+07 6.68E+08 21.16 3.43E+10 1.86E+10 'I.16E+10 7.84E+09 Future ,3.42E+08 1.01E+09 32.00 3.71E+10 2.2813+10 1.47E+10 9.811E+09 Future 5.0513+08 1.511E+09 48.00 3.71E+10 2.28E+10 1.47E+10 9.811E+09 Future 1.89E+08 1.70E+09 54.00- 3.711E+10 2.2813+10 1.47E+ 10 9.811E+09 Turkey Point Unit 3 Capsule X

6-18 Table 6-2 cont'd Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation [n/cm 2]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 150 300 450 1 3.62E+07 3.62E+07 1.15 2.19E+ 18 1.01E+18 5.59E+17 3.65E+17 2 2.45E+07 6.06E+07 1.92 3.64E+ 18 1.66E+ 18 9.17E+ 17 5.90E+ 17 3 2.40E+07 8.46E+07 2.68 5.OOE+ 18 2.37E+ 18 1.34E+ 18 8.71E+17 4 2.46E+07 1.09E+08 3.46 6.40E+ 18 3.04E+ 18 1.68E+18 1.13E+ 18 5 2.45E+07 1.34E+08 4.24 8.17E+ 18 3.80E+ 18 1.99E+ 18 1.32E+ 18 6 1.59E+07 1.50E+08 4.74 9.22E+ 18 4.27E+ 18 2.18E+ 18 1.44E+ 18 7 2.90E+07 1.79E+08 5.66 1.07E+ 19 4.96E+ 18 2.56E+18 1.70E+ 18 8 4.30E+07 2.22E+08 7.03 1.30E+19 5.97E+18 3.12E+18 2.05E+18 9 3.26E+07 2.54E+08 8.06 1.40E+ 19 6.53E+18 3.53E+ 18 2.35E+ 18 10 3.95E+07 2.94E+08 9.31 1.53E+ 19 7.22E+1 8 4.OOE+1 8 2.66E+18 11 4.13E+07 3.35E+08 10.62 1.66E+19 7.95E+18 4.50E+18 2.98E+18 12 3.91E+07 3.74E+08 11.86 1.79E+ 19 8.64E+ 18 4.96E+ 18 3.29E+ 18 13 3.95E+07 4.14E+08 13.11 1.91E+19 9.33E+18 5.41E+18 3.61E+18 14 3.93E+07 4.53E+08 14.36 2.05E+ 19 1.01E+19 5.90E+ 18 3.96E+18 15 4.18E+07 4.95E+08 15.68 2.19E+ 19 1.09E+ 19 6.39E+ 18 4.29E+ 18 16 4.38E+07 5.39E+08 17.07 2.34E+ 19 1.18E+19 7.OOE+ 18 4.65E+ 18 17 4.12E+07 5.80E+08 18.37 2.48E+19 1.25E+ 19 7.46E+ 18 4.96E+ 18 18 4.65E+07 6.26E+08 19.85 2.62E+ 19 1.34E+ 19 8.05E+ 18 5.35E+ 18 19 (Pjt.) 4.15E+07 6.68E+08 21.16 2.76E+19 1.42E+19 8.54E+18 5.67E+18 Future 3.42E+08 1.01E+09 32.00 4.03E+19 2.19E+ 19 1.35E+ 19 9.03E+18 Future 5.05E+08 1.51E+09 48.00 5.91E+19 3.33E+ 19 2.09E+ 19 1.40E+ 19 Future 1.89E+08 1.70E+09 54.00 6.61E+19 3.76E+19 2.37E+19 1.58E+19 Turkey Point Unit 3 Capsule X

6-19 Table 6-2 cont'd Calculated Azimuthal Variation Of Fast Neutron Exposure Rates And Iron Atom Displacement Rates At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacement Rate Cycle Irradiation Irradiation [dpa s]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 150 300 450 I 3.62E+07 3.62E+07 1.15 1.05E-10 4.63E- I 1 2.63E-11 1.66E-l I 2 2.45E+07 6.06E+07 1.92 1.02E-10 4.49E- I1 2.49E-11 1.52E-11 3 2.40E+07 8.46E+07 2.68 9.74E- 11 4.88E-11 2.98E-11 1.9 1E-11 4 2.46E+07 1.09E+08 3.46 9.79E- I1 4.59E-1 I 2.35E- I I 1.69E-] I 5 2.45E+07 1.34E+08 4.24 1.24E-10 5.22E-11 2.18E-l ! 1.31E-11 6 1.59E+07 1.50E+08 4.74 1.15E-10 4.85E-11 2.01E-I I 1.27E-11 7 2.90E+07 1.79E+08 5.66 8.77E-11 4.00E-11 2.25E- 1 1.45E-11 8 4.30E+07 2.22E+08 7.03 9.08E-1I 4.OOE- 11 2.21E-I I 1.40E- 11 9 3.26E+07 2.54E+08 8.06 5.56E-11 3.1OE- 11 2.15E-11 1.50E-11 10 3.95E+07 2.94E+08 9.31 5.33E-I1 3.08E-11 2.01E-I 1 1.3 1E-11 S1I 4.13E+07 3.35E+08 10.62 5.69E-I 1 3.07E-1 2.04E-11 1.30E-11 12 3.91E+07 3.74E+08 11.86 5.56E-1 3.07E-I1 2.00E-11 1.29E-1 1 13 3.95E+07 4.14E+08 13.11 5.45E-1I 2.94E- 1 1.91E-1I 1.33E- 11 14 3.93E+07 4.53E+08 14.36 5.76E-I I 3.17E- 11 2.10E-lI1 1.46E-11 15 4.18E+07 4.95E+08 15.68 5.85E-11 3.12E-1I 1.99E-11 1.32E-11 16 4.38E+07 5.39E+08 17.07 6.OOE-I1 3.51E-I 1 2.39E-11 1.35E-11 17 4.12E+07 5.80E+08 18.37 5.88E-1l I 2.98E-1 1.88E-11 1.22E-11 18 4.65E+07 6.26E+08 19.85 5.14E-1I 3.06E- 1 2.16E-I 1 1.36E-11 19 (Pit.) 4.15E+07 6.68E+08 21.16 5.89E-1I 3.07E-I1 1.96E-1I 1.28E-II Future 3.42E+08 1.01 E+09 32.00 6.39E-11 3.76E-11 2.49E-11 I1.60E- I I Future 5.05E+08 1.51E+09 48.00 6.39E- I1 3.76E- I1 2.49E-11 1.60E-I I Future 1.89E+08 1.70E+09 54.00 6.39E- I1 3.76E-I I 2.49E- I I 1.60E-I I Turkey Point Uni 3 Capsule X

6-20 Table 6-2 cont'd Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacements Cycle Irradiation Irradiation [dpa]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 150 300 450 1 3.62E+07 3.62E+07 1.15 3.78E-03 1.67E-03 9.51E-04 5.99E-04 2 2.45E+07 6.06E+07 1.92 6.28E-03 2.76E-03 1.56E-03 9.67E-04 3 2.40E+07 8.46E+07 2.68 8.62E-03 3.93E-03 2.28E-03 1.43E-03 4 2.46E+07 1.09E+08 3.46 .1OE-02 5.06E-03 2.85E-03 1.84E-03 5 2.45E+07 1.34E+08 4.24 1.41 E-02 6.32E-03 3.39E-03 2.16E-03 6 1.59E+07 1.50E+08 4.74 1.59E-02 7.09E-03 3.71E-03 2.36E-03 7 2.90E+07 1.79E+08 5.66 1.84E-02 8.25E-03 4.36E-03 2.78E-03 8 4.30E+07 2.22E+08 7.03 2.23E-02 9.92E-03 5.3 1E-03 3.37E-03 9 3.26E+07 2.54E+08 8 06 2.42E-02 1.08E-02 6.01E-03 3.85E-03 10 3.95E+07 2.94E+08 9.31 2.63E-02 i .20E-02 6.81 E-03 4.36E-03 11 4.13E+07 3.35E+08 10.62 2.86E-02 1.32E-02 7.65E-03 4.88E-03 12 3.91E+07 3.74E+08 11.86 3.08E-02 1.43E-02 8.43E-03 5.39E-03 13 3.95E+07 4.14E+08 13.11 3.29E-02 1.55E-02 9.19E-03 5.90E-03 14 3.93E+07 4.53E+08 14.36 3.52E-02 1.67E-02 1.00E-02 6.47E-03 15 4.18E+07 4.95E+08 15.68 3.77E-02 1.80E-02 1.08E-02 7.03E-03 16 4.38E+07 5.39E+08 17.07 4.03E-02 1.95E-02 1.19E-02 7.61E-03 17 4.12E+07 5.80E+08 18.37 4.27E-02 2.08E-02 1.27E-02 8.11 E-03 18 4 65E+07 6.26E+08 19.85 4.5 1E-02 2.22E-02 1.37E-02 8.75E-03 19 (Pjt.) 4.15E+07 6.68E+08 21.16 4.75E-02 2.35E-02 1.45E-02 9.28E-03 Future 3.42E+08 1.01 E+09 32.00 6.94E-02 3.62E-02 2.30E-02 1.48E-02 Future 5.05E+08 1.5 1E+09 48.00 1.02E-01 5.50E-02 3.55E-02 2.29E-02 Future 1.89E+08 1.70E+09 54.00 1.14E-01 6.21E-02 4.02E-02 2.59E-02 Turkey Point Unit 3 Capsule X

6-21 t

Table 6-3 Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Intermediate Shell Course to Lower Shell Course Girth Weld Clad/Base Metal Interface Cumulative Cumulative Neutron Flux (E > 1.0 MeV)

Cycle Irradiation Irradiation [n/cm 2-s]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] '00 150 300 450 1 3.62E+07 3.62E+07 1.15 5.90E+ 10 2.71E+10 1.50E+ 10 9.84E+09 2 2.45E+07 6.06E+07 1.92 5.75E+ 10 2.62E+ 10 1.42E+10 9.01E+09 3 2.40E+07 8.46E+07 2.68 . 5.46E+ 10 2.84E+ 10 1.69E+ 10 1.13E+ 10 4 2.46E+07 1.09E+08 3.46 5.52E+10 2.68E+ 10 1.34E+ 10 1.00E+ 10 5 2.45E+07 1.34E+08 4.24 6.84E+ 10 2.98E+ 10 1.22E+10 7.59E+09 6 1.59E+07 1.50E+08 4.74 6.45E+10 2.82E+ 10 1.14E+10 7.49E+09 7 2.90E+07 1.79E+08 5.66 4.96E+ 10 2.35E+ 10 1.29E+10 8.64E+09 8 4.30E+07 2.22E+08 7.03 4.86E+ 10 2.22E+10 1.20E+ 10 7.88E+09 9 3.26E+07 2.54E+08 8.06 2.08E+ 10 1.50E+ 10 1.22E+ 10 8.88E+09 10 3.95E+07 2.94E+08 9.31 1.99E+ 10 1.46E+10 I.11E+10 7.58E+09 11 4.13E+07 3.35E+08 10.62 2.14E+10 1.50E+10 1.15E+10 7.59E+09 12 3.91E+07 3.74E+08 11.86 2.09E+10 1.49E+10 1.15E+10 7.70E+09 13 3.95E+07 4.14E+08 13.11 2.05E+10 1.42E+10 1.08E+I0 7.86E+09 14 3.93E+07 4.53E+08 14.36 2.17E+10 1.52E+10 1.17E+10 8.44E+09 15 4.18E+07 4.95E+08 15.68 2.18E+ 10 1.48E+10 1.11 E+10 7.67E+09 16 4.38E+07 5.39E+08 17.07 2.24E+ 10 1.69E+ 10 1.33E+'10 7.82E+09 17 4.12E+07 5.80E+08 18.37 2.21E+10 1.41 E+ 10 1.06E+ 10 7.13E+09 18 4.65E+07 6.26E+08 19.85 1.94E+ 10 1.51E+10 1.23E+ 10 8.05E+09 19 (Pjt.) 4.15E+07 6.68E+08 21.16 2.22E+10 1.46E+10 1.10E+10 7.49E+09 Future 3.42E+08 1.01E+09 32.00 3.55E+10 2.18E+ 10 1.40E+ 10 9.38E+09 Future 5.05E+08 1.51E+09 48.00 3.55E+ 10 2.18E+ 10 1.40E+10 9.38E+09 Future 1.89E+08 1.70E+09 54.00 3.55E+ 10 2.18E+ 10 1.40E+10 9.38E+09 Turkey Point Unit 3 Capsule X

6-22 Table 6-3 cont'd Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Intermediate Shell Course to Lower Shell Course Girth Weld Clad/Base Metal Interface Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation [n/cm2 ]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 150 300 450 1 3.62E+07 3.62E+07 1.15 2.13E+18 9.79E+17 5.44E+17 3.56E+17 2 2.45E+07 6.06E+07 1.92 3.54E+ 18 1.62E+ 18 8.90E+ 17 5.76E+17 3 2.40E+07 8.46E+07 2.68 4.85E+ 18 2.30E+ 18 1.30E+ 18 8.47E+ 17 4 2.46E+07 1.09E+08 3.46 6.21 E+ 18 2.96E+ 18 1.63E+ 18 1.09E+ 18 5 2.45E+07 1.34E+08 4.24 7.89E+ 18 3.69E+ 18 1.93E+ 18 1.28E+18 6 1.59E+07 1.50E+08 4.74 8.91E+18 4.14E+18 2.11E+18 1.40E+ 18 7 2.90E+07 1.79E+08 5.66 1.04E+ 19 4.82E+18 2.48E+ 18  !.65E+18 8 4.30E+07 2.22E+08 7.03 1.24E+19 5.78E+18 2.99E+18 1.99E+18 9 3.26E+07 2.54E+08 8.06 1.3 1E+19 6.27E+18 3.39E+18 2.28E+18 10 3.95E+07 2.94E+08 9.31 1.39E+ 19 6.85E+18 3.83E+ 18 2.58E+18 11 4.13E+07 3.35E+08 10.62 1.48E+19 7.47E+18 4.31E+18 2.89E+18 12 3.91E+07 3.74E+08 11.86 1.56E+19 8.05E+18 4.76E+18 3.19E+18 13 3.95E+07 4.14E+08 13.11 1.64E+19 8.61E+18 5.18E+18 3.50E+ 18 14 3.93E+07 4.53E+08 14.36 1.73E+19 9.20E+ 18 5.64E+ 18 3.84E+ 18 15 4.18E+07 4.95E+08 15.68 1.82E+19 9.82E+ 18 6.11E+18 4.16E+18 16 4.38E+07 5.39E+08 17.07 1.92E+ 19 1.06E+ 19 6 69E+18 4.50E+ 18 17 4.12E+07 5.80E+08 18.37 2.01E+19 1.1IE+19 7.12E+18 4.79E+18 18 4.65E+07 6.26E+08 19.85 2.10E+19 1.19E+19 7.69E+18 5.17E+18 19 (Pjt.) 4.15E+07 6.68E+08 21.16 2.19E+19 1.25E+19 8.15E+18 5.48E+ 18 Future 3.42E+08 1.01E+09 32.00 3.40E+19 1.99E+19 1.29E+19 8.68E+18 Future 5.05E+08 1.51E+09 48.00 5.20E+ 19 3.09E+ 19 2.OOE+ 19 1.34E+ 19 Future 1.89E+08 1.70E+09 54.00 5.87E+19 3.50E+19 2.27E+19 1.52E+19 Turkey Point Unit 3 Capsule X

6-23 Table 6-3 cont'd Calculated Azimuthal Variation Of Fast Neutron Exposure Rates And Integrated Exposures At The Intermediate Shell 'Course to Lower Shell Course Girth Weld Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacement Rate Cycle 'Irradiation Irradiation [dpa/s]

Length Time Time Cycle [EFPS] [EFPS] [EFPY] 00 150 300 450 1 3.62E+07 3.62E+07 1.15 1.02E-10 4.52E-11 2.57E-11 .62E- 1I 2 2.45E+07 6.06E+07 1.92 9.95E- I! 4.37E-I I 2.42E- 1I 1.48E-11 3 2.40E+07 8.46E+07 2.68 9.44E-1 I 4.73E-1 I 2.89E- I1 1.85E-I I 4 2.46E+07 1.09E+08 3.46 9.54E-1I 4.47E-Il 2.29E-1I 1.65E-1 I 5 2.45E+07 1.34E+08 4.24 1.18E-10 4.98E- I I 2.08E-I I 1.25E-I I 6 1.59E+07 1.50E+08 4.74 1.12E-10 4.71E-11 1.95E-I I 1.24E-11 7 2.90E+07 1.79E+08 5.66 8.58E-11 3.92E-11 2.20E- I I 1.42E- 11 8 4.30E+07 2.22E+08 7.03 8.40E- 1I 3.70E- 11 2.04E- I I 1.30E- 1I 9 3.26E+07 2.54E+08 8.06 3.59E- I I 2.48E- II 2.07E- I I 1.45E-I I 10 3.95E+07 2.94E+08 9.31 3.45E-11 2.42E-11 1.89E-1I 1.24E-I I II 4.13E+07 3.35E+08 10.62 3.70E- 11 2.48E- 11 1.95E-I1 1.24E- 11 12 3.9 1E+07 3.74E+08 11.86 3.61E-1I 2.46E- I I 1.95E-Il 1.26E-11 13 3.95E+07 4.14E+08 13.11 3.55E- I I 2.34E-I1 1.84E-II 1.29E-11 14 3.93E+07 4.53E+08 14.36 3.75E-11 2.51E-I1- 1.99E-Il 1.38E-1 I 15 4.18E+07 4.95E+08 15.68 3.77E-11 2.45E-I1 I1.89E-1 I 1.26E- 11 16 4.38E+07 5.39E+08 17.07 3.87E-11 2.79E- I1 2.26E--11 1.29E- 11 17 4.12E+07 5.80E+08 18.37 3.82E-1I 2.33E-I l 1.80E-11 1.17E-I1 18 4.65E+07 6.26E+08 19.85 3.36E-1I 2.49E-II 2.08E--11 1.32E-11 19 (Pjt.) 4.15E+07 6.68E+08 21.16 3.84E-11 2.42E-I l 1.87E-I I1!.23E-11 Future 3.42E+08 1.01 E+09 32.00 6.13E- 11 3.61E-11 2.38E-11 I1.54E-11 Future 5.05E+08 1.511E+09 48.00 6.13E-1 1 3.611E-11 2.38E-Il 1.54E-Il Future 1.89E+08 1.70E+09 54.00 6.13E-11 3.61E-11 2.38E- I I .54E- 1 Turkey Point Unit 3 Capsule X

6-24 Table 6-3 cont'd Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Intermediate Shell Course to Lower Shell Course Girth Weld Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacements Cycle Irradiation Irradiation [dpa]

Length Time Time Cycle [EFPSI [EFPS] [EFPY] 00 150 30- 450 I 3.62E+07 3.62E+07 1.15 3.69E-03 1.63E-03 9.29E-04 5.85E-04 2 2.45E+07 6.06E+07 1.92 6.133E-03 2.70E-03 1.52E-03 9.48E-04 3 2.40E+07 8.46E+07 2.68 8.39E-03 3.84E-03 2.22E-03 1.39E-03 4 2.46E+07 1.09E+08 3.46 1.07E-02 4.94E-03 2.78E-03 1.80E-03 5 2.45E+07 1.34E+08 4.24 1.36E-02 6.16E-03 3.29E-03 2.11 E-03 6 1.59E+07 1.50E+08 4.74 1.54E-02 6.91 E-03 3.60E-03 2.30E-03 7 2.90E+07 1.79E+08 5.66 1.79E-02 8.05E-03 4.24E-03 2.71 E-03 8 4.30E+07 2.22E+08 7.03 2.15E-02 9.64E-03 5.12E-03 3.27E-03 9 3.26E+07 2.54E+08 8.06 2.27E-02 1.04E-02 5.79E-03 3.75E-03 10 3.95E+07 2.94E+08 9.31 2.41E-02 1.14E-02 6.54E-03 4.24E-03 II 4.13E+07 3.35E+08 10.62 2.56E-02 1.24E-02 7.34E-03 4.75E-03 12 3.9 11E+07 3.74E+08 11.86 2.70E-02 1.34E-02 8.11 E-03 5.24E-03 13 3.95E+07 4.14E+08 13.11 2.84E-02 1.43E-02 8.83E-03 5.75E-03 14 3.93E+07 4.53E+08 14.36 2.99E-02 1.53E-02 9.61E-03 6.29E-03 15 4.18E+07 4.95E+08 15.68 3.14E-02 1.63E-02 1.04E-02 6.82E-03 16 4.38E+07 5.39E+08 17.07 3.31E-02 1.75E-02 1. 1413-02 7.38E-03 17 4.12E+07 5.80E+08 18.37 3.47E-02 1.85E-02 1.2 1E-02 7.86E-03 18 4.65E+07 6.26E+08 19.85 3.63E-02 1.97E-02 1.3 1 E-02 8.47E-03 19 (Pjt.) 4.15E+07 6.68E+08 21.16 3.79E-02 2.07E-02 1.39E-02 8.98E-03 Future 3.42E+08 1.01E+09 32.00 5.88E-02 3.30E-02 2.20E-02 1.42E-02 Future 5.05E+08 1.5 1E+09 48.00 8.98E-02 5.12E-02 3.41E-02 2.20E-02 Future 1.89E+08 1.70E+09 54.00 1.01 E-0 I 5.80E-02 3.86E-02 2.49E-02 Turkey Point Unit 3 Capsule X

6-25 Table 6-4 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV)

Within The Reactor Vessel Wall RADIUS AZIMUTHAL ANGLE (cm) 00 150 300 450 197.881 1.000 1.000 1.000 1.000 202.802 0.591 0.596 0.600 0.595 207.724 0.308 0.311 0.318 0.311 212.645 0.152 0.155 0.161 0.158 217.566 0.064 0.069 0.076 0.079 Note: Base Metal Inner Radius = 197.881 cm Base Metal 1/4T = 202.802 cm Base Metal 1/2T = 207.724 cm Base Metal 3/4T = 212.645 cm Base Metal Outer Radius = 217.566 cm Table 6-5 Relative Radial Distribution Of Iron Atom Displacements (dpa)

Within The Reactor Vessel Wall RADIUS AZIMUTHAL ANGLE (cm) 00 150 300 450 197.881 1.000 1.000 1.000 1.000 202.802- 0.678 0.696 0.692 0.691 207.724 0.430 0.451 0.452 0.453 2 12.645 0.256 0.280 0.287 0.296 217.566 0.120 - 0.145 0.164 0.187 Note: Base Metal Inner Radius = 197.881 cm Base Metal l/4T = 202.802 cm Base Metal l/2T = 207.724 cm Base Metal 3/4T = 212.645 cm Base Metal Outer Radius = 217.566 cm Turkey Point Unit 3 Capsule X

6-26 Table 6-6 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Turkey Point Unit 3 Irradiation Time Fluence (E > 1.0 MeV) Iron Displacements Capsule [EFPY] [n/cm 2] [dpa]

T 1.15 5.99E+ 18 1.01E-02 S 3.46 1.27E+19 2.17E-02 V 8.06 1.22E+19 2.OOE-02 X 19.85 2.90E+ 19 4.82E-02 Table 6-7 Calculated Surveillance Capsule Lead Factors Capsule ID And Location Status Lead Factor T (00) Withdrawn EOC 1 2.73 S (100) Withdrawn EOC 4 1.99 V (200) Withdrawn EOC 9 0.87 X (400 Yp 00) Withdrawn EOC 18 1.11 U (300) In Reactor 0.78 Y (300) In Reactor 0.78 W (400) In Reactor 0.53 Z (40-) In Reactor 0.53 Notes: (I) Capsule X was irradiated at a 400 location for Cycles I through 1I, and at a 00 location during Cycles 12 through 18, after which it was removed from service.

(2) Lead factors for capsules remaining in the reactor are based on cycle specific exposure calculations through the current operating fuel reload, i.e., Cycle 19.

Turkey Point Unit 3 Capsule X

7-1 7 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 Turkey Point Unit 3 and Unit 4 reactor vessels.

Table 7-1 Turkey Point Unit 3 and Unit 4 Reactor Vessel Surveillance Capsule Withdrawal Schedule Capsule Withdrawal EFPY(a) Fluence (n/cm2)

T3 1.15 7.39 x 10'" (b)

T4 1.17 7.08 x 1018 (b)

S3 3.46 1.72 x 10'9 (b)

S4 3.41 1.43 x 10'9 (b)

V3 8.06 1.53 x 1019 (b)

X3 19.85 2.90 x 1019 (b)

X4 35(c) 5.91 x 1019 (c)

U3 Standby -

Y3 Standby -

W3 Standby -

Z3 Standby -

U4 Standby -

V4 Standby -

Y4 Standby -

W4 Standby -

Z4 Standby -

Notes (a) Effective Full Power Years (EFPY) from plant startup.

(b) Plant Specific Evaluation (c) Capsule X4 will reach a fluence of 5.91 x 1019 at 35 EFPY (peak EOL vessel fluence).

Turkey Point Unit 3 Capsule X

8-1 8 REFERENCES

1. WCAP-7656, "FloridaPowerand Light Co. Turkey Point Unit No. 3 Reactor Vessel Radiation Sur'eillance Program," S. E. Yanichko, May 1971.

2. Code of Federal Regulations, IOCFR50, Appendix 3, FractureToughness Requirements, U.S.

Nuclear Regulatory Commission, Washington, D.C.

3. Regulatory Guide 1.99, Revision 2, May 1988, Radiation Embrittlement of Reactor Vessel Materials

4.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, Fracture Toughness Criteria fir ProtectionAgainst Failure

5. ASTM E208, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperatureof FerriticSteels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.

6. ASTM E185-82, StandardPracticefor Conducting Surveillance Tests for Light- Water Cooled Nuclear Power Reactor Vessels, E706 (IF), in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.

7. WCAP- 15092 Revision 3, "Turkey Point Units 3 and 4 WOG Reactor Vessel 60-Year Evaluaion Minigroup Heatup and Cooldown Limit Curves for Normal Operation" T. J. Laubham and J. H.

Ledger, May 2000.

8. ASTM E23-98, Standard Test Method for Notched Bar Impact Testing of Metallic Materials.

9. ASTM A370-97a, Standard Test Methods and Definitions for Mechanical Testing of Steel Products

10. ASTM E8-99, Standard Test Methods for Tension Testing of Metallic Materials

11. ASTM E21-92 (1998), Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials

12. ASTM E83-93, Standard Practice for Verification and Classification of Extensometers

13. ASTM Designation El 85-66, Surveillance Tests on Structural Materials in Nuclear Reactors

14. Regulatory Guide RG-1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

15. WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 1996.

16. WCAP- 15557, Revision 0, "Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology," August 2000.

Turkey Point Unit 3 Capsule X

8-2

17. RSICC Computer Code Collection CCC-650, "DOORS 3.1, One, Two- and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System," August 1996.

18. RSIC Data Library Collection DLC-1 85, "BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," March 1996.

19. BAW-2312, Revision 1 "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of Turkey Point Units 3 and 4 for Extended Life Through 48 Effective Full Power Years" (FTI Document No. 77-2312-01) transmitted by letter FTI-00-3072 dated 12/18/2000 to E.

Thompson.

20. Licensing Letter L-85-66, J. W. Williams to D. G. Eisenhut "Turkey Point Units 3 and 4 Proposed License Amendment, Reactor Plant Surveillance Material Program" dated 2/8/1985.

21. NRC letter dated 4/22/1985, Safety Evaluation by NRR Related to Amendment No. 112 to Facility Operating License No. DPR-31 and Amendment No. 106 to Facility Operating License No. DPR-41, D. GCMacdonald to J. W. Williams Turkey Point Unit 3 Capsule X

A-1 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS Turkey Point Unit 3 Capsule X

A-2 A.1 Neutron Dosimetry Comparisons of measured dosimetry results to both the calculated and least squares adjusted values for all surveillance capsules withdrawn from service to date at Turkey Point Unit 3 are described herein. The sensor sets from these capsules have been analyzed in accordance with the current dosimetry evaluation methodology described in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.'IA'll One of the main purposes for presenting this material is to demonstrate that the overall measurements agree with the calculated and least squares adjusted values to within +/- 20% as specified by Regulatory Guide 1.190, thus serving to validate the calculated neutron exposures previously reported in Section 6.2 of this report. This information may also be useful in the future, in particular, as least squares adjustment techniques become accepted in the regulatory environment:

A.1.1 Sensor Reaction Rate Determinations In this section, the results of the evaluations of the four neutron sensor sets withdrawn to date as a part of the Turkey Point Unit 3 Reactor Vessel Materials Surveillance Program are presented. The capsule designation, location within the reactor, and time of withdrawal of each of these dosimetry sets were as follows: I Azimuthal Withdrawal Irradiation Capsule ID Location Time -Time [EFPY1 T 00 . End of Cycle 1 1.15 S 100- End of Cycle 4 3A6 V 200 End of Cycle 9 8.06 X* 40* - 00 End of Cycle 18 19.85

• Capsule X was irradiated at a 40° location during Cycles I through II followed by irradiation at a 00 location during Cycles 12 through 18 when it was subsequently removed from service.

The azimuthal locations included in the above tabulation represent the first octant equivalent azimuthal angle of the geometric center of the respective surveillance capsules.

The passive neutron sensors included in the evaluations of Surveillance Capsules T, S, V, and X are summarized as follows:

Reaction Sensor Material Of Interest Capsule T Capsule S Capsule V Capsule X 63 Copper Cu(n,a)60Co X X X X 54 Iron Fe(n,p) 54Mn X X X X 58 Nickel Ni(np) 5t Co X N/A N/A N/A 238 Uranium-238 U(nf)137 cs X - N/A X X 23 7 Neptunium-237 Np(nf) 137 Cs X N/A X X 59 Cobalt-Aluminum* Co(n,7) 6°Co N/A X X- X The cobalt-aluminum measurements for this plant include both bare wire and cadmium-covered sensors.

Turkey Point Unit 3 Capsule X

A-3 The copper, iron, nickel, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several radial locations within the test specimen array. As a result, gradient corrections were applied to these measured reaction rates in order to index all of the sensor measurements to the radial center of the respective surveillance capsules. Since the cadmium-shielded uranium and neptunium fission monitors were accommodated within the dosimeter block centered at the radial, azimuthal, and axial center of the material test specimen array, gradient corrections were not required for the fission monitor reaction rates. Pertinent physical and nuclear characteristics of the passive neutron sensors are listed in Table A-I.

The use of passive monitors such as those listed above does not yield a direct measure of the energy dependent neutron flux at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:

"* the measured specific activity of each monitor,

"* the physical characteristics of each monitor,

"* the operating history of the reactor,

"* the energy response of each monitor, and

"* the neutron energy spectrum at the monitor location.

The radiometric counting of the neutron sensors from Capsule T was carried out at the Westinghouse Analytical Services Laboratory at the Waltz Mill Site.1 A'21 The radiometric counting of the sensors from Capsules S and V were performed by the Southwest Research Institute.[A' 3 andA-41 The radiometric counting of the sensors from Capsule X was completed at the Antech Analytical Laboratory, also located at the Waltz Mill Site. In all cases, the radiometric counting followed established ASTM procedures.

Following sample preparation and weighing, the specific activity of each sensor was determined by means of a high-resolution gamma spectrometer. For the copper, iron, nickel, and cobalt-aluminum sensors, these analyses were performed by direct counting of each of the individual samples. In the case of the uranium and neptunium fission sensors, the analyses were carried out by direct counting preceded by dissolution and chemical separation of cesium from the sensor material.

It is worthwhile noting that the majority of measured reaction rates for Capsules S and V were determined to be statistically different than similar measurement data obtained from the Westinghouse 3-loop, thermal-shield reactor plant database for 10' and 200 surveillance capsules. In addition, Reference A-5 documents that detector calibration problems existed at the laboratory that performed the Capsule S counting reported in Reference A-3. Furthermore, Reference A-4 states that the counting "data is inconclusive for computing fluence rate" for all Capsule V dosimetry measurements except for iron.

As a result, the Capsule S and V measurements were not utilized in the least squares adjustment calculation for these capsules.

The irradiation history of the reactor over the irradiation periods experienced by Capsules T, S, V, and X was based on the reported monthly power generation of Turkey Point Unit 3 from initial reactor startup through the end of the dosimetry evaluation period. For the sensor sets utilized in the surveillance capsules, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. The irradiation history applicable to Capsules T, S, V, and X is given in Table A-2.

Turkey Point Unit 3 Capsule X

A-4 Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full-power operation were determined from the following equation:

A R=

No F Y 'C,[Ii- e*i Ile- I Pr where:

R = Reaction rate averaged over the irradiation period and referenced to operation at a core power level of Pef (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.

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

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

Cj= Calculated ratio of 4(E > 1.0 MeV) during irradiation period j to the time weighted average

• (E > 1.0 MeV) over the entire irradiation period.

= Decay constant of the product isotope (1/sec).

tj = Length of irradiation period j (sec).

td = Decay time following irradiation period j (sec).

and the summation is carried out over the total number of monthly intervals comprising the irradiation period.

In the equation describing the reaction rate calculation, the ratio [Pi]/[P1 ,f] accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio C,, which was calculated for each fuel cycle using the transport methodology discussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in flux level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, C, is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those employing low leakage fuel management, the additional Cj 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.

The fuel cycle specific neutron flux values along with the computed values for C, are listed in Table A-3.

These flux values represent the cycle dependent results at the radial and azimuthal center of the respective capsules at the axial elevation of the active fuel midplane.

Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 238U measurements to account for the presence of 215U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.

Corrections were also made to the 238U and 237Np sensor reaction rates to account for gamma ray induced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the Turkey Point Unit 3 fission sensor reaction rates are summarized as follows:

Turkey Point Unit 3 Capsule X

A-5 Correction Capsule T Capsule S* Capsule V* Capsule X 2 35

- U Impurity/Pu Build-in 0.861 N/A 0.837 0.780 238 U(y,f) 0.959 N/A 0.960 0.957 Net 238U Correction 0.826 N/A 0.804 0.746 237Np(yf) 0.985 N/A 0.984 0.982

• Type I capsules (e.g., S) do not contain 218U and 2"Np

- 7 sensors ,hereas Type 11capsules (e.g, T. V, and X) contain fission monitors. As a result, the aforementioned corrections are not applicable to Type I capsules. Also recognize that most of the Capsule S and V measured reaction rates were determined to be statistically different than the corresponding data obtained from the Westinghouse 3-loop, thermal-shield plant database for 10' and 20' surveillance capsules. This is consistent with historical documentation that descnbes detector calibration problems at the Southwest Research Institute laboratory that analyzed Capsule S (see Reference A-5) and a statement in the Southwest Research Institute dosimetry analysis report for Capsule V that suggests the counting "data is inconclusive for computing fluence rate" for all measured results except for iron (see Reference A-4). Therefore. the Capsule S and V measurement results were not used in the subsequent least squares adjustment calculation in%olving these capsules.

These factors were applied in a multiplicative fashion to the decay corrected uranium and neptunium fission sensor reaction rates.

Results of the sensor reaction rate determinations for Capsules T, S, V, and X are given in Table A-4. In Table A-4, the measured specific activities, decay corrected saturated specific activities, and computed reaction rates for each sensor indexed to the radial center of the capsule are listed. The fission sensor reaction rates are listed both with and without the applied corrections for 218U impurities, plutonium build-in, and gamma ray induced fission effects.

A.1.2 Least Squares Evaluation of Sensor Sets Least squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations resulting in a Best Estimate neutron energy spectrum with associated uncertainties. Best Estimates for key exposure parameters such as ý(E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least squares methods, as applied to surveillance capsule dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrum within their respective uncertainties. For example, R, _Jo,)R (0o+/-+ go) relates a set of measured reaction rates, R,, to a single neutron spectrum, q, through the multigroup dosimeter reaction cross-section, a,,, each with an uncertainty 3. The primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the least squares evaluation of the Turkey Point Unit 3 surveillance capsule dosimetry, the FERRET code1 A-61 was employed to combine the results of the plant specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters Turkey Point Unit 3 Capsule X

A-6

(ý(E > 1.0 MeV) and dpa) along with associated uncertainties for the four in-vessel capsules withdrawn to date.

The application of the least squares methodology requires the following input:

I - The calculated neutron energy spectrum and associated uncertainties at the measurement location.

2 - The measured reaction rates and associated uncertainty for each sensor contained in the multiple foil set.

3 - The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set.

For the Turkey Point Unit 3 application, the calculated neutron spectrum was obtained from the results of plant specific neutron transport calculations described in Section 6.2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section A. 1.1.

The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section library[^A-. The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard El018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E 706 (IIB)".

The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum were input to the least squares procedure in the form of variances and covariances.

The assignment of the input uncertainties followed the guidance provided in ASTM Standard E 944, "Application of Neutron Spectrum Adjustment.Methods in Reactor Surveillance."'

The following provides a summary of the uncertainties associated with the least squares evaluation of the Turkey Point Unit 3 surveillance capsule sensor sets.

Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates inchldes components due to the basic measurement process, irradiation history corrections, and corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM National Consensus Standards for reaction rate determinations for each sensor type.

After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least squares evaluation:

Reaction Uncertainty 6-Cu(n,a) Co 5%

'4Fe(n,p)-Mn - 5%

85Ni(n,p)' 8 Co "238 5%

37 U(n,f) 1 Cs 10%

237 Np(n,f) 137Cs 10%

59 5%

Co(n,y)6°Co These uncertainties are given at the la level.

Turkey Point Unit 3 Capsule X

A-7 Dosimetry Cross-Section Uncertainties The reaction rate cross-sections used in the least squares evaluations were taken from the SNLRML library. This data library provides reaction cross-sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross-sections and uncertainties are provided in a fine multigroup structure for use in least squares adjustment applications. These cross-sections were compiled from the most recent cross-section evaluations and they have been tested with respect to their accuracy and consistency for least squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as in the fluence and energy characterization of 14 MeV neutron sources.

For sensors included in the Turkey Point Unit 3 surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.

63 Reaction Uncertainty Cu(n.o)U"Co 4.08-4.16%

"5

-Fe(np) 54Mn 58Ni(n.p) 58 3.05-3.11%

Co 4.494.56%

2 8 3 U(n.f) 137 Cs 0.54-0.64%

237 Np(n,f) 37 1 Cs 10.32-10.97%

"95Co(n,,y) 60Co 0.79-3.59%

These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in LWR irradiations.

Calculated Neutron Spectrum The neutron spectra input to the least squares adjustment procedure were obtained directly from the results of plant specific transport calculations for each surveillance capsule irradiation period and location. The spectrum for each capsule wa%input in an absolute sense (rather than as simply a relative spectral shape). Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements.

While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks and the dosimetry cross-section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:

Mg.=R2+Rj *Rg: *P.

where Rn specifies an overall fractional normalization uncertainty and the fractional uncertainties R. and Rg specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:

Pq. = [I]- 0],. + 0 e-"

where 2

H (g - g')

2 2y Turkey Point Unit 3 Capsule X

A-8 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range y (0 specifies the strength of the latter term). The value of 8 is 1.0 when g = g', and is 0.0 otherwise.

The set of parameters defining the input covariance matrix for the Turkey Point Unit 3 calculated spectra was as follows:

Flux Normalization Uncertainty (Rn) 15%

Flux Group Uncertainties (R., Rgd)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV)29%

(E < 0.68 eV)52%

Short Range Correlation (0)

(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV)0.5 (E < 0.68 eV) 0.5 Flux Group Correlation Range (y)

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV)3 (E < 0.68 eV) 2 A.1.3 Comparisons of Measurements and Calculations Results of the least squares evaluations of the dosimetry from the Turkey Point Unit 3 surveillance capsules withdrawn to date are provided in Tables A-5 and A-6. In Table A-5, measured, calculated, and best-estimate values for sensor reaction rates are given for each capsule. Also provided in this tabulation are ratios of the measured reaction rates to both the calculated and least squares adjusted reaction rates.

These ratios of M/C and M/BE illustrate the consistency of the fit of the calculated neutron energy spectra to the measured reaction rates both before and after adjustment. In Table A-6, comparison of the calculated and best estimate values of neutron flux (E > 1.0 MeV) and iron atom displacement rate are tabulated along with the BE/C ratios observed for each of the capsules.

The data comparisons provided in Tables A-5 and A-6 show that the adjustments to the calculated spectra are relatively small and well within the assigned uncertainties for the calculated spectra, measured sensor reaction rates, and dosimetry reaction cross-sections. Further, these results indicate that the use of the least squares evaluation results in a reduction in the uncertainties associated with the exposure of the surveillance capsules. From Section 6.4 of this report, it may be noted that the uncertainty associated with the unadjusted calculation of neutron fluence (E > 1.0 MeV) and iron atom displacements at the surveillance capsule locations is specified as 12% at the Ia level. From Table A-6, it is noted that the corresponding uncertainties associated with the least squares adjusted exposure parameters have been reduced to 6% for neutron flux (E > 1.0 MeV) and 7% for iron atom displacement rate. Again, the uncertainties from the least squares evaluation are at the I a level.

Further comparisons of the measurement results with calculations are given in Tables A-7 and A-8.

These comparisons are given on two levels. In Table A-7, calculations of individual threshold sensor Turkey Point Unit 3 Capsule X

A-9 reaction rates are compared directly with the corresponding measurements. These threshold reaction rate comparisons provide a good evaluation of the accuracy of the fast neutron portion of the calculated energy spectra. In Table A-8, calculations of fast neutron exposure rates in terms of 4(E > 1.0 MeV) and dpa/s are compared with the best estimate results obtained from the least squares evaluation of the capsule dosimetry results. These two levels of comparison yield consistent and similar results with all measurement-to-calculation comparisons falling well within the 20% limits specified as the acceptance criteria in Regulatory Guide 1.190.

In the case of the direct comparison of measured and calculated sensor reaction rates, the M/C comparisons for fast neutron reactions range from 0.83-1.35 for the 9 samples included in the data set.

The overall average M/C ratio for the entire set of Turkey Point Unit 3 data is 1.10 with an associated standard deviation of 14.7%.

In the comparisons of best estimate and calculated fast neutron exposure parameters, the corresponding BE/C comparisons for the capsule data sets range from 0.92-1.16 for neutron flux (E > 1.0 MeV) and from 0.91 to 1.13 for iron atom displacement rate. The overall average BE/C ratios for neutron flux (E > 1.0 MeV) and iron atom displacement rate are 1.04 with a standard deviation of 16.5% and 1.02 with a standard deviation of 15.5%, respectively.

Based on these comparisons, it is concluded that the calculated fast neutron exposures provided in Section 6.2 of this report are validated for use in the assessment of the condition of the materials comprising the beltline region of the Turkey Point Unit 3 reactor pressure vessel.

Turkey Point Unit 3 Capsule X

A-10 Table A-I Nuclear Parameters Used In The Evaluation Of Neutron Sensors Target 90% Response Fission Monitor Reaction of Atom Range Product Yield Material Interest Fraction (MEV) Half-life 071) 63 Copper CU (n,cx) 0.6917 4.9-11.8 5.271 y Iron "4Fe (n,p) 0.0585 2.1-8.3 312.3 d Nickel "5SNi (n,p) 0.6808 1.5-8.1 70.82 d 238 Uranium-238 U (n,f) 0.9996 1.2-6.7 30.07 y 6.02 23 7 Neptunium-237 Np (n,f) 1.0000 0.4-3.5 30.07 y 6.17 Cobalt-Aluminum " Co (n,y) 0.0015 non-threshold 5.271 y Notes: The 90% response range is defined such that, in the neutron spectrum characteristic of the Turkey Point Unit 3 surveillance capsules, approximately 90% of the sensor response is due to neutrons in the energy range specified with approximately 5% of the total response due to neutrons with energies below the lower limit and 5% of the total response due to neutrons with energies above the upper limit.

The counting results determined by the Southwest Research Institute for the "9Co (n,y) reactions from Capsules S and V were reported based on the weight of Co in the sample rather than the total weight of the dosimeter material. As a result, the target atom fraction used in the 59Co (n,y) analysis of Capsules S and V was set to unity.

Turkey Point Unit 3 Capsule X

A-11 Table A-2 Monthly Thermal Generation During The First Eighteen Fuel Cycles Of The Turkey Point Unit 3 Reactor (Reactor Power of 2200 MWt through October 11, 1996, and 2300 MWt thereafter)

Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 1972 11 93785 1976 1 1540335 1979 1 707 1972 12 191555 1976 2 1177919 1979 2 0 1973 1 508292 1976 3 1340004 1979 3 0 1973 2 847407 1976 4 1529615 1979 4 449651 1973 3 745477 1976 5 1277770 1979 5 1299648 1973 4 759432 1976 6 1085183 1979 6 811321 1973 5 1154965 1976 7 1629884 1979 7 1556985 1973 6 1304071 1976 8 1012048 1979 8 1329137 1973 7 1053495 1976 9 1525830 1979 9 1440462 1973 8 726401 1976 10 1614699 1979 10 1242241 1973 9 1101797 1976 11 684046 1979 11 1569305 1973 10 799316 1976 12 0 1979 12 2229 1973 11 1141315 1977 1 512809 1980 I 0 1973 12 540365 1977 2 1395687 1980 2 1033604 1974 1 1198875 1977 3 1629974 1980 3 1612394 1974 2 930609 1977 4 1244675 1980 4 1513229 1974 3 915164 1977 5 1599328 1980 5 1458855 1974 4 1422477 1977 6 1543264 1980 6 1566925 1974 5 1494789 1977 7 1387483 1980 7 1612271 1974 6 1072047 1977 8 1626729 1980 8 1613090 1974 7 1453392 1977 9 1391626 1980 9 1558594 1974 8 1550349 1977 10 1628892 1980 10 257027 1974 9 948103 1977 11 1084071 1980 !1 1192309 1974 10 138875 1977 12 0 1980 12 1285346 1974 11 0 1978 1 0 1981 1 1560412 1974 12 587671 1978 2 521921 1981 2 1464865 1975 1 1590161 1978 3 1420093 1981 3 0 1975 2 1404960 1978 4 1537183 1981 4 0 1975 3 1245352 1978 5 1449975 1981 5 0 1975 4 1457721 1978 6 1443357 1981 6 0 1975 5 1511021 1978 7 1405115 1981 7 0 1975 6 1564961 1978 8 1209429 1981 8 0 1975 7 1281697 1978 9 1492193 1981 9 0 1975 8 1565298 1978 10 1479800 1981 10 0 1975 9 1506518 1978 11 1481730 1981 11 0 1975 10 1238155 1978 12 1551960 1981 12 0 1975 11 0 1975 12 264818 Turkey Point Unit 3 Capsule X

I A-12 Table A-2 cont'd Monthly Thermal Generation During The First Eighteen Fuel Cycles Of The Turkey Point Unit 3 Reactor (Reactor Power of 2200 MWt through October 1I, 1996, and 2300 MWt thereafter)

Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 1982 1 0 1985 1 1410411 1988 1 518370 1982 2 0 1985 2 1429004 1988 2 370671 1982 3 0 1985 3 1468951 1988 3 888882 1982 4 905576 1985 4 0 1988 4 1557479 1982 5 1476098 1985 5 0 1988 5 1638874 1982 6 1538557 1985 6 -0 1988 6 1546625 1982 7 1067596 1985 7 232394 1988 7 1625539 1982 8 1237180 1985 8 1524164 1988 8 1621379 1982 9 1534227 1985 9 1575804 1988 9 1574465 1982 10 1640558 1985 10 906410 1988 10 46931 1982 11 1201773 1985 11 1226293 1988 11 0 1982 12 1618792 1985 12 1302579 1988 12 0 1983 1 1506295 1986 1 1066580 1989 1 0 1983 2 1365136 1986 2 1349846 1989 2 932196 1983 3 1615637 1986 3 212780 1989 3 1461857 1983 4 1586657 1986 4 1108434 1989 4 0 1983 5 1588610 1986 5 1568694 1989 5 0 1983 6 1577299 1986 6 1497417 1989 6 247865 1983 7 1627091 1986 7 737454 1989 7 1611920 1983 8 1631788 1986 8 975435 1989 8 1554544 1983 9 1531474 1986 9 1512236 1989 9 1457916 1983 10 40693 1986 10 1643601 1989 10 1461571 1983 11 0 1986 11 1577530 1989 11 1565712 1983 12 0 1986 '12 1389881 1989 12 1547473 1984 1 1049166 1987 1 1215136 1990 1 1638267 1984 2 943756 1987 2 1234335 1990 2 152915 1984 3 1516127 1987 3 312127 1990 3 0 1984 4 1303305 1987 4 0 " 1990 4 0 1984 5 994815 1987 5 0 1990 5 0 1984 6 1520181 1987 6 0 1990 6 805329 1984 7 1386143 1987 7 0 -1990 7 1607741 1984 8 1535986 1987 8 10 1990 8 1613090

-1984 .9 1562568 1987 9 69494 1990 '9 1559480 1984 10 1619414 1987 -'10 0 1990 10 1630270 1984 11 1577093 1987 11 0 1990 11 1580537 1984 12 632365 1987 12 127764 1990 12 609211 Turkey Point Unit 3 Capsule X

A-13 Table A-2 cont'd Monthly Thermal Generation During The First Eighteen Fuel Cycles Of The Turkey Point Unit 3 Reactor (Reactor Power of 2200 MWt through October I1, 1996, and 2300 MWt thereafter)

Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 1991 I 0 1994 I 1583340 1997 1 1581388 1991 2 0 1994 2 1390840 1997 2 1536699 1991 3 0 1994 3 1637020 1997 3 55821 1991 4 0 1994 4 133320 1997 4 651475 1991 5 0 1994 5 401478 1997 5 1707980 1991 6 0 1994 6 1520860 1997 6 1653585 1991 7 0 1994 7 1592118 1997 7 1477428 1991 8 0 1994 8 1634380 1997 8 1655793 1991 9 0 1994 9 1543300 1997 9 1653263 1991 10 1215537 1994 10 1638780 1997 10 1709889 1991 11 1483685 1994 11 1574760 1997 11 1617245 1991 12 1639202 1994 12 1375880 1997 12 1708256 1992 1 1558758 1995 1 1615460 1998 1 1707773 1992 2 1414833 1995 2 1475540 1998 2 1308539 1992 3 1422730 1995 3 1598520 1998 3 1708394 1992 4 1198455 1995 4 1471140 1998 4 1650825 1992 5 912262 1995 5 1628660 1998 5 1666603 1992 6 1384481 1995 6 1537800 1998 6 1653401 1992 7 1420554 1995 7 1634820 1998 7 1679943 1992 8 840627 1995 8 1620960 1998 8 1707957 1992 9 0 1995 9 134200 1998 9 1047259 1992 10 0 1995 10 933900 1998 10 64009 1992 11 0 1995 11 1575200 1998 11 1628653 1992 12 1184161 1995 12 1634820 1998 12 1707727 1993 1 1272201 1996 I 1631740 1999 1 1707819 1993 2 1476438 1996 2 998580 1999 2 1516114 1993 3 1557074 1996 3 1531860 1999 3 1708348 1993 4 1579380 1996 4 1574606 1999 4 1650549 1993 5 1635920 1996 5 1632246 1999 5 1651561 1993 6 1547898 1996 6 1538328 1999 6 1563425 1993 7 1595000 1996 7 1632444 1999 7 1708325 1993 8 1637020 1996 8 1576146 1999 8 1708118 1993 9 1520200 1996 9 1387012 1999 9 1602410 1993 10 1188638 1996 10 1652661 1999 10 1709015 1993 11 1583120 1996 11 1651745 1999 11 1541276 1993 12 1636580 1996 12 1704231 1999 12 1707612 Turkey Point Unit 3 Capsule X

A-14 Table A-2 cont'd Monthly Thermal Generation During The First Eighteen Fuel Cycles Of The Turkey Point Unit 3 Reactor (Reactor Power of 2200 MWt through October 11, 1996, and 2300 MWt thereafter)

Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-h) 2000 1' 1708601 2000 8 1708003 2001 3 1662463 2000 2 1432210 2000 9 1652619 2001 4 1651101 2000 3 213900 2000 10 1710533 2001 5 1667178 2000 4 1651285 2000 11 1614508 2001 6 1653125 2000 5 1700988 2000 12 1708049 2001 7 1656966 2000 6 1607884 2001 1 1708072 2001 8 1320200 2000 7 1708670 2001 2 1542794 2001 9 1557376 Turkey Point Unit 3 Capsule X

A-15 Table A-3 Calculated Cj Factors at the Surveillance Capsule Center Core Midplane Elevation Fuel O(E > 1.0 MeV) [n/cm 2-s] C, Cycle Capsule T Capsule S Capsule V Capsule X* T S V X*

I 1.66E+11 1.16E+ll 5.14E+I0 2.59E+10 1.000 1.000 1.068 0.560 2 1.13E+lI 4.88E+10 2.37E+10 0.970 1.016 0.512 3 1.19E+II 5.73E+10 3.03E+10 1.024 1.191 0.655 4 1.17E+11 4.91E+10 2.64E+10 1.008 1.021 0.571 5 5.20E+ 10 2.07E+10 1.082 0.447 6 4.81E+10 1.99E+10 1.000 0.430 7 4.55E+10 2.27E+10 0.946 0.491 8 4.56E+10 2.20E+10 0.949 0.476 9 3.90E+10 2.35E+10 0.812 0.509 10 2.05E+ 10 0.443 11 1.98E+10 0.427 12 7.16E+10 1 548 13 7.02E+ 10 1.518 14 7.46E+10 1.612 15 7.55E+ 10 1.632 16 7.71E+10 1.666 17 7.62E+ 10 1.646 18 6.64E+ 10 1.435 Average 1.66E+11 I1.16E+11 4.81E+10 4.63E+10 1.000 1.000 1.000 1.000

• Note: Cj factors based on the ratio of the cycle specific fast (E > 1.0 MeV) neutron flux divided by the average flux over the total irradiation period were deemed unsuitable for Capsule X since reaction rates did not vary by constant values as a function of azimuthal position for this capsule. To a large extent, this was due to moving Capsule X from a 40' to 0' location following the eleventh fuel cycle. As a result of this observation, the Cj terms that were utilized in the final Capsule X analysis were based on the individual reaction rates determined from the synthesized transport calculations. The final Cj terms for Capsule X, which are based on individual reaction rates, are reported on the next page of this table.

Turkey Point Unit 3 Capsule X

A-16 Table A-3 cont'd Calculated Cj Factors at the Surveillance Capsule Center Core Midplane Elevation (Capsule X only)

Fuel Capsule X Reaction Rates [rps/atom]

6 TCu 238U (n,f) 237Np (n,f) 59Co 59 (n,y) Cd Cycle (n,cx) -"Fe (n,p) (ny) Co I 2.27E-17 2.18E-15 9.55E-15 6.60E-14 9.57E-13 4.88E-13 2 2.10E-17 2.01E-15 8.75E-15 6.03E-14 8.76E-13 4.46E-13 3 2.65E-17 2.55E-15 1.12E-14 7.71E-14 1.12E-12 5.71E-13 4 2.34E-17 2.23E-15 9.75E-15 6.72E-14 9.71E-13 4.95E-13 5 1.93E-17 1.79E-15 7.68E-15 5.24E-14 7.55E-13 3.84E-13 6 1.86E-17 1.72E-15 7.40E-15 5.05E-14 7.27E-13 3.70E-13 7 2.09E-17 1.95E-15 8.42E-15 5.76E-14 8.29E-13 4.22E-13 8 2.03E-17 1.90E-15 8.18E-15 5.59E-14 8.04E-13 4.101E-13 9 2.15E-17 2.02E-15 8.73E-15 5.97E-14 8.55E-13 4.36E-13 10 1.91E-17 1.77E-15 7.62E-15 5.19E-14 7.40E-13 3.78E-13 11 1.84E-17 1.71E-15 7.35E-15 5.011E-14 .7.18E-13 3.66E-13 12 4.37E-17 5.03E-15 2.50E-14 1.92E-13 3.24E-12 1.68E-12 13 4.30E-17 4.94E-15 2.45E-14 1.89E-13 3.18E-12 1.65E-12 14 4.58E-17 5.25E-15 2.60E-14 2.003E-13 3.36E-12 1.74E-12

'15 4.59E-17 5.30E-15 2.63E-14 2.03E-13 3.42E-12 1.77E-12 16 4.69E- 17 5.41E-15 2.69E- 14 2.07E- 13 3.49E- 12 1.811E-12 17 4.66E- 17 5.36E- 15 2.66E- 14 2.04E- 13 3.44E- 12 1.78E-12 18 4.08E- 17 4.67E- 15 2.32E-14 1.78E-13 3.00E- 12 1.56E-12 Average 3.19E-17 3.44E-15 1.64E-14 1.22E-13 1.99E-12 1.0313-12 Fuel Capsule X j 6 2 8 2 7 59Co (n,-y) 59 Co (n,-y) Cd Cycle 3Cu (n,x) '4Fe (n,p) 1 U (n,f) 3 Np (n,f)

I 0.711 0.634 0.582 0.539 0.482 0.476 2 0.659 0.583 0.533 0.493 0.441 0.435 3 0.830 0.741 0.681 0.630 0.564 0.557 4 0.732 0.649 0.594 0.549 0.489 0.483 5 0.605 0.520 0.468 0.428 0.380 0.375 6 0.581 0.500 0.451 0.413 0.366 0.361 7 0.653 - 0.567 0.513 0.471 0.418 0.412 8 0.636 0.552 0.499 0.457 0.405 0.399 9 0.674 0.587 0.532 0.488 0.431 0.425 10 0.598 0.515 0.464 0.424 0.373 0.369 11 0.576 0.497 0.448 0.410 0.362 0.357 12 1.367 1.462 1.523 1.571 1.634 1.639 13 1.346 1.437 1.495 1.540 1.599 1.605 14 1.433 1.527 1.587 1.634 1.694 1.701 15 1.438 1.540 1.605 1.656 1.721 1.728 16 1.469 1.572 1.638 1.690 1.759 1.765 17 1.458 1.558 1.620 1.670 1.732 1.739 18 1.277 1.359 1.412 1.455 1.512 1.517 Average 1.000 1.000 1.000 1.000 1.000 1.000 Turkey Point Unit 3 Capsule X

A-17 Table A-4 Measured Sensor Activities And Reaction Rates Surveillance Capsule T Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate Reaction Location (dps/2) (dps/) (rps/atom)

(dps/&)

63 Cu (n,a) 6°Co Top 9.52E+04 7.67E+05 7.99E+05 1.22E-16 Bottom 9.19E+04 7.40E+05 7.71 E+05 1.18E-16 Average 1.20E-16 54Fe 54Mn (n,p) H- 17 Charpy 3.34E+06 8.78E+06 9.16E+06 1.45E- 14 H-23 Charpy 3.23E+06 8.49E+06 8.86E+06 1.41E-14 R-63 Charpy 2.90E+06 7.62E+06 7.96E+06 1.26E-14 W-18 Charpy 3.5 1E+06 9.22E+06 7.93E+06 1.26E-14 W-24 Charpy 3.7 1 E+06 9.75E+06 8.38E+06 1.33E-14 P-58 Charpy 3.51 E+06 9.22E+06 7.93E+06 1.26E-14 Average 1.33E-14 58 Ni (np) 58Co Middle 2.56E+07 1.18E+08 1.23E+08 1.76E-14 Average 1.76E-14 238 U (nf) 13 7 Cs (Cd) Middle 2.92E+05 1.19E+07 1.19E+07 238U (n,) 7.80E-14

'"7Cs (Cd) Including 2v)Pu, and yfission corrections:

235U, 6.44E-14 237 Np (n,0 137Cs (Cd) 237 Middle 2.16E+06 8.78E+07 8.78E+07 5.60E- 13 Np (n,f) 137Cs (Cd)

Including 7,fission correction: 5.52E-13 Notes: 1) Measured specific activities are indexed to a counting date of February 3, 1975.

2) The average 238U (n,f) reaction rate of 6.44E-14 includes a correction factor of 0.861 to account for plutonium build-in and an additional factor of 0.959 to account for photo-fission effects in the sensor.

3) The average 237 Np (n,f) reaction rate of 5.52E-13 includes a correction factor of 0.985 to account for photo-fission effects in the sensor.

Turkey Point Unit 3 Capsule X

A-18 Table A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule S Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate Reaction Location (dpslg) (dps65) (dps71E) (rps/atom) 6 3Cu (n,a) 6°Co Top 2.25E+05 6.85E+05 7.13E+05 1.09E- 16 Bottom 2.24E+05 6.82E+05 7.1013+05 1.08E-16 Average 1.09E-16 54 1.05E-14 Fe (n,p) 54Mn P-9 Charpy 4.76E+06 6.36E+06 6.64E+06 P-7 Charpy 4.64E+06 6.20E+06 6.47E+06 1.03E- 14 R-6 Charpy 4.39E+06 5.87E+06 6.12E+06 9.71E-15 P-5 Charpy 4.84E+06 6.47E+06 6.75E+06 1.07E-14 P-1 Charpy 4.50E+06 6.01E+06 6.28E+06 9.95E-15 S-9 Charpy 6.13E+06, 8.199E+06 7.04E+06 1.12E-14 S-7 Charpy 5.41E+06 7.23E+06 6.22E+06 9.86E- 15 R-8 Charpy 4.91E+06 6.56E+06 5.64E+06 8.94E- 15 S-5 Charpy 5.64E+06 7.54E+06 6.48E+06 1.03E- 14 S-I Charpy 5.47E+06 S7.31 E+06 6.29E+06 9.96E-15 Average 1.01E-14 59 6.68E+07 6.54E- 15 Co (ny) 6 0Co Top 2.34E+07 7.12E+07 Middle 2.07E+07 6.30E+07 5.9 1E+07 5.78E- 15 Bottom 2.01 E+07 6.12E+07 5.74E+07 5.62E- 15 Average 5.98E-15 59

" Co (n,y) 6°Co (Cd) Top 1.1IE+07 3.38E+07 2.93E+07 2.86E-15 Bottom 9.40E1+06 2.86E+07 2.48E+07 2.43E- 15 Average 2.64E-15 Notes: 1) Measured specific activities are indexed to a counting date of November 23, 1977.

2) Measured reaction rates for iron as originally reported in Reference A-3 were subsequently identified in Reference A-5 as being biased high by 15% due to detector calibration issues. As a result, the Capsule S measured iron reaction rates listed above have been reduced to 85% of the original Reference A-3 values.

Turkey Point Unit 3 Capsule X

A-19 Table A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule V Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate Reaction Location (dpslg) (dOps/g) (dps/g) (rps/atom) 63 Cu (n,x) 6°Co Top 3.80E+04 7.80E+04 8.12E+04 1.24E- 17 Bottom 3.81 E+04 7.82E+04 8.14E+04 1.24E- 17 Average 1.24E-17 "54Fe (n,p) 54 Mn R-48 Charpy 1.58E+06 2.55E+06 2.66E+06 4.22E-15 R-42 Charpy 1.62E+06 2.62E+06 2.73E+06 4.33E-15 H-2 Charpy 1.47E+06 2.37E+06 2.48E+06 3.93E- 15 S-58 Charpy 1.90E+06 3.07E+06 2.65E+06 4.19E-15 S-52 Charpy 1.87E+06 3.02E+06 2.60E+06 4.13E-15 W-2 Charpy 1.75E+06 2.83E+06 2.44E+06 3.86E-15 Average 4.11E-15 238 U (n,f) "'Cs (Cd) Middle 2.52E+05 2 38 1.64E+06 1.64E+06 1.08E-14 U (n.f) '"7Cs (Cd) 23 5U, 239 Including pu, and y,fission corrections: 8.64E-15 237 Np (n,f) 13Cs (Cd) 237 Middle 1.23E+05 7.99E+05 7.99E+05 5.IOE-15 Np (n,f) 137Cs (Cd) Including yfission correction: 5.02E-15 59 Co (n,y) ')Co Top 3.11 E+06 6.38E+06 6.11 E+06 5.98E-16 Middle 2.82E+06 5.79E+06 5.54E+06 5.42E- 16 Bottom 3.01 E+06 6.18E+06 5.9 1E+06 5.79E-16 Average 5.73E-16 "59Co (n,y) 6°Co (Cd) Top 1.53E+06 3.14E+06 2.77E+06 2.71E-16 Middle 3.17E+04 6.5 1E+04 5.74E+04 5.62E- 18 Bottom 1.52E+06 3.12E+06 2.75E+06 2.69E- 16 Average 1.82E-16 Notes: I) Measured specific activities are indexed to a counting date of March 30, 1985.

2) The average 238 U (n,f) reaction rate of 8.64E-15 includes a correction factor of 0.837 to account for plutonium build-in and an additional factor of 0.960 to account for photo-fission effects in the sensor.

3) The average 237Np (n,f) reaction rate of 5.02E-15 includes a correction factor of 0.984 to account for photo-fission effects in the sensor.

Turkey Point Unit 3 Capsule X

A-20 Table A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule X Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate Reaction Location (dps/gt (d ps/ ) (dps/g) (rps/atom) 63 60 Cu (n,() Co Top 2.28E+05 2.50E+05 2.60E+05 3.97E- 17 Unspecified 1.58E+05 1.73E+05 1.80E+05 2.75E-17 Average 3.36E-17 54 Fe (n,p) 54Mn R-56 Charpy 1.72E+06 2.08E+06 2.17E+06 3.44E-15 R-50 Charpy 1.46E+06 1.77E+06 1.84E+06 2.92E-15 S-66 Charpy 2.03E+06 2.46E+06 2.11 E+06 3.35E-15 S-60 Charpy 1.77E+06 2.14E+06 1.84E+06 2.92E-15 Average 3.16E-15 23 RU (n,f) ' 37Cs (Cd) Middle 9.94E+05 2.77E+06 2.77E+06 1.82E-14 238 U (n,f) 137Cs (Cd) pu, and y,fission correctionis:

2 23 9 Inclucling 35U, 1.36E-14 237 Np (n,f) 13 7Cs (Cd) Middle 7.07E+06 1.96E+07 1.96E+07 1.25E-13 237 Np (n,f) 137Cs (Cd) Including yfission correction: 1.23E-13 59Co (n,y) 6°Co 3.58E+07 2.34E-I 2 Top 4.0 1E+07 3.8 1E+07 Middle 3.04E+07 2.89E+07 2.72E+07 1.77E-12 Middle 2.39E+07 2.27E+07 2.14E+07 1.39E-12 Bottom 2.22E+07 2.1 IE+07 1.98E+07 1.29E- 12 Average 1.70E-12 "59Co (n,y) 6°Co (Cd) Top 1.78E+07 1.69E+07 1.47E+07 9.56E- 13 Bottom 1.05E+07 9.95E+06 8.65E+06 5.64E-13 Average 7.60E-13 Notes: I) Measured specific activities are indexed to'a counting date of May 1, 2002.

2) The average 23SU (n,f) reaction rate of 1.36E-14 includes a correction factor of 0.780 to account for plutonium build-in and an additional factor of 0.957 to account for photo-fission effects in the sensor.

3) The average 237Np (n,f) reaction rate of 1.23E- 13 includes a correction factor of 0.982 to account for photo-fission effects in the sensor.

"Turkey Point Unit 3 Capsule X

A-21 Table A-5 Comparison of Measured, Calculated, and Best Estimate Reaction Rates At The Surveillance Capsule Center CapsuleT Reaction Rate [rus/atomi Best Reaction Measured Calculated Estimate M/C M/BE 6 3Cu(n,a)60Co 1.20E-16 8.91E-17 1.15E-16 1.35 1.04 54Fe(n,p) 54 Mn 1.33E-14 1.IOE-14 1.33E-14 1.21 1.00 5 8Ni(n,p),.Co 1.76E-14 1.52E-14 1.81E-14 1.16 0.97 238U(n,f)137 Cs (Cd) 6.44E-14 5.71E-14 6.67E-14 1.13 0.97 237Np(n,f)137 Cs (Cd) 5.52E-13 4.52E-13 5.33E-13 1.22 1.04 Capsule S*

Reaction Rate [rps/atom]

Best Reaction Measured Calculated Estimate M/C M/BE 6ACu(n,a)6°Co Rejected 6.51 E- 17 N/A NIA N/A 54Fe(np)14 Mn Rejected 7.84E- 15 N/A N/A N/A 59Co(n,y)6°Co Rejected 5.68E- 12 N/A NIA NIA 59Co(n,y)6Co (Cd) Rejected 2.94E- 12 N/A N/A N/A

• Notes: I) Measured reaction rates for Capsule S were rejected since they were incongruent with analogous results for similar Westinghouse plant designs. Furthermore, Reference A-5 reported that detector calibration problems existed at the laboratory that performed the Capsule S counting analysis.

2) The Capsule S calculated results reported above for the individual reaction rates were taken from the synthesized transport calculations at the core midplane after the fourth fuel cycle.

Turkey Point Unit 3 Capsule X

A-22 Table A-5 cont'd Comparison of Measured, Calculated, and Best Estimate Reaction Rates At The Surveillance Capsule Center Capsule V*

Reaction Rate Irws/atoml Best "Reaction Measured Calculated Estimate M/C M/BE 63Cu(n,a) 60 Co Rejected 3.15E-17 N/A N/A N/A 54Fe(n,p)5 4Mn Rejected 3.16E-15 N/A N/A N/A 238U(n,f)a 37 cs (Cd) Rejected 1.42E-14 N/A N/A N/A 237Np(n,f) 137 Cs (Cd) Rejected 1.OOE-13 N/A N/A N/A 59Co(n,7)60CO Rejected 1.49E-12 N/A -N/A N/A 5 ?Co(n,y)6°Co (Cd) Rejected 7.61E-13 N/A N/A N/A

• Notes: 1) Measured reaction rates for Capsule V were rejected since they were incongruent with analogous results for similar Westinghouse plant designs. Furthermore, Table X of Reference A-4 indicates that: "Data inconclusive for computing fluence rate" for all measured counting results except for iron from this capsule.

2) The Capsule V calculated results reported above for the individual reaction rates were taken from the synthesized transport calculations at the core midplane after the ninth fuel cycle.

Capsule X Reaction Rate [res/atom]

Best Reaction Measured Calculated Estimate M/C M/BE 63 Cu(n,a) 60 Co 3.36E-17 3.19E-17 3.25E-17 1.05 1.03 54Fe(np) 54 Mn 3.16E-15 3.41E-15 3.22E-15 0.93 0.98 8

3 U(n,f)13TCs (Cd) 1.36E-14 1.64E-14 1.51E-14 0.83 0.90 237Np(n,f)137 Cs (Cd) 1.23E-13 1.23E-13 1.17E-13 1.00 1.05 59 Co(n,Y) 60 Co 1.70E-12 1.96E-12 1.70E-12 0.87 1.00 59Co(nY) 60 Co (Cd) 7.60E-13 9.85E-13 7.67E-13 0.77 0.99 Turkey Point Unit 3 Capsule X

A-23 Table A-6 Comparison of Calculated and Best Estimate Exposure Rates At The Surveillance Capsule Center O(E > 1.0 MeV) [n/cm 2-s]

Best Uncertainty Capsule ID Calculated Estimate (1c) BE/C T 1.66E+11 1.92E+ 1 6% 1.16 S 1.16E+11 N/A N/A N/A V 4.81E+10 N/A N/A N/A X 4.63E+ 10 4.24E+ 10 6% 0.92 Notes: 1) Best estimate results are not reported for Capsules S and V since all measured reaction rates were rejected.

2) Calculated results are based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period.

Iron Atom Displacement Rate [dpa/s]

Best Uncertainty Capsule ID Calculated Estimate (1() BE/C T 2.81E-10 3.17E-10 7% 1.13 S 1.99E-10 N/A N/A N/A V 7.87E- 11 N/A N/A N/A X 7.70E- 11 6.99E- I11 7% 0.91 Notes: 1) Best estimate results are not reported for Capsules S and V since all measured reaction rates were rejected.

2) Calculated results are based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period.

Turkey Point Unit 3 Capsule X

A-24 Table A-7 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions M/C Ratio Reaction Capsule T Capsule S Capsule V Capsule X 63Cu(n,ca)6°Co 1.35 N/A N/A 1.05 54Fe(n,p)5 4 Mn 1.21 N/A N/A 0.93 8

5 58

" Ni(n,p) Co 1.16 N/A N/A 238U(n,p) 13 7Cs (Cd) 1.13 237Np(n,'1 3 7 N/A N/A 0.83 Cs (Cd) 1.22 N/A N/A 1.00 Average 1.21 N/A N/A 0.95

% Standard Deviation 6.9 N/A N/A 10.2 Notes: 1) All measured reaction rates for Capsules S and V were rejected.

2) The overall average M/C ratio for the set of 9 sensor measurements is 1.10 with an associated standard deviation of 14.7%.

Table A-8 Comparison of Best Estimate/Calculated (BE/C) Exposure Rate Ratios BE/C Ratio Capsule ID 4(E > 1.0 MeV) dpa/s T 1.16 1.13 S N/A N/A V N/A N/A X 0.92 0.91 Average 1.04 1.02

% Standard Deviation 16.5 15.5 Note: Best estimate results were not determined for Capsules S and V since all measured reaction rates were rejected.

Turkey Point Unit 3 Capsule X

A-25 Appendix A References A-1. Regulatory Guide RG-1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

A-2. WCAP-863 1, Revision 0, "Analysis of Capsule T from the Florida Power and Light Company Turkey Point Unit No. 3 Reactor Vessel Radiation Surveillance Program," December 1975.

A-3. E. B. Norris, "Reactor Vessel Material Surveillance Program, Capsule S - Turkey Point Unit No. 3, Capsule S - Turkey Point Unit No. 4," Final Report SwRI Project No. 02-5131 and SwRI Project No. 02-5380, Southwest Research Institute, May 1979.

A-4. P. K. Nair and E. B. Norris, "Reactor Vessel Material Surveillance Program for Turkey Point Unit No. 3: Analysis of Capsule V," Final Report SwRI Project No. 06-8575, Southwest Research Institute, August 1986.

A-5. WCAP-14044, Revision 0, "Westinghouse Surveillance Capsule Neutron Fluence Reevaluation,"

April 1994.

A-6. A. Schmittroth, FERRET Data Analsis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

A-7. RSIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section Compendium", July 1994.

Turkey Point Unit 3 Capsule X

B-1 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS Turkey Point Unit 3 Capsule X

B-2 0

-j 0 1 2 3 4 5 6 Time (1) 0 F S62, O 1 2 3 4 5 6 Time (1)

S60, 25°F Turkey Point Unit 3 Capsule X

B-3 0

0 1 2 3 4 5 6 Tine (1)

S63, 50-F 0

-. 1 0 1 2 3 5 6 Time (1)

S65, 90°F Turkey Point Unit 3 Capsule X

B-4 5000 4000' 0

0 12 3 4 56 Time (1)

-000 S66, 1300 F 5000 3000 0

0 1 2 3 4 5 6 Time (1)

S64, 150°F Turkey Point Unit 3 Capsule X

B-5 600c 4000 3000 0

2000 1000 0 1 2 3 4 5 6 Time (1)

S61, 160°F 0

-J 0 1 2 3 4 5 6 Time (1)

S59, 180°F Turkey Point Unit 3 Capsule X

B-6 5000" 4000 3000.

o 2000{

1000 0

0 1 2 3 4 5 6 Time (1)

R51, 60°F 5000 1 40001 3000+

2000 10001 0

0 1 2 3 4 5 6 Time (1)

R50, 100°F Turkey Point Unit 3 Capsule X

B-7 5000 4000 3000 2000 01 2 3 4 5 6 Time (1)

R49, 1500 F 5000.

4000 S3000-0 2 3 4 56 Time (1)

R55, 200 0 F Turkey Point Unit 3 Capsule X

B-8 C

C

-J 0 1 2 3 4 5 6 "Time (1)

R54, 225°F C

C

-J 0 2 3 5 6 Time (1)

R56, 2500 F Turkey Point Unit 3 Capsule X

B-9 n

a 0..

0 2 3 4 5 6 Tsme (1)

R52, 275°F

• 3000 2000 1000 0 2 3 4 5 6 Time (1)

R53, 3253F Turkey Point Unit 3 Capsule X

B-10 5000 4000 o

C 3000k

-J 20001 1000.

0 0 2 3 4 6 Time (1)

W12, 80 0 F 5000 4000

"* 30001 2000 10001 0

0 1 2 3 4 5 6 Time (1)

W13, 20 0 F Turkey Point Unit 3 Capsule X

B-11 03

-J 0 1 2 3 4 S 6 Time (1)

WIO, 125-F 5000k 4000+

k 3000+

0 0 1 2 3 4 5 6 Time (1)

W14, 150°F Turkey Point Unit 3 Capsule X

B-12 5000.

4000' V 3000" 2000.

1000.

0 1 2 3 4 5 6 "Time (1)

W1l, 175°F V*

o 0 2 3 6 "Time (1)

W16, 225°F Turkey Point Unit 3 Capsule X

B-13 5000 4000 D3000 0

-j 2000 1000 0.

0 1 2 3 4 5 6 Time (1)

W15, 2500 F 5000 4000 0

2000, 1000, 0,

0 1 2 3 4 56 Time (1)

W9, 3250 F Turkey Point Unit 3 Capsule X

B-14 5000 4000 3000 0

-J 2000 1000 0 r..

0 1 2 3 4 5 6 Time (1)

H9, -30°F 5000 4000

~3000 2000 1000 0

0 1 2 3 4 6 Time (1)

H12,-30°F Turkey Point Unit 3 Capsule X

B-15 5000 4000

~3000 3o0 2000 1000 0

0 1 2 3 4 5 6 Time (1)

H11, -10F S40001 S3000 2000 1000 0 1 2 3 4 5 6 Time (1)

H10, 20°F Turkey Point Unit 3 Capsule X

B-16 63 0

-J 0 1 2 3 4 5 6 Tme (1)

H14, 25°F

• 0 63 C

-J 0 2 3 4 5 6 Time (1)

H16, 75°F Turkey Point Unit 3 Capsule X

B-17 5000 4000 S3000

-J 2000 1000 0

0 1 2 3 4 5 6 Time (1)

H15, 100-F 5000 4000

• ' 3000 2000 1000' 0 1 2 3 4 5 6 Time (1)

H13, 1300 F Turkey Point Unit 3 Capsule X