ML023460493

Analysis of Capsule U from American Electric Power Company D.C. Cook Unit 1 Reactor Vessel Radiation Surveillance Program

ML023460493
Person / Time
Site:	Cook
Issue date:	12/31/2002
From:	Gresham J Westinghouse
To:	Office of Nuclear Reactor Regulation
References
WCAP-12483, Rev 1
Download: ML023460493 (94)

Text

Westinghouse Non-Proprietary Class 3 WCAP-12483 December 2002 Revision 1 Analysis of Capsule U from the American Electric Power Company D. C. Cook Unit 1 Reactor Vessel Radiation Surveillance Program k Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-12483, Revision 1 Analysis of Capsule U from the American Electric Power Company D. C. Cook Unit I Reactor Vessel Radiation Surveillance Program J. H. Ledger E.T. Hayes December 2002 Approved: PA)

J. A. Gresham, Manager L.V'-L/

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

©2002Westinghouse Electric Company LLC All Rights Reserved Cook Unit I Capsule U

TABLE OF CONTENTS LIST OF TABLES ........................................................................................................................................ iv LIST OF FIGURES ..................................................................................................................................... vii PREFACE ......................................................................................................................................... viii RECORD OF REVISION ......................................................................................................................... viii EXECUTIVE SUMM ARY (OR) A BSTRACT ....................................................................................... ix I SUM M ARY OF RESULTS .......................................................................................................... 1-1 2 INTRODUCTION ........................................................................................................................ 2-1 3 BACKGROUN D ........................................................................................................................ 3-1 4 D ESCRI rTION OF PROGRAM ............................................................................................. 4-1 5 TESTING OF SPECIMENS FROM CAPSULE U ................................................................ 5-1 5.1 OVERVIEW .................................................................................................................... 5-1 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS .......................................................... 5-3 5.3 TEN SILE TEST RESULTS ............................................................................................ 5-4 5.4 COM PACT TENSION TESTS ....................................................................................... 5-4 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY ....................................................... 6-1

6.1 INTRODUCTION

........................................................................................................ 6-1 6.2 D ISCRETE ORD INATES ANALYSIS ........................................................................... 6-2 6.3 NEUTRON DOSIM ETRY .............................................................................................. 6-4 6.4 PROJECTIONS OF REACTOR VESSEL EXPOSURE .............................................. 6-13 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE ........................................................ 7-1 8 REFERENCES ............................................................................................................................. 8-1 APPENDIX A CORE POWER DISTRIBUTIONS USED IN THE TRANSPORT CALCULATIONS FOR D.C. COOK UNIT I .................................................... A-0 Cook Unit I Capsule U

iv LIST OF TABLES Table 4-1 Chemical Composition and Heat Treatment of the D. C. Cook Unit 1 Reactor Vessel Surveillance M aterials ............................................................................ 4-2 Table 4-2 Heat Treatment of the D. C. Cook Unit 1 Reactor Vessel Surveillance Materials .......... 4-3 Table 5-1 Charpy V-Notch Data for the D. C. Cook Unit 1 Intermediate Shell Plate B4406-3 Irradiated at 550TF Fluence 1.837 x 10'9 n/cm2 (E > 1.0 MeV) ...................................... 5-5 Table 5-2 Charpy V-notch Data for the D. C. Cook Unit 1 Intermediate Shell Plate B4406-3 Irradiated at 550T Fluence 1.837 x 1019 n/em2 (E > 1.0 MeV) ...................................... 5-6 Table 5-3 Charpy V-notch Data for the D. C. Cook Unit 1 ASTM Correlation Monitor Material Irradiated at 550TF Fluence 1.837 x 10'9 n/cr 2 (E> 1.0 MeV) ....................................... 5-7 Table 5-4 Instrumented Charpy Impact Test Results for the D. C. Cook Unit 1 Reactor Vessel Shell Plate B 4406-3 ................................................................................................................ 5-8 Table 5-5 Instrumented Charpy Impact Test Results for the D. C. Cook Unit 1 Reactor Vessel Weld Metal and HAZ Metal ............. ...................................................... 5-9 Table 5-6 Instrumented Charpy Impact Test Results for the D. C. Cook Unit 1 ASTM Correlation Monitor Material ........................................................................................................... 5-10 Table 5-7 The Effect of 550TF Irradiation at 1.837 x 1019 (E>1.0 MeV) on the Notch Toughness Properties of the D. C. Cook Unit I Reactor Vessel Surveillance Capsule Materials.. 5-11 Table 5-8 Comparison of the D. C. Cook Unit 1 Reactor Vessel Surveillance Capsule Charpy Impact Test Results with Regulatory Guide 1.99 Revision 2 Predictions .................... 5-12 Table 5-9 Tensile Properties for D. C. Cook Unit 1 Reactor Vessel Material Irradiated to 1.837 x 1019 n/cm 2 (E> 1.0M eV) ....:............................................................................. 5-13 Table 5-10 D. C. Cook Unit 1 Reactor Vessel Beltline Region Material Properties .......... 5-14 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures Rates at the Surveillance Capsule Center ......................................................................................... 6-16 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 Relative Radial Distribution of Neutron Fluence (E> 1.0 MeV) within the Reactor Vessel Wall .................................................................................................................... 6-22 Cook Unit I Capsule U

v LIST OF TABLES (Cont.)

Table 6-4 Relative Radial Distribution of Iron Atom Displacements (dpa) within the Reactor Vessel Wall .................................................................................................................... 6-22 Table 6-5 Nuclear Parameters used in the Evaluation of Neutron Sensors ................................... 6-23 Table 6-6 Monthly Thermal Generation During the First Eight Fuel Cycles of the D. C. Cook Unit 1 Reactor (Reactor Power of 3250 MWt) ............................................................. 6-24 Table 6-7 Calculated 4(E > 1.0 MeV) and Cj Factors at the Surveillance Capsule Center core M idplane Elevation ............................................................................................... 6-26 Table 6-8 Measured Sensor Activities and Reaction Rates

- Surveillance Capsule T ............................................................................... 6-27

- Surveillance Capsule X .............................................................................. 6-28

- Surveillance Capsule Y ............................................................................... 6-29

- Surveillance Capsule U .............................................................................. 6-30 Table 6-9 Comparison of Measured, Calculated and Best Estimate Reaction Rates at the Surveillance Capsule Center ............................................................................... 6-31 Table 6-10 Comparison of Calculated and Best Estimate Integrated Neutron Exposure Rates at the Surveillance Capsule Center ..................................................... 6-33 Table 6-11 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions ............................................................. 6-34 Table 6-12 Comparison of Best Estimate/Calculated (BE/C) Exposure Rate Ratios .......................... 6-34 Table 6-13 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from D . C . Cook Unit 1.......................................................................................................... 6-34 Table 6-14 Calculated Maximum Fast Neutron Exposure of the D.C. Cook Unit 1 Reactor Pressure Vessel at the Clad/Base Metal Interface ............................................ 6-35 Table 6-15 Calculated Surveillance Capsule Lead Factors ............................................................. 6-36 Table 7-1 D. C. Cook Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule ............. 7-1 Cook Unit I Capsule U

vii LIST OF FIGURES Figure 4-1 Original Arrangement of Surveillance Capsules in the D. C. Cook Unit 1 Reactor V essel ................................................................................................................. 4-4 Figure 4-2 Capsule U Diagram Showing the Location of Specimens, Thermal Monitors, and D osimeters ..................................................................................................................... 4-5 Figure 5-1 Charpy V-Notch Impact Data for D. C. Cook Unit 1 Reactor Vessel Shell Plate B4406-3 (Transverse Orientation) ................................................................................ 5-15 Figure 5-2 Charpy V-Notch Impact Data for D. C. Cook Unit 1 Reactor Vessel Shell Plate B4406-3 (Longitudinal Orientation) ............................................................................. 5-16 Figure 5-3 Charpy V-Notch Impact Data for D. C. Cook Unit 1 Reactor Vessel Weld Metal ........ 5-17 Figure 5-4 Charpy V-Notch Impact Data for D. C. Cook Unit 1 Reactor Vessel Heat-Affected-Zone Material ........................................................................................ 5-18 Figure 5-5 Charpy V-Notch Impact Data for D. C. Cook Unit 1 ASTM Correlation Material ...... 5-19 Figure 5-6 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 1 Reactor Vessel Shell Plate B4406-3 (Longitudinal Orientation) ................................................ 5-20 Figure 5-7 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 1 Reactor Vessel Shell Plate B4406-3 (Transverse Orientation) ................................................... 5-21 Figure 5-8 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 1 Reactor Weld M etal .................................................................................................................... 5-22 Figure 5-9 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 1 Reactor Vessel Weld HAZ M etal ........................................................................................................... 5-23 Figure 5-10 Charpy Impact Specimen Fracture Surfaces for ASTM Correlation Material .............. 5-24 Figure 5-11 Tensile Properties for D. C. Cook Unit 1 Reactor Vessel Shell Plate B4406-3 (Longitudinal Orientation) ............................................................................................ 5-25 Figure 5-12 Tensile Properties for D. C. Cook Unit 1 Reactor Vessel Weld Metal .......................... 5-26 Figure 5-13 Fractured Tensile Specimens for D. C. Cook Unit 1 Reactor Vessel Shell Plate B4406-3 (Longitudinal Orientation) ............................................................................. 5-27 Figure 5-14 Fractured Tensile Specimens from D. C. Cook Unit 1 Reactor Vessel Surveillance W eld M etal .................................................................................................................... 5-28 Figure 5-15 Typical Stress-Strain Curve for Tension Specimens ..................................................... 5-29 Cook Unit I Capsule U

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

Reviewer:

Sections 1 through 5, 7, 8, Appendices A and B T. J. Laubham Section 6 S. L. Anderson RECORD OF REVISION Revision 1: In addition to text and font changes, the following was changed from Revision 0:

This Revision incorporates the analyses for 40 and 60 years. Revised Section 6.0 for updated fluence methodology (Per Regulatory Guide 1.190) and to include the calculated fluence projections. Removed Figure 6-1 "Plan View of a Reactor Vessel Surveillance Capsule" from Revision 1 and changed titles for Tables 6-5 through 6-15. The Heatup and Cooldown Curves were removed from Appendix A of this report and new Curves were incorporated in WCAP-15878 Revision 0. Updated the Surveillance Capsule Removal Schedule found in Section 7.

Cook Unit I Capsule U

ix EXECUTIVE

SUMMARY

The purpose of this report is to document the results of the testing of surveillance Capsule U specimens and dosimeters from the D. C. Cook Unit I reactor vessel. Capsule U was removed at 9.17 EFPY and post irradiation mechanical testing of the Capsule U Charpy V-notch and tensile specimens was performed along with a fluence evaluation that included effects of a 10.8% uprating. The surveillance capsule U fluence was 1.837 x 10'9 n/cm 2 after 9.17 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.

Cook Unit I Capsule U

1-1 1

SUMMARY

OF RESULTS be The analysis of the reactor vessel materials contained in surveillance Capsule U, the fourth capsule to removed from the D. C. Cook Unit 1 reactor pressure vessel, resulted in the following conclusions:

Capsule U was pulled at 9.17 EFPY and had a lead factor of 3.50. With a lead factor of 3.50 at 9.17 EFPY capsule U received an average fast neutron fluence (E > 1.0 MeV) of 1.837 x 10'9 n/cm2 at the geometric center of the capsule.

• Irradiation of the reactor vessel intermediate shell plate B4406-3 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the maj or working direction of the plate 2

(longitudinal orientation), to 1.837 x 1019 n/cm (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 115TF and a 50 ft-lb transition temperature increase of 1251F. This results in an irradiated 30 ft-lb transition temperature of 1201F and an irradiated 50 ft-lb transition temperature of 155TF for the longitudinally oriented specimens.

Irradiation of the reactor vessel intermediate shell plate B4406-3 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major working direction of the plate 2

(transverse orientation), to 1.837 x 1019 r/cm (E> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 1 15°F and a 50 ft-lb transition temperature increase of 120°R This results in an irradiated 30 ft-lb transition temperature of 130TF and an irradiated 50 ft-lb tranisition temperature of 1851F for transversely oriented specimens.

2 Irradiation of the weld metal Charpy specimens to 1.837 x 1019 n/cm (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 205TF and a 50 ft-lb transition temperature increase of 2451F. This results in an irradiated 30 ft-lb transition temperature of 1151F and an irradiated 50 ft-lb transition temperature of 175T.

2

• Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 1.837 x 10'9 n/cm (E> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 1751F and a 50 ft-lb transition temperature increase of 1901F. This results in an irradiated 30 ft-lb transition temperature of 701F and an irradiated 50 ft-lb transition temperature of 120TF.

2 Irradiation of the Correlation Material Charpy specimens to 1.837 x 10'9 n/cm (E> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 120OF and a 50 ft-lb transition temperature increase of 120TF. This results in an irradiated 30 ft-lb transition temperature of 1651F and an irradiated 50 ft-lb transition temperature of 200TF.

• The average upper shelf energy of the intermediate shell plate B4406-3 (longitudinal orientation) 2 resulted in an average energy decrease of 17 ft-lb after irradiation to 1.837 x 10i"n/cm (E> 1.0 MeV). This results in an irradiated average upper shelf energy of 113 ft-lb for the longitudinally oriented specimens.

• The average upper shelf energy of the intermediate shell plate B4406-3 (transverse orientation) 2 resulted in an average energy decrease of 1 ft-lb after irradiation to 1.837 x 1019 n/cm (E> 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 95 ft-lb for the transversely oriented specimens.

Cook Unit I Capsule U

1-2 The average upper shelf energy of the weld metal Charpy specimens resulted an average energy decrease of 16 ft-lb after irradiation to 1.837 x 1019 n/cm 2 (E> 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 94 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 9 ft-lb after irradiation to 1.837 x 10'9 n/cm2 (E> 1.0MeV). This results in an irradiated average upper shelf energy of 117 ft-lb for the weld HAZ metal.

The average upper shelf energy of the Correlation Material Charpy specimens resulted in an average energy decrease of 10 ft-lb after irradiation to 1.837 x 1019 n/cm 2 (E> 1.0MeV). This results in an irradiated average upper shelf energy of 110 ft-lb for the weld HAZ metal

• The surveillance capsule U test results indicate that the intermediate shell plate B4406-3 (transverse and longitudinal) and the surveillance capsule weld metal 30 ft-lb transition temperature shift is in good agreement with the Regulatory Guide 1.99 Revision 2 predictions.

• The surveillance capsule materials exhibit a more than adequate upper shelf energy level for continued safe plant-operation and are expected to maintain an upper shelf energy of greater than 50 ft-lb throughout the life (32 EFPY) and life extension (48 EFPY) of the vessel as required by IOCFR50, Appendix GC 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 D. C. Cook Unit 1 reactor vessel using the Regulatory Guide 1.99, Revision 2 attenuation formula (ie. Equation # 3 in the guide; f(dh x1)=

fsurface

• e (-o24x)) is as follows:

Calculated (32 EFPY): Vessel inner radius* = 1.802 x 10'9 n/cm 2 2

Vessel 1/4 thickness = 1.082 x 1019 n/cm Vessel 3/4 thickness = 3.902 x lI0" n/cm2 2

Calculated (48 EFPY): Vessel inner radius* = 2.831 x l0ol n/cm 2

Vessel 1/4 thickness = 1.70 x 1019 n/cm 2

Vessel 3/4 thickness = 6.13 x 10" n/cm

• Clad/base metal interface Cook Unit 1 Capsule U

2-1 2 INTRODUCTION This report presents the results of the examination of Capsule U, the fourth capsule removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the American Electric Power Company D. C. Cook Unit I reactor pressure vessel materials under actual operating conditions.

The surveillance program for the D. C. Cook Unit I 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-8047, entitled "American Electric Power Service Corp. Donald C. Cook Unit No. 1 Reactor Vessel Radiation Surveillance Program" by S. E. Yanichko, D. J. Legsil. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM El185-70, "Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors"'3 . Westinghouse Energy System personnel were contracted to aid in the preparation of procedures for removing Capsule U from the reactor and its shipment to the Westinghouse Science and Technology Center Hot Cell Facility, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.

This report summarizes the re-analysis of the post-irradiation data obtained from surveillance Capsule U removed from the American Electric Power Company D. C. Cook Unit I reactor vessel and discusses the 1321 33 re-analysis of the data. The data is also compared to capsules T , X" " and Y14 which were previously removed from the reactor.

Cook Unit 1 Capsule U

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 SA533 Grade B Class 1 (base material of the D. C. Cook Unit 1 reactor pressure vessel shell plate) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.

A method for 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 Code14]. The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RT DT).1 RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-20815) or the temperature 60°F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the plate. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KIc curve) which appears in Appendix G to the ASME Code. The K1 , curve is a lower bound of static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the K1 c 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.

RTN*,T 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, such as the D. C. Cook Unit I reactor vessel radiation surveillance programE3, 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 to adjust the RTNDT for radiation embrittlement. This RTNDT (RTNDT initial +

ARTNDT) is used to index the material to the KI, 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.

Cook Unit I Capsule U

4-1 4 DESCRIPTION OF PROGRAM Eight surveillance capsules for monitoring the effects of neutron exposure on the D. C. Cook Unit I 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 U (Figure 4-2) was removed after 9.17 Effective Full Power Years (EFPY) of plant operation.

This capsule contained Charpy V-notch, tensile specimens and IX-WOL fracture mechanics specimens from the reactor vessel intermediate shell Plate B4406-3, weld metal representative of the beltline region weld seams, Charpy V-notch specimens from weld heait-affected zone (HAZ) material and Charpy V notch specimens from ASTM correlation material. All heat-affected zone specimens were obtained from the weld heat-affected zone of Plate B4406-3.

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 '/4 thickness location. Test specimens represent material taken at least one plate thickness from the quenched end of the plate. All base material Charpy V-notch impact and tensile specimens were oriented with the longitudinal axis of the specimen both normal to (transverse orientation) and parallel to (longitudinal orientation) the principal working direction of the plate. 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 normal to the welding direction. The I X-WOL fracture crack in the specimen would propagate parallel to the major working direction and the major surfaces of the shell plate. All specimens were fatigue precracked per ASTM E399-70T.

Capsule U contained dosimeters of pure Iron, Copper, Nickel and Aluminum-Cobalt wire (cadmium shielded and unshielded), and Neptunium (Np 237) and Uranium (U 238) 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 579'F (304°C) 1.75% Ag, 0.75% Sn, 97.5% Pb Melting Point 5901F (310 0 F)

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

Cook Unit I Capsule U

4-2 Table 4-1 Chemical Composition of the D. C. Cook Unit 1 Reactor Vessel Surveillance Materials Element Intermediate Shell Plate B4406-3 Weld Metal C 0.24 0.26 S 0.015 0.014 N2 0.008 0.010 Co <0.001 <0.001 Cu 0.14 0.27 Si 0.25 0.18 Mo 0.46 0.44 Ni 0.49 0.74 Mn 1.40 1.33 Cr 0068 0022 V <0.001 0001 P 0.009 0023 Sn 0.010 0.006 Ti <0.001 <0.001 Al 0.024 0.006 Zn <0.001 0.002 As 0.010 0.009 B <0.003 <0.003 Sb 0.001 0.001 Cook Unit 1 Capsule U

4-3 Table 4-2 Heat Treatment of the D. C. Cook Unit I Reactor Vessel Surveillance Materials Material Temperature (°F) Time (hr) Coolant 1600 4 Water quenched Intermediate Shell Course Plate B4406-3 1150 40 Furnace Cooled Weldment 1150 40 Furnace Cooled Cook Unit I Capsule U

4-4 4-4 X (220*)

Y (320*)

Z (3560)

+ 0.

/-v (176')

/ 90° Reactor Vessel Thermal Shield Car. Barrel Figure 4-1 Original Arrangement of Surveillance Capsules in the D. C. Cook Unit 1 Reactor Vessel Cook Unit I Capsule U

SPECIMEN NUMBERING CODE A - PLATE B4406-3 (LONGITUDINAL DIRECTION)

AT - PLATE B4406-3 (TRANSVERSE DIRECTION)

R - ASTM CORRELATION MONITOR MATERIAL H - WELD HEAT- AFFECTED ZONE W - WELD METAL SURVEILLANCE CAPSULE U Np237 23 8 U

I DOSIMETER WOL WOL WOL WOL CHARPY CHARPY BLOCK C:HARPY CHARPY CHARPY CHARPY CHARPY CHARPY CHARPY CHARPY A A7 A6 A5 I W4JJ6W-44 Wd4 EI6Ik8 A60 58 N+59 A59 A1 Co Co(Cd) 590OF -Cu Cu Ni, MONITOR Fe Fe Fe Fe CENTER REGION OF VESSEL TO TOP OF VESSEL TO BOTTOM OF VESSEL Figure 4-2 Capsule U Diagram Showing the Location of Specimens, Thermal Monitors, and Dosimeters I,

Cook Unit I Capsule U Ii

5-1 5 TESTING OF SPECIMENS FROM CAPSULE U 5.1 OVERVIEW The post-irradiation mechanical testing of the Charpy V-notch and tensile sliecimens was performed at the Westinghouse Science and Technology Center Laboratory with consultation by Westinghouse Energy 21 Systems personnel. Testing was performed in accordance with IOCFR50, Appendices G and Ht , ASTM Specification El 85-82161, and Westinghouse Remote Metallographic Facility (RMF) Procedure RMF 8402, Revision I 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-8047' . No discrepancies were found.

Examination of the two low-melting point 304'C (579°F) and 31 0°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).

261 The Charpy impact tests were performed per ASTM Specification E23-86' and RMF Procedure 8103, Revision I, on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy impact test machine is instrumented with an Effects Technology model 500 instrumentation system. 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 (PGY), the time to general yielding (tNy), the maximum load (PrI), and the time to maximum load (tM) 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 (PF), and the load at which fast fracture terminated is identified as the arrest load (PA).

The energy at maximum load (EM) 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 (Em).

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

Ou=(Pay*L) I[B * (W - a) 2

• C] M) 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 (0), 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 4= -

450 and p = 0.010 inch, Equation 1 is valid with C = 1.21. Therefore, (for L = 4W),

Cook Unit 1 Capsule U

5-2 o,=(Pcy*L)I [B*(W-a)*l21] = (3.33*PGy *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 A370-88 271 1 . 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 E8-83128 ] and E21-791291 , and RMF Procedure 8102, Revision 1. All pull rods, grips, and pins were made of Inconel 718 hardened to Rc45. 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.

Deflection 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 E83-67'"01.

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.

Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperatures. Chromel-Alumel thermocouples were inserted in shallow holes in the center and each end of the gage section of a dummy specimen and in each grip. In test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower grip and controller temperatures was developed over the range room temperature to 550'F (288°C). The upper grip was used to control the furnace temperature. During the actual testing, the grip temperatures were used to obtain desired specimen temperatures. Experiments have indicated that this method is accurate to +2°F.

The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were Cook Unit 1 Capsule U

5-3 determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.

5.2 CHARPY V-NOTCH IMPACT TEST RESULTS The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule U irradiated to approximately 1.837 x10'9 n/cm at 550'F are presented in Tables 5-1 through 5-6 and 2

Figures 5-1 through 5-5. The transition temperature increases and upper shelf energy decreases for the Capsule U material are shown in Table 5-7.

Capsule U was pulled at 9.17 EFPY and had a lead factor of 3.50. With a lead factor of 3.50 9at 9.17 EFPY capsule U received an average fast neutron fluence (E > 1.0 MeV) of 1.837 x 10' n/cm2 at the geometric center of the capsule.

Irradiation of the reactor vessel intermediate shell plate B4406-3 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction of the plate 2

(longitudinal orientation), to 1.837 x 1019 n/cm (E> 1.0MeV) resulted in a 30 ft-lb0transition temperature increase of 115'F and a 50 ft-lb transition temperature increase of 125 F. This

-results in an irradiated 30 ft-lb transition temperature of 1201F and an irradiated 50 ft-lb transition temperature of 1551F for the longitudinally oriented specimens.

Irradiation of the reactor vessel intermediate shell plate B4406-3 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major working direction of the plate 2

(transverse orientation), to 1.837 x 1019 n/cm (E> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 1151F and a 50 ft-lb transition temperature increase of 1201F. This results in an irradiated 30 ft-lb transition temperature of 1301F and an irradiated 50 ft-lb transition temperature of 185°F for transversely oriented specimens.

2 Irradiation of the weld metal Charpy specimens to 1.837 x 10'9 n/cm (1> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 205'F and a 50 ft-lb transition temperature increase of 245°1F. This results in an irradiated 30 ft-lb transition temperature of 1151F and an irradiated 50 0

ft-lb transition temperature of 175 F.

2 Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 1.837 x 1019 n/cm (E> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 175°F and a 50 ft-lb transition temperature increase of 190 F. This results in an irradiated 30 ft-lb transition 0

temperature of 701F and an irradiated 50 ft-lb transition temperature of 120'F.

2 Irradiation of the Correlation Material Charpy specimens to 1.837 x 10'9 n/cm (E>, 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 1201F and a 50 ft-lb transition temperature increase of 1201F. This results in an irradiated 30 ft-lb transition temperature of 165 0F and an irradiated 50 ft-lb transition temperature of 2001F.

The average upper shelf energy of the intermediate shell plate B4406-3 (longitudinal orientation) 2 resulted in an average energy decrease of 17 ft-lb after irradiation to 1.837 x 1019 n/cm (E> 1.0 MeV). This results in an irradiated average upper shelf energy of 113 ft-lb for the longitudinally oriented specimens.

Cook Unit I Capsule U

5-4 0 The average upper shelf energy of the intermediate shell plate B4406-3 (transverse orientation) resulted in an average energy decrease of 1 ft-lb after irradiation to 1.837 x 1019 n/rcm 2 (E> 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 95 ft-lb for the transversely oriented specimens.

• The average upper shelf energy of the weld metal Charpy specimens resulted an average energy decrease of 16 ft-lb after irradiation to 1.837 x 1019 n/cm 2 (E> 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 94 ft-lb for the weld metal specimens.

0 The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an average energy decrease of 9 ft-lb after irradiation to 1.837 x 1019 n/cm 2 (E> 1.0MeV). This results in an irradiated average upper shelf energy of 117 ft-lb for the weld HAZ metal.

0 The average upper shelf energy of the Correlation Material Charpy specimens resulted in an average energy decrease of 10 ft-lb after irradiation to 1.837 x 10'9 n/cm2 (E> 1.0MeV). This results in an irradiated average upper shelf energy of 110 ft-lb for the weld HAZ metal 0 The Fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-6 through 5-10 and shows an increasingly ductile or tougher appearance with increasing test temperature.

0 A comparison of the 30 ft-lb transition temperature increase and upper shelf energy decreases for the various D. C. Cook Unit I surveillance materials with predicted values using the methods of NRC Regulatory Guide 1.99, Revision 2t31 is presented in Table 5-8. This comparison indicates that the transition temperature increases and the upper shelf energy decreases of the intermediate shell plate B4406-3 (longitudinal and transverse) and the surveillance weld resulting from irradiation to 1.837 x 10"9 n/cm2 (E> 1.0MeV) are less than the Regulatory Guide 1.99 Revision 2 predictions.

5.3 TENSILE TEST RESULTS The results of the tensile tests performed on Plate B4406-3 (longitudinal orientation) and weld metal irradiated to 1.837 x 1019 n/cm2 are shown in Table 5-9 and Figures 5-11 and 5-12, respectively. These results show that irradiation produced a 12 to 15 ksi increase in 0.2 percent yield strength for Plate B4406-3 and 18 to 20 ksi increase for the weld metal. Fractured tension specimens for each of the materials are shown in Figures 5-13 and 5-14. A typical stress-strain curve for the tension specimens is shown in Figure 5-15.

5.4 WOL SPECIMENS Per the surveillance capsule testing contract with American Electric Power Company, the 1X-WOL Fracture Mechanics specimens will not be tested and will be stored at the Hot Cell at the Westinghouse Science and Technology Center.

Cook Unit 1 Capsule U

5-5 Table 5-1 Charpy V-notch Impact Data for the D. C. Cook Unit I Reactor Vessel Shell Plate B4406-3 Irradiated at 550%F, Fluence 1.837 x 1019 n/cm 2 (E> 1.0 MeV)

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

Longitudinal Orientation A60 82 28 16.0 21.5 13.0 0.33 10 A54 100 38 37.0 50.0 24.0 0.61 15 A56 125 52 43.0 58.0 33.0 0.84 25 A52 150 66 34.0 46.0 30.0 0.76 35 A55 175 79 63.0 85.5 43.0 1.09 40 A58 225 107 121.0 164.0 78.0 1.98 100 A51 250 121 100.0 135.5 77.0 1.96 100 A57 300 149 124.0 168.0 82.0 2.08 100 A53 300 149 113.0 '153.0 81.0 2.06 100 A59 350 177 110.0 149.0 73.0 1.85 10o Transverse Orientation AT60 50 10 19.0 26.0 15.0 0.38 10 AT57 82 28 29.0 39.5 20.0 0.51 15 AT59 100 38 22.0 306.0 22.0 0.56 20 AT58 125 52 32.0 43.5 21.0 '0.53 20 AT54 150 66 37.0 50.0 32.0 0.81 25 AT51 200 93 56.0 76.5 42.0 '1.07 75 AT56 240 116 82.0 111.0 65.0 1.65 100 AT52 250 121 95.0 129.0 71.0 1.80 100 AT55 300 149 95.0 129.0 61.0 1.55 100 AT53 350 177 103.0 139.5 64.0 1.63 100 Cook Unit 1 Capsule U

5-6 Table 5-2 Charpy V-notch Impact Data for the D. C. Cook Unit 1 Reactor Vessel Weld Data and HAZ Metal Irradiated at 550%F, Fluence 1.837 x 101' n/cm2 (E> 1.0 MeV)

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

Weld Metal W47 50 10 21.0 28.5 14.0 0.36 15 W42 74 23 53.0 71.8 35.0 0.89 65 W44 82 28 63.0 85.5 41.0 1.04 60 W45 125 52 38.0 51.5 31.0 0.79 65 W43 150 66 48.0 65.0 35.0 0.89 70 W48 200 93 61.0 82.5 50.0 1.27 100 W46 250 121 90.0 122.0 63.0 1.60 100 W41 350 177 95.0 129.0 68.0 1.73 100 HAZ Metal H45 0 -18 10.0 13.5 10.0 0.25 20 H47 50 10 48.0 65.0 36.0 0.91 35 H46 82 28 75.0 101.5 42.0 1.07 70 H44 150 66 64.0 87.0 460 1.17 80 H48 200 93 95.0 129.0 72.0 1.83 100 H43 250 121 121.0 164.0 67.0 1.70 100 H41 275 135 112.0 152.0 72.0 1.83 100 H42 350 177 78.0 106.0 63.0 1.60 100 Cook Unit I Capsule U

5-7 Table 5-3 Charpy V-notch Impact Data for the D. C. Cook Unit 1 ASTM Correlation Monitor 2

Material Irradiated at 550%F, Fluence 1.837 x 10'9 n/cm (E> 1.0 MeV)

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

100 38 14.0 19.0 12.0 0.30 10 R44 R42 175 79 26.0 35.5 20.0 0.51 25 R43 200 93 31.0 42.0 25.0 0.64 30 107 68.0 92.5 49.0 1.24 90 R48 225 R45 250 121 81.0 110.0 59.0 1.50 100 R46 275 135 89.0 120.5 67.0 1.70 100 R47 300 149 112.0 152.0 70.0 1.78 100 R41 350 177 108.0 146.5 74.0 1.88 100 Cook Unit I Capsule U

5-8 Table 5-4 Instrumented Charnv Inmact Test Ref*ults for D_ CP Cz iLI;t 1 . 1 Vr....I v-.. QI, ,It DI.,4.~ DAAnt2 Normalized Energies Sample Test Charpy Normline) Yield Time to Maximum Time to Fracture Arrest Yield Flow NSbe Temp Energy (ftlb/in') Load Yield Load Maximum Load Load Stress umer (OF) (ft-lb) Charpy Maximum Stress Prop (kips) (gsec) (kips) (Qisec) (kips) (kips) (ksi) (ksi)

Ed/A Em/A Ep/A Longitudinal Orientation A60 82 16.0 129 102 26 3.10 100 4.20 260 4.15 0.20 102 121 A54 100 37.0 298 189 108 2.75 75 4.25 435 4.15 -- 91 116 A56 125 43.0 346 233 113 3.40 125 4.60 530 4.55 0.30 113 132 A52 150 34.0 274 301 -27 4.45 90 6.00 500 6.00 1.40 148 173 A55 175 63.0 507 313 195 2.85 80 4.55 670 4.50 1.80 95 122 A58 225 121.0 974 293 682 2.55 100 4.25 690 .... 84 112 A51 250 100.0 805 286 519 2.70 125 4.30 695 -- 90 116 A53 300 113.0 910 382 528 4.05 100 5.75 660 .... 133 162 A57 300 124.0 998 284 715 2.55 80 4.05 675 .... 84 109 A59 350 110.0 886 250 636 2.70 110 4.15 605 .... 89 113 Transverse Orientation AT60 50 19.0 153 101 52 2.65 70 3.90 260 3.90 -- 87 108 AT57 82 29.0 234 164 69 3.00 85 4.60 370 4.45 0.25 99 125 AT59 100 22.0 177 207 -30 4.45 135 5.75 395 5.65 0.25 151 171 AT58 125 32.0 258 146 111 2.95 100 4.25 355 4.15 0.95 97 118 AT54 150 37.0 298 174 124 2.40 90 4.05 445 3.90 1.20 80 106 AT51 200 56.0 451 220 230 2.95 85 4.30 505 4.20 1.90 98 120 AT56 240 82.0 660 288 372 3.90 105 5.65 510 .... 129 159 AT52 250 95.0 765 249 516 2.50 95 4.05 600 -- 83 109 AT55 300 95.0 765 240 525 2.50 100 3.90 600 .... 82 106 AT53 350 103.0 829 211 619 2.60 110 4.10 525 .... 86 I10 Cook Unit 1 Capsule U

5-9 Table 5-5 Instrumented Charpy Impact Test Results for D. C. Cook Unit 1 Reactor Vessel Weld Metal and HAZ Metal Test Charpy Normalized Energies Yield Time to Maximum Time to Fracture Arrest Yield Flow Sampl Temp Energy (ft-lb/in) Load Yield Load Maximum Load Load Stress Stress Number (F) (ft-lb) Charpy Maximum Prop (kips) (psec) (kips) (11see) (kips) (kips) (ksi) (ksi)

Ed/A Em/A Ep/A I _

Weld Metal W47 50 21.0 169 160 9 - 4.35 80 5.80 275 5.75 0.15 144 168 W42 74 53.0 427 Computer Malfunction .......... .. -

W44 82 63.0 507 244 263 3.15 95 4.60 520 4.35 1.25 103 128 W45 125 38.0 306 308 -2 4.40 90 5.90 505 5.70 2.50 145 171 W43 150 48.0 387 197 198 2.75 70 4.35 435 4.35 1.35 91 118 W48 200 61.0 491 244 248 4.45 125 4.70 450 .... 148 169 W46 250 90.0 725 206 518 2.60 85 3.90 510 - 86 108 W41 350 95.0 765 -200 565 2.55 75 3.75 505 - - 85 104

,,____ HAZ Metal -_

H45 0 10.0 81 52 29 5.20 140 5.50 155 5.50 1.60 172 177 H47 50 48.0 387 384 2 4.86 125 6.60 605 6.25 0.40 161 190 H69 82 75.0 604 245 359 3.00 45, - 5.00 460 4.50 2.00 99 132 H44 " 150 64.0 515 246 269 3.2 90 4.60 520 4.45 2.25 105 129 H48 200 95.0 765 307 458 2.35 95 6.10 505 -- 144 174 H43 250 121.0 '974 218-.. 756 2.7 75 4.15 1505- 89 113 H41 275 112.0 902 350 552 4.3' 125 6.00 6410 -- -- 142 170 H42 350 78.0 628 ' 211 417 2.55 115 3.80 545 - - 84 105 Cook Unit I Capsule U

5-10 Table 5-6 Instrumented Chiarnv Impact Test Results for D. C. Cook Unit 1 ASTM Correlation Monitor Material Sa Test Charpy Normalized Energies Yield Tiime to Maximum Time to Fracture Arrest Yield Flow amnpe Temp Energy (ftlb/iii) Load Yield Load Maximum Load Load Stress Stress Number (OF) (ft-lb) Charpy Maximum Prop (kips) (psec) (kips) (lPsec) (kips) (kips) (ksi) (ksi)

Ed/A Em/A Ep/A R44 100 14.0 113 98 15 4.00 115 5.05 220 4.95 0.50 132 149 R42 175 26.0 209 108 101 2.90 125 3.80 310 3.80 3.80 96 11 0 R43 200 31.0 250 54 54 3.80 100 5.05 380 4.85 4.85 126 147 R48 225 68.0 548 126 126 4.25 130 5.50 775 5.50 5.50 141 162 R45 250 81.0 652 435 435 2.7 95 4.15 520 -- -- 89 113 R46 275 89.0 717 436 436 3.80 95 5.45 510 - - 126 154 R47 300 112.0 902 539 539 3.70 85 5.35 665 - - 123 150 R41 350 108.0 870 624 624 2.30 40 3.80 605 1 76 101 Cook Unit I Capsule U

5-11 2 Properties of the D. C. Cook Unit 1 Table 5-7 The Effect of 550°F Irradiation at 1.837 x 1019 n/cm (E>1.0 MeV) on the Notch Toughness Reactor Vessel Surveillance Capsule Materials Average 30 (ft-lb) Average 35 mil Lateral Average 50 ft-lb Average Energy Absorption 0 at Full Shear (ft-lb)

Material Transition Temperature (*F) Expansion Temperature (*F) Transition Temperature ( F)

AT' Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiated' Irradiated 120 30 155 125 130 113 -17 Plate B4406-3 5 120 115 20 140 (Longitudinal) I I 115 65 185' 120 96 95 -1 Plate B4406-3 15 130 115 25 140' (Transverse) 205 -70 175 245 110 94 -16 Weld Metal -90 115 205 -80 125' 165 -65 120 185 126 117 -9 HAZ Metal -100 70 175 -100 65 130 80g 200 120 120 110 -10 Correlation 45 '165 120 60 190 Material 111 Note: All unirradiated data presented here was taken from WCAP-8047 .

Cook Unit I Capsule U

I 5-12 Table 5-8 Comparison of the D. C. Cook Unit I 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 n/cm 2) (0 F) (a) (CF) (%) (a) (%) (b)

Plate B4406-3 T 0.267 66.7 60 18 18 (Longitudinal) X 0.831 98.6 90 23 21 Y 1.195 109.1 105 25 21 U 1.837 121.4 115 27 13 Plate B4406-3 T 0.267 66.7 70 18 14 (Transverse) X 0.831 98.6 110 23 19 Y 1.195 108.1 115 25 21 U 1.837 121.4 115 27 1 Weld Metal T 0.267 132.3 80 33 27 X 0.831 195.7 165 44 33 Y 1.195 216.5 200 45 37 U 1.837 240.9 205 48 15 HAZMetal T 0.267 -- 120 -- 25 X 0.831 160 36 Y 1.195 165 38 U 1.837 170 7 Correlation Material T 0.267 - - 60 -- 15 X 0.831 - - 100 - - 33 Y 1.195 -- 110 -- 26 U 1.837 - - 120 - - 8 Notes:

(a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material. Changes in the predicted values here versus Revision 0 are due to the changes in capsule fluence values.

(b) Values are based on the definition of upper shelf energy given in ASTM E185-82.

Cook Unit I Capsule U

5-13 2

Table 5-9 Tensile Properties for D. C. Cook Unit I Reactor Vessel Material Irradiated to 1.837 x 10i"n/cm (E > 1.0 McV)

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 A3 74 83.0 103.9 3.40 199.8 69.3 11.3 23.1 65 Plate A4 600 70.8 95.7 3.20 162.6 65.2 10.5 21.6 60 Weld WIl 74 83.5 97.8 3.30 167.7 67.3, 12.8 23.9 60 Weld W12 600 79.5 96.8 3.85 171.6 78.4 10.5 19.4 54 Cook Unit I Capsule U

I 5-14 Table 5-10 D. C. Cook Unit I Reactor Vessel Beltline Region Material Properties Component Plate No. Cu (Wt %) (Ni %) Initial RTNDT (OF)

Intermediate Shell Plate B4406-1 0.12 0.52 5 Intermediate Shell Plate B4406-2 0.15 0.50 33 Intermediate Shell Plate B4406-3 0.15 0.49 40 Lower Shell Plate B4407-1 0.14 0.55 28 Lower Shell Plate B4407-2 0.12 0.59 -12 Lower Shell Plate B4407-3 0.14 0.50 38 Longitudinal Welds 0.28(a) 0.74(a) -56(a)

Circumferential Welds 0.28() 0.74(a -56(a)

Closure Head Flange -- --- 60 Vessel Flange ...... 28 Notes:

(a) Reference 33 Cook Unit I Capsule U

5-15 (CC)

-150 -100, -50 -0 50 100 80 60 V) 40 20 0

100. Sr "r '7 I --- I -- 2.5 A I in

.80, 2.0 L5 60 E LU 1.0 40 0.5 20 0

! I  !  ! I I I I 0 I I 200 180 240 160 200 "140 S120 160

~100 120 080 80 60 40 40 20 0 0 100 200 Temperature (0 F)

Shell Plate B4406 Figure 5-1 Charpy V-Notch Impact Data for D. C. Cook Unit 1 Reactor Vessel "3(Transverse Orientation)

Cook Unit I Capsule U

5-16

(°C)

-153 -100 -50 0 50 100 150 200 250 100 80 S60 v40 20 0.

100 0 2.5 9 80 -. O

-1.5 200 160

~20 - .5 01 1 200 180 -- 240 160 Unirradiated o 2 p120 - -160

~100 a,,~ (

- *80 Irradiated at 550°F ~~120120

" 60 1.837x1019 n/cm 2 a

80 40 -15 20 - 0 40 0 I!II1I!I 02 0

-200 -100 0 ,100 200 300 400 500 Temperature (OF)

Figure 5-2 Charpy V-Notch Impact Data for D. C. Cook Unit 1 Reactor Vessel Shell Plate B4406 3 (Longitudinal Orientation)

Cook Unit I Capsule U

5-17 (CC)

-150 -100 -50 -0 50 100 80 LO 60 40 20 0

100 2.5 80 2.0 L.5,

,60 E E

LO UJ 40 0.5 20 0

-J 0

200 180 240 160 200 140 160 120 L&

100 120 so 80 60 40 40 20 0 0

-200 -100 0 100 -. 200 300 400 - 500 Temperature (OF)

Figure 5-3 "Charpy V-Notch Impact Data for D. C. Cook Unit I Reactor Vessel Weld Metal Cook Unit I Capsule U

5-18

( 0 C)

-150 -100 -50 50 100 80 1

60 40 20 0

100 2.5 E 80 60 L5 40 LO Ia 20 0.5 mi 0 0 200 180 240 160 140 o 1200 0

120 Unirradiated 0 160

~100 -A 120

,80 60 1850 F Irradiated at 550'F 80 40 175 0F 1.837 x 10'9 n/cm2 40 20 2 o I ( I I I I I I I 0 0

-200 -100 0 100 200 300 400 500 Temperature (OF)

Figure 5-4 Charpy V-Notch Impact Data for D. C. Cook Unit 1 Reactor Vessel Weld Heat Affected-Zone MIetal Cook Unit I Capsule U

5-19 (0c)

-150- -100 -50 0 50 100 150 200 250

-3 100

• 80 S 460 -

20 20 - I I "3

-0 100 - I I *I I I T 1 1 2.5

-80 - - 2.0 E

60 - 0 L-5 180 240 160 205 "140-2 120 Unirradiated 9 160 00- 100 i;; 120w 380 0 120o F Irradiated at 550 80 atF 60

-50

- 01200 F 1.837Ux 10 n.ca 203 2 *4 0 0

-200 -100 0 100 200 300 400 500 Temperature (IF)

Figure 5-5 Charpy V-Notch Impact Data for D. C. Cook Unit 1 ASTM Correlation Material Cook Unit I Capsule U

5-20 AJ

• A60 A54 A56 A52 A55 A58 A51 A53 A57 A59 Figure 5-6 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit I Reactor Vessel Shell Plate B4406-3 (Longitudinal Orientation)

Cook Unit 1 Capsule U

5-21 ATS5 AT58 AT54 ATBO AT57 AT53 AT52 AT55 AT51 AT56 Unit 1 Reactor Vessel Figure 5-7 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Shell Plate B4406-3 (Transverse Orientation)

Cook Unit I Capsule U

I 5-22 W47 . . W42 W44 W45 in W43 W48 W48 W41 Figure 5-8 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit I Reactor Vessel Weld Metal Cook Unit 1 Capsule U

5-23 1145 147 1146 1144 148 1143 H4i 142 Figure 5-9 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit I Reactor Vessel Weld HAZ Metal Cook Unit 1 Capsule U

I 5-24

<4

-R48 R45 R46 R47 R41-Figure 5-10 Charpy Impact Specimen Fracture Surfaces for ASTM Correlation Material Cook Unit 1 Capsule U

5-25 0C

-50 50 100 150 200 250 300 120i 0

I I I I - I I 1_ 800 1101 le Strenath - 700 100 In 90

%A,80 500

• 70 60 400 50 300.

I I I i I I I -

40 I I I i, Code :

Open Points - Unirradiated x 1019 n/cm2 Closed Points - Irradiated at 5500F 1. 88 80 70 60 "Reduction in Area 50 0 -, Total Elongation

=

0 10- -w--UniformElongation 0  !  !  ! '

-100 0 100 200 '300 *400 500 600 Temperature (°F)

Figure 5-11 Tensile Properties for D. C. Cook Unit 1 Reactor Vessel Shell Plate B4406-3 (Longitudinal Orientation)

Cook Unit I Capsule U

5-26 oC

-50 0 50 100 150 200 250 300 1 qct I-- I II IiI 5-26 800 110 100 700 90 600 Ultimate Tensile Strength 80 L.r 70 500

0. 2 %Yield Strength 60 400 50 40 I I I

I iI I I I I

I II I 300 Code:

Open Points - Unirradiated 2

Closed Points - Irradiated at 550 0 F L 8x 1019 n/cm 80 0

70 60 50 Reduction in Area 40

= 30 Total Elongation 0

20 10 Unifqrm Eloggation 0

-100 500 600 0 100 200 300 400 Temperature (OF)

Figure 5-12 Tensile Properties for D. C. Cook Unit I Reactor Vessel Weld Metal Cook Urut I Capsule U

5-27 Specimen A3 74°F Specimen A4 600F Figure 5-13 Fractured Tensile Specimens for D. C. Cook Unit I Reactor Vessel Shell Plate B4406-3 (Longitudinal Orientation)

Cook Unit 1 Capsule U

I 5-28 ITZ' 4'A

""'-S emn II* "* ' 74"F

-. Specimen Wi2 6007F Figure5-14 Fractured Tensile Specimens for D.C. Cook Unit I Reactor Vessel Surveillance Weld Metal Cook Unit 1 Capsule U

5-29 120 110 100 90 370 0260 E- 50 40 30 20 SPEC A4 10 0 0.04 0.08 0.12 0.18 0.2 024 STRAIN, INAN Fiigure 5-15 Typical Stress-Strain Curve for Tension Specimens Cook Unit I Capsule U

6-1 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6.1 INTRODUCTION

This section describes a discrete ordinates S, transport analysis performed for the D. C. Cook Unit 1 reactor to determine the neutron radiation environment within the reactor pressure vessel and surveillance capsules. In this evaluation, 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 for the first seventeen reactor operating cycles. In addition, neutron dosimetry sensor sets from Surveillance Capsu-les T, X, Y, and U withdrawn from the D. C. Cook Unit I reactor at the conclusion of fuel cycles 1, :4, 6, and 10 weie analyzed using ciirrent dosimetry evaluation methodology.

Comparisons of the results of these dosimetry evaluations with the analytical predictions provided a validation of the plant specific neutron transport calculations. These validated calculations were then used to provide projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 54 Effective Full Power Years (EFPY). These projections conservatively account for an assumed plant uprating, from 3250 MWt to 3600 MWt, beginning with the operation of the eighteenth fuel cycle. All of the neutron transport calculations and dosimetry evaluations described in this section meet the requirements of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."'71 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 damagd trend curves as well as for the implementation of trend curve data to assess the condition of the vessel. In recent years; however, as discussed in Regulatory Guides 1.190 and 1.99, Revision 2, 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 this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, "Analysis and Interpretation of Light-Water Reactor Surveillance Results," recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a 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 were based on the latest available nuclear cross-section data derived from ENDF/B-VI and made use of the latest available calculational tools. 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.t] The specific calculational methods applied are also consistent with those Cook Unit I Capsule U

I 6-2 described in WCAP-15557, "Qualification of the Westinghouse Pressure Vessel Neutron Fluence t9 Evaluation Methodology." g It should be noted that because the neutron dosimetry sets from the first four surveillance capsules were re-analyzed herein using current dosimetry evaluation methodology, the results do not exactly match those reported in Reference 11.

6.2 Discrete Ordinates Analysis A plan view of the D. C. Cook Unit I 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 40, 1760,1 840, 3560 (40 from the core cardinal axes) and 40', 1400, 2200, 3200 (400 from the core cardinal axes) as shown in Figure 4-1. The stainless steel specimen containers are I-inch square and approximately 38 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 3 feet of the 12-foot high reactor core. Capsules T, X, Y, and U were originally loaded into the 400 azimuthal location, while Capsules S, V, W, and Z were positioned at 40. Capsules T, X, Y, and U were withdrawn for analysis at the conclusion of the first, fourth, sixth, and tenth fuel cycles, respectively. After fuel cycle fourteen, Capsule S was re-positioned from the 40 to the 40' location. At this time, the capsule was re-designated as Capsule W, while the original Capsule W was re-designated Capsule S (Reference 35). The irradiation history of the eight surveillance capsules is summarized as follows:

Capsule Location Irradiation History T 400 Cycle I (withdrawn for analysis)

X 400 Cycles 1-4 (withdrawn for analysis)

Y 400 Cycles 1-6 (withdrawn for analysis)

U 400 Cycles 1-10 (withdrawn for analysis)

SMt) 40 Cycles 1-17 (in reactor)

V 40 Cycles 1-17 (in reactor)

WM 40/400 Cycles 1-17 (in reactor)

Z 40 Cycles 1-17 (in reactor)

Note:

1. Capsule W was formally located at the 40 position and known as Capsule S. Capsule S was formally known as Capsule W. These changes are documented in Reference 35.

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.

The fast neutron exposure evaluations for the D. C. Cook Unit I surveillance capsules and reactor vessel were based on a series of fuel cycle specific forward transport calculations that were combined using the following three-dimensional flux synthesis technique:

• A(rz) = [re) * [4(r,z)]/[M(r)]

"CkLnit I Capsule U

6-3 where O(rO,z) is the synthesized three-dimensional neutron flux distribution, *,(rO) is the transport solution in r,0 geometry, O(rz) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and 0(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the re two-dimensional calculation. This synthesis procedure was carried out for each operating cycle at D. C. Cook Unit 1.

For the D. C. Cook Unit I calculations, one r,0 model was developed since the reactor is octant symmetric. This re model includes the core, the reactor internals, the thermal shield -- including explicit representations of the surveillance capsules at 40 and 400, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. The symmetric rO model was utilized to perform both the surveillance capsule dosimetry evaluations, and subsequent comparisons with calculated results, and to generate the maximum fluence levels at the pressure vessel wall. 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 re geometric mesh description of the reactor model consisted of 170 radial by 67 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 r,z model used for the D. C. Cook Unit I calculations 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 I-foot below the active fuel to I-foot above the active fuel. As in the case of the r,0 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 153 radial by 90 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 rz calculations was also set at a value of 0.001.

The one-dimensional radial model used in the synthesis procedure consisted of the same 153 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 for the first seven fuel cycles used in the plant specific transport analysis were obtained from Reference 10 (which was the input used in the previous surveillance capsule analysis, documented in Reference 11). The core power distributions for cycles eight through seventeen "were taken from the appropriate D. C. Cook Unit I fuel cycle design reports (References 12 through 21).

The data extracted from the references 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.

Cook Unit 1 Capsule U

6-4 The individual cycle lengths used in these calculations differ slightly from the AEP provided design input. These differences were evaluated and found to be insignificant.

In constructinm 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. The spatial power distributions used in the transport analyses are provided in Appendix A to this report.

All of the transport calculations supporting this analysis were carried out using the DORT discrete ordinates code Version 3. [122]and the BUGLE-96 cross-section library.t 231 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 S16 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 6-4. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the Capsules T, X, Y, and U irradiation and provide the calculated neutron exposure of the pressure vessel wall for the first seventeen fuel cycles. In Table 6-1, 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 two azimuthally symmetric surveillance capsule positions (4' and 40'). These data, representative of the axial midplane of the active core, are meant to establish the exposure of the surveillance capsules withdrawn to date and to provide an absolute comparison of measurement with calculation. Similar information is provided in Table 6-2 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. Again, both fluence (E > 1.0 MeV) and dpa data are provided. It is important to note that the data for the vessel inner radius were taken at the clad/base metal interface, and thus, represent the maximum calculated exposure levels of the vessel plates and welds.

Radial gradient information applicable to 4O(E > 1.0 MeV) and dpa/sec are given in Tables 6-3 and 6-4, respectively. The data, based on the Cycles I through 17 cumulative fluence, 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 6-3 and 6-4.

6.3 Neutron Dosimetry 6.3.1 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 D. C. Cook Unit I 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:

Cook Unit I Capsule U

6-5 Azimuthal Withdrawal Irradiation Capsule ID Location Time Time [EFPY1 T 400 End of Cycle 1 1.27 X 400 End of Cycle 4 3.48 Y 400 End of Cycle 6 4.95 U 400 End of Cycle 10 9.17 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, X, Y, and U are summarized as follows:

Reaction Sensor Material of Interest Capsule T Capsule X Capsule Y Capsule U 6

Copper ICu(n,t)6°Co X X X X X X**

Iron "-Fe(n,p)V4Mn X X "5'Ni(n,p)5 8Co X X X X Nickel, 238 Uranium-238 U(nf) 137Cs X X ° _ X X

237 SX X Neptunium-237 Np(n,f) 137Cs X 5"gCo(n,y)°Co -X X X X Cobalt-Aluminum*

• The cobalt-aluminum measurements for this plant include both bare wire and cadmium-covered sensors.

• • One of the five iron sensors in Capsule U was not recovered 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 6-5.

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:

Cook Unit I Capsule U

I 6-6

"* 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 Capsules T, X, and Y was carried out at the Southwest Research Institute (SwRI). The radiometric counting of the sensors from Capsule U was completed at the Westinghouse Analytical Laboratory, 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.

The irradiation history of the reactor over the irradiation periods experienced by Capsules T, X, Y, and U was based on the reported monthly power generation of D. C. Cook Unit I 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, X, Y, and U is given in Table 6-6.

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 No FY X P Cj [IeA" [e'--"]

where:

R = Reaction rate averaged over the irradiation period and referenced to operation at a core power level of P~f (rps/nucleus).

A = Measured specific activity (dps/gm).

No = Number of target element atoms per gram of sensor.

F = Weight fraction of the target isotope in the sensor material.

Y = Number of product atoms produced per reaction.

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

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

Cook Unit 1 Capsule U

6-7 CJ Calculated ratio of cO(E> 1.0 MeV) during irradiation period j to the time weighted average O(E > 1.0 MeV) over the entire irradiation period.

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

t = 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 [Pj]/[Pr] 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, Cj 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 Cj are listed in Table 6-7.

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, 235 additional corrections were made to the U measurements to account for the presence of U impurities 238 in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.

The corrections used in the D. C. Cook Unit I dosimetry evaluations were obtained using the ORIGEN code to develop a correlation defining the U-238(n,f) contribution to the total integrated fissions in the dosimeter as a function of fluence experienced by the sensor. The specific corrections used in the evaluation of the D. C. Cook Unit 1 U-238 sensors are summarized as follows:

Calculated Fluence Fractional Capsule ID (E > 1.0 MeV) U-238 And Location I[n/cm 2] Contribution T (40 Degrees) 2.667e+ 18 0.874 X (40 Degrees) 8.313e+*18 0.853 Y (40 Degrees) 1.195e+19 0.838 U (40 Degrees) 1.837e+19 0.815 7

Corrections were also made'to the "U and " Np sensor reaction rates to account for gamma ray induced fission reactions that occurred over the course of the capsule irradiations. The corrections were determined for each capsule location from the results of the ENDF/B-VI transport calculations using the BUGLE-96 library. The transport calculations were completed for the entire 67 group structure included in the BUGLE-96 library. From these calculations, the ratio of the gamma ray induced fission to neutron induced fission was obtained for both of the fission sensors. Based on these calculated ratios, the correction factors associated with the D. C.

Cook Unit 1 capsules were determined as follows.

Cook Unt 1 Capsule U

6-8 (y,f) Correction Capsule ID Ratio I And Location [U-238(y,f)]/[U-238(n,f)] (1+Ratio)

T (40 Degrees) 0.0439 0.958 X (40 Degrees) 0.0439 0.958 Y (40 Degrees) 0.0439 0.958 U (40 Degrees) 0.0439 0.958 (y,f) Correction Capsule ID Ratio 1 And Location [Np-237(y,f)]/[Np-237(n,f)] (1+Ratio)

T (40 Degrees) 0.0156 0.985 X (40 Degrees) 0.0156 0.985 Y (40 Degrees) 0.0156 0.985 U (40 Degrees) 0.0156 0.985 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 fdr Capsules T, X, Y, and U are given in Table 6-8. In Table 6-8, 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 23SU impurities, plutonium build-in, and gamma ray induced fission effects.

In regard to the data listed in Table 6-8, it should be noted that the reaction rates obtained for the fission monitors from Capsule T were significantly higher than would be expected for this capsule configuration and irradiation history. Similarly, the reaction rates for obtained for the fission monitors from Capsule U were significantly lower than would be expected for this capsule configuration and irradiation history.

These observations are based on comparison of the reaction rates for these fission monitors with data obtained from other 40* surveillance capsule irradiations at 4-loop plants. Because of this, the data for these fission monitors was not used in the least squares analyses of Capsules T and U.

6.3.2 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 O(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, Cook Unit 1 Capsule U

6-9 R,+3R=(o- ) +S) ---

g relates a set of measured reaction rates, Ri, to a single neutron spectrum, 4,, through the multigroup dosimeter reaction cross-section, (Yg, each with an uncertainty 8. 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 D. C. Cook Unit I surveillance capsule dosimetry, the FERRET code 124] 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

(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:

1- 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 D. C. Cook Unit I 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 6.3.1.

The dosimetry reaction cross-sections and uncertainties were obtained from the Sandia National Laboratory Radiation Metrology Laboratory (SNLRML) dosimetry cross-section library'2 1 . The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard El01 8,-"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 D. C. Cook Unit I surveillance capsule sensor sets.

Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes 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 Cook Unit I Capsule U

6-10 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 "5436Cu(n,a)'Co 54 5%

Fe(n,p) Mn 5%

58 Ni(n,p)5 SCo 5%

23 8 U(nf)13 7Cs 10%

2 37 Np(n,f) 13 7Cs 10%

59Co(n,y)60Co 5%

These uncertainties are given at the Ia level.

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 D. C. Cook Unit I surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.

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 Cook Unit I Capsule U

6-11 location. The spectrum for each capsule was 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:

M. =R+Rg *Rg *Pgg where R, 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:

P~g= [R- 015,g, + 0 e'H where 2

(g-g,)

2 2y 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 strengih 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 D. C. Cook Unit I calculated spectra was as follows:

Flux Normalization Uncertainty (R.) 15%

Flux Group Uncertainties (R., Re-)

(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 Cook Unit I Capsule U

I 6-12 (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 6.3.3 Comparisons of Measurements and Calculations Results of the least squares evaluations of the dosimetry from the four D. C. Cook Unit I surveillance capsules withdrawn to date are provided in Tables 6-9 and 6-10. In Table 6-9, measured, calculated, and best-estimate values for sensor reaction rates are given for each capsule. The best estimate values represent the adjusted values resulting from the least squares evaluation of the calculations and measurements. Also provided in this tabulation are ratios of the measured reaction rates to both the calculated and least squares adjusted reaction rates (best estimate). 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 6-10, 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 6-9 and 6-10 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. 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 I Ylevel. From Table 6-10, it is noted that the corresponding uncertainties associated with the least squares adjusted exposure parameters have been reduced to 6-7%

for neutron flux (E > 1.0 MeV) and 7-9% for iron atom displacement rate. Again, the uncertainties from the least squares evaluation are at the Ia level.

Further comparisons of the measurement results with calculations are given in Tables 6-11 and 6-12.

These comparisons are given on two levels. In Table 6-1I, calculations of individual threshold sensor 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 6-12, calculations of fast neutron exposure rates in terms of cO(E > 1.0 MeV) and dpals are compared with the best estimate results obtained from the least squares evaluation of the four capsule dosimetry results. These two levels of comparison yield consistent and similar results with all Cook Unit I Capsule U

6-13 measurement-to-unadjusted 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, only one of the measurement foil reaction rates differs from the corresponding calculated value by more than 20%

(Cu-63 from Capsule T). The M/C comparisoni for the other fifteen fast neutron reactions range from 0.89-1.17. The overall average M/C ratio for the entire set of D. C. Cook Unit 1 data is 1.06 with an associated standard deviation of 8.7%.

In the comparisons of best estimate and calculated fast neutron exposure parameters, the corresponding BE/C comparisons for the four capsule data set range from 0.93-1.11 for neutron flux (E > 1.0 MeV) and from 0.93 to 1.08 for iron atom displacement rate. The overall average BE/C ratios for neutron flux (E > 1.0 MeV) and iron atom displacement rate are 1.03 with a standard deviation of 7.2% and 1.02 with a standard deviation of 6.5%, respectively.

Based on these comparisons, it is concluded that the calculated fast neutron exposures provided in Section 6.4 of this report are validated for use in the assessment of the condition of the materials comprising the beltline region of the D. C. Cook Unit I reactor pressure vessel.

The uncertainty associated with the calculated neutron exposure of the D. C. Cook Unit I 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.

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 D. C. Cook Unit I 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'aipply 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 Cook Unit I Capsule U

6-14 uncertainty applicable to the D. C. Cook Unit 1 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with D. C. Cook Unit 1 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. It should be noted that the measured reaction rates and adjusted values of neutron flux (E > 1.0 MeV) and iron atom displacement rate have been used only to validate the calculated results and associated calculational uncertainty. They have not been used to modify the calculated results in any way.

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 9.

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 provided in Tables 6-11 and 6-12 support these uncertainty assessments for D. C. Cook Unit 1.

6.4 Projections of Reactor Vessel Exposure The final results of the fluence evaluations performed for the four surveillance capsules withdrawn from the D. C. Cook Unit 1 reactor are provided in Table 6-13. These assigned neutron exposure levels are based on the plant and fuel cycle specific neutron transport calculations performed for the D. C. Cook Unit 1 reactor. As shown by the comparisons provided in Tables 6-11 and 6-12, the validity of these calculated fluence levels is demonstrated both by a direct comparison with measured sensor reaction rates as well by comparison with the least squares evaluation performed for each of the capsule dosimetry sets.

The corresponding calculated fast neutron fluence (E > 1.0 MeV) and dpa exposure values for the D. C. Cook Unit 1 pressure vessel are provided in Table 6-14. As presented, these data represent the maximum exposure of the clad/base metal interface at azimuthal angles of 0, 15, 30, and 45 degrees relative to the core cardinal axes. The data tabulation includes the plant and fuel cycle specific calculated fluence at the end of cycle sixteen (the last cycle completed at the D. C. Cook Unit 1 plant at the time this analysis was performed), a projection to the end of cycle seventeen (the current operating cycle at the time this analysis was performed) and further projections for future operation to 25, 32, 36, 48, and 54 effective full power years.

Cook Unit 1 Capsule U

6-15 The projection to the completion of cycle 17 was based on the cycle 17 design power distribution provided in Appendix A, continued operation at a core power level of 3250 MWt, and a design cycle length of 1.45 effective full power years. Projections beyond the end of cycle 17 were based on the a

assumption that future operation would continue to make use of low leakage fuel management and that representative power distribution burnup averaged over cycles 15 through 17 would be typical of future operating cycles. In addition, to provide a degree of conservatism in the projected fluence, a positive bias of 10% was applied to the neutron source in all fuel assemblies located on the core periphery. It was further assumed that, for cycles 18 and beyond, the core power level would be uprated to 3600 MWt.

Therefore, the fluence projections for future operation at D. C. Cook Unit 1 are based on the assumption of a constant neutron flux at the surveillance capsule and pressure vessel locations for the operating as period between 16.68 and 54 effective full power years. As required by Regulitory Guide 1.190 and mentioned in the previous section, no bias or uncertainty is applied to the results.

Updated lead factors for the D. C. Cook Unit 1 surveillance capsules are provided in Table 6-15. 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 clad/base metal interface. In Table 6-15, the lead factors for capsules that have been withdrawn from the reactor (T, X, Y, and U) were based on the calculated fluence values for the irradiation period corresponding to the time of withdrawal for the individual capsules.

Cook Unit I Capsule U

6-16 Table 6-I Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance Capsule Center Neutrons (E > 1.0 MeV)

Total Neutron Flux (E > 1.0 MeV) Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation [n/cm 2 -s] [n/cm 2]

Length Time Cycle [EFPY] [EFPY] 4 Degrees 40 Degrees 4 Degrees 40 Degrees 1 1.27 1.27 2.099E+10 6.674E+10 8.386E+17 2.667E+18 2 0.78 2.04 2.453E+10 8.360E+I0 1.440E+18 4.718E+18 3 0.70 2.75 2.349E+ 10 8.069E+ 10 1.963E+ 18 6.512E+ 18 4 0.73 3.48 2.270E+10 7.776E+ 10 2.488E+ 18 8.313E+ 18 5 0.74 4.22 2.287E+10 8.106E+10 3.023E+ 18 1.021 E+ 19 6 0.72 4.95 2.290E+10 7.634E+10 3.545E+18 1.195E+19 7 0.73 5.67 2.301E+I0 7.689E+10 4.072E+ 18 1.371E+19 8 1.12 6.80 2.266E+10 4.301E+10 4.876E+18 1.524E+19 9 1.19 7.98 2.183E+10 4.214E+10 5.696E+18 1.682E+19 10 1.19 9.17 1.936E+10 4.151E+10 6.422E+18 1.837E+19 11 1.14 10.32 1.934E+10 4.215E+10 7.119E+ 18 1.989E+19 12 1.19 11.51 1.844E+10 4.306E+10 7.812E+18 2.151E+19 13 1.18 12.68 1.612E+10 4.472E+10 8.411E+18 2.317E+19 14 1.04 13.72 1.480E+10 4.146E+10 8.897E+18 2.453E+19 15 1.16 14.88 1.466E+10 5.711E+10 9.433E+ 18 2.662E+19 16 0.35 15.23 1.341E+10 5.220E+10 9.579E+18 2.719E+19 17 1.45 16.68 1.592E+10 4.815E+10 1.031E+19 2.940E+19 Cook Unit 1 Capsule U

6-17 Table 6-1 cont'd Calculated Neutron Exposure Rates and Integrated Exposures at the Surveillance Capsule Center IRON ATOM DISPLACEMENTS Total Displacement Rate Displacements Cycle Irradiation [dpa/s] [dpa]

Length Time Cycle [EFPY] [EFPY] 4 Degrees 40 Degrees 4 Degrees 40 Degrees 1 1.27 1.27 3.381E-!1 1.125E-10 1.351E-03 4.494E-03 2 0.78 2.04 3.952E-11 1.411E-10 2.320E-03 7.955E-03 3 0.70 2.75 3.784E- II 1.362E-10 3.162E-03 1.098E-02 4 0.73 3.48 3.658E-1 I 1.312E- 10 4.009E-03 1.402E-02 5 0.74 4.22, 3.685E-11 1.368E- 10 4.87 1E-03 1.722E-02 6 0.72 4.95 3.690E-11 1.288E- 10 5.712E-03 2.015E-02 7 0.73, 5.67 3.707E-11 1.297E- I0 6.560E-03 2.313E-02 8 1.12 6.80 3.649E-1 I 7.225E-I I 7.856E-03 2.569E-02 9 1.19 7.98 3.514E-1 I 7.075E- I1 9.174E-03 2.834E-02 10 1.19 9.17 3.119E-1I 6.973E- 11 1.034E-02 3.096E-02 11 1.14 10.32 3.114E-J1 7.075E-11 1.147E-02 3.35 1E-02 12 1.19 11.51- 2.964E- 11 7.209E- 11 1.258E-02 3.622E-02 13 1.18 12.68 2.591E-11 7.489E-11 1.354E-02 3.900E-02 14 1.04 13.72 2.379E-11 6.939E-11 1.432E-02 4.128E-02 15 1.16 14.88 2.358E-11 9.585E-11 1.519E-02 4.479E-02 16 0.35 15.23 2.157E-11 8.751E-11 1.542E-02 4.574E-02 17 1.45 16.68 2.560E-1 1 8.076E-I I 1.659E-02 4.944E-02 Cook Unit I Capsule U

I 6-18 Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface Total Neutron Flux (E > 1.0 MeV) [n/cm 2-s]

Cycle Irradiation Length Time Cycle [EFPY] [EFPY] 0 Degrees 15 Degrees 30 Degrees 45 Degrees I 1.27 1.27 6.258E+09 9.924E+09 1.254E+ 10 1.901E+l0 2 0.78 2.04 7.406E+09 1.195E+10 1.491E+10 2.416E+I0 3 0.70 2.75 6.983E+09 1.109E+10 1.413E+10 2.292E+I0 4 0.73 3.48 6.755E+09 1.089E+10 1.375E+10 2.212E+10 5 0.74 4.22 6.820E+09 1.090E+10 1.410E+10 2.312E+10 6 0.72 4.95 6.826E+09 1.095E+10 1.366E+10 2.173E+10 7 0.73 5.67 6.856E+09 1.102E+10 1.374E+10 2.188E+10 8 1.12 6.80 6.798E+09 9.609E+09 9.233E+09 1.237E+ 10 9 1.19 7.98 6.663E+09 8.406E+09 8.819E+09 1.222E+10 10 1.19 9.17 5.725E+09 9.229E+09 9.344E+09 1.182E+10 11 1.14 10.32 5.822E+09 8.829E+09 9.465E+09 1.206E+ 10 12 1.19 11.51 5.478E+09 8.195E+09 8.953E+09 1.210E+10 13 1.18 12.68 4.791E+09 7.179E+09 9.018E+09 1.256E+10 14 1.04 13.72 4.387E+09 6.675E+09 8.524E+09 1.159E+10 15 1.16 14.88 4.500E+09 7.845E+09 1.074E+10 1.679E+10 16 0.35 15.23 3.985E+09 6.423E+09 9.334E+09 1.480E+ 10 17 1.45 16.68 4.829E+09 7.703E+09 9.378E+09 1.402E+ 10 Cook Unit I Capsule U

6-19 Table 6-2 cont'd Calculated Azimuthal Variation of Maximum Exposure Rates And Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface 2

Total Neutron Fluence (E > 1.0 MeV) [n/cm ]

Cycle Irradiation Length Time Cycle [EFPY] [EFPY] 0 Degrees 15 Degrees 30 Degrees 45 Degrees 1 1.27 1.27 2.501E+17 3.965E+17 5.009E+17 7.596E+17 2 0.78 2.04 4.299E+17 6.868E+17 8.631E+17 1.346E+18 3 0.70 2.75 5.851E+17 9.333E+17 1.177E+18 1.856E+18 4 0.73 3.48 7.416E+17 1.185E+18 1.496E+18 2.368E+18 5 0.74 4.22 9.01]E+17 1.440E+18 1.826E+18 2.909E+18 6 0.72 4.95 1.057E+18 1.690E+18 2.137E+18 3.404E+18 0.73 5.67 1.214E+18 1.942E+18 2.451E+18, 3.905E+18

.7 8 1.12 6.80 1.455E+18 2.283E+18 2.779E+18 4.343E+18 9 1.19 7.98 1.705E+18 2.598E+18 3.110E+18 4.802E+18 10 1.19 9.17 1.919E+18 2.944E+18 3.460E+18- 5.245E+18 11 1.14 10.32 2.129E+18 --3.263E+18 3.802E+18 5.680E+18 12 1.19 11.51 2.334E+18 3.569E+18 4.136E+18 6.132E+18 13 1.18 12.68 2.512E+18 3.836E+18 4.471E+18 6.599E+18 14 1.04 13.72 2.656E+18 4.055E+18 4.751E+18 6.980E+18 15 1.16 14.88 2.814E+18 4.330E+18 5.128E+18 7.568E+18 16 0.35 15.23 2.857E+18 4.400E+ 18 5.229E+18 7.730E+18 17' 1.45 16.68 3.071E+18 4.741E+18 5.645E+18 8.351E+18 Note: At the end of Cycle 17, the maximum fast (E > 1.0 MeV) neutron fluences at the pressure vessel wall occur at an axial elevation 15.2 cm above the midplane of the active fuel for the 00, 15', 30', and 450 azimuths.

Cook Unit 1 Capsule U

6-20 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 Total Iron Atom Displacement Rate [dpa/s]

Cycle Irradiation Length Time Cycle [EFPY] [EFPY] 0 Degrees 15 Degrees 30 Degrees 45 Degrees 1 1.27 1.27 1.014E-11 1.588E-I I 2.020E-11 3.068E-I1 2 0.78 2.04 1.200E-11 1.913E-I1 2.406E-11 3.900E- 11 3 0.70 2.75 1.131E-1I 1.774E-I1 2.279E-11 3.699E-l1 4 0.73 3.48 1.094E-11 1.742E-II 2.219E-I 1 3.569E- 11 5 0.74 4.22 1.105E-1 1 1.745E-I I 2.275E- II 3.730E-! I 6 0.72 4.95 1.106E-I I 1.752E- I I 2.204E-I I 3.508E- 11 7 0.73 5.67 1.111E-1I 1.763E-I I 2.216E- I1 3.531E-11 8 1.12 6.80 1.099E-11 1.534E-II 1.487E- 1 1.996E-11 9 1.19 7.98 1.076E-11 .344E- I I 1.420E- I I 1.972E-I I 10 1.19 9.17 9.286E-12 1.474E- I1 i.505E- I I 1.909E-11 11 1.14 10.32 9.426E-12 1.411E-I 1 1.523E-11 1.948E-11 12 1.19 11.51 8.848E-12 1.307E-11 1.438E-11 1.948E-11 13 1.18 12.68 7.748E-12 1.147E-I I 1.449E-11 2.024E-11 14 1.04 13.72 7.097E-12 1.067E-11 1.369E-11 1.867E-11 15 1.16 14.88 7.290E-12 1.253E-1I 1.726E- 11 2.701E-11 16 0.35 15.23 6.453E-12 1.027E-1I 1.501E-11 2.382E-I 1 17 1.45 16.68 7.817E-12 1.230E-1 I 1.508E-I1 2.257E-11 Cook Unit 1 Capsule U

6-21 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 Total Iron-Atom Displacements [dpa]

Cycle Irradiation Length Time Cycle [EFPY] [EFPY] 0 Degrees 15 Degrees 30 Degrees 45 Degrees 1 1.27 1.27 4.051 E-04 6.347E-04 8.073E-04 1.226E-03 2 0.78 2.04 6.964E-04 1.099E-03 1.391E-03 2.173E-03 3 0.70 2.75 9.478E-04 1.494E-03 1.898E-03 2.995E-03 4 0.73 -3.48 1.201E-03 1.897E-03 2.412E-03 3.822E-03 5 0.74 4.22 1.460E-03 2.305E-03 2.944E-03 4.694E-03 6 0.72 4.95 1.712E-03 2.704E-03 3.446E-03 5.493E-03 7 0.73 5.67 1.966E-03 3.108E-03 3.954E-03 6.302E-03 8 1.12 6.80 2.356E-03 3.652E-03 4.481E-03 7.01OE-03 9 1.19 7.98 2.760E-03 4.156E-03 5.014E-03 7.750E-03 10 1.19 9.17 3.108E-03 4.710E-03 5.578E-03 8.466E-03 11 1.14 10.32 3.448E-03 5.219E-03 6.128E-03 9.169E-03 12 1.19 11.51 3.779E-03 5.707E-03 6.665E-03 9.896E-03 13 1.18 12.68 4.067E-03 6.134E-03 7.204E-03 1.065E-02 14 1.04 13.72 4.300E-03 6.484E-03 7.653E-03 1.1 26E-02 15 1.16 14.88 4.555E-03 6.924E-03 8.259E-03 1.221E-02 16 0.35 15.23 4.626E-03 7.036E-03 8.422E-03 1.247E-02 17 1.45 16.68 4.973E-03 7.581E-03 9.091E-03 1.347E-02 Note: At the end of Cycle 17, the maximum iron atom displacements at the pressure vessel wall occur at an axial elevation 15.2 cm above the midplane of the active fuel for the 00, 150, 30° and 450 azimuths.

Cook Unit 1 Capsule U

6-22 Table 6-3 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV) within the Reactor Vessel Wall RADIUS AZIMUTHAL ANGLE (cm) 00 150 300 450 220.35 1.000 1.000 1.000 1.000 225.87 0.544 0.546 0.551 0.540 231.39 0.262 0.263 0.267 0.256 236.90 0.121 0.121 0.124 0.116 242.42 0.056 0.054 0.056 0.049 Note: Base Metal Inner Radius = 220.35 cm Base Metal 1/4T = 225.87 cm Base Metal 1/2T = 231.39 cm Base Metal 3/4T = 236.90 cm Base Metal Outer Radius = 242.42 cm Table 6-4 Relative Radial Distribution of Iron Atom Displacements (dpa) within the Reactor Vessel Wall RADIUS AZIMUTHAL ANGLE (cm) 00 150 300 450 220.35 1.000 1.000 1.000 1.000 225.87 0.641 0.639 0.652 0.638 231.39 0.395 0.390 0.406 0.387 236.90 0.238 0.235 0.249 0.227 242.42 0.136 0.133 0.142 0.117 Note: Base Metal Inner Radius = 220.35 cm Base Metal 1/4T = 225.87 cm Base Metal 1/2T = 231.39 cm Base Metal 3/4T = 236.90 cm Base Metal Outer Radius = 242.42 cm Cook Unit I Capsule U

6-23 Table 6-5 Nuclear Parameters Used in the Evaluation of Neutron Sensors Target 90% Response Fission Atom RANGE Product Yield Monitor Reaction of Fraction (MEV) Half-life (17)

Material Interest 63Cu (n,ca) 0.69 17 " 4.9-11.8 5.271 y Copper Iron ""4Fe (n,p) 0.0585 2.1-8.3 312.3 d 5"8Ni 0.6808 1.5-8.1 70.82 d Nickel (n,p) 23 SU (n,) 0.9996 1.2-6.7 30.07 y 6.02 Uranium-238 237 1.0000 0.4-3.5 30.07 y Neptunium-237 Np (n,f) 6.17 59Co 0.0015 non-threshold 5.271 y Cobalt-Aluminum (ny)

Note: The 90% response range is defined such that, in the neutron spectrum characteristic of the D. C. Cook Unit I 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.

Cook Unit 1 Capsule U

6-24 Table 6-6 Monthly Thermal Generation during the First Ten Fuel Cycles of the D. C. Cook Unit I Reactor (Reactor Power of 3250 MWt)

Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MWt-hr) Year Month (MWt-hr) 75 I 1,040 i 1 78 1,923,060 81 1,894,240 75 2 18,300 78 2 1,953,625 81 2 2,176,463 75 3 572,000 78 3 2,328,300 81 3 2,411,013 75 4 1,627,271 78 4 438,253 81 4 2,336,161 75 5 1,933,490 78 5 0 81 5 2,254,975 75 6 1,702,931 78 6 230,413 81 6 0 75 7 670,334 78 7 2,256,038 81 7 0 75 8 1,849,116 78 8 2,226,565 81 8 1,790,135 75 9 1,883,324 78 9 2,147,300 81 9 2,333,230 75 10 1,547,637 78 10 2,290,261 81 10 2,411,900 75 11 964,571 78 11 2,306,615 81 11 1,279,657 75 12 1,823,818 78 12 1,681,168 81 12 2,248,410 76 1 1,682,310 79 1 2,299,406 82 1 1,569,517 76 2 1,684,470 79 2 2,091,567 82 2 0 76 3 1,871,720 79 3 2,139,350 82 3 2,024,073 76 4 920,020 79 4 411,763 82 4 2,243,660 76 5 1,099,840 79 5 0 82 5 2,192,908 76 6 2,278,280 79 6 0 82 6 2,329,332 76 7 1,850,650 79 7 762,273 82 7 145,572 76 8 2,400,200 79 8 2,396,428 82 8 0 76 9 1,625,940 79 9 2,313,408 82 9 12,755 76 10 2,417,250 79 10 2,067,631 82 10 1,965,827 76 11 1,913,200 79 11 1,522,934 82 11 2,275,411 76 12 1,740,158 79 12 1,723,754 82 12 2,180,316 77 I 0 80 1 980,650 1 83 2,371,584 77 2 122,284 80 2 2,209,403 83 2 2,154,510 77 3 1,854,264 80 3 2,410,741 83 3 2,215,613 77 4 1,690,059 80 4 2,212,354 83 4 2,298,454 77 5 1,361,490 80 5 2,305,783 83 5 2,075,555 77 6 1,206,310 80 6 0 83 6 2,182,703 77 7 1,553,625 80 7 0 83 7 943,698 77 8 1,746,161 80 8 1,501,561 83 8 0 77 9 1,255,066 80 9 2,100,551 83 9 0 77 10 1,594,097 80 10 2,312,061 83 10 323,110 77 11 1,051,487 80 11 2,293,722 83 11 1,476,269 77 12 2,068,940 80 12 1,833,874 83 12 737,558 Note: Monthly power generation data were obtained from NUREG-0020. *Licensed Operating Reactors Status Summary Report." for the time period spanning January 1975 though March 1989.

Cook Unit 1 Capsule U

6-25 2,,

Table 6-6 Cont'd Monthly Thermal Generation during the First Ten Fuel Cycles of the D. C. Cook Unit I Reactor (Reactor Power of 3250 MWt)

Thermal Thermal Generation' .Generation (MWt-hr) (Mwt-hr),

Year Month Year Month 1 1,743,010 87 2,149,465 84 2 1,940,152 87 2 1,913,061 84 3 2,279,795 87 3 2,099,286 84 4 1,855,391 87 4 969,378 84 5 2,398,744 87 5 2,056,583 84 6 1,861,180 87 6 1,635,814 84 2,096,975 87 7 0 84 7 1,349,657 87 8 0 84 8 2,229,693 87 9 0 84 9 10 2,030,349 87 10 1,326,293 84 11 2,187,769 87 11 2,112,159 84 12 2,115,145 87 12 2,096,851 84 1 699,676 88 1 2,076,972 85 2 2,075,841 88 2 1,997,614 85 3 2,301,237 88 3 2,007,159 85 4 341,767 88 4 2,089,788 85 5 0 88 5 2,156,579 85 6 0 88 6 1,987,237 85 7 0 88 7 2,182,564 85 8 0 88 8 2,300,287 85 9 0 .88 9 1,448,245 85 10 0 88 10 1,854,946 85 11 297,751 88 11 1,922,271 85 12 1,109,907 88 12 2,183,754 85 1 2,182,101 89 1 1,846,055 86 2 1,967,273 89 2 1,563,768 86 3 2,180,320 89 3 723,239 86 86 4 1,901,472 86 5 2,110,806 86 6 0 86 7 819,019 86 8 1,986,925 86 9 2,108,718 86 10 2,181,226 86 11 2,004,755 86 12 2,194,212 Reactors Status Note:* Monthly power generation data were obtained from NUREG-0020. "Licensed Operating Summary Report," for the time period spanning January 1975 though March 1989.

Cook Unit I Capsule U

6-26 Table 6-7 Calculated O(E > 1.0 MeV) and Cj Factors at the Surveillance Capsule Center Core Midplane Elevation Fuel O(E > 1.0 MeV) [n/cm2 -s] C Cycle Capsule T Capsule X Capsule Y Capsule U T X Y U I 6.674E+10 6.674E+10 6.674E+10 6.674E+I0 1.000 0.882 0.872 1.052 2 8.360E+10 8.360E+10 8.360E+10 1.105 1.092 1.317 3 8.069E+I 0 8.069E+10 8.069E+ 10 1.067 1.054 1.272 4 7.776E+10 7.776E+10 7.776E+10 1.028 1.016 1.225 5 8.106E+10 8.106E+10 1.059 1.277 6 7.634E+10 7.634E+10 0.997 1.203 7 7.689E+10 1.212 8 4.301E+10 0.678 9 4.214E+10 0.664 10 4.151E+10 0.654 Average 6.674E+10 7.564E+10 7.655E+10 6.345E+10 1.000.00000 1.000 1.000 Cook Unit I Capsule U

6-27 1

Table 6-8 Measured Sensor Activities and Reaction Rates Surveillance Capsule T Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate Reaction Location (dpslgi) (dps/g) (dns/g) (rps/atom) 63 Cu (nc') 60 CO Top Middle 5.140E+04 3.435E+05 3.280E+05 5.004E- 17 Middle 5.270E+04 3.521E+05 3.363E+05 5.130E-17 Bottom Middle 6.040E+04 4.036E+05 3.854E+05 5.880E-17 Average 5.338E-17 54 5.616E-15 Fe (n,p) 54Mn 'Top 1.930E+06 3.374E+06 3.543E+06 Top Middle 1.690E+06 2.955E+06 3.102E+06 4.91 8E-15 Middle 1.690E+06 2.9555E+06 3.102E+06 4.918E-15 Bottom Middle 1.690E+06 2.955E+06 3.102E+06 4.91 8E- 15 Bottom 1.800E+06 3.147E+06 3.304E+06 5.238E-15 Average 5.122E-15 5 Top Middle 3.830E+07 4.511 E+07 5.220E+07 7.473E-15 "Ni (n,p) 1Co Middle 3.770E+07 4.441E+07 5.138E+07 7.356E-15 Bottom Middle 3.950E+07 4.653E+07 5.383E+07 7.707E- 15 Average 7.512E-15 238U (n,f) 37Cs (Cd) 4.190E+07 4.190E+07 ,,2.751E-13 Middle 1.200E+06, 23 8 U (n,f) 137Cs (Cd) Including 23SU, 2 39 pu, and yfission corrections: 2.303E-13 27Np (n,f) 137Cs (Cd) Middle 4.530E+06 1.582E+08 1.582E+08 1.009E-12 237 9.935E-13 Np (n,f) 137Cs (Cd) Including yfission correction:

Top 3.254E+ 10 3.170E+10 3.102E-12 59

" Co (n,y) 6°Co 4.870E+09 Bottom 5.030E+09 3.361E+10 3.274E+ 10 3.204E-12 Average 3.153E-12 5"9Co (n,y) Top 1.830E+09 1.223E+10 1A09E+10 1.379E-12 6°Co (Cd)

Bottom 1.640E+09 1.096E+ 10 1.262E+10 1.235E-12 Average 1.307E-12 Note: Measured specific activities are corrected to a shut down date of December 23, 1976.

Cook Unit 1 Capsule U

I 6-28 Table 6-8 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 (d10s/g) 3p.30E+ (dps38) (rps/atom) 63 Cu (n,a) 6°Co Top Middle 1.140E+05 3 .370E+05 3.218E+05 4.910E-17 Middle I. 150E+05 3.399E+05 3.247E+05 4.953E-17 Bottom Middle 1.180E+05 3.488E+05 3.331 E+05 5.082E- 17 Average 4.982E-17 54 Fe (n,p) 54 Mn Top 2.21 OE+06 3.01 OE+06 3.160E+06 5.01OE-15 Top Middle 2.280E+06 3.105E+06 3.260E+06 5.169E- 15 Middle 2.230E+06 3.037E+06 3.189E+06 5.055E-15 Bottom Middle 2.340E+06 3.187E+06 3.346E+06 5.305E-15 Bottom 2.41 OE+06 3.282E+06 3 446E+06 5.463E- 15 Average 5.200E-15 58 Ni (n,p) 5"Co Top Middle 3.980E+07 4.471 E+07 5.173E+07 7.406E- 15 Middle 4.020E+07 4.516E+07 5.225E+07 7.481 E-15 Bottom Middle 4.180E+07 4.696E+07 5.433E+07 7.778E-15 Average 7.555E-15 238 U (n,f) 13 7 Cs (Cd) Middle 238 3 640E+05 4.795E+06 4.795E+06 3.148E-14 U (n,f) 13 7 Cs (Cd) Including 2 3 5U, 2 39 pu, and 7,fission corrections: 2.572E-14 237 Np (n,f) 37 1 Cs (Cd) 7 Middle 2.570E+06 3.385E+07 3.385E+07 2.160E-13 21 Np (n,f) 137Cs (Cd) Including yfission correction: 2.127E-13 "59Co (n,y) 6°Co Top 1.440E+10 4.257E+10 4.146E+10 4.057E-12 Bottom 1.400E+I0 4.139E+10 4.031E+10 3.945E-12 Average 4.001E-12 "59Co (n,y) 6°Co (Cd) Top 5.520E+09 1.632E+10 1.880E+10 1.840E-12 Bottom 5.530E+09 1.635E+10 1.883E+10 1.843E-12 Average 1.842E-12 Note: Measured specific activities are corrected to a shut down date of May 30, 1980.

Cook Unit I Capsule U

6-29 Table 6-8 cont'd Measured Sensor Activities and Reaction Rates Surveillance Capsule Y Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate (dps35) (dpsW~ (rps/atom)

Reaction Location (d1s70) 63Cu 3.280E+05 5.004E- 17 (n,a) 6°Co Top Middle 1.470E+05 3.435E+05 Middle I.480E-i05 3.302E+05 5.038E-1 7 3.458E+05 1.540E+05 3.598E+05 3.436E+05 5.242E- 17 Bottom Middle Average 5.095E-17 54Fe Top 2.320E+06 3.133E+06 3.290E+06 5.215E-15 (n,p) "Mn 2.390E+06 3.228E+06 3.389E+06 5.372E-15 Top Middle 2.41 OE+06 3.255E+06 3.417E+06 5.417E-15 Middle 2.460E+06 3.322E+06 3A88E+06 5.529E-15 Bottom Middle 2.400E+06 3.241 E+06 3.403E+06 5.395E-15 Bottom Average 5.386E-15 58 3.800E+07 4.654E+07 5.385E+07 7.709E-1 5

" Ni (n,p) "8 Co Top Middle 3.81 OE+07 4.666E+07 "5.399E+07 7.729E- 15 Middle 4.030E+07 4.936E+07 5.711 E+07 8.175E-15 Bottom Middle Average 7.871E-15 238U 5.260E+05 4.991E+06 4.991E+06 3.277E-14 (n,f) '37Cs (Cd) Middle 23 239pu, SU (n,f) 37Cs (Cd) Including 235 U, and yfission corrections: 2.630E-14 217Np (n,f) '7Cs (Cd) Middle 3.990E+06 3.786E+07 3.786E+07 2.415E-13 237 Np (n,f) 37Cs (Cd) Including ,,fission correction: 2.378E-13 59Co Top 1.690E+ 10 3.949E+ 10 3.846E+10 3.764E-12 (n,7) 6°Co Bottom 1.660E+ I0 3.879E+10 3.778E+10 3.697E- 12 Average 3.731E-12 59 60 Top 6.800E+09 1.589E+10 1.830E+10 1.791E-12

" Co (nY) CO (Cd)

Bottom 7.070E+09 1.652E+10 1.903E+10 1.862E-12 Average 1.827E-12 Note: Measured specific activities are corrected to a counting date of July 3, 1982.

Cook Unit I Capsule U

6-30 Table 6-8 cont'd Measured Sensor Activities and Reaction Rates Surveillance Capsule U Radially Radially Adjusted Adjusted Measured Saturated Saturated Reaction Activity Activity Activity Rate Reaction Location (dps/g) (dps/g) (dps_) (fps/atom) 6

-Cu (n,a) 6OCo Top Middle 1.300E+05 2.868E+05 2.739E+05 4.179E-17 Middle 1.260E+05 2.780E+05 2.655E+05 4.050E- 17 Bottom Middle 1.330E+05 2.935E+05 2.803E+05 4.275E- 17 Average 4.168E-17 4Fe (n,p) "Mn Top 7.100E+05 2.343E+06 2.460E+06 3.899E- 15 Top Middle 7.340E+05 2.422E+06 2.543E+06 4.031E-15 Middle 7.31 OE+05 2.412E+06 2.533E+06 4.015E-15 Bottom 7. 1OOE+05 2.343E+06 2.460E+06 3.899E-15 Average 3.961E-15 "58Ni (n,p) -SCo Top Middle 2.300E+06 3.543E+07 4.099E+07 5.868E-15 Middle 2.270E+06 3.497E+07 4.045E+07 5.792E- 15 Bottom Middle 2.360E+06 3.635E+07 4.206E+07 6.021E-15 Average 5.894E-15 23 SU (n,f) 137Cs (Cd) Middle 4.490E+05 2.568E+06 2.568E+06 1.686E-14 23 8U (n,f) '37Cs (Cd)

Including 235 U, 23 9 p1 , and yfission corrections: 1.316E-14 237Np (n,f) 137Cs (Cd) 237 Middle 2.870E+06 1.641E+07 1.641E+07 1.047E-13 Np (n,f) '7Cs (Cd)

Including yfission correction: 1.031E-13 "59Co (n,y) 'Co Top 2.060E+07 4.545E+07 4.427E+07 2.888E-12 Bottom 1.930E+07 4.259E+07 4.148E+07 2.706E-12 Average 2.797E-12 59

" Co (n,7) 6OCo (Cd) Top 8.660E+06 1.911 E+07 2.20 1E+07 1.436E-12 Bottom 8.01 OE+06 1.767E+07 2.036E+07 1.328E-12 Average 1.382E-12 Note: Measured specific activities are indexed to a counting date of October 10, 1989.

Cook Unit I Capsule U

6-31 Table 6-9 Comparison of Measured, Calculated, and Best Estimate Reaction Rates at the Surveillance Capsule Center Capsule T Capsule X Reaction Rate [rps/atom]

Best Reaction Measured Calculated Estimate M/C M/BE 63Cu(n,a)6oCo 4.98E-17 4.65E-17 4.91E-17 1.07 1.01 54Fe(n,p)S4Mn 5.20E-15 5.24E-15 5.33E-15 0.99 0.98 58

" Ni(np)"'Co 7.55E- 15 7.23E- 15 7.45E-15 -- 1.04 1.01 238U(n,f) 37 1 Cs (Cd) 2.57E- 14 2.62E- 14 2.65E-14 0.98 0.97 237Np(n,f)137 Cs (Cd) 2.13E-13 2.06E-13 2.11E-13 1.03 1.01 59

". Co(n,'y)6Co 4.OOE- 12 3.06E- 12 3.97E- 12 1.31 1.01 59 Co(n,-y)6°Co (Cd) 1.83E-12 1.60E- 12 1.83E-1 2 1.14 1.00 Cook Unit 1 Capsule U

6-32 Table 6-9 cont'd Comparison of Measured, Calculated, and Best Estimate Reaction Rates at the Surveillance Capsule Center Capsule Y Reaction Rate [rps/atom]

Best Reaction Measured Calculated Estimate M/C M/BE 63Cu(n,a)6°Co 5.09E- 17 4.70E- 17 5.03E- 17 1.08 1.01

-Fe(nP) 54Mn 5.39E- 15 5.30E- 15 5.53E- 15 1.02 0.97 5"8Ni(n,p)5 238U(n,f)137 "Co 7.87E- 15 7.31E-15 7.73E- 15 1.08 1.02 Cs (Cd) 2.63E-14 2.65E-14 2.77E-14 0.99 0.95 237 Np(n,f)137Cs (Cd) 2.38E-13 2.08E- 13 2.28E- 13 1.14 1.04 5 9Co(ny)6°Co 3.73E-12 3.1OE-12 3.71E-12 1.20 1.01 59 Co(n,y)6Co (Cd) 1.83E-12 I1.62E-12 1.83E-12 1.13 1.00 CAPSULE U Reaction Rate [rps/atom]

Best Reaction Measured Calculated Estimate M/C M/BE 63Cu(n,a)6oCo 4.17E- 17 3.98E- 17 4.04E- 17 1.05 1.03 54Fe(n,p)-Mn 3.96E- 15 4.43E- 15 4.16E- 15 0.89 0.95 58

" Ni(n,p) 58 Co 5.89E-15 6.11E-15 5.82E-15 0.96 1.01 59Co(n,y)6°Co 59Co(n,y)6°Co 2.80E-12 2.55E-12 2.79E-12 1.10 1.00 (Cd) 1.38E- 12 1.33E- 12 1.38E-12 1.04 1.00 Cook Unit 1 Capsule U

6-33 Table 6-10 Comparison of Calculated and Best Estimate Exposure Rates at the Surveillance Capsule Center O(E > 1.0 MeV) [n/cm2?-s]

Best Uncertainty Capsule ID Calculated Estimate (la) BE/c T 6.674E+10 7.377E+10 7% 1.105 X 7.564E+10 7.680E+10 6% 1.015 Y 7.655E+10 8.062E+10 6% 1.053 U 6.345E+10 5.899E+10 7% 0.930 Iron Atom Displacement Rate [dpa/s]

Best Uncertainty Capsule ID Calculated Estimate (icy) BE/C T 1.125E-10 1.217E-10 8% 1.082 X 1.276E-10 1.287E-10 7% 1.009 Y 1.291E-10 1.350E-10 7% 1.046 U 1.069E-10 9.929E-11 _ 9% 0.928 Cook Unit I Capsule U

I 6-34 Table 6-11 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios including all Fast Neutron Threshold Reactions M/C Ratio Reaction Capsule T Capsule X Capsule Y Capsule U 63 Cu(n,at)6Co 1.28 1.07 1.08 1.05 "4Fe(n,p) 54Mn 1.10 0.99 1.02 0.89 58

" Ni(n,p) 5"Co 238 37 1.17 1.04 1.08 0.96 U(n,p)1 Cs (Cd) --- 0.98 0.99 237 --

Np(n,f) 137Cs (Cd) __ 1.03 1.14 _

Average 1.19 1.02 1.06 0.97

% Standard Deviation 7.8 3.6 5.6 8.0 Note: The overall average M/C ratio for the set of 16 sensor measurements is 1.06 with an associated standard deviation of 8.7%.

Table 6-12 Comparison of Best Estimate/Calculated (BE/C) Exposure Rate Ratios BE/C Ratio Capsule ID O(E > 1.0 MeV) dpa/s T 1.11 1.08 X 1.02 1.01 Y 1.05 1.05 U 0.93 0.93 Average 1.03 1.02

% Standard Deviation 7.2 6.5 Table 6-13 Calculated Fast Neutron Exposure of Surveillance Capsules withdrawn from D. C. Cook Unit 1 Irradiation Time Fluence (E > 1.0 MeV) Iron Displacements Capsule [EFPY] [n/cm 2] [dpa]

T 1.27 2.667E+ 18 4.494E-03 X 3.48 8.313E+ 18 1.402E-02 Y 4.95 1.195E+19 2.015E-02 U 9.17 1 837E+19 3.096E-02 Cook Unit 1 Capsule U

6-35 Table 6-14 Calculated Maximum Fast Neutron Exposure of the D. C. Cook Unit I Reactor Pressure Vessel at the Clad/Base Metal Interface Neutron Fluence [E > 1.0 MeV]

Cumulative Neutron Fluence [n/cm2 ]

Operating Time

-[EFPY] ý0.0 Degrees 15.0 Degrees 30.0 Degrees 45.0 Degrees 15.23 (EOC 16) 2.857E+18 4.400E+ 18 5.229E+18 7.730E+ 18 16.68 (EOC 17) 3.071E+18 4.741E+18 5.645E+18 8.351E+18 S25.00 4.698E+ 18 7.399E+18 9.096E+18 1.357E+19 32.00 6.069E+ 18 9.652E+18 1.204E+19 1.802E+19 36.00 6.870E+ 18 1.096E+19 1.373E+19 2.059E+19 48.00 9.273E+18 1.489E+19 1.883E+19 -2.831E+19 54.00 1.047E+19 1.685E+19 2.138E+19 3.216E+ 19 Iron Atom Displacements Cumulative Iron Atom Displacements [dpa]

Operating Time

[EFPY] 0.0 Degrees 15.0 Degrees 30.0 Degrees 45.0 Degrees 15.23 (EOC 16) 4.626E-03 7.036E-03 8.422E-03 1.247E-02 16.68 (EOC 17) 4.973E-03 7.581E-03 9.091E-03 1.347E-02 25.00 7.613E-03 1.1 84E-02 1.466E-02 ,2.191 E-02 "32.00 9.835E-03 1.544E-02 1.939E-02 2.908E-02 36.00 1.1 13E-02 1.754E-02 2.212E-02 3.324E-02 48.00 1.503E-02 2.382E-02 3.034E-02 4.570E-02 54.00 1.698E-02 2.697E-02 3.445E-02 5.192E-02 Note: For future projections through 25.00 EFPY, the maximum fast (E >: 1.0 MeV) neutron fluences and iron atom displacements at the pressure vessel wall occur at an axial elevation 15.2 cm above the midplane of the active fuel for the 00, 15', 30' and 450 azimuths. These peaks at the pressure vessel wall shift such that they occur at an axial elevation of 88.4 cm below the midplane of the active fuel at 36.00 EFPY through 54.00 EFPY for all four azimuthal locations. The future projections also account for a plant uprating from'3250 MWt to 3600 MWt at the onset of cycle 18.

Cook Unit I Capsule U

6-36 Table 6-15 Calculated Surveillance Capsule Lead Factors Note:

1. Capsule W was formally located at the 40 position and known as Capsule S. Capsule S was formally known as Capsule W. These changes are documented in Reference 35.

Cook Unit I Capsule U

7-1 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following surveillance capsule removal schedule meets the requirements of ASTM El 85-82 and is recommended for future capsules to be removed from the D. C. Cook Unit 1 reactor vessel.

Table 7-1 D. C. Cook Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule 2)

Capsule Capsule Location Lead Factor Withdrawal EFPY°a) Fluence (n/crm T 400 3.51 1.27 (EOC 1) 0.267 x 1019 X 400 3.51 3.48 (EOC 4) 0.831 x 1019 y 400 3.51 4.95 (EOC 6) 1.195 x 1019 U 400 3.50 9.17 (EOL 10)(c) 1.837 x 10i9 V 40 1.23 Standby(c) EOL)

W~d) 400 1.65 Standby(c) EOL)

Z 40 1.23 Standby(c) EOL(b)

S(d) 40 1.23 Standby(c) EOL(b)

Notes (a) Effective Full Power Years (EFPY) from Plant Startup.

(b) It is recommended to remove the standby capsules at 32 EFPY (EOL) and place the standby's in storage. Alternative fluence measuring techniques must be applied.

It (c) For EOL of 32 EFPY, Capsule U satisfies the criteria for the 4 hand 5th Capsule withdrawal.

should be noted that if license renewal is obtained for 48 EFPY, then Capsule U will no longer satisfy the criteria for the 5"h Capsule withdrawal. This is due to EOL going from 32 to 48 EFPY and the Capsule U fluence will then be less then the peak EOL vessel fluence. With that in mind the four standby Capsules should be withdrawn at EOL (48 EFPY).

(d) Capsule W was formerly located at the 40 position and known as Capsule S. Capsule S was formerly known as Capsule W. These changes were documented in AEP Safety Review Screening Checklist CE-95-0309, dated 9/19/95.

Cook Unit 1 Capsule U

8-1 8 REFERENCES

1. WCAP-8047, American Electric PowerService Corp. Donald C. Cook Unit No. I Reactor Vessel Radiation Surveillance Program,S. E. Yanichko and D. J. Lege, March, 1973.

2. Code of Federal Regulations, 10CFR50, Appendix G, FractureToughness Requirements, U.S.

Nuclear Regulatory Commission, Washington, D.C.

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

4.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix GQ FractureToughness Criteria for ProtectionAgainst Failure,Dated December 1995, through 1996 Addendum.

5.' ASTM E208, StandardTest Methodfor Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of FerriticSteels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.

6. ASTM E185-82, Standard Practicefor ConductingSurveillance Testsfor Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF), in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.

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

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

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

10. NTSD/SI-598/88, "Cook Rerating, AEPSC Input," February 1988.

11. WCAP-1 2483, Revision 0, "Analysis of Capsule U from the American Electric Company D. C. Cook Unit I Reactor Vessel Surveillance Program," January 1990.

12. WCAP-10376, "Core Physics Characteristics of the Donald C. Cook Station Nuclear Plant (Unit 1, Cycle 8)," July 1983.

13. WCAP-10862, "Core Physics Characteristics of the Donald C. Cook Station Nuclear Plant (Unit 1, Cycle 9)," August 1985.

14. WCAP-I 1586, "Nuclear Parameters and Operations Package for the Donald C. Cook Station Nuclear Plant (Unit 1, Cycle 10)," October 1987.

15. WCAP-12153, "Nuclear Parameters and Operations Package for the Donald C. Cook Station Nuclear Plant (Unit 1, Cycle II)," June 1989.

8-2

16. WCAP-12797, "Nuclear Parameters and Operations Package for the Donald C. Cook Nuclear Plant (Unit I, Cycle 12)," December 1990.

17. WCAP- 13482, "Nuclear Parameters and Operations Package for the Donald C. Cook Nuclear Plant (Unit 1, Cycle 13)," August 1992.

18. WCAP-13955, "Nuclear Parameters and Operations Package for the Donald C. Cook Nuclear Plant (Unit ], Cycle 14)," April 1994.

19. WCAP- 14465, Revision 1, "Nuclear Parameters and Operations Package for the Donald C. Cook Nuclear Plant (Unit 1, Cycle 15)," October 1995.

20. WCAP-14852, "Nuclear Parameters and Operations Package for the Donald C. Cook Nuclear Plant (Unit I, Cycle 16)," March 1997.

21. WCAP-15594, "Nuclear Parameters and Operations Package for the Donald C. Cook Nuclear Plant (Unit 1, Cycle 17)," October 2000.

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

23. RSIC Data Library Collection DLC-185, "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.

24. A. Schmittroth, FERRET DataAnalysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

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

26. ASTM E23-86, StandardTest Methodsfor Notched Bar Impact Testing of Metallic Materials

27. ASTM A370-88, StandardTest Methods and Definitionsfor MechanicalTesting of Steel Products

28. ASTM E8-83, StandardTest Methodsfor Tension Testing of Metallic Materials

29. ASTM E21-79 (1988), StandardTest Methodsfor Elevated Temperature Tension Tests of Metallic Materials

30. ASTM E83-67, StandardPracticefor Verification and Classificationof Extensometers

31. ASTM Designation El 85-70, Recommended Practicefor Surveillance Tests on StructuralMaterials in NuclearReactors.

32. "Reactor Vessel Material Surveillance Program for Donald C. Cook Unit No. I Analysis of Capsule T," Final Report SWRI Project 024770, December 1977.

8-3

33. "Reactor Vessel Material Surveillance Program for Donald C. Cook Unit No. I Analysis of Capsule X," Final Report SWRJ Project 02-6159, June 1981.

34. "Reactor Vessel Material Surveillance Program for Donald C. Cook Unit No. I Analysis of Capsule Y," Final Report SWRI Project 06-7244-001, January 1984.

35. AEP Safety Review Screening Checklist CE-95-0309, September 1995.

A-1 APPENDIX A CORE POWER DISTRIBUTIONS USED IN THE TRANSPORT CALCULATIONS FOR D. C. COOK UNIT I

I A-2 Average Radial Core Power Distribution Cycle 1 Relative Power 1.141 1.152 1.151 1.164 1.164 1.186 1.022 0.743 1.154 1.139 1.094 1.148 1.159 1.125 1.009 0.796 1.152 1.099 1.129 1.142 1.134 1.145 0.962 0.682 1.167 1.149 1.143 1.133 1.070 1.067 0.979 0.589 1.170 1.160 1.135 1.070 1.204 0.942 0.854 1.162 1.120 1.146 1.062 0.941 0.997 0.513 1.017 1.006 0.964 0.974 0.842 0.511 0.743 0.792 0.681 0.586 Average Radial Core Power Distribution Cycle 2 Relative Power 1.106 0.823 0.902 0.847 0.872 1.127 1.055 0.973 0.820 0.867 1.094 0.996 0.897 1.121 1.244 0.994 0.902 1.095 0.877 0.895 1.076 0.984 1.203 0.952 0.857 0.999 0.895 0.877 1.044 1.030 1.171 0.736 0.861 0.897 1.074 1.040 1.058 1.104 1.017 1.097 1.112 0.992 1.041 1.106 1.184 0.737 1.025 1.234 1.192 1.161 1.008 0.740 0.970 0.979 0.951 0.733 Average Radial Core Power Distribution Cycle 3 Relative Power 0.951 0.894 1.075 0.939 1.100 0.870 0.930 0.907 0.901 0.940 1.158 1.163 1.108 0.893 1.224 0.922 1.089 1.160 0.950 1.172 1.171 1.076 0.913 0.853 0.947 1.169 1.168 0.971 1.168 0.971 1.157 0.675 1.095 1.109 1.170 1.163 0.955 0.964 0.963 0.848 0.886 1.078 0.965 0.964 1.110 0.683 0.904 1.205 0.912 1.138 0.952 0.684 0.897 0.916 0.847 0.668

A-3 Average Radial Core Power Distribution Cycle 4 Relative Power 0.929 1.077 0.985 1.057 1.061 0.823 0.908 0.870 1.091 0.978 1.152 1.014 1.143 0.951 1.144 0.908 0.991 1.131 1.016 1.187 1.147 1.120 0.923 0.833 1.066 1.024 1.185 1.016 1.151 1.010 1.118, 0.660 1.060 1.129 1.146 1.146 1.089 0.966 0.926 0.855 0.950 1.115 0.999 0.973 1.082 0.678 0.922 1.169 0.927 1.103 0.914 0.667 0.875 0.894 0.835 0.679 Average Radial Core Power Distribution Cycle 5 Relative Power 1.103 0.848 0.951 1.093 0.974 0.860 0.941 0.877 0.853 0.915 1.117 1.027 1.121 0.920 1.176 0.898 0.955 1.117 1.018 1.199 1.119 1.094 0.937 0.824 1.197 1.190 1.178 0.986 1.129 0.657 1.096 1.009 0.964 1.128 1.118 1.178, 0.977 1.069 0.955 0.841 0.918 1.104 0.997 1.066 1.135 0.692 0.945 1.187 0.944 1.117 0.945 0.690 0.877 0.896 0.825 0.668 Average Radial Core Power Distribution Cycle 6 Relative Power 1.019 1.027 1.007 1.149 0.978 0.898 0.917 0.879 1.028 0.925 1.147 1.136- 1.140 0.938 1.187, 0.909 1.002 1.148 0.959 1.135- 1.167 1.067 0.950 0.843 1.147 1.139 1.138 1.006 1.172 1.003 1.143 0.669 0.980 1.140 1.170 1.173 0.938 0.950 0.910 0.949 1.055 0.658 0.876 0.940 1.067 1.002 0.917 1.193 0.951 1.143-, 0.911 0.656 0.879 0.908 0.844 0.673

A-4 Average Radial Core Power Distribution Cycle 7 Relative Power 0.883 1.043 0.962 1.130 0.920 0.871 0.923 0.891 1.041 0.996 1.156 1.107 1.115 0.917 1.190 0.909 0.963 1.153 0.988 1.114 1.176 1.119 0.952 0.853 1.130 1.108 1.114 0.996 1.185 0.993 1.149 0.679 0.919 1.116 1.181 1.189 0.978 0.945 0.912 0.857 0.917 1.121 0.993 0.944 1.057 0.663 0.925 1.191 0.951 1.150 0.912 0.663 0.891 0.909 0.853 0.680 Average Radial Core Power Distribution Cycle 8 0.842 1.027 0.938 1.091 1.142 1.212 0.995 0.957 1.027 1.072 1.193 0.995 1.218 1.008 1.218 0.931 0.938 1.193 1.008 1.186 1.110 1.250 1.129 0.880 1.091 0.995 1.186 0.969 1.148 1.110 1.040 0.421 1.142 1.218 1.110 1.148 1.084 1.110 0.631 1.212 1.008 1.250 1.110 1.110 0.753 0.319 0.995 1.218 1.129 1.040 0.631 0.319 0.957 0.931 0.880 0.421 Average Radial Core Power Distribution Cycle 9 Relative Power 1.022 1.101 1.071 1.237 0.949 1.111 1.109 0.931 1.111 1.163 1.263 0.996 1.248 1.075 1.258 0.920 1 079 1.268 1.006 1.274 1.095 1.278 1.052 0.670 1.265 0.993 1.281 1.027 1.270 1.031 1.049 0.419 1.086 1.282 1.101 1.271 0.921 1.094 0.580 1.141 1.085 1.263 1.025 1.125 0.726 0.281 1.116 1.241 0.946 1.012 0.627 0.293 0.923 0.898 0.625 0.344

A-5 Average Radial Core Power Distribution Cycle 10 Relative Power 0.980 1.041 1.019 0 993 1.036 1.028 0.971 0.641 1.041 1.068 1.240 1.042 1.271 1.161 1.244 0.881 1.013 1.235 1.058 1.262 1.127 1.288 1.106 0.792 0.994 1.045 1.259 1.035 1.267 1.059 1.079 0.396 1.031 1.270 1.121 1.269 1.091 1.124 0.617 1.020 1.164 1.292 1.066-- 1.134 0.577 0.243 0.949 1.232 1.115 1.096 0.630 0.304 0.540 0.854 0.796 0.436 Average Radial Core Power Distribution Cycle 11 Relative Power 0.906 1.236 0.963 1.272, 1.101 1.266 0.961 0.775 1.236 1.081 1.236 1.078- 1.267 0.948 1.186 0.659 0.964 1.236 1.057 1.115 1.091 1.251 1.053 0.764 1.271 1.078 1.117 1.085 1.284 1.157 1.063 0.415 1.097 1.264 1.090 1.286, 1.095 1.173 0.694 1.259 0.942 1.249 1.160 1.175 0.652 0.289 0.975 1.190 1.049 1.063 0.699 0.290 0.814 0.669 0.764 0.411 Average Radial Core Power Distribution Cycle 12 Relative Power 0.999 1.262 1.013 1.260 1.006 1.259 1.173 0.903 1.259 1.005 1.051 1.063 1.257 0.955 1.228 0.718 1.008 1.047 1.021 1.254 1.087 1.249 0.996 0.793 1.241 1.054 1.254 0.961 1.260 1.057 1.072 0.446 0.945 1.247' 1.086 1.263 1.046 1.174 0.730 1.254 0.950 1.248 1.058 1.178 0.749 0.309 1.164 1.224 0.994 1.073 0.735 0.333 0.899 0.715 0.792 0.446

A-6 Average Radial Core Power Distribution Cycle 13 Relative Power 0.919 1.301 1.047 1.120 1.063 1.003 1.267 0.614 1.304 0.967 1.230 1.085 1.230 1.023 1.197 0.691 1.050 1.233 0.970 1.253 1.045 1.071 1.234 0.557 1.121 1.086 1.253 1.087 1.265 1.093 1.127 0.385 1.065 1.231 1.042 1.265 1.072 1.234 0.827 1.008 1.024 1.070 1.094 1.234 0.725 0.334 1.269 1.198 1.234 1.127 0.829 0.340 0.616 0.691 0.557 0.385 Average Radial Core Power Distribution Cycle 14 Relative Power 0.918 1.293 1.018 1.064 1.099 1.084 1.260 0.573 1.296 0.968 1.257 1.047 1.271 1.054 1.193 0.614 1.022 1.260 1.067 1.286 1.150 1.124 1.212 0.514 1.064 1.047 1.284 1.001 1.298 1.042 1.101 0.337 1.099 1.270 1.148 1.296 1.132 1.223 0.814 1.084 1.055 1.125 1.044 1.224 0.695 0.318 1.261 1.195 1.220 1.113 0.817 0.308 0.574 0.616 0.521 0.364 Average Radial Core Power Distribution Cycle 15 Relative Power 0.869 1.230 0.922 1.274 0.917 1.087 1.186 0.578 1.232 0.895 1.195 1.046 1.281 1.111 1.258 0.585 0.922 1.195 0.932 1.269 0.892 1.027 1.250 0.604 1.275 1.045 1.267 0.988 1.012 1.244 1.156 0.410 0.917 1.281 0.892 1.012 1.244 1.105 0.878 1.092 1.111 1.025 1.248 1.104 1.058 0.505 1.177 1.256 1.262 1.175 0.877 0.485 0.535 0.581 0.618 0.482

A-7 Average Radial Core Power Distribution Cycle 16 Relative Power 0.688 0.916 0.970 1.252 1.125 1.120 1.225 0.504 0.909 1.010 1.209 1.039 1.274 1.097 1.218 0.498 0.966 1.208 1.029 1.277 1.120 1.070 1.189 0.513 1.232 1.037 1.279 1.084 1.296 1.089 1.105 0.335 1.122 1.269 1.120 1.301 1.155 1.225 0.887 1.110 1.091 1.068 1.089 1.225 1.057 0.397 1.219 1.214 1.186 1.104 0.887 0.397 0.503 0.496 0.511 0.335 Average Radial Core Power Distribution Cycle 17 Relative Power 0.956 1.162 0.934 1.178 1.223 1.173 1.190 0.577 1.161 0.983 1.278 0.954 0.969 0.975 1.198 0.724 0.933 1.276 0.988 1.177 1.218 1.178 1.251 0.691 1.177 0.954 1.176 1.288 1.016 1.187 1.044 0.411 1.222 0.967 1.214 1.013 1.258 1.097 0.819 1.172 0.974 1.175 1.181 1.088 0.939 0.401 1.190 1.197 1.248 1.034 0.816 0.400 0.577 0.723 0.690 0.409 Radial Core Power Distribution for Fluence Projections Cycles 15 through 17 Average with 10% Peripheral Bias 0.890 1.159 0.934 1.224 1.092 1.133 1.193 0.625 0.951 1.237 1.000 1.128 1.042 1.224 0.707 1.158 1.236 0.971 1.225 1.079 1.106 1.243 0.700 0.933 1.000 1.224 1.147 1.047 1.197 1.205 0.442 1.222 1.091 1.126 1.077 1.047 1.240 1.116 0.935 1.134 1.041 1.104 1.196 1.110 1.099 0.485 1.188 1.222 1.246 1.207 0.933 0.476 0.607 0.705 0.705 0.471

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