ML022100438

WCAP-13515, Revision 1, Analysis of Capsule U from Indiana Michigan Power Company D. C. Cook, Unit 2 Reactor Vessel Radiation Surveillance Program.

ML022100438
Person / Time
Site:	Cook
Issue date:	05/31/2002
From:	Gresham J, Hayes E, Laubham T Westinghouse
To:	Office of Nuclear Reactor Regulation
References
AEP:NRC:2349-01, FOIA/PA-2005-0108 WCAP-13515, Rev 1
Download: ML022100438 (128)

Text

Westinghouse Non-Proprietary Class 3 Analysis of Capsule U from the Indiana Michigan Power Company D. C.

Cook Unit 2 Reactor Vessel Radiation Surveillance Program Westinghouse Electric Company LLC

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-13515, Revision 1 Analysis of Capsule U from the Indiana Michigan Power Company D. C. Cook Unit 2 Reactor Vessel Radiation Surveillance Program T.J. Laubham E. T. Hayes May 2002 Approved: vý Equipment & Matexials Technology Westinghouse Electnc Compamy LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355

©2002 West--nghouse Electric Company L.LC All Rights Reserved Cook Unit 2 Capsule U

111 TABLE OF CONTENTS LIST OF TABLES ..................................................... ....... IV vii LIST OF F IG U R E S ......................................................................................................

P REFA CE ................................................................................................................................... Vill Vi RE C OR D O F R EVIS IO N .....................................................................................................................

ix EX ECUTIVE SU NINL-kRY (O R) ABSTRA CT .......................................................................................

1 SU-MMN,A R Y O F R E S ULT S ...................................................................................................... 1-1 2-1 2 IN TR O D U C T ION ...................................................................................................................

1 3 BA C K G R OU ND ...................................................................................................................

4 DESCRIPTION OF PROGRAM ............................................................ 4-1 5 TESTING OF SPECIM ENS FROM CAPSULE U .................................................................. 5-1 5 .1 O V ERVIE W ................................................................................................................ 5- 1 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS ........................................................ 5-3 5.3 T EN SIL E T E ST RE S ULT S ..........................................................................................

5.4 WEDGE OPENING LOADING SPECIMENS ............. ........... 5-5 6 RADLdTION ANALYSIS AND NEUTRON DOSIMETRY ................................................ 6-1 6 .1 IN T R O D UC T IO N .................................................................................................... 6- 1 6.2 DISCRETE ORDINATES ANALYSIS ........................................... 6-2 6 .3 N E UT R O N D O S IM E T RY ......................................................................................... 6-5 6.4 PROJECTIONS OF REACTOR VESSEL EXPOSURE ............................................ 6-14 7 SURVEILLANCE CAPSULE REMO'VAL SCHEDULE .......................................................... 7-1 8 R E FE R E N C ES ............................................................................................................. .......... 8- 1 AdPPENDLX A LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS ........................... A-0 AkPPENDIX B PHOTOGRAPHS OF CHARPY, TENSILE AND WOL SPECIMENS P RIO R T O T E STIN G B -0 B................................

A.PPENDIC C CORE POWER DISTRIBUTIONS USED IN THE TRANSPORT CALCLULATIONS FOR D.C. COOK UNIT 2 .... .. ......................... C-0 Cook Unit 2 Capsule U

iv LIST OF TABLES Table 4-1 Chemical Composition and Heat Treatment of the D. C. Cook Unit 2 R eactor Vessel Surveillance M aterials .......................................................................... 4-3 Table 5-1 Charpy V-Notch Data for the D. C. Cook Unit 2 Intermediate Shell Plate C5521-2 Irradiated to a Fluence of 1.583 x 10'9 n/cmr (E > 1.0 MeV)

(L o ng itu d in a l Orien tatio n) ............................................................................................ 5 -6 Table 5-2 Charpy V-notch Data for the D. C. Cook Unit 2 Intermediate Shell Plate C552 1-2 Irradiated to a Fluence of 1.583 x 10'9 n/cm 2 (E> 1.0 MeV)

(Tran sverse O rientatio n ) ............................................................................................... 5 -7 Table 5-3 Charpy V-notch Data for the D. C. Cook Unit 2 Surveillance Weld Metal Irradiated to a Fluence of 1.583 x 1019 n/cm 2 (E> 1.0 M eV) .......................................... 5-8 Table 5-4 Charpy V-notch Data for the D- C. Cook Unit 2 Heat-Affected-Zone Material Irradiated to a Fluence of 1.583 x 1019 n/cm" (E> 1.0 MeV) ............................ 5-9 Table 5-5 Instrumented Charpy Impact Test Results for the D. C. Cook Unit 2 Intermediate Shell Plate C5521-2 Irradiated to a Fluence of 1.583 x 10'9 n/cmý (E> 1.0 MeV)

(L o ngitu din al Orientation ) .......................................................................................... 5-10 Table 5-6 Instrumented Charpy Impact Test Results for the D. C. Cook Unit 2 Intermediate Shell Plate C5521-2 Irradiated to a Fluence of 1.583 x 1019 n/cm2 (E> 1.0 MeV)

(Transverse Orientation) ........................................... 5-11 Table 5-7 Instrumented Charpy Impact Test Results for the D. C. Cook Unit 2 Surveillance 2

Weld Metal Irradiated to a Fluence of 1.583 x 10'9 n/cm (E> 1.0MeV) ...................... 5-12 Table 5-8 Instrumented Charpy Impact Test Results for the D. C. Cook Unit 2 Heat-Affected Zone (HAZ) Metal Irradiated to a Fluence of 1.583 x 1019 rVcm2(E> 1.0MeV) ..... 5-13 Table 5-9 Effect of Irradiation to 1.583 x 1019 n/cm 2 (E> 1.0 MeV) on the Notch Toughness Properties of the D. C. Cook Unit 2 Reactor Vessel Surveillance Materials................ 5-14 Table 5-10 Comparison of the D. C. Cook Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99.

R ev isio n 2 , Pred ictio n s ............................. .......................................... 5-15 Cook Unit 2 Capsule U

LIST OF TABLES (Cont.)

Table 5-11 Tensile Properties of the D. C. Cook Unit 2 Reactor Vessel Surveillance Materials Irradiated to 1.583 x 10'9 n/cm (E > 1.0M e'v) ............................................................ 5-16 Table 6-1 Calculated Fast Neutron Exposure Rates and Integrated Exposures Rates at the Surveillance C apsule Center ................................................................................. 6- 6 Table 6-2 Calculated Azimuthal Variation of Fast Neutron Exposure Rates and Iron Atom Displacement Rates at the Reactor Vessel Clad/Base Metal Interface ................. 6-17 Table 6-3 Relative Radial Distribution of 4(E> 1.0 MeV) within the Reactor Vessel Wa ll ....................................................................................................... . . . 6-19 Table 6-4 Relative Radial Distribution of Iron Atom Displacements (dpa) within the Reactor Ve sse l Wa ll .......................................................................................... ..................... 6- 19 Table 6-5 Nuclear Parameters used in the Evaluation of Neutron Sensors .................................. 6-20 Table 6-6 Monthly Thermal Generation During the First Eight Fuel Cycles of the D. C. Cook U nit 2 Reactor (Reactor Pow er of 3411 MWt) ............................................................ 6-21 Table 6-7 Calculated f(E > 1.0 MeV) and Cj Factors at the Surveillance Capsule Center co re M idp lane E levatio n ............................................................................................. 6-2 3 Table 6-8 Measured Sensor Activities and Reaction Rates

- Surv eillance C ap su le T ............................................................................. 6-24

- Surveillance C apsule Y ...................................................................... 6-25

- Surveillance C ap su le X ..................................... .... .................... .......... 6-26

- S urv eillan ce C ap su le U 2................................

6-27 Table 6-9 Comparison of Measured, Calculated and Best Estimate Reaction Rates at the Surveillan ce C apsule C enter ............................................................................. 6-28 Table 6-10 Comparison of Calculated and Best Estimate Integrated Neutron Exposure Rates at the Surveillance Capsule Center ..................................................... 6-30 Table 6-11 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions ...................... ............................... 6-31 Table 6-12 Comparison of Best Estimate/Calculated (BE./C) Exposure Rate Ratios ............... 6-31 Table 6-13 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from D . C. C o o k Un it 2 .......................................................... ...................... ................... 6-31 Cook Unit ' Capsule U

Vi LIST OF TABLES (Cont.)

Table 6-14 Calculated Maximum Fast Neutron Exposure of the D.C. Cook Unit 2 Reactor Pressure Vessel at the Clad/Base M etal Interface ............................................ 6-32 Table 6-15 Calculated Surveillance Capsule Lead Factors ............................................................ 6-33 Table 7-1 D. C. Cook Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedu le ........... ........................................................................................................ 7- 1 Cook Unit 2 Capsule U

.9 vii LIST OF FIGURES Figure 4-1 Arrangement of Surveillance Capsules i the D. C. Cook Unit 2 Reactor Vessel ............ 4-6 Figure 4-2 Capsule U Diagram Showing the Location of Specimens, Thermal Monitors, and D os im ete rs ................................................................................................................. 4 -6 Figure 5-1 Charpy V-Notch Impact Properties for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Longitudinal Orientation) ....................................... 5-17 Figure 5-2 Charpy V-Notch Impact Properties for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation) ......................................... 5-18 Figure 5-3 Charpy V-Notch Impact Properties for D. C. Cook Unit 2 Reactor Vessel We ld M eta l ............................................................................................................... 5- 19 Figure 5-4 Charpy V-Notch Impact Properties for D. C_ Cook Unit 2 Reactor Vessel H eat-A ff ected-Z one M aterial ...................................................................................... 5-20 Figure 5-5 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Longitudinal Orientation) ............................ 5-21 Figure 5-6 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation) ............................... 5-22 Figure 5-7 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 2 Reactor Vesse l W eld M etal .................................................................................................. 5-2 3 Figure 5-8 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 2 Reactor Vessel W eld H eat-A-ffected-Zone M etal ...................................................................... 5-24 Figure 5-9 Tensile Properties for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521 -2 (Transverse O rientation) ..................................................................... 5-25 Figure 5-10 Tensile Properties for D. C. Cook Unit 2 Reactor Vessel Surveillance Weld Metal ....... 5-26 Figure 5-11 Fractured Tensile Specimens from D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C552 1-2 (Transverse Orientation) ....................... 5-27 Figure 5-12 Fractured Tensile Specimens from D. C. Cook Unit 2 Reactor Vessel Weld Metal ........ 5-28 Figure 5-13 Engineering Stress-Strain Curve for Intermediate Shell Plate C5521-2 Tensile Specimen NIT1 1 and MT 12 (Transverse Orientation) .............................................. 5-29 Figure 5-14 Engineenng Stress-Strain Curves for Weld Metal Tensile Specimens MW II and M W 12 .. .. .................................................................................. .5-30 Cook Uni 2 Ccppsule U

.iii viii PREFACE This report has been technically reviewed and verified by-Reviewer:

Sections 1 through 5, 7, 8, Appendices A and B J.H. Ledger -I Section 6 S.L. Anderson cJ' LN )?-L */Ce/Z RECORD OF REVISION Revision 1: In addition to text and font changes, the followxing was changed from Revision 0:

Revised Section 6.0 for updated fluence methodology (Per Regulatory Guide 1 190) and to include the calculated fluence projections. The Heatup and Cooldown Curves, as well as Table 5-7., were removed from this report and incorporated in WCAP-13517 Rev. 1 (for the RTý4D-) and WCAP-15047 Rev. 2 (for the USE). Tables in Section 5.0 were added to account for "Longitudinal" and Transverse" directions individually as opposed to a combined table. Associated text and tables in were updated to reflect the changes in the capsule fluences and vessel fluence projections that were made in Section 6.0.

Cook Unit 2 Capsule U

EXECUTIVE

SUMMARY

The purpose of this report is to document the results of the testing of the surveillance Capsule U specimens and dosimeters from the D. C. Cook Unit 2 reactor vessel. Capsule U was removed at 8.65 EFPY and post irradiation mechanical testing of the Capsule U Charpy V-notch and tensile specimens was performed, along with a fluence evaluation. The surveillance Capsule U fluence was 0 0"58' n/cm- after 8.65 EFPY of plant operation. A brief summary of the Charpy V-notch testing results can be found in Section I and the updated capsule removal schedule can be found in Section 7.

Cook Uint - -,Ipsule Ik

1-1

SUMMARY

OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule U, the fourth capsule to be removed from the D. C. Cook Unit 2 reactor pressure vessel, led to the following conclusions:

The capsule received an average fast neutron fluence (E> 1.0 MeV) of 1.583 x 1019 n/cm: after 8.65 effective full power years (EFPY) of plant operation.

Irradiation of the reactor vessel intermediate shell plate C5521-2 Charpy specuinens, oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal orientation), to 1.583 x 1019 n/cm 2 (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 95'F and a 50 ft-lb transition temperature increase of 110°F. This results in an irradiated 30 ft-lb transition temperature of 120'F and an irradiated 50 ft-lb transition temperature of 165'F for the longitudinally oriented specimens.

Irradiation of the reactor vessel intermediate shell plate C5521-2 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major rolling direction of the plate (transverse orientation), to 1.583 x l0'9 n/cm' (E> 1.0 MNIeV) resulted in a 30 ft-lb transition temperature increase of 130'F and a 50 ft-lb transition temperature increase of 135'F- This results in an irradiated 30 ft-lb transition temperature of 160'F and an irradiated 50 ft-lb transition temperature of 205'F for transversely oriented specimens.

Irradiation of the weld metal Charpy specimens to 1.583 x 1019 n/cm- (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 75'F and a 50 ft-lb transition temperature increase of 40'F. This results in an irradiated 30 ft-lb transition temperature of 85'F and an irradiated 50 ft-lb transition temperature of 1I00 F.

Irradiation of the weld Heat-A-ffected-Zone (I-LAZ) metal Charpy specimens to 1.583 x 10"'n/cm:

(E> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 105"F and a 50 ft-lb transition temperature increase of 110°F. This results in an irradiated 30 ft-lb transition 0

temperature of 45'F and an irradiated 50 ft-lb transition temperature of 80 F.

0 -The average upper shelf energy of the intermediate shell plate C5521-2 (longitudinal orientation) resulted in an average energy decrease of 16 ft-lb after irradiation to 1.583 x 1019 n/cm: (E> 1.0 MeV). This results in an irradiated average upper shelf energy of 111 ft-lb for the longitudinally oriented specimens.

The average upper shelf energy of the intermediate shell plate C5521-2 (transverse orientation) resulted in an average energy decrease of 14 ft-lb after irradiation to 1.583 x 109 n*cm (E> 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 72 ft-lb for the transversely oriented specimens.

The average upper shelf energy of the weld metal Charpy specimens resulted an average energy decrease of 6 ft-lb after irradiation to 1.583 x 10'9 ni'cm: (E> 10 NMeV). Hence, this results in an irradiated average upper shelf energy of 71 ft-lb for the weld metal specimens.

Cook Unit ' Capsule U

1-2 The average upper shelf energy of the weld FILZ metal Charpy specimens resulted in an average energy decrease of 33 ft-lb after irradiation to 1-583 x 1019 n/cm- (E> 1.0MeV). This results in an irradiated average upper shelf energy of 82 ft-lb for the weld HAZ metal.

The surveillance capsule U test results indicate that the intermediate shell plate C5521-2 (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. However, comparison of the 30 ft-lb transition temperature increase for the intermediate shell plate C5521-2 (transverse) is 33'F greater than the Regulatory Guide 1.99 Revision 2 predictions. Regulatory Guide 1.99 Revision 2 requires a 2 sigma allowance of 34°F for base metal to be added to the predicted reference transition temperature to obtain a conservative upper bound value. Thus, the reference transition temperature increase for the intermediate shell plate C552 1-2 (transverse) is bounded by the 2 sigma allowance for shift prediction.

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 10CFR50, Appendix G 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 2 reactor vessel using the Regulatory Guide 1.99, Revision 2 attenuation formula (ie. Equation # 3 in the guide; f(depth,*)

fsuface e*

• X.) is as follows:

Calculated (32 EFPY): Vessel inner radius* = 1.625 x 1019 n/cm 2 Vessel 1/4 thickness = 9. 758 x 10 "n/cm' "vessel 3/4 thickness = 3 519 x 10 " n/cm2 Calculated (48 EFPY): Vessel inner radius* = 2.457 x 1019 n/cm:

Vessel 1/4 thuckness = 1.475 x 1019 n/cm2 "Vessel 3/4 thickness = 5.320 x 10" nicm

• Clad/base metal interface

%ookUnit 2 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 Indiana Michigan Power Company D. C. Cook Unit 2 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the D. C. Cook Unit 2 reactor pressure vessel materials was designed and recorrmended by the Westinghouse Electric Corporation. A description of the surveillance program and the 12, entitled preirradiation mechanical properties of the reactor vessel materials is presented in WCAP-85 "American Electric Power Company Donald C. Cook Unit No. 2 Reactor Vessel Radiation Surveillance Program" by J. A. Davidson, et. alf . The surveillance program was planned to cover the 40-year design 1

life of the reactor pressure vessel and was based on ASTM E 185-73, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels". Westinghouse Power 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 testing of and the post-irradiation data obtained from surveillance Capsules U removed from the Indiana Michigan Power Company D. C. Cook Unit 2 reactor vessel and discusses the analysis of the data.

Cook UCt 2 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 fast reactor pressure vessel is the most critical region of the vessel because it is subjected to sigmficant 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 I (base material of the D. C. Cook Unit 2 reactor pressure vessel shell plate) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.

A method for 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 Code*41. The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RT.,T).

RTT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-2081 61) 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 matenial is used to index that material to a reference stress intensity factor the curve (KI, curve) which appears in Appendix G to the ASME Code. The Kic 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, 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.

RT'NDT 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 such properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program, in which a surveillance capsule is as the D. C. Cook Unit 2 reactor vessel radiation surveillance programill, periodically removed from the operating nuclear reactor and the encapsulated specimens tested. The temperature (ART,,'T) due to irradiation is added to the increase in the average Charpy V-notch 30 ft-lbembrittlement.

initial RTN to adjust the RTO.MT for radiation This RTNIDT (RTN'v initial "- ARTMT) is used to index the matenal to the K;. 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 2 Capsule U

4-1 4 DESCRIPTION OF PROGRAM In accordance with WCAP-8512f'1 eight surveillance capsules for monitoring the effects of neutron exposure on the D. C. Cook Unit 2 reactor pressure vessel core region materials were inserted in the reactor the vessel prior to imtial plant start-up. The eight capsules were positioned in the reactor vessel between thermal shield and the vessel wall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core..

Capsule U was removed after 8.65 Effective Full Power Years (EFPY) of plant operation. This capsule C5521-2 ad contained Charpy V-notch and tensile specimens (Figure 4-2) from the intermediate shell plate Charpy V-notch tensile and wedge load opening (WOL) specimens from submerged arc weld metal specimens representative of that used in the original fabrication. Capsule U also contained Charpy V-notch from from the weld Heat-affected-Zone (HAZ) material. All heat-affected zone specimens were obtained within the I--AZ of the intermediate shell plate C5521-2.

forming Test material obtained from intermediate shell plate C5521-2 (after the thermal heat treatment and test of the plate) were taken at least one forging thickness from the quenched ends of the plate. All after performing a simulated specimens were machined from the ',/4 and '/ thickness locations of the forging post-weld stress-relieving treatment on the test material. Specimens from weld metal and heat-affected and zone metal were machined from a stress-relieved weldment joining intermediate shell plate B8605-2 heat adjacent lower shell plate B8806-1. All heat-affected-zone specimens were obtained from the weld affected-zone of intermediate shell plate B8605-2.

All test specimens were machined from the ',4 thickness location of the plate after a simulated postweld taken stress-relieving treatment on the test material was performed. The test specimens represent material at least one plate thickness from the quenched ends.

Base metal Charpy V-notch impact specimens were machined both in the longitudinal orientation (longitudinal axis of the specimen parallel to the major working direction) and transverse orientation and (longitudinal axis of the specimen perpendicular to the major working direction). Charpy V-notch the specimens transverse tensile specimens from the weld metal were oriented with the longitudinal axis of to the weld direction to the The WOL specimens were machined from the weldment such that the long dimension was parallel propagation in the specimen was weld direction. The notch was machined such that the direction of crack in the welding direction.

The chemical composition and heat treatment of the surveillance material is presented in Table 4-1.

0 Capsule U contained dosimeter wires of pure copper, iron, nickel, and aluminum- .15 weight percent cobalt wire (cadmium-shielded and unshielded). In addition, cadmium shielded dosimeters of neptunium at specific neutron (Np=37 ) and uranium (U, ) were placed in the capsule to measure the integrated flux 28 energy levels.

Cook Unit Cpsue (2 U

4-2 in The capsules contained thermal monitors made from two low-melting-point eutectic alloys and sealed attained by the test Pyrex tubes. These thermal monitors were used to define the maximum temperature points are as specimens dunng irradiation. The composition of the two eutectic alloys and their melting follows:

2.5% Ag, 97.5% Pb Melting Point: 5791F (304'C) 1.75% Ag, 0 75% Sn, 97.5% Pb Melting Point: 590'F (3 10°C)

Cook Unit 2 Capsuie U

4-3 4-3 Table 4-1 Chemical Composition and Heat Treatment of the D. C. Cook Unit 2 Reactor Vessel Surveillance Materials 't Intermediate Shell Plate C5521-2 Weld Metal Element Westinghouse Lukens Steel Westinghouse Chicago Bridge &

Analysis Analysis Analysis Iron Analysis 0.21 0.110 0.08 C 0.220 0.015 0.012 0.016 S 0.014 I I i 0.006 N 0.014 I _________________ .1

- - - 0.032 Co 0.016 0.055 0.05 Cu 0.110 0.14 0.36 0.270 0.16 0.440 Si 0.550 0.50 0.540 Mo 0.96 0.580 0.58 0.970 Ni 1.42 1.280 1.29 1.330 Mn

. - - 0.068 0.07 Cr 0.072

--- 0.001 AV 0.001 0.019 0.017 0.013 0.022 P

.--_ 0.006 Sn 0.013 J

Heat Treatment History Material Temperature ('F) Time (Hr) Coolant Intermediate Shell 1650-1750 4 A, Water Quenched Plate, C5521-2 1550-1650 5 Water Quenched 1200-1300 411, Air Cooled 1150 225 51 1/2 Furnace cooled Weld Metal 1140 _25 9 Furnace Cooled Notesn I. WCAP-8512. Table A-2.

Cook LUnt _2Capsule U

4-4 270' X(220"1 Y (320")"

REACTOR VESSEL THERMAL SHIELD CORE BARREL Figure 4-1 Arrangement of Surveillance Capsules in the D. C. Cook Unit 2 Reactor Vessel Cook Urni 2 Capsule U

4-5 COSIMETER CHARPY CHARPY CHARPY CnARPY ILOCX CHARPY CHARP'I CHARPY CHARPY CMARPY TEmSILE CHARPY CHARPY WOL WOL WOL TENS ILE WOL CW-U2MT-52 H"-771ML-48 mm-7 -146 "t-68 ML-W44 m-66 L-41 MHE 2

,MM w--72 mT-72 Ewis-70MTE-70 HW68 MW-68 I MV-66 MT-66 M*-94 MT-64 CAPSULE U [ýJ.T-7 jj[1 j*jj.9TI, [-EEjjj jjjLuiii 73fj 6j6.

Lj.[L~jjjj-jJ SPECIMEN COOE: MT - PLATE C5521-2 (TRANSVERSE)

ML - PLATE C5521-2 (LONGITUDINAL)

MW - WELD METAL MH - WELD HEAT AFFECTED ZONE Figure 4-2 Capsule I Diagram Showing the Location of Specimens.

Thermal -Monitors. and Dosimeters Cook Umt - ('apsuieU

5-1 5 TESTING OF SPECIMENS FROM CAPSULE U 5.1 OVERvIEW The post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens was performed at the Westinghouse Science and Technology Center hot cell with consultation by Westinghouse 21 Power Systems personnel. Testing was performed in accordance with 10CFR50, Appendices G and HE ,

ASTM Specification E 185-82[6", and Westinghouse Remote Metallographic Facility (RKMF) Procedure RMF 8402, Revision 2 as modified by RIMF Procedures 8102, Revision 1, and 8103, Revision 1.

Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-8512[11. No discrepancies were found.

Examination of the two low-melting point 579°F (304'C) and 590'F (3 10'C) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 579°F (304'C).

The Charpy impact tests were performed per ASTM Specification E23-88(71 and RMF Procedure 8103, Revision 1. on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy impact test machine is instrumented with a GRC 8301 instrumentation system. feeding information into an IBM XT computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED). From the load-time curve (Appendix A), the load of general yielding (Poy). the time to general yielding (ty),.the maximum load (P* 1), and the time to maximum load (t[) 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 (PA), 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 (E,) is the difference between the total energy to fracture (ED) and the energy at maximum load (Ek1).

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

(yy=(Pr*L)1/[B *(W -a)

• C] (1) where: L distance between the specimen supports in the impact machine B = the width of the specimen measured parallel to the notch W height of the specimen, measured perpendicularly to the notch a = notch depth The constant C is dependent on the notch flank angle (4), notch root radius (p) and the type of loading (i-e.. pure bending or three-point bending). In three-point bending, for a Charpy specimen in which 4ý= 45' and p = 0 010 inch, Equation t is valid with C = J.21. Therefore, (for L = 4W),

Cook Umit 2 Capsule U

5-2 o7,=(Pr*L)/[B(W - a) 2*1.21]= (333`*Pr *W) / [B*(W- a) (2)

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

r=3,, r, (3) where o> is in units of psi and P3 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 A370-891i1. 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 E8-89B19 1and E21-79 11 ], and RMF Procedure 8102, Revision I All pull rods, grips, and pins were made of Inconel 7 18 hardened to HRC45. The upper pull rod was connected through a universal Joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.

Extension measurements were made with a linear variable displacement transducer (LVDT) extensometer.

The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure.

The extensometer gage length was 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83 85i]1.

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 positioned at the center and at each end of the gage section of a dummy specimen and in each tensile machine griper. In the test configuration, with a slight load on the specimen, a plot of specimen temperature was developed over the range from room temperature to 550OF (288°C). 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 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.

Cook Unit 2 Capsule U

.9

.2 CHARPY V-NOTCH IIPACT TEST RESULTS The results of the Charpy V-notch impact tests per-formed on the various materials contained in Capsule U, which received a fluence of 1.583 x 10

9 nicm(E> 1.0 MeV) in 8.65 EFPY of operation, are presented in 1

Tables 5-1 through 5-8 and are compared with unirradiated resultsil as shown in Figures 5-1 through 5-12.

The transition temperature increases and upper shelf energy decreases for the Capsule U materials are summarized in Table 5-9. These results led to the following conclusions:

Irradiation of the reactor vessel intermediate shell plate C5521-2 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major rolling direction of the plate (longitudinal orientation), to 1.583 x 1019 n/cm 2 (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 957F and a 50 ft-lb transition temperature increase of I I0°F. This results in an irradiated 30 ft-lb transition temperature of 120'F and an irradiated 50 ft-lb transition temperature of 165°F for the longitudinally oriented specimens.

Irradiation of the reactor vessel intermediate shell plate C552 1-2 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major rolling direction of the plate (transverse orientation), to 1.583 x t0*9 n/cm2 (E> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 1307F and a 50 ft-lb transition temperature increase of 135°F. This results in an irradiated 30 ft-lb transition temperature of 160'F and an irradiated 50 ft-lb transition temperature of 205'F for transversely oriented specimens.

• Irradiation of the weld metal Charpy specimens to 1.583 x 10'9 n/cm- (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 75°F and a 50 ft-lb transition temperature increase of 40'F. This results in an irradiated 30 ft-lb transition temperature of 85°F and an irradiated 50 ft-lb transition temperature of 1100 F.

• Irradiation of the weld Heat-Affected-Zone (-I-kZ) metal Charpy specimens to 1.583 x 10:9 n/cm:

(E> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 105'F and a 50 ft-lb transition temperature increase of 1I0°F. This results in an irradiated 30 ft-lb transition temperature of 45'F and an irradiated 50 ft-lb transition temperature of 80'F.

The average upper shelf energy of the intermediate shell plate C552 1-2 (longitudinal orientation) resulted in an average energy decrease of 16 ft-lb after irradiation to 1.583 x 1019 n/cm2 (E> 1.0 MeV). This results in an irradiated average upper shelf energy of 111 ft-lb for the longitudinally oriented specimens, The average upper shelf energy of the intermediate shell plate C552 1-2 (transverse orientation) resulted in an average energy decrease of 14 ft-lb after irradiation to 1.583 x 1019 rilcm (E> 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 72 ft-lb for the transversely oriented specimens.

The average upper shelf energy of the weld metal Charpy specimens resulted an average energy decrease of 6 ft-lb after irradiation to 1.533 x 10;9 n/cm' (E> 1.0 MeV). Hence. this results in an irradiated average upper shelf energy of 71 ft-lb for the weld metal specimens.

Cook Unit 2 Capsule U

5-4 The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an average energy decrease of 33 ft-lb after irradiation to 1.583 x I019 n/cm' (E> 1.0MeV). This results in an irradiated average upper shelf energy of 82 ft-lb for the weld I-HkAZ metal.

The fracture appearance of each irradiated Charpy specimen from the various materials is shown in Figures 5-5 through 5-8 and shows an increasingly ductile or tougher appearance with increasing test temperature.

A comparison of the 30 ft-lb transition temperature increase and upper shelf energy decreases for the various D. C. Cook Unit 2 surveillance materials with predicted values using the methods of NRC Regulatory Guide 1.99, Revision 2131 is presented in Table 5-10. Comparison of the 30 ft-lb transition temperature increase for the intermediate shell plate C5521-2 (transverse) is 33°F greater than the Regulatory Guide 1.99 Revision 2 predictions. Regulatory Guide 1.99 Revision 2 requires a 2 sigma allowance of 34°F for base metal to be added to the predicted reference transition temperature to obtain a conservative upper bound value. Thus, the reference transition temperature increase for the intermediate shell plate C5521-2 (transverse) is bounded by the 2 sigma allowance for shift prediction. This comparison indicates that the transition temperature increases and the upper shelf energy decreases of the intermediate shell plate C552 1-2 2

(longitudinal) and the surveillance weld resulting from irradiation to 1.583 x 109 n/cm (E> 1.0MeV) are less than the Regulatory Guide 1.99 Revision 2 predictions. This comparison also indicates that the upper shelf energy decrease of the intermediate shell plate C5521-2 9

(transverse) resulting from irradiation to 1.583 x 10' n/cm2 (E> 1.0MeV) is less than the Regulatory Guide predictions.

Photographs of the Charpy and tensile specimens before testing are shown in Appendix B.

Cook Unit 2 Capsule U

5-5 5.3 TENSILE TEST RESULTS The results of the tensile tests performed on the various materials contained in Capsule U irradiated to 1.583 x 10'9 nlcm' (E> 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated results°*

as shown in Figures 5-9 and 5-10 The results of the tensile tests performed on the intermediate shell plate C552 1-2 (transverse orientation) indicated that irradiation to 1.583 x 1019 n/cm (E> 1.0 MeV) caused an approximate increase of 18 ksi in 2

the 0.2 percent offset yield strength and approximately a 16 ksi increase in the ultimate tensile strength 1

when compared to unirradiated data 'l (Figure 5-18).

The results of the tensile tests performed on the surveillance weld metal indicated that irradiation to 1.583 x 10i'9 n/cm (E> 1.0 MeV) caused approximately a 9 ksi increase in the 0.2 percent offset yield strength and approximately a 8 ksi increase in the ultimate tensile strength when compared to unirradiated dataý1' (Figure 5-19).

The fractured tensile specimens for the intermediate shell plate C5521-2 material are shown in Figure 5-11, while the fractured tensile specimens for the surveillance weld metal are shown in Figure 5-12. The engineering stress-strain curves for the tensile tests are shown in Figures 5-13 and 5-14 5.4 WEDGE OPENLNG LOADLNG SPECIaENS Per the surveillance capsule testing program with Indiana Michigan Power Company, the WOL Specimens were not tested and are being stored at the Westinghouse Science and Technology Center Hot Cell facility.

Cook Uit 2 Capsule U

5-6 Table 5-1 Charpy V-notch Data for the D. C. Cook Unit 2 Intermediate Shell Plate C5521-2 9 2 Irradiated to a Fluence of 1.583 x 10' nlcm (E> 1.0 MeV) (Longitudinal Orientation)

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

ML45 75 24 14 19 10 0.25 10 ML48 100 38 27 37 23 0.58 20 ML42 125 52 35 47 27 0.69 30 ML44 175 79 62 84 42 1.07 50 ML41 200 93 55 75 42 1.07 50 ML43 225 107 103 140 50 1.27 90 ML46 250 121 114 155 79 2.01 100 ML47 300 149 115 156 82 2.08 100 Cook Unit 2 Capsule U

5-7 Table 5-2 Charpy V-notch Data for the D. C. Cook Unit 2 Intermediate Shell Plate C5521-2 Irradiated to a Fluence of 1.583 x 10'9 n/cm2 (E> 1.0 MeV) (Transverse Orientation)

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

MT62 25 -4 5 7 2 0.05 5 MT61 50 10 14 19 12 0.30 10 MT66 75 24 17 23 10 0.25 15 MT71 125 52 19 26 16 0.41 30 MT64 150 66 25 34 21 0.53 35 N1T72 175 79 33 45 28 0.71 40 NIT70 200 93 37 50 30 0.76 45 MT69 215 102 39 53 33 0.84 65 N1T63 225 107 63 85 52 1.32 95 MT68 250 121 67 91 54 1.37 100 MT67 275 135 72 98 50 1.27 100 MT65 300 149 77 104 58 1.47 100 Cook Unit 2 Capsule U

5-8 3-8 Table 5-3 Charpy V-notch Data for the D. C. Cook Unit 2 Surveillance Weld Metal 9 2 Irradiated to a Fluence of 1.583 x 10' n/cm (E> 1.0 MeV)

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

MW70 -10 -23 23 31 16 0.41 25 MMW64 0 -18 29 39 26 0.66 35 M-W7I 25 -4 17 23 14 0.36 30 MvW68 50 10 21 28 21 0.53 40 MNW63 75 24 26 35 20 0.51 60 MVW61 100 38 42 57 37 0.94 80 MW72 125 52 60 81 50 1.27 85 MNV65 150 66 71 96 57 1.45 100 MW66 175 79 47 64 39 0.99 85 NW62 185 85 62 84 49 1.24 100 MW67 200 93 78 106 62 1.57 100 N[W69 250 121 74 100 62 1.57 100 Cook Unit 7 Capsule U

5-9 5-9 Table 5-4 Charpy V-notch Data for the D. C. Cook Unit 2 Heat Affected Zone Material Irradiated to a Fluence of 1.583 x 10' 9 n/cm2 (E> 1.0 MeV)

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

MiH67 -25 -32 4 5 2 0.05 10 M-63 25 -4 21 28 10 0.25 35 MCH71 50 10 48 65 30 0.76 55 NIH69 65 18 18 24 19 0.48 30 NM-172 75 24 64 87 44 1.12 90 MM70 100 38 39 53 32 0.81 50 MN1-162 125 52 118 160 71 1.80 100 MH64 150 66 72 98 55 1.40 95 MIH61 175 79 83 113 61 1.55 100 MI-H65 200 93 103 140 71 1.88 100 MI-I68 225 107 74 100 53 1.35 100 NH66 250 121 69 94 66 1.68 100 Cook Unit 2 Capsule U

10 Table 5-5 lInstruniented Charpy Impact Test Results for the D. C. Cook Unit 2 Intermediate Shell Plate C5521-2 2

Irradiated to a Fluence of 1.583 x 10"9 n/cm (E>1.0 MeV) (Longitudinal Orientation)

________________________________________________________ 1 T T F 5-10 Normalized Energies 2

(fl-lb/in )

Charp) Yield 'rime to Time to Fast Load Yield IGy Max. Max. Frac. Arrest Yield Flow 11 "restl Energy Load PM Load PF Load PA Stress Stress Temp. Charpy Max. Prop. pcy (ni1ec) Tmu Samnple EL) Sy (ksi) (ksi)

(o1F) (kt- Ib) EF)A (Ib) (Ih) (reset) (Ib) (Ib)

No. EL/A EM,/A

_____~~~~( __ _ I4 ++)t 127 Nll 45 75 14 113 6,1 48 3722 0.14 3929 0.20 3929 638 124 167 50 3670 0,14 4600 0.38 4600 858 122 137 M L4 8 100 27 217 282 124 157 3945 0.29 4258 0.38 4258 384 131 136 MIL42 125 35 499 241 258 3467 0.14 4735 0,53 4563 792 115 136 M114 175 62 443 122 321 3731 0.28 4139 0.38 3887 1340 124 131 MtL41 200 55 103 829 308 521 3262 0.14 4507 0.54

• 108 129 M1,43 225 918 234 684 3265 0.14 4546 054 108 130 M,46 250 114 N1L47 300 115 926 147 779 3126 0.14 4071 038

• 104 120

• Fully Dluctile Fracture. No Arrest Load.

Cook Ilim 2 (Calsulc U

5-11 Table 5-6 lnstritimented Charpy Impact Test Results for the I). C. Cook Unit 2 Intermediate Shell Plate C5521-2 Irradiated to a Fluence of 1.583 x 1019 II/ciin2 (F>1.0 MeV)(Transverse Orientation)

Normalized Energies (ft-I h/in 2)

Charpy Yield Time to Time to Fast Test Energy Load Yield tm, Max. Max. Fract. Arrest Yield Flow Sample Temp. FD) CharpN' Max. Prop. PGV (rnsec) Load ProI tr' LMad PF Load PA Stress S. Stress No. ('F) (ft-11)) FD/A Ejý,/A Ep/A (11) (b) (Insec) (Ib) (lb) (ksi) (k-si)

MT62 25 5 40 12 28 1107 0.10 1319 0.14 1319 55 37 40 N,1T61 50 14 113 77 35 3600 0.14 4154 0.22 4154 126 120 129 NIT66 75 17 137 88 49 3825 0.14 4273 0.25 4273 513 127 134 MTTI 125 19 153 48 105 2969 0.21 3187 0.24 3187 202 99 102 MT64 150 25 201 124 77 3381 0.14 4194 0.32 4194 1577 112 126 N1T72 175 33 266 156 110 3438 0.13 4273 0.38 4273 2083 114 128 N1T70 200 37 298 128 170 3681 0.28 4096 0.39 4096 2151 122 129 NIT69 215 39 314 153 161 3289 0.14 4158 0.38 4158 2533 109 124 MT63 225 63 507 236 272 3359 0.15 4457 0.54 4225 3236 112 130 MT68 250 07 540 221 319 3288 0.14 4212 0.54 *

• 109 125 MT 67 275 72 580 222 358 3051 0.16 4249 0.54 *

• 101 121 77 620 223 397 3153 0.14 4251 054

• 105 123 MTh5 300

• Frilly Ductile Fracture. No Arrest Load.

Couk Unl 2 (Capsilc 1

5-12 "Table5-7 Instruniented Charpy Impact Test Results for the D. C. Cook Unit 2 Surveillance Weld Metal 2

Irradiated to a Fluence of 1.583 x 10'9 n/ci (E>1.0 MeV) r I Normalized Energies (ft-lb/in 2)

Yield Time to Time to Fast Charpy Yield Load Yield tcy Max. Fract. Arrest Flow Energy Max.

Test Load PA Stress Sy St resi Pt. (msec) Load PrM Load P1V Sample Temp. ED Charpy Max. Prop. (ksi) (ksi)

(Ib) (mnsec) (lb) (Ib)

(ft-lh) E0 /A Er,/A F,,p/A (Ib)

No.

031 4573 778 129 141 185 133 52 3898 0.14 4576 M4\V70 - 10 23 0.28 4482 467 129 139 234 106 127 3891 0.14 4482 M\W6-1 29 0.23 4105 978 125 131 25 137 83 54 3776 0.14 4105 M\V71 17 95 4021 0.30 4021 941 128 131 169 85 84 3847 0.26 M\W68 50 21 0.32 4430 1618 121 134 75 209 131 79 3655 0.14 4430 MW63 26 4- + t t I 1 463 123 134 338 151 187 3712 0.26 4359 0.42 4359 MW61 100 42 I 4 I I- -1 1 r 1 4581 1252 110 131 60 483 239 244 3323 0.14 4581 0.54 NIW72 125 I I I I-4535 0.54 117 134 572 235 337 3511 0.14 NMIW65 150 71 4404 0.50 4404 810 113 130 378 221 158 3397 0.14 NMW66 75 47 4232 0.54 109 125 62 499 231 269 3280 0.13 NIW62 185 378 4440 0.62 *

• 4 125 136 628 250 3754 0.28 MW67 200 78

[ _ _ 11II111111 1

&IXIVAO) 9 So1 74 596 231 365 3283 0.14 4273 0.54 109 12-5

_____ 25 74_ _ _ _ _ __ __ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _

• Fully D)uctile Fracture. No Arrest Load.

Cook Cook Uinit 2 Capd 1 ei Ui

5-13 Table 5-8 Instrumented Charpy Impact Test Results for the D. C. Cook Unit 2 Heat-Affected-Zone (IIAZ) Metal 2

Irradiated to a Fluence of 1.583 x 10'9 n/c 1 (E>1.0 MeV)

Normalized Energies (ft-lb/in2) charpy Yield Time to Time to Fast Load Yield t[Gy Max. Max. Fract. Arrest Yield Flow Test Energy Max. Prop. l,;y (nrsec) Load PIm tMt Load PF Load Stress S" Stress Sample Temp. ED Charpy ED/A Em/A Eu/A (Ib) (Ib) (resec) (I)) PA (Ib) (ksi) (ksi)

No. (°F) (ft-lb) 32 15 17 964 0.10 1133 0.17 1133 1,16 32 35 N'1167 -25 -4 169 47 122 3327 0.21 3486 0.23 3486 858 II1 113 NIH63 25 21 48 387 166 221 3797 0.14 4557 0.38 4420 2126 126 139 NMi171 50 18 145 58 87 3653 0.14 3834 020 3834 971 121 124 MH69 65 6 515 162 354 3910 0.14 4530 0.37 4500 3802 130 140 fMvH72 75 100 39 314 114 200 2970 0.21 3618 0.38 3618 954 99 109 M1170 N11162 125 118 950 301 6419 3352 0.16 4355 0.53 *

• 111 128 72 580 240 340 3991 0.22 4756 0.54 4756 4056 133 145 Ni 11641 150 175 83 668 243 425 3622 0.14 4662 0.54

• 120 138 NIlI61 200 103 829 253 576 3642 0.15 4799 0.54 *

• 121 140 NIH65 N,1l168 225 74 596 187 409 3066 0.17 3981 0.50 *

• 102 117 556 228 328 3330 0.14 4287 0.52 *

• 111 126 NIH6o 250 69 Fully D)uctile Fracture. No Arrest Load.

Cook UtLi( 2 ('apsilc U

5-14 Effect of Irradiation to 1.583 x 10'9 n/cin (E>1.0 MeV) on the Notch Toughness Properties of tile D. C. Cook Unit 2 2

Table 5-9 Reactor Vessel Surveillance Materials I I . J,.',

Average 35 ruil Lateral(') Average 50 tt-l'(a) Average Energy Absorption'"

Average 30 (ft-lb)(')

Transition 'l'emperatu're ('F) at Full Shear (ft-lb)

Material 'Transition remperature ('F) Expansion Temperature ('F)

+~ I r- Irradiated Irradiated AT I Jniiadiated Inadiated AT Unirradiated Irradiated AT Unirnadiated Iihnrrattiaied Unirdiate

_____ _ 4 1I I -16 95 50 150 100 55 165 111 127 III Inter. Shell 25 120 Plate C552 1 2 (Longitudinal)

I _ I__ I t I 205 135 86 72 1 30 160 130 70 190 120 70 Inter. Shell Plate C5521-2 (Tiansveise) 10 85 75 50 150 705 110 -10 77 71 -6 IIAZ Metal -6 1 1 85 1125 -30 80 14t 115 1 82 -3

a. "Average" is defined as the value read from the curve fit through the data points of the Charpy tests (see Figures 5-1 tmrough .-- 4).

Couk Unit 2 Cap1 auc Ii

5-15 Table 5-10 Comparison of the D. C. Cook Unit 2 Surweillanec Material 30 ft-lb Transition Temperature 5-A Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy I. Temperature Shift Decrease Predicted Measured Material Capsule Fluence_ ~ ~ Predicted

~.. aN ~Measured 0a, o (C) fo/ '(b)

(x 10'"ncn) '-' k(OF) %j I T 0 238 55.3 55 17 13 Intermediate Shell Y 0.664 80.0 90 22 19 P~ate C5521-2 1.019 90.9 95 24 19 X

U 1.583 101.9 95 27 13 (Longitudinal)

T 0,238 55.3 80 17 14 Intermediate Shell Plale C5521-2 Y 0.664 80.0 100 22 20 X 1.019 90.9 103 24 19 (Transverse) U 1.583 101.9 130 27 16 T 0.238 45.9 40 14 4 Weld Metal Y 0.664 66.4 50 17 9 X 1.019 754 70 19 10 U 1.583 84.5 75 J 21 8 PAZ Metal T 0.239 -- 50 -- 13 Y 0.664 70 21 X 1.019 72 12 U 1.583 105 - 29 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 E185-12.

Cook Unit Z Cacsule U

5-16 Table 5-11 Tensile Properties of the D).C. Cook Unit 2 Reactor Vessel Surveillance Materials Irradiated to 1.583 x 10'9 n/cm' (E > 1.0 MeV)

Test 0.2% Ultimate Fracture Fracture Fracture Uniform Total Reduction Material Sample Temp. Yield Strength Load Stress Strength Elohgation Elongation in Area Number

('F) Strength (ksi) (kip) (ksi) (ksi) (%) (U) (%)

(ksi) 3.60 172.0 73.3 10.5 20.1 57 Lower Shell Plate MTI 1 150 79.5 97.4 3.65 131.8 74.4 9.6 16.7 44 C5521-2 MTI2 550 70.3 93.7 3.20 173.9 65.2 9.0 19.5 63 Weld Metal NMV 4 125 82.5 96.8 92.7 3.50 190.2 713 8 I 17.1 58 NIV 5 550 759 (Cook Lim 2 C apsuic H

5-17 (0 C)

-150 -100 -50 0 50 100 150 200 100 r 60 LJ S40 20 0

1U[

lf 21 2

1.0 40

r 20 0.5

.=J 0 0 o UNIRRADIATED 2

• IRRADIATED (550"F), FLUENCE 1.58x 1O n/c, 160 200 14O 0 :60 120 * .

p100 120 v0 2

L.*J Z:

60 0 40 40 0

20 0 I- I I 0 0

- C00 -i00 0 100 200 300 410 0

TEMPERATJRE ( r)

Figure 5-I Charpy V-Notch Impact Properties for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Longitudinal Orientation)

Cook Umt 2 Capsule U

18 5-18 (0 C)

-150 -100 -50 0 50 100 150 2C0 SIII L I I I 100

>4 60 0

LU-. 60 2 0 0 _

40 20 3 0

2.5 80 2.0 60 1.5 E:

LU 40 1.0 20 0.5

_-J 0

0 o UNIRRADIATED 19

• IRRADIATED (550'F), FLUENCE1.58x 10 n/rl 160 120 100 12 80 S

.Q, 80 60 LUj z

LU 40 © 0 I./o 40 0FL

)D i n 0

-PJO -we 0 100 200 300 400 i00 TEMPERATURE (F)

Figure 5-2 Charpy V-Notch Impact Properties for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation)

Cook Unit 2 Capsule U

5-19 (oC)

-150 -100 -50 0 50 100 150 200 100 80 LQJ 60 Uj--

40 20 0

100

-; 80 20 1.5 60 E L.0 40 20 0.5

-_J 0

0 o UNIRRADMATED 19

• [RRAD[ATED (550'F), rLUEN(E 158 x 10 n/Cm 160 120 100 120 80 S

LUj 40 40 20 too I 2I0 0 a 0 100 200 000 400 500

- 100 -100 TEMPEATURE ("F)

Figure 5-3 Charpy V-Notch Impact Properties for D. C. Cook Unit 2 Reactor Vessel Surveillance Weld Metal Cook UCt 2 Capsuie U

5-20 i5-?0 (0C)

-150 -100 -70 0 50 100 150 200 100 K 0 80

.- )- 2 _Z 8 60 S

40 0

20 I I i ,

0 100 2.5 LA 2.0 80 1.5 60

-4 1.0 40

-J 0.5 20

-4 0

0 o UNIRRADIATED 19

• IRRADIATED (50'F), FLUENCE 1.5Bx 10lInlcm 160 200 140 160 120 t00 120 80 124 60 so 40 0

_200 -100 0 100 290 300 400 TEMPERAT.jRF ,GF Figure 5-4 Charpy V-Notch Impact Properties for D. C. Cook Unit 2 Reactor Vessel Weld Heat Affected-Zone Metal Cook Utl f[ Capsule U

5-21 I-2 ML48 ML42 MLA4 MLA5 ML43 MML6 MLA7 MlA1 Figure 5-5 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Longitudinal Orientation)

Cook Unit 2 Capsule U

5-22 MT62 MT61 MIT66 MT71 MT64 MT72 MT70 MT69 1'~ 't.4.. . . ...

S~rt.V . ~

MT63 MT68 MT67 MI 65 Figure 5-6 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation)

Cook Unit 2 Capsule U

5-23 MW70 MW64 MW71 MW6S MW63 MW61 MW72 MW65 MW66 MW62 MWv67 MTW69 Figure 5-7 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 2 Reactor Vessel Weld Metal Cook Umt ' Capsule U

MH67 MH63 MH71 MH69 Mb. i i ma MH72 MH70 MH62 MH64 MH61 MH65 MI-i68 MHG6 Figure 5-8 Charpy Impact Specimen Fracture Surfaces for D. C. Cook Unit 2 Reactor Vessel Weld Heat-Affected-Zone Metal Cook Urul 2 Capsule T"

5-25 5-25 (0C) 0 50 100 150 200 250 300 120 800 I I i-

[10 700 100 90 600 SULTIMATE TENSILE STRE4GTH U, I=

80 0-500 70 A

.27 YIELD STRENGTH 2A 60

-A 4Or 50 300 40 A 0 UN[RRADIATED A

• 9 IRRADIATED AT 550'F, FLUENCE L58 x4DI rVcm2 80 70 REDUCTION IN AREA 60 40

=-J- 40

(-)

S30 UTOTAL ELO]NGATION 20 I I UNIFORM EýON.ATIGN 0

0 100 200 300 400 500 TEMPERATURE (OF)

Figure 5-9 Tensile Properties for D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation)

Cook Umct 2 Capsule U

5-26

(°C) 0 100 150 200 250 300 120 e00 i I I I I 1 -

110 700 100 ULTIMATE TENSILE STRENGTH L T S90 600 LU a-0

- 80 500

• 70 0.27. YIELD STRENGTH 60 400 50 300 40 A 0 UN[RRADIATED 9

A 0 [RRADIATED AT 550'F, FLUENCE L58 xIOg n/cr2 80 70 60 50

-U 40 D 30 20 10 0

0 t00 200 300 400 500 EMPERATURE (OF)

Figure 5-10 Tensile Properties for D. C. Cook Unit 2 Reactor Vessel Surveillance Weld Metal Cook Umt 2 Capsule U

5 r>.

8 9 aUi ;4§ 7;a

• U It NoC3GR a 9 1 2. 3 4 6 7 Specimen MT1I 150'F 1 2.

• -~.

Specimen MT12 550'F Figure 5-11 Fractured Tensile Specimens from D. C. Cook Unit 2 Reactor Vessel Intermediate Shell Plate C5521-2 (Transverse Orientation)

Cook Uni 2 Capsule U

5-2 5-28 3- 71" 8 S{ I:.... ...

34~7 Specimen MW11 125" F Specimen MNW12 550"F Figure 5-12 Fractured Tensile Specimens from D. C. Cook Unit 2 Reactor Vessel Surveillance Weld Metal Cook Unmt ' Capsule U

5-29 100.00 90.00 80.00 70.001 60.00 LU 50.00 40.00 30.00 MT1 1 20.00 150 F 10.00 0.000.0 a0 0.10 0.20 STRAIN, IN/IN 1

(I crr U.

C,)

0.08 STRAIN, IN/IN Figure 5-13 Engineering Stress-Strain Curves for Plate C5521-2Tensile Specimens MT1l and MT12 (Transverse Orientation)

Cook Unit_ Capsule U

5-30 100.00 80.00 70.00

""60.00 C6, U) 50.00 Lii (r-) .- 40.00 30.004 20.00- MW11 10.00- 125 F 0.00 0.00 0.64 0.060 0o.2 0.o16 0.20 STRAIN, IN/IN 100.00 90.00 80.00 70.00 U5 Z 60.00 C/) 50.00 LU

. 40.00 30.00 MW12 20.00 10.00 550 F 0.00- .

0.00 0.04 0.08 0.12 0.16 0.20 STRAIN, IN/IN Figure 5-14 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens MRVll and MWV12 Cook Unit 2 Capsule U

6-1 6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY 6.1 [NTRODUCTION This section describes a discrete ordinates S, transport analysis performed for the D. C. Cook Unit 2 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 twelve reactor operating cycles. In addition, neutron dosimetry sensor sets from Surveillance Capsules T, Y, X. and U withdrawn from the D. C. Cook Unit 2 reactor at the conclusion of fuel cycles 1, 3, 5, and 8 were analyzed using current 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 3411 MWt to 3800 MRVt. beginning with the operation of the thirteenth 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 Dosimetrv Methods for Determining Pressure Vessel Neutron Fluence_'[: l The use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the matenial has traditionally been accepted for the development of damage trend curves as well as for the implementation of trend curve data to assess the condition of the vessel. In recent years; however, 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 dosimetrv 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 WCA_P-14040-NP-A., "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves.,"

January 1996 ý:61 The specific calculational methods applied are also consistent with those described in WCAP-15557. 'Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology."l -

Cook Urut ' Capsule U

6.2 DISCRETE ORDINATES ANALYSIS A plan view of the D. C. Cook Unit 2 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, 176', 184', 356' (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 specumen containers are 1-inch square and approximately 38 inches in height. The containers are positioned axially such that the test specimens are centered on the core rnidplane, thus spanning the central 3 feet of the 12-foot high reactor core.

From a neutronic standpoint, the surveillance capsules and associated support structures are significant.

The presence of these materials has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus between the thermal shield and the reactor vessel. In order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analyvtical model.

The fast neutron exposure evaluations for the D. C. Cook Unit 2 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:

(¢(r,0,z) = L¢(r,0)] * [4(r~z)]/[4(r)]

where 4(r,0,z) is the synthesized three-dimensional neutron flux distnbution, ý(r.0) is the transport solution in r.0 geometry, )(r,z) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and ý(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the r,0 two-dimensional calculation. This synthesis procedure was carried out for each operating cycle at D. C. Cook Unit 2.

For the D. C. Cook Unit 2 calculations. one r,9 model was developed since the reactor is octant symnmetric This r,9 model includes the core, the reactor internals, the thermal shield -- including explicit representations of the surveillance capsules at 40 and 40', the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. The symmetric r.0 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 r,0 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 r.0 calculations was set at a value of 0.001.

The r,z model used for the D. C. Cook Unit 2 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 Cook Unit 2 Capsule U

• 1e 6-3 elevation 1-foot below the active fuel to 1-foot above the active fuiel. As in the case of the r.e 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 rz geometric mesh description of the reactor model consisted of 153 radial by 90 axial intervals. As in the case of the r,0 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 cnterion utilized in the r.z calculations was also set at a value of 0. 00 1.

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 geometr.

The core power distributions for the first seven fuel cycles used in the plant specific transport analysis were obtained from Reference 18 (which was the input used in the previous surveillance capsule analysis.

documented in Reference 19). The core power distributions for cycles eight through twelve were taken from the appropriate D. C. Cook Unit 2 fuel cycle design reports (References 20 through 24). The data extracted from the references represented cycle dependent fuel assembly enrichments, bumups, and axial power distributions. This information was used to develop spatial and energy dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle averaged neutron flux, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distnbutions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and bumup 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 power distributions used in the transport analyses are provided in Appendix C to this report.

All of the transport calculations supporting this analysis were carried out using the DORT discrete 1 61 ordinates code Version 3.1It21I and the BUGLE-96 cross-section librarv. 2 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 S'6 order of angular quadrature. Energy and space dependent core power distributions, as well as system operating temperatures. were treated on a fuel cycle specific basis.

Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-4. The data listed in these tables establish the means for absolute compansons of analysis and measurement for the Capsules T. Y X, and U irradiation and provide the calculated neutron exposure of the pressure vessel wall for the first twelve 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 symnmetric surveillance capsule positions (40 and 400). These data, representative of the axial midplane of the active core, are meant to establish the exposure of the surveillance capsules wvithdrawn to date and to provide an absolute comparison of measurement with calculation. Similar Cook Unit 2 Capsule U

6-4 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 azmnuthal 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 4(E > 1.0 MeV) and dpa/sec are given in Tables 6-3 and 6-4.,

respectively. The data, based on the Cycles 1 through 12 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.

Cook Unit 2 Capsule U

,p,,

6-5 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 2 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:

Azimuthal Withdrawal Irradiation Capsule ID Location Time Time [EFPY1 T 400 End of Cycle I 1.08 Y 400 End of Cycle 3 3.22 x 400 End of Cycle 5 5.25 400 End of Cycle 8 8.65 U

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, Y, X, and U are sumnmanzed as follows:

Reaction Sensor Material of Interest CapsuleT CapsuleY Capsule X Capsule U 63 Copper Cu(n,x)-°Co x x X x 54Fe(n.p)",itn Iron x x x x 5Ni(n,p)-8 Co x x Nickel x x Uranium-238 38tU(n,f) 13 Cs x x x X*x Neptunium-237 :3 Np(n,f) :37 Cs x x x x Cobalt-Aluminum* f 9Co(n.y) 6°C0 The cobalt-aluminum measurements for this plant include both bare wire and cadiroum-covered sensors Both of the cadmium covered cobalt-aluminum sensors in Capsules T and Y were not recovered. One of the tvwo cadmium covered cobalt-aluminum sensors in Capsules X and U were not recovered.

Cook Urut 2 Capsule U

V 6-6 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:

"* 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, Y, and X 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, Y, X. and U was based on the reported monthly power generation of D. C. Cook Unit 2 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, Y X, 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:

Cook Unit 2 Capsule U

6-7 A

R No F Y C, [I Fref where:

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

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

CJ = Calculated ratio of 4(E > 1.0 Me'v) during irradiation periodj to the time weighted average 4(E > 1.0 MeV) over the entire irradiation period.

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

t = Length of irradiation period j (sec).

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

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

In the equation describing the reaction rate calculation, the ratio [P]/[Pmed 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 C1, 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 bv variations in flux level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, C, is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those employing low leakage fuel management, the additional C, 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 C1 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.

additional corrections were made to the "TU measurements to account for the presence of :-1sU impurities in the sensors as well as to adjust for the build-m of plutonium isotopes over the course of the irradiation.

Cook Unit 2 Capsule U

.9 6-8 37 Corrections were also made to the "'U and 2 Np sensor reaction rates to account for gamma ray induced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the Dý C. Cook Unit 2 fission sensor reaction rates are summarized as follows:

Correction Capsule T Capsule Y Capsule X Capsule U 23'U Impurity/Pu Build-in 0.875 0.859 01845 0.823 21U(.V,) 0.958 0.958 0.958 0.958 Net '3'U Correction 0.838 0.823 0.810 0.788 2%VNp(yf) 0.985 0.985 0.985 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 for Capsules T, Y, X, 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 238 reaction rates are listed both with and without the applied corrections for U impurities, plutonium build in. and gamma ray induced fission effects.

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 ý(E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least squares methods, as applied to surveillance capsule dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections. and the calculated neutron energy spectrum within their respective uncertainties. For example, R;c5 )(0 5 Cook Unit 2 Capsule U

6-9 relates a set of measured reaction rates, R,, to a single neutron spectrum, (0,, through the multigroup dosimeter reaction cross-section, (7g, each with an uncertainty 5. 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 2 surveillance capsule dosimetry, the FERRET code1271 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 (4(E > 1.0 MeV) and dpa) along with associated uncertainties for the four rn-vessel capsules withdrawn to date.

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

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

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

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

For the D. C. Cook Unit 2 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 SNLRML dosimetry cross section library'2". The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard E 1018, "Application of ASTNI Evaluated Cross-Section Data File, Matrix E 706 (JIB).'

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 vanances 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 2 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 lexel of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM National Consensus Standards for reaction rate determinations for each sensor type.

Cook Unit 2 Capsuie U

6-10 After combining all of these uncertainty components, the sensor reaction rates denved from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least squares evaluation:

These uncertainties are given at the lc7 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 2 surveillance program. the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.

Reaction Uncertainty 63Cu(n,<()6 0 Co 4.08-4.16%

ý4Fe(np)-4Mn 3.05-3.11%

isNi(n.p)58Co 4.49-4.56%

' 38U(n,f)'37 Cs 0.54-0464%

2

'Np(n.f)137Cs 10.32-10.97%

59Co(n.,)O6Co 0.79-3.599%

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

Calculated Neutron Spectrum The neutron spectra input to the least squares adjustment procedure were obtained directly from the results of plant specific transport calculations for each surveillance capsule irradiation period and location. The spectrum for each capsule 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.

Cook Unit 2 Capsule U

6-11 While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks and the dosimetrv cross-section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:

M.gg = R2 + R *Rg *P where R, specifies an overall fractional normalization uncertainty and the fractional uncertainties R. and R,. specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:

Pg = [I-0](g,+ 0e where (g - gI):

27<

The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range y (0 specifies the strength of the latter term).

The value of 5 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 2 calculated spectra was as follows:

Flux Normalization Uncertainty (R,,) 15%

Flux Group Uncertainties (R,. R.)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 29%

(E < 0.68 eV) 52%

Short Range Correlation (0)

(E > 00055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Flux Group Correlation Range (y)

(E > 0.0055 NMeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 Cook irut 2 Capsule U

6-12 6.3.3 Comparisons of Measurements and Calculations Results of the least squares evaluations of the dosimetry from the four D. C. Cook Unit 2 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 Mi/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. From Section 6.4 of this report, it may be noted that the uncertainty associated with the unadjusted calculation of neutron fluence (E > 1.0 MeV) and iron atom displacements at the surveillance capsule locations is specified as 12% at the lIc level. From Table 6-10, it is noted that the corresponding uncertainties associated with the least squares adjusted exposure parameters have been reduced to 6% for neutron flux (E > 1.0 MeV) and 7% for iron atom displacement rate. Again, the uncertainties from the least squares evaluation are at the Icy 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 6-11, 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 4(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 sumlar results with all measurement-to-unadjusted calculation comparisons falling well within the 20%o lunits specified as the acceptance criteria in Regulatory Guide 1. 190.

In the case of the direct comparison of measured and calculated sensor reaction rates. the NL'C comparisons for fast neutron reactions range from 0.85-1.16 for the 20 samples included in the data set. The overall average N/C ratio for the entire set of D. C. Cook Unit 2 data is 1.0 1 with an associated standard deviation of 8.2%"

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 0.92-1.04 for neutron flux (E > 1.0 MeV) and from 0.93 to 1.02 for iron atom displacement rate. The overall average BE/C ratios for neutron flux (E > 1.0 MeV) and iron atom displacement rate are 0.98 with a standard deviation of 4.8% and 0.97 with a standard deviation of 4.2%, respectively.

Cook Unit 2 Capsule U

6-13 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 2 reactor pressure vessel.

The uncertainty associated with the calculated neutron exposure of the D. C. Cook Unit 2 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:

1 - 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 2 surveillance program.

The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operational or geometric variables that impact power reactor calculations. The second phase of the qualification (H. B. Robinson comparisons) addressed uncertainties in these additional areas that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations. The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant specific input parameters. The overall calculational uncertainty applicable to the D. C. Cook Unit 2 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 2 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 17.

Cook ULit 2' Capsude U

'V 6-14

_ Capsule [ Vessel IR PCA Compansons 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 1 12% 13% i 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 2.

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 UL't 2 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 2 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 2 pressure vessel are provided in Table 6-14. As presented, these data represent the maximum exposure of the cladibase 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 eleven (the last cycle completed at the D. C. Cook Unit 2 plant), a projection to the end of cycle twelve (the current operating cycle) and further projections for future operation to 25, 32, 36, 48, and 54 effective full power years.

The projection to the completion of cycle 12 was based on the cycle 12 design power distribution provided in Appendix C, continued operation at a core power level of 3411 MvWt, and a design cycle length of 1 4 effective full power years. Projections beyond the end of cycle 12 were based on the assumption that future operation would continue to make use of low leakage fuel management and that a representative power distribution averaged over cycles 10 through 12 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 13 and beyond, the core power level would be uprated to 3800 /Mt.Therefore, the fluence projections for future operation at D. C. Cook Unit 2 are based on the assumption of a constant neutron flux at the surveillance capsule and pressure vessel locations for the operating period between 13.6 and 54 effective full power years. As required by Regulatory Guide 1.190 and as mentioned in the previous section, no bias or uncertainty is applied to the analytical results.

Cook Unit 2 Capsule U

6-15 Updated lead factors for the D C Cook Unit 2 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, Y. X, and U) were based on the calculated fluence values for the irradiation period corresponding to the time of withdrawal for the individual capsules. For the capsules remaining in the reactor (S. V. W, and Z),

the lead factors correspond to the calculated fluence values at the conclusion of cycle twelve, the last fuel cycle for which fuel cycle specific transport calculations have been completed.

Cook Unit 2 Capsuie U

.I 6-16 Table 6-1 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-]

Length Time Cycle [EFPY] [EFPY] 4 Degrees 40 Degrees 4 Degrees 40 Degrees 1 108 1.08 2.210E+10 7 009E-10 7.517E-17 2.384E-18 2 091 1.99 2.411E-10 7.022E-10 1 445E-18 4.404E-18 1.23 322 2 .297E+10 5.758E-10 2.338E-18 6 640E- 18 4 0,92 4 14 2-318E+10 5.881E-10 3.013E-18 8.354E -18 5 1.11 5.25 2.360E+10 5.248E--10 3.841E-18 1.019E-19 6 1.17 6.42 2.042E-10 5.511E+10 4.593E-18 1.223Eý-19 7 1.12 7.54 2.256E -10 5.069E-10 5388E-18 1 401E+19 8 1.12 8.65 1.784E+-10 5.156E- 10 6.017E-18 1 583E+19 9 1.16 9.82 1.679E+0 5. 210E- 10 6.634E-r-18 1.774E+19 10 1.14 10.96 1.380E-+10 4.656E'10 7.132E+18 1.942E+19 11 1.23 12.20 1.345E+10 5.601EE+ 10 7.655E+ 18 2.160E+19 12 1.41 13.60 1.276E-,10 3.836E+I10 8 221E+18 2.330E+ 19 IRON ATOM DISPLACEMENTS Total Displacement Rate Displacements Cycle Irradiation [dpa"s] [dpa]

Length Time Cycle [EFPY] [EFPYI 4 Degrees 40 Degrees 4 Degrees 40 Denrees 1.08 1.08 3.562E-1 1 182E-10 1 212E-03 4.021E-03 2 0.91 1.99 3.885E-11 1. 184E-10 21330E-03 7.429E-03 3 1.23 3.22 3.702E-11 9.696E-11 3 767E-03 1.119E-02 4 0.92 4.14 3.738E-11 9.896E-I I 4.857E-03 I 408E-02 1.11 5.25 3.8.4E-11 8.829E-11 6.190E-03 1 717E-02 6 117 6.42 3,-93E-11 9 281E-I1 7.404E-03 2.059E-02 7 1.12 7.54 3.638E-I1 8 .522E-1 1 8.685E-03 2.360E-02 8 L12 8.65 2.874E-11 8.666E-11 9.699E-03 2.665E-02 9 1.16 9.82 2.706E- l I 8 758E-11 1.069E-02 2. 987E-02 10 1.14 10.96 2.223E-11 7 820E-1 I. 15OE-02 3.269E-02 11 1.23 1220 2.168E-11 9413E-1I 1 234E-02 3.635E-02 12 1.41 1360 2 056E-I1 6 443E-1l 1.325E-02 921E-02 Cook Unit 2 Capsule U

6-17 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--s]

Cycle Irradiation

!Le~ngth Time Cycle [EFPY] I[EFPY] 0 Degrees 15 Degrees 30 Degrees 45 Degrees 1 1.08 1.08 6.632Eý-09 1.047E-110 1 320E+10 2.013E-+10 2 0.91 1.99 7.370E-09 1.095E--10 1 322E+10 2.050E-l10 3 1.23 3 22 6.926E-09 9.882E-09 1.125E-10 1.656E-10 4 0.92 4.14 6,992E+09 1.112E--10 1.227E-10 1 718E- 10 5 111 5.25 7.136E+09 1.004E+10 1.076E-,10 1.530E+-10 6 1.17 6.42 6.087E+09 1.009E+ 10 1.097E+10 1.612E+10 7 1.12 7.54 6.705E-09 1.068E+10 1.058E+10 1.494E+10 8 1 12 8.65 5.490E-09 7.691E-09 1.064E+ 10 1.499E-10 9 1.16 9.82 5.070E+09 8.569E+09 1.148E+-10 1.483E-10 10 1.14 10.96 4.215E+09 6.790E+09 1.001E+10 1.338E-+-10 11 1.23 12.20 4. 156E+09 6.858E+09 1.092E+10 1+659E+10 12 1.41 13.60 3.946E+-09 6.197E+09 7.883E+09 1.141E+10 Total Neutron Fluence (E > 1.0 MeV) [n/cmj]

Cycle Irradiation Length Time Cycle [EFPY] [EFPY] 0 Degrees 15 Degrees 30 Degrees 45 Degrees 1 1.08 1.08 2.255E+17 3.562E[-7 4.488E-17 6.846E-17

. 0.91 1.99 4. 355E+17 6 682E+-17 S.256E-17 1 269E-, 1 12 3 22 7.043E-17 1.052E-18 1.262E+18 1.912E+18 4 092 4.14 9.081E-17 1376E- 18 1.620E+18 2.412E- 18 1 5125 1.158E-18 1.728E-18 1.997E-18 2.949E-18 6 1.17 6.42 1.383E+18 2. 100E-r18 2.401E-1S 3 543E-18 7 112 7.54 1.619E+ 18 2.476E+ 18 2.774E-18 4.069E- 18 8 1.12 8.65, 1.812E-18 2.747E+ 18 3.149E-18 4.597E+18 9 1.16 9 82 1.998E--18 3.062E-18 3 571E+13 5. 141E+13 10 1.14 10.96 2- 150E-18 3.305E-18 3 931E+1I 5 622E-18 11 123 t2.2.0 2.309E+18 3.568E-18 4.348E-18 6.256E-18 12 1.41 13.60 2.481 E+18 3.838E-18 4.692E-18 6 755E-18 Note At the end of Cycle 12. the maximum fast (E > 1.0 MeV) neutron fluences at the pressure vessel wall occur at an axial elevation 15 2 cm above the midpiane of the active fuel for the 0', 150, 300, and 450 azimuths.

(ook Unit 2 -apsuie U

'e~

6-18 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 45 Degrees Clcte [EFPY] [EFPY] 0 Degrees 15 Degrees 30 Degrees 1 1.08 1.08 1 075E-11 1 677E-1II 2.128E-I1 3.251E-11

?2 0.91 1.99 1. 194E-11 1.753E-11 2.134E- 1 3.312E-1l 3 1.23 322- 1.121E-1 I 1.581E-11 1.814E-II 2.675E-11 4 092 414 1.133E-11 1.779E-1 1.978E-11 2.775E-I1 5 1.11 5.25 1.155E-11 1.606E-11 1.735E-11 2.472E-11 6 1.17 6.42 9.885E-12 1.614E-11 .771E-11 2 603E-11 7 1.12 7.54 1 089E-11 1.707E-I1 1.707E-11 2.412E-1I 8 1.12 8.65 8.886E-12 1.234E-11 1,713E-II 2.422E-11 9 1 16 9.82 8.232E-12 1-373E-11 1.847E-11 2 398E-I1 10 1.14 1096 6834E-12 I.088E-11 1.608E-11 2.162E-11 11 1-23 12.2.0 6.744E-12 IOOE-11 1.7_56E-I 2.677E-11 12 1.41 13.60 6.400E-12 9.917E-12 1.269E-11 1.842E-11 Total Iron Atom Displacements [dpa]

Cycle Irradiation iL ength Time C,,cle [EFPY] [EFP'Y] 0 Deg;rees 15." Degarees "30Degrees 45 Degrees 1 1.08 1 08 3.6-55E -04 5704E-04 7.2-37E-04 1. 106E-03-3 1 0.91 1.99 7 T05E-04 1.070E-03 1.332,E-03 2.049E-03 1.23 3.22 I.141E-03 1.683E-03 2 036E-03 3.087E-03 4 0. 92 4.14 1.47 1E-03 2.202E-03 2.612E-03 3.896E-03 5 1.11 5.25 1,876E-03 2.765E-03 3.220E-03 4.762E-03 6 1.17 6.42 2.240E-03 3 359E-03 3.873E-03 5.722E-03 7 1.12 754 2.623E-03 3.960E-03 4.474E-03 6.57 1E-03 8 1.12 8.65 2.937E-03 4.396E-03 5 078E-03 7 425E-03 9 1.16 9. 82 , 2 3 -03

3. -,,E 4.899E-03 5.75E-03 8.305E-03 10 1.14 10.96 3.485E-03 S 290E-03 6 3.4E-0. 9.083E-0 3.743E-03 5. 711E-03 7 005E-03 1 01IE-02 1 123 12.20 12 1,41 13,60 4,022E-03 6.144E-03 7 560E 1.091E-02 Note At the end of Cycle 12. the maximum iron atom displacements at the pressure vessel wall occur at an axial elevation I52 cm above the midplane of the active fuel for the (0, 15-1 30° and 45' azimuths.

Cook Unit 2 ajpsule :

6-19 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 0.539 231.39 0.262 0.262 0.266 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 0.650 0.638 231.39 0.395 0.390 0.404 0.387 236.90 0.239 0.236 0.247 0.227 242. 42 0. 136 0.134 0.141 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 Umt 2 Capsule U

6-20 Table 6-5 Nuclear Parameters Used In The Evaluation Of Neutron Sensors Target 90% Response Fission Monitor Reaction of Atom Range Product Yield Matenral Interest Fraction (MeV)ý Half-life 63 Copper Cu (na) 0.6917 4.9- 11.8 5.271 y 54 Iron Fe (n,p) 0.0585 2.1 -8.3 312.3 d Nickel 58Ni (n,p) 0.6808 1.5 -8 1 70.82 d 238U (n,f) 1.2 - 6.7 30.07 y 6.02 Uranium-238 37 0.9996 2 -.- p 6.17 Neptunium-237 (nf) 1.0000 0.4-3,5 30.07 y Cobalt-Aluminum 59Co (n,;) 0.0015 non-threshold 5.271 y Note: The 90% response range is defined such that, in the neutron spectrum characteristic of the D. C. Cook Unit 2 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 Urut 2 Capsule U

6-21 Table 6-6 Monthly Thermal Generation During The First Eight Fuel Cycles Of The D C. Cook umt 2 Reactor (Reactor Power of 3411 M-vWt)

Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MvVt-hr) Year Month (NMVt-hr) 78 3 53096 80 12 1488758 83 9 2269321 78 4 521821 81 1 2505373 83 10 1188161 78 653969 81 2 2271684 83 11 453959 78 6 1365478 81 3 1053877 83 12 2383687 78 7 1247083 81 4 0 84 1 2435731 78 8 1529472 81 5 449803 84 2 2235476 78 9 2178779 81 6 2374202 84 3 733977 78 10 2231119 81 7 1775877 84 4 0 78 11 848238 81 8 2338703 84 5 0 78 12 2476056 81 9 2430714 84 6 0 79 1 2240714 81 10 284784 84 7 1417277 79 2 2220562 81 11 2435848 84 8 2341526 79 3 2483455 81 12 2517865 84 9 2325725 79 4 2164269 82 1 2295944 84 10 2423846 79 5 1449347 82 2 2196190 84 11 2056182 79 6 0 82 3 833555 84 12 1059199 79 7 2258164 82 4 2391274 85 1 1361108 79 8 2513690 82 5 2516937 85 2 2271484 79 9 2266726 82 6 2331168 85 3 2461488 79 10 1522346 82 7 2496782 85 4 2449523 79 11 0 82 8 1011517 85 2532441 79 12 0 82 9 2241332 85 6 2451623 80 1 584404 82 10 2293400 85 7 1049002 80 2 2209403 82 11 1575311 85 8 60639 80 3 2418799 82 12 0 85 9 0 80 4 2354329 83 1 341534 85 10 163249 80 5 2483250 83 2 2242228 85 11 1372641 80 6 2187611 83 3 2533602 85 12 2019347 80 7 1408949 83 4 2428234 86 1 2043640 80 8 2496594 83 5 2461540 86 2 1360957 80 9 2393783 83 6 1851461 86 3 0 80 10 1414143 83 7 1711373 86 4 0 80 11 0 83 8 2343637 86 5 0 Cook '-mtr.2 Capsule U

6-22 Table 6-6 Cont'd Monthly Thermal Generation Dunng The First Eleven Fuel Cycles Of The D. C. Cook Unit 2 Reactor (Reactor Power of 3411 Nv[Wt)

Thermal Thermal Thermal Generation Generation Generation Year Month (MWt-hr) Year Month (MNWt-hr) Year Month (MNWt-hr) 86 6 0 89 3 775260 91 12 2474282 7 980980 89 4 2288800 92 1 2442527 86 8 2034055 89 5 2530450 92 2 1219168 86 86 9 1973118 89 6 1297315 86 10 2014579 89 7 2508038 86 11 1974208 89 8 2155830 86 12 2039056 89 9 2452143 87 1 2039325 89 10 2493553 87 2 1776049 89 11 2355817 87 3 159603 89 12 2454307 87 4 456573 90 1 861130 87 2080553 90 2 2282581 87 6 1849770 90 3 2534067 87 7 1409763 90 4 2452883 87 8 1485651 90 5 2165049 87 9 0 90 6 1767222 87 10 1258092 90 7 0 87 11 1973726 90 8 0 87 12 1995088 90 9 0 88 1 2039814 90 10 0 88 2 1900060 90 11 1338296 88 3 2038466 90 12 1887935 88 4 1432639 91 1 2533040 88 5 0 91 2 2110364 88 6 0 91 3 2241782 88 7 0 91 4 2447136 88 8 0 91 5 2495857 88 9 0 91 6 2410166 88 10 0 91 7 2435150 88 11 0 91 8 590907 88 12 0 91 9 2357045 89 1 0 91 10 2519131 89  ? 0 91 11 1912720 Noteý Monthly power generation data were obtained from NU-REG-0020, "Licensed Operating Reactors Status Summarv Report' for the time period spanning March 1978 through February 1992.

Cook Unit 2 Capsule U

6-23 Table 6-7 Calculated 4(E > 1.0 MeV) and C, Factors at the Surveillance Capsule Center Core Midplane Elevation Fuel r(E > 1.0 MeV) [n/cm 2 -s] C, Cycle CapsuleT CapsuleY Capsule X Capsule U T Y X U I 7.009E-10 7.009E+10 7.009E-10 7.009E+10 1.000 1.072 1.1401 1.210 2 7.022E+10 7.022E+10 7.022E+10 1.074 1.143 1.212 3 5.758E+10 5.758E+10 5.758E--10 0.881 0.937 0,994 4 5.88 LE+10 5.881E-10 0.957 1.015 5 5.248E+10 5.248E-10 0.854 0.906 6 5.511EE+-10 0.951 7 5.069E- 10 0.875 8 5.156E+10 0.890 Average 7,009E-;-10 6.535E+10 6.146E-*10 5.794E+10 1.000 1.000 1.000 1.000 Cook UTmt 2 Capsuie U

6-24 Table 6-8 Measured Sensor Activities And Reaction Rates Surveillance Capsule T Radially Radially Adjusted Adjusted Saturated Reaction Measured Saturated Activity Activity Activity Rate (dps/q) (dps/g) (dps/g) (ms/atom)

Reaction Location 3.495E-05 3.338E-05 5.092E-17 6'3Cu (ncx) 60Co Top Middle 4.530E-04 4,430E+04 3.418E-05 3.264E+05 4.979E-17 Middle 4+510E+04 3-479E-05 3.323E-05 5.069E-17 Bottom Middle 5.047E- 17 Average 1.590E-06 3.009E+06 3.159E+06 5.008E-15

'-Fe (np) 5 4Mn Top 1.620E+06 3.066E+06 3.2 19E+06 5.103E-15 Top Middle 2.97 1E-06 3.120E+06 4.945E-15 Middle 1.570E+06 1.630E-.-06 3.085E+06 3.239E-06 5.134E-15 Bottom Middle 1.600E+06 3.028E+06 3.179E+06 5.040E-15 Bottom 5.046E-15 Average 3 590E+07 4.368E+07 5.054E-07 7.236E-15

"_qNi (n,p) 58Co Top Middle 3.500E-.-07 4.259E+07 4.928E+07 7.054E-15 Middle 3.570E+07 4.344E+07 5.026E+07 7.195E-15 Bottom Middle 7.162E-15 Average Middle 1.060E -05 4.335E+06 4.335E-06 2.847E-14

-3*U (n~f) 1'7Cs (Cd) 238U (nif) 137Cs (Cd) Including 235U, 239Pu, and y,.fission corrections: 2.386E-14 8.220E+05 3.362E+07 3.362E+07 2.145E-13 237Np (nf) 137Cs (Cd) Middle 2.112E-13 Including -y,fission correction:

59Co (n,/) 50Co Top 4.960E-0)6 3. 827E-07 3.727E-07 2,432E-12 6.310E+06 4.868E+-07 4.741E+07 3.093E-12 Bottom 2.763E-12 Average Note: Measured specific activities are corrected to a shut down date of October 19, 1979.

Cook '-nit 2 Capsule U

.9 6-25 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 Reaction Location (dps/g) (dps/2) (dps.'g) (rps/atom) 63Cu 9.360E+04 2.947E-05 2-814E+05 4.294E-17 (nca) 6°Co Top Middle Middle 9.590E+04 3.020E-05 2.884E+05 4.399E-17 Bottom Middle 9.460E+04 2.979E-05 2.8455E+05 4.339E-17 Average 4.344E-17 54 54 Fe (n,p) Mn Top 1.760E4-06 2.558E+06 2.685E+06 4.257E-15 Top Middle 1.800E+06 2.616E+06 2.746E+06 4.354E-15 Middle 1.840E+06 2.674E+06 2.807E-06 4.450E-15 Bottom Middle 1.830E+06 2.659E+06 2.792E+06 4.426E-1 5 Bottom 1.871OE-06 2.630E+06 2.762E-06 4.378E-15 Average 4.373E-15 58Nl (n.p) 58 Co Top Middle 3.853E+07 4.458E+07 6.383E-15 2.870E+07 Middle 2.880E+07 3 867E+07 4.474E+07 6.405E-15 Bottom Middle 2.920E+07 3.921E+07 4 536E+07 6.494E-15 Average 6.427E-15 8

23 U (nif) 13 7Cs (Cd) Middle 3.000E+05 4.260E-06 4 260E+06 2 797E-14 2"U (nf) ICs (Cd) 3 Including 2 "U, ,3o" . and -,fission corrections: 2.301E-14

.Np(nf) !3Cs (Cd) Middle 1.780E+06 2.528E-07 2.528E--+07 1.613E-13 3

Np (nf) 1Cs (Cd) Including y,fission correction: 1.588E-13

9Co (n.y) 60Co Top 1.020E-07 3.212E-07 3.128E+07 2.04 1E-12 Bottom 1.020E+07 3.212E-07 3. 128E-+07 2.041E-12 Average 2.041E-12 Note: Measured specific activities are corrected to a shut down date of November 21. 1982.

(XjOk Unit 2 caipsuie U

6-26 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 (dps/g) (dps/_) (dps/_) (ms/atom)

Top Middle 1200E+05 2.902E+05 2.77 1E+05 4.227E-17 63 Cu (nox) 6Co Middle 1.200E+05 2.902E-05 2.77 1E-05 4.227E-17 Bottom Middle 1.220E+05 2.95OE-05 2.817E+05 4 298E-17 Average 4.251E-17 54 Fe (n.p) 5 4NMn Top 1.380E+06 2.555E+06 2.682E+06 4.252E-15 Top Middle 1.4 10E-06 2.6 10OE+06 2.74 1E+06 4.345E-15 Middle 1.400E+06 2.592E+06 2.72 1E+06 4.314E-15 Bottom Middle 1.420E+06 2.629E+06 2.760E-06 4.375E-15 Bottom 1.370E+06 2.536E+06 2.663E+06 4.22 1E-15 Average 4.3011E-13 58Ni (n,p) ý8Co Top Middle 1.840E+07 3.699E-07 4.280E+07 6.128E-15 Middle 1.810E+07 3.639E+07 4.2 1OE+07 6.028E-15 Bottom Middle 1.840E-07 3.699E+07 4 280E+07 6.12 8E-15 Average 6.095E-15

'3'U (nf) 1 37Cs 7

(Cd) Middle 3.760E+05 3.4[6E+06 3.416E-06 2.24SE-14

-- U (n.f) :3 Cs (Cd) Including "3U, 239 Pu, and yfission corrections: 1.816E-14 "3*7N 137CS 3"7Np (nf) (Cd) Middle 3.140E+06 2. 853 E+07 2. 853E+07 1.820E-13

37N\p (n,f) 37Cs (Cd) Including -,fission correction: 1.792E-13 59Co (n,y) Co Top 1.550E-07 3,748E+07 3.650E-'07 2.382E-12 Bottom 1.520E+07 3.675E+07 3.530E-07 2 336E-12 Average 2.359E-12 4'Co (ny) 6°Co (Cd) Top 6.480E+06 1.567E-07 1.805E+07 1. 178E-12 Average 1.178E-12 Noteý Measured specific activities are corrected to a counting date of February 2, 1986 Cook L~rut 2 Capsule U

6-27 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/2) (dps/_) (dps/g) (rps/atorn) 63 Cu (n,cc) 6°Co Top Middle 1 230E+05 2.662E--05 2.542E4-05 3.879E-17 Middle 1.230E+05 2.662E-05 2.542E+05 3.879E-17 Bottom Middle 1.240E+05 2.684E+-05 2.563E÷05 3.910E-17 Average 3.889E-17 4Fe 5 (n,p) s"IMn Top 9.190E+05 2.117E+06 2.222E+06 3.523E-15 Top Middle 9.450E+05 2.177E+06 2.285E4-06 3.623E-15 Middle 9.500E+05 2.188E+06 2.297E-06 3.642E-15 Bottom Middle 8.590E+05 1.978E+06 2.077E+06 3.293E-15 Bottom 9.300E+05 2.142E+06 2.249E+06 3.565E-15 Average 3.529E-15 58Ni (n,p) 58Co Top Middle 3.930E+06 3.153E+07 3.648E+07 5.223E-15 Middle 3.930E--06 3. 153E-+07 3.648E+07 5?223E-15 Bottom Middle 3.960E-06 3 177E+07 3.676E+07 5.263E-15 Average 5.236E-15 23U (n.f) '3'7Cs (Cd) Middle 5 910E-05 3.546E+06 3.546E+06 2.329E-14

'3'U (n.f) 13,Cs (Cd) Including 23TU 39Pu, and 7.fission corrections: 1.836E-14 3 Np (n.f) 3' Cs (Cd) 1.631E-13 Middle 4.260E-06 2.556E-07 2.556E+07

37Np (n.f) 137Cs (Cd) Including y.fission correction: 1.606E-13

ý9Co (n./) 6°Co Top 1.880E-07 4.069E-07 3.963E+07 2.586E-12 Bottom 1.740E-07 3+766E+07 3.668E+07 2.393E-12 Average 2.490E-12 9Co (n.!) '°Co (Cd) Bottom 7.320E-06 1.584E-07 1.825E+07 1.191E-12 Average 1.191E-12 Note: Measured specific activities are indexed to a counting date of August 24. 1092.

Cook .nit 2 Capsule J

6-28 6-28 Table 6-9 Comparison of Measured, Calculated, and Best Estimate Reaction Rates At The Surveillance Capsule Center Capsule T Reaction Rate [rps/atoml I Best Reaction Measured Calculated Estimate NIC UMBE 63Cu(n,ct)6°Co 5.05E-17 4.35E-17 4.90E-17 1.16 1.03 54 Fe(np)54 Mn 5.05E-15 4.87E-15 5.16E-15 1.04 0.98 8

5 Ni(np)5 Co 8

7.16E-15 6.71E-15 7,14E-15 1.07 1.00 238 U(n,f)137Cs (Cd) 2.39E-14 2.42E-14 2.52E-14 0.98 0.95 37 37 2 Np(n,f) Cs (Cd) 2.11E-13 1.90E-13 2-03E-13 1.11 1.04 59Co(ny)06Co 2.76E-12 2.83E-12 2 77E-12 0.98 1.00 Capsule Y Reaction Rate [rps/atom]

Best Reaction Measured Calculated Estimate MIC NI/B E 3 4 34E-17 4.1OE-17 4.26E-17 1.06 1.02 6 Cu(n, c)a°Co 4Fe(n.p)55Mtn 4.37E-15 4.56E-15 4.52E-15 0.96 0.97 5" NNi(n.p)5 8Co 6.43E-15 6.29E-15 6.32E-15 1.02 1.02 23'U(n,f) ý Cs (Cd) 2.30E-14 2.26E-14 2.22E-14 1.02 1.04 3

2 7Np(n,t)137Cs (Cd) 1.59E-13 1.77E-13 1.65E-13 0.90 0.96 59Co(n.7)6Co 2.04E-12 2.63 E-12 2.06E-12 0.78 0.99 Cook Urut 2 Capsule U

6-29 Table 6-9 cont'd Comparison of Measured, Calculated, and Best Estimate Reaction Rates At The Surveillance Capsule Center Capsule X Reaction Rate [rps/atom]

Best i Reaction Measured Calculated Estimate VVC NIBE 63Cu(n,aQ)OCo 4.25E-17 3.92E-17 4.16E-17 1.08 1.02 5"Fe(n.p)5 "Mn 4.30E-15 4.32E-15 4.35E-15 1.00 0.99

"*8Ni(n,p)58Co 6.09E-15 5.95E-15 6.01E-15 1.02 1.01 238U (njf)'3 7Cs (Cd) 1.82E-14 2.13E-14 2.1OE-14 0.85 0.87 1 37 237Np(n,f)O Cs (Cd) 1.79E-13 1.67E-13 1.70E-13 1.08 1*05 5 9Co(n,y)S5Co 2.36E-12 2.46E-12 2.36E-12 0.96 1.00 59Co(n,y) 60Co (Cd) 1.18E-12 1.29E-12 1.18E-12 0.92 1.00 Capsule U Reaction Rate [rps/atom]

Best .

Reaction Measured Calculated Estimate INFC NUBE 63UC(n~a)6PCo 3.89E- 17 3.74E-17 3. 73E- 17 1.04 1.04

ýaFe(np)5 Mn 3.53E-15 4.09E-15 3.74E-15 0.86 0.94 58Ni(n.p)58Co 5.24E-15 5.63E-15 5.23E-15 0.93 1.00 23"U(n, f)37Cs (Cd) 1.84E-14 2.0 1E-14 1.85E-14 0.91 0.99 37 Np(n,f)13 Cs (Cd) 1L61E-13 1.57E-13 1.53E-13 1.02 1.05 9Co(ny)

, vCo 2.49E-12 2.3 1E-12 2.48E-12 1.08 100 Co(n,!)5Co (Cd) 1 19E-12 1.21E-12 1.19E-12 0.99 1 00 Cook Unit 2 Capsule U

6-30 Table 6-10 Comparison of Calculated and Best Estimate Exposure Rates At The Surveillance Capsule Center VbE > 1.0 MeV) [n/cmz-s]

Best ]Uncertainty Capsule ID Calculated Estimate (1a) BE/C T 7 009E-10 7.258E-+-10 6% 1.035 Y 6.535E+10 6.345E+10 6% 0.971 X 6.146E+ 10 6.011E+10 6% 0.978 U 5 794E+10 5.338E+10 6% 0.921 Iron Atom Displacement Rate [dpa/s] _

Best Uncertainty Capsule ID Calculated Estimate (1a) BE/C T 1.182E-10 1.210E- 10 7% 1.023 Y 1.IOIE-10 1.045E-10 7% 0.949 X 1.036E-10 1.004E-10 7% 0.970 U 9.755E- 11 9.066E- 11 7% 0.929 Cook Umut 2 Capsule L-

6-31 Table 6-11 Comparison of MeasuredlCalculated (NtJC) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions Mi/C Ratio Reaction Capsule T Capsule Y Capsule X Capsule U "63Cu(n.ct) 64°Co 1.16 1.06 1 1.08 1.04 54Fe(n,p)V Mn 1.04 0.96 1.00 0U86 "5 8Ni(n.p)f8Co 1.07 1.02 1.02 0.93 "U 3 8(n,p) 3 'Cs (Cd) 0.98 1.02 0.85 0.91 3 (Cd) 37,Np(nf): 7Cs 1.11 0.90 1.08 1.02 Average 1.07 0.99 1.01 0.95

%Standard Deviation 6.4 6.3 9.4 8.0 Note: The overall average M/C ratio for the set of 20 sensor measurements is 1.01 with an associated standard deviation of 8 2S .

Table 6-12 Comparison of Best Estimate/Calculated (BE/C) Exposure Rate Ratios BE/C Ratio Capsule ID (E > 1.0 MeV) dpaIs T 1 04 1.02 y 0.97 0.95 X 0.98 0.97 U 0.92 0.93 Average 0.98 0.97 "iloStandard Deviation 4 8 4.2 Table 6-13 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from D. C. Cook Unit 2 Irradiation Time Fluence (E > 1.0U eV) Iron Displacements Capsule i [EFPY] [n/cml] [dpa]

T 1L08 2.384E+18 4.02 1E-03 Y.2 6.640E- 18 1. 119E-02 X 525 1.019E-19 1.717E-02 U 8.65 1.583E-19 2.66- E-0" Cook U-nit 2 Capsule U

6-32 Table 6-14 Calculated Maximum Fast Neutron Exposure of the D C. Cook Unit 2 Reactor Pressure Vessel at the Clad/Base Metal Interface Neutron Fluence [E > 1.0 MeV]

Cumulative Neutron Fluence [nrcm:]

Operating Time

[EFPYI 0.0 Degrees 15.0 Degrees 30.0 Degrees 45.0 Degrees 12.20 (EOC 11) 2.309E+18 3.568E-18 4.348E+18 6 256Et-18 13.60 (EOC 12) 2.481E+18 3.838E-18 4.692E+18 6.755E-+18 25.00 4.211EE+1 8 6.645E-+18 8.776E+18 1.264E-19 32.00 5.274E+ 18 8 368E-÷18 1.128E+ 19 1.625E&19 36.00 5.881EE- 18 9.353E+18 1.272E+19 1.832E-19 48.00 7.705E+ t 8 1.232E+ 19 1,706E+ 19 2.457E+ 19 54.00 8.626E+ 18 1.381E+ 19 1.923E-19 2.770E+19 Iron Atom Displacements Cumulative Iron Atom Dis Dlacements [dpa]

Operating [,,

Time

[EFPY] 0.0 Degrees 15.0 Degrees 30.0 Degrees 45.0 Degrees 12.20 (EOC 11) 3.743E-03 5.711E-03 7.005E-03 1.011E-02 13.60 (EOC 12) 4.022E-03 6.144E-03 7. 560E-03 1.091E-02 225. 00 6.83 1E-03 1.064E-02 1.413E-02 2.042E-02 32.00 8-556E-03 1.341E-02 i 817E-02 2.625E-02 36.00 9541E-03 1.499E-02' 2.047E-02 2.959E-02 48.00 1.) 50E-02 1.973E-02 2.745E-02 3.968E-022 54.00 1.399E-02 2.212E-02 3 095E-02 4.473E-02 Note For future projections through 36.00 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 48.00 EFPY and 54.00 EFPY for all four azimuthal locations. The future projections also account for a plant uprating from 34 11 MWt to 3800 MNVt at the onset of cycle 13.

Cook nrut 2 Capsule U

6-33 Table 6-15 Calculated Surveillance Capsule Lead Factors Capsule ID And Location Status Lead Factor T (400) Withdrawn EOC 1 3.48 Y (400) Withdrawn EOC 3 3.47 X (400) Withdrawn EOC 5 3.46 U (400) Withdrawn EOC 8 3.44 S (40) In Reactor 1.22 V (40) In Reactor 1 22 W (40) In Reactor 1.22 Z (40 ) In Reactor 1.22 Note: Lead factors for capsules remaining in the reactor are based on cycle specific exposure calculations through fuel cycle 12.

Cook Umt 2 Capsul e

7-1 7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following surveillance capsule removal schedule meets the requirements of ASTM E185-82 and is recommended for future capsules to be removed from the D. C. Cook Unit 2 reactor vessel.

Table 7-1 D. C. Cook Unit 2 Reactor Vessel Surveillance Capsule Withdrawal Schedule Removal Time Fluence Capsule Location Lead Factor(') (EFPY)e (n/cm2,E>1.0 MeV)*(o T 400 3.48 1.08 (Removed) 2.384 x 10"s (c) y 3200 3.47 3.22 (Removed) 6.64 x 10" (c)

X 2200 3.46 5.25 (Removed) 1.019 x I019 (c)

U 1400 3.44 8.65 (Removed) 1.583 x 10'9 (C)

S 40 1.22 32(d) 1.983 x 10'9" Z 3560 1.22 Standby --

W 1840 1-22 Standby --

V 1760 1.22 Standby --

Notes:

(a) Updated in Capsule U dosimetr' analysis, see Section 6.0 of this report.

(b) Effective Full Power Years (EFPY) from plant startup.

(c) Plant specific evaluation.

(d) If license renewal is obtained, Capsule S should then be withdrawn at 48 EFPY at a fluence equal to approximately 2,99 x 10". If license renewal data is desired prior to reaching 48 EFPY. then it is recommended that this capsule be relocated to a higher lead factor location.

Cook LUnit 2 Capsule U

8-1 8 REFERENCES

1. WCAP-85 12, American Electric Power Company Donald C. Cook Unit No. 2 Reactor P~ssel Radiation Surveillance Program, J. A. Davidson, November, 1975.

2. Code of Federal Regulations, 10CFR50, Appendix G; Fracture Toughness Requirements., U.S. Nuclear Regulatory Commission, Washington, D C.

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

4.Section XI of the ASME Boiler and Pressure Vessel Code., Appendix Q. Fracture Toughness Criteria for Protectnon Against 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, StandardPracticefor Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF), in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA. 1993.

7. ASTM E23-88, Standard Test Methods for Notched Bar Impact Testing of Metallic Materials

8. ASTM A370-89, Standard Test Methods and Definitionsfor Mechanical Testing of Steel Products

9. ASTM E8-89b, StandardTest Methods for Tension Testing of Metallic Materials

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

11. ASTM E83-85, StandardPracticefor J4erification and Classification of Extensometers 12- SwRI Project No. 02-5928, Reactor Pessel Material Surveillance Programfor Donald C. Cook Unit No. 2 Analysis of Capsule T, E. B. Norris, September 16, 1981.

13. SwRl Project No. 06-7244-002, Reactor Pessel MaterialSurveillance Programfor Donald C. Cook UnitNo. 2Analysis of Capsule Y, E. B. Norris, February 1984.

14. SwRI Project No. 06-8888, Reactor Vessel Material Surveillance Programfor Donald C. Cook Unit No. 2Analysis of Capsule X, P. K. Nair and M. L. Williams, May 1987.

15- 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.

Cook Unit 2 Capsule U

3-2

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

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

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

19. RSAC-AMP-769, "Analysis of Neutron Dosimetry from D. C. Cook Unit 2 Surveillance Capsule U,"

October 1992.

20. WCAP- 12651, "Nuclear Parameters and Operations Package for the Donald C. Cook Nuclear Plant (Unit 2, Cycle 8),." October 1990.

21. WCAP- 13231, "Nuclear Parameters and Operations Package for the Donald C. Cook Nuclear Plant (Unit 2, Cycle 9)," June 1992.

22. WCAP-14193, "Nuclear Parameters and Operations Package for the Donald C. Cook Nuclear Plant (Unit 2, Cycle 10)," December 1994.

23. WCAP-14614, "Nuclear Parameters and Operations Package for the Donald C. Cook Nuclear Plant (Unit 2, Cycle 1),"April 1996.

24. WCAP-14985, Revision 1, "Nuclear Parameters and Operations Package for the Donald C. Cook Nuclear Plant (Unit 2, Cycle 12)," January 2000.

25. RSICC Computer Code Collection CCC-650, "DOORS 3.1, One, Two- and Three-Dimensional Discrete Ordinates NeutrorvPhoton Transport Code System," August 1996.

26, 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.

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

28. RSIC Data Librarv Collection DLC-178. "SNLRML Recommended Dosimetry Cross-Section Compendium". July 1994 Cook Unit 2 Capsule U

II A-0 APPENDIX A LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS Cook Unit 2 Cipsule U

LOAD r-C~

C-;

N hi C

I' p

-4 m

C

-4 m F

0 0 p

C C 1.

.E 02 "J" OC1L45 ML45

.0.8 1.6 2.4 3.2 4.0 TIME (MSEC)

-J

':;E42 '"= DCM-I.48 rM.48 CU

.D .8 1.6 2.4 3.2 4.0 TIME ( EC Figure A-2. Load-time records for Specimens NIL45 and NML48

C.

3.2 4.0

-rE TIM 0

_3 C -

1.6 TIME ( M=C )

Figure A-3. Load-time records for Specimens MIA2 and MLA4

.9 U

-.J a .8 3.2 4. 0 1.6 2.4 TIME

(<MSC >

C" U

T.6

( MSEC )

Figure A-4. Load-time records for Specimens MvLA and ML43

0

'0 CU 1.6 TIME

( MSEC )

-J

'0 .8 1.6 TIME lSEC M

Figure A-5. Load-time records for Specimens ML46 and ML47

.9 c'J

.8 1.6 2.4 3.2 4.0 TI*E MSEC

.E 02 -U- DCT*6T MT61 a!

S,;

M

-J

.D .8 1.6 2.4 3.2 4.0 TIME ( MSEC )

Figure A-6. Load-time records for Specimens MT62 and MT61

0 Ix -2 "T" mT" cc 9

_J

.8 1.6 2.4 4. 0 TI*E ( ?9SEC )

=

fi 0

'0

_j 0

1.6 TIME ( MSEC )

Figure A-7. Load-time records for Specimens MT66 and MT7I

I.E02 "J" -CMT64 mT64 N

o 1 N

0 '0.

-J V

0 R'

0

.8 1.6 2.4 3.2 4.0 TIME MSEC )

<X

_j 0

1.6 TIME ( MNSC )

Figure A-8. Load-time records for Specimens MT64 and MT72

IE 02 "U" DCr T70 MT7-?

C.

.8 1.6 2.4 3.2 4.0 TIME (MSEC) 0, E 2 'U' DCMT69 MT69 o

• C.,

2.4 3.2 4.0

.8 1.6 TIME ( MrC )

Figure A-9. Load-time records for Specimens MT'70 and MT69

o *d U

C-

'I.

.* B .G2.4 3.2 4.0 TIME ( MSEC )

"*E*2 "U" 0C0T6a MT68 SC!

"-J 0 U C3

?* L.6 2.4 3.2 TI*E ( MrSEC '

Figure A-10. Load-time records for Specimens MT63 and MT68

0 U

C, o

tu

.8 T 1.6 2.4 3.2 4.0 TIME ( M*C )

U CO

-j 0-4.0

.8 1.6 TIME ( MSEC )

Figure A-1I. Load-time records for Specimens MT67 and MT65

I M1L7 0 S

0 0

-J 1.6 TI1*E (CNSC )

M1464 U

0 U

-J 0

0

-J 3.2 4.0

.8 1.6 2.4 TIME ( MSC Figure A-12. Load-time records for Specimens MfW70 and MNW64

0:

.B 1.6 2.4 4. 0 TIME MZEC )

0 1.6 2.4 3.2 4.0 TIME~ (MSEC )

Figure A-13. Load-time records for Specimens MW71 and MW68

( LB ) m10mm3 LOAD CLB ) 910*N3 LOAD 2.4 3.6 4.8 6.0 1.2 2.4 0j 60 I fn

'3D

-4 C' r

Li

ýj C!'

0EA Li 0'

r

'U Li I I I I

LOAO ( LB ) ulOa3 LOAD C LB ) nu0om3 6.0 1.2 2.4 3.6 4.8 2.4 3.6 4..8 0

(1 N

C)'

--4

-4 a

0 1

U) iru N

NJ (4

Ot 0 I I I

LOAD ( LB ) ni0mu3 LOAO ( LB ) &10043 2.4 3.6 4.8 2.4 3.6 4.8 6.0 d0 P1 N

U a,

CD

-4

-4 a' a' a'

0

K P1

-1C0 C]

C) N CA 0)]

3

0) C.)

N A.)

i

J 1.6 2. 3. 4

-U69

.E 2 -U-.M6 0

1.62.4 3.2 4.(

.8 TIME ( MSEC) o L r l

• -4 C.,

-4 T~IE ( Mr1JEC)

Figure A-17. Load-time records for Specimens MWV67 and MWV69

LOAD ( LB ) xl0mm3 LOAD ( LB ) w10-53 4.8 6.0 1.2I 2.4I 3.6I 4.8 I 6.0 1.2 2.4 3.6 *,0 I I I I-

'3 0

-4 In. C)

(-3 I

0 0

U-I a..

I ('1 PC C) 0*

U C)

N A

N A

IL I

Cl

(-3 0o N

A F A

I I I I

'a o

'0

-J On 0

-J (U

C TIME ( MSEC )

OCMH69 MH69

'0 C,

N C U

o N

C

-J C,

'N o

0

-J 0 4.0

.0 1.6 2.4 3.2 TI ME ( MS*ZC)

Figure A-19. Load-time records for Specimens M.-71 and MH69

0C*H72 mR72

%ý "U, OCIH7?

L C

1.6 2.4 3.2 4. 0

.8 TIME ( I*SEC )

CO C.j

• 470 C.

.1 rC

-J

-j

.8 1.6 2.4 3.2 TIME MC )SE NMH70 Figure A-20. Load-time records for Specimens NIH72 and

0*

S'0

-.J 0I DTIME MSEC )

CU ri 64 MH64 Fiur2 AU- LCot

.B .6 2.4 3.2 I-

.4J TIME ( MSEC )

and M1-i64 Figure A-21. Load-time records for Specimens MH62

C3

.8 1.6 2.4 3.2 4.0 TIME ( MSEC )

0~

<C C3,

_3 1.6 2.4 4.0

.8 TIME ( MEC )

Figure A-22. Load-time records for Specimens vfI-H61 and NvIH65

'0 P'ICO o (U

.°I .81.6 2'.4 3.2::

g '0 TIME (MSEC)

C" CU 0E2 ' U" 0C .M H6 6 M6 ci'_

pciesH66 n M6 Figue A-23 Lod-im records fo

.o] .8 1.6 2.4 3.2 TIME co MSEC

B-0 APPENDIX B PHOTOGRAPHS OF CHARPY, TENSILE AND WOL SPECIMENS PRIOR TO TESTING Cook Unia 2 Capsule U

Figure B-1. Charpy impact specimens ML45, ML48. ML42, and ML44 from Intermediate Shell Plate C5521-2 (lonritudinai orientation) before testing.

B-1 RM-28359

Fig~ure B-2. Chary =mpact specimens ML41. NML43, 'M146, 3--d MNL47 from In-ermedi ate Shell Pla~e C51521 Iorztudinal ornerentation,' before testing.

B- 2 RM-28 360

Figure B-3. Charpy impact specimens MT62, MT61, MT66, MT71, MT64, and MTz2 from Intermediate Shell Plate C5521- ' transverse orientation) before testing.

B-3 RM-28361

Figure 3-4. Charpy impact specimens NMT70. MNT 69. MT63, MT68, MT67, and NMT65 from Intermediate Shell Plate C5521-2 (transverse orientation) before testing.

B-4 RM-28362

.9 Fig-ure B-5. Char.y impact specimens .WT70, N*W64, MTV71, MW68. MW63. and MW5V61 from the weld metal, before testing.

B-5 RM-29363

.I Figure E-6. Charpy impact specimens MWV72, MW65, NWV66, MW62. MWV67, and MW69 from the we'd metal, before testing.

B-6 RM-28364

.I Figure B-7. Char-py impact specimens mH67, IF-163, MH71, MH69, MH72, and MIH70 from the heat-affected zone (-kZ!, before testing.

B-7 RM-28365

99ýsz-N8

,ýUl Sal @IOj@q 'ýI-\,-H @Uoz P`;`TTU-lua'ý '9-4" 1,10-11 99HI,ý C, sua=macs iz)rd= ýd-m'qD Puv'29HPý'29HI,'C19KI`ý 4ý91EAý U'

1.

Figure B-9. Tensile snecinmens M'IT11 and MT12 from D. C. Cook Unit 2 reactor vessel Intermediate Shell Plate C5521-2 !transverse orentation) before testing.

B-9 RM*

-29367

Figure B-10. Tensile specimens MW11 and M-W,12 from D. C. Cook Unit 2 reactor vessel weld before testing.

B-lO RM-28368

Figure B-11. WOL specimens NF*5, MW6, NI'W7 and MWVS, from D. C. Cook Unit 2 reactor vessel. The specimens were not tested, but stored for future reference.

B-11 RM-28369

C-1 APPENDLX C CORE POWER DISTRIBUTIONS USED IN THE TRANSPORT CALCULATIONS FOR D. C. COOK UNIT 2

C-2 Average Radial Core Power Distribution Cycle 1 Relative Power 1.106 1.146 1.145 1.187 1.140 1.124 1.023 0.741 1.135 1.130 1.163 1.143 1.178 1.098 1.069 0.767 1.121 1.160 1.144 1.189 1.123 1.121 0.971 0.688 1.179 1.118 1.175 1.133 1.125 1.046 0.998 0.568 1.099 1.142 1.102 1.123 1.160 1.004 0.844 1.096 1.073 1.113 1.054 0.994 0.979 0.518 0.963 1.056 0.962 0.988 0.821 0.505 0.735 0.767 0.684 0.560 Average Radial Core Power Distribution Cycle 2 Relative Power 0.857 1.016 0.969 1.133 1.009 1.064 0.988 0.949 1.019 0.965 1.047 1.071 1.222 1.100 1.169 0.922 0.965 1.050 1.013 1.238 1.019 1.199 0.926 0.832 1.140 1.062 1.228 1.023 1.201 0.961 1.083 0.561 1.004 1.226 1.020 1.211 0.923 0.984 0.902 1.045 1.098 1.191 0.961 0.994 1.044 0.550 0.986 1.154 0.922 1.091 0.907 0.547 0.941 0.922 0.824 0.561 Average Radial Core Power Distribution Cycle 3 Relative Power 0.861 1.059 1.130 1.222 1.136 1.079 1.087 0.853 1.059 1.087 1.215 1.129 1.184 1.148 1.110 0.845 1.138 1.208 1.116 1.200 1.124 1.183 1.045 0.743 1.219 1.127 1.201 0-983 1.146 1.041 1.012 0.435 1.134 1.182 1.120 1.144 1.056 0.956 0.775 1.073 1.145 1.185 1.040 0.958 0.875 0.409 1.060 1.111 1.040 1.012 0.783 0.405 0.850 0.841 0.739 0.430

.I Average Radial Core Power Distribution Cycle 4 Relative Power 0.781 0.902 0.923 0.970 0.934 1.007 0.956 0.906 0.902 1.001 1.277 1.006 1.276 0.956 1.194 0.709 0.920 1.277 1.023 1.020 1.063 1.317 0.977 0.893 0.977 1.008 1.039 1.040 1.347 1.015 1.205 0.527 0.938 1.277 1.058 1.349 0.978 1.279 0.728 0.997 0.937 1.312 1.008 1.274 1.091 0.388 0.963 1.195 0.961 1.202 0.736 0.385 0.905 0.715 0.887 0.527 Average Radial Core Power Distribution Cycle 5 Relative Power 0.832 1.069 1.116 1.197 1.103 1.086 0.824 0.867 1.068 1.110 1.165 0.972 1.201 1.078 1.076 0.868 1.118 1.167 1.171 1.235 1.144 1.175 1.012 0.784 1.200 0.970 1.236 1.164 1.193 1.107 1.046 0.378 1.112 1.210 1.155 1.193 1.131 1.122 0.715 1.092 1.080 1.174 1.112 1.122 0.938 0.325 0.823 1.077 1.018 1.046 0.726 0.339 0.865 0.871 0.786 0.381 Average Radial Core Power Distribution Cycle 6 Relative Power 0.816 1.030 1.084 1.227 1.065 1.031 0.912 0.524 1.034 1.035 1.201 1.160 1.209 1.095 1.094 0.809 1.073 1.206 1.115 1.192 1.140 1.204 1.048 0.788 1.224 1.159 1.195 1.113 1.225 1.149 1.062 0.457 1.067 1.211 1.144 1.223 1.099 1.101 0.601 1.040 1.092 1.204 1.148 1.101 0.942 0.392 0.913 1.095 1.052 1.072 0.603 0.386 0-524 0.807 0.788 0.455

C-4 Average Radial Core Power Distribution Cycle 7 Relative Power 0.926 0-979 1.196 1.040 1.075 1.030 0.680 0.893 0.995 1.034 1.023 1.273 1.045 1.217 0.967 0.921 1.032 0.888 1.245 1.088 1.285 1.029 0.903 0.641 1.255 0.956 1.305 1.061 1.123 0.409 1.211 1.031 1.097 1.296 1.095 1.297 1.100 1.207 0.638 1.129 1.059 1.291 1.064 1.207 1.031 0.321 1.039 1.209 1.025 1.117 0.610 0.323 0.533 0.925 0.891 0.407 Average Radial Core Power Distribution Cycle 8 Relative Power 1.124 1.211 0.915 1.088 1.146 0.630 0.931 1.242 1.139 1.074 1.120 1.296 1.109 1.139 0.604 1.242 0.875 1.274 1.144 1.183 1.175 0.423 1.123 1.076 1.274 0.954 1.265 1.216 1.101 0.348 1.215 1.122 0.916 1.298 1.144 1.264 1.156 1.243 0.833 1.085 1.108 1.193 1.220 1.243 0.794 0.326 1.145 1.138 1.173 1.099 0.832 0.322 0.629 0.603 0.415 0.346 Average Radial Core Power Distribution Cycle 9 Relative Power 0.987 1.307 1.041 0.982 1.223 0.486 0.786 1.323 1.076 1.033 1.110 1.318 1.079 1.241 0.488 1.327 1.033 1.000 1.315 1.067 1.115 1.247 0.568 0.990 1.109 1.313 1.149 1.335 1.007 1.166 0.385 1.307 1.041 1.318 1.067 1.335 1.029 1.215 0.926 1.078 1.115 1.006 1 213 0.610 0.325 0.982 1.223 1.238 1.243 1.164 0.923 0.319 0.487 0.484 0.553 0.384

C-5 Average Radial Core Power Distribution Cycle 10 Relative Power 0.981 1.322 1.064 1.298 1.142 1.118 1.247 0.416 1.322 0.993 1.042 1.049 1.336 1.126 1.241 0.411 1.063 1.042 1.010 1.332 1.153 1.181 1.228 0.397 1.298 1.050 1.333 1.104 1.347 1.044 1.143 0.308 1.143 1.336 1.154 1.348 1.119 1.255 0.864 1.119 1.126 1.182 1.045 1.260 0.662 0.285 1.249 1.243 1.229 1.146 0.869 0.288 0.419 0.412 0.398 0.308 Average Radial Core Power Distribution Cycle 11 Relative Power 0.820 1.345 1.021 1.034 1.332 1.107 1.227 0.399 1.347 1.045 1.304 0.994 1.076 1.107 1.223 0.412 1.021 1.303 1.052 1.317 1.089 1.326 1.177 0.406 1.034 0.994 1.315 1.120 1.086 1.017 1.134 0.311 1.332 1.075 1.097 1.035 1.118 1.322 0.929 1.107 1.108 1.325 1.025 1.322 1.105 0.355 1.228 1.223 1.178 1.134 0.929 0.356 0.399 0.412 0.407 0.316 Average Radial Core Power Distribution Cycle 12 Relative Power 0.776 1.120 1.318 1.371 1.364 1.137 1.155 0.396 1.311 1.076 1.111 1.070 1.347 1.210 0.397 1.120 1.318 1.077 1.340 1.321 1.157 1.351 1.136 0.448 1.371 1.111 1.324 1.140 1.151 1.322 0.782 0251 1.364 1.070 1.158 1.153 1.361 1.193 0.595 1.137 1.348 1.352 1.323 1.193 0.701 0.250 1.154 1.211 1.136 0.782 0.595 0.250 0.395 0.397 0.448 0.251

C-6 Radial Core Power Distribution for Fluence Projections Cycles 10 through 12 Average with 10% Peripheral Bias Relative Power 1.257 1.141 1.236 1.286 1.121 1.207 0.443 0.853 1.125 1.142 1.053 1.152 1.200 1.224 0.447 1.257 1.144 1.323 1.133 1.291 1.177 0.460 1.141 1.142 1.053 1.323 1.122 1.188 1.135 1.110 0.317 1.236 1.286 1.152 1.136 1.172 1.206 1.255 0.867 1.121 1.200 1.292 1.139 1.257 0.908 0.325 1.208 1.224 1.178 1.111 0869 0.327 0.444 0.447 0.461 0-319

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Attachment 4 to AEP:NRC:2349-01 WCAP-15047, Revision 2 "D. C. Cook Unit 2 WOG Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldown Limit Curves for Normal Operation" Dated May 2002