ML20093G337
| ML20093G337 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/31/1995 |
| From: | Osborne M WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20093G335 | List: |
| References | |
| WCAP-12795, WCAP-12795-R03, WCAP-12795-R3, NUDOCS 9510190023 | |
| Download: ML20093G337 (256) | |
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WESTINGHOUSE.NON-PROPRIETARY CLASS 3 __WCAP-12795, Rev. 3 Reactor Cavity Neutron Measurement Program for Wisconsin Electric Power Company Point Beach Unit 2 Stanwood L. Anderson August 1995 B APPROVED: n# A. P. Chwu=. M. P. Osborne, Manager Fluid Systems and Radiation Engineering l I Prepared by Westinghouse for the Wisconsin Electric Power Company ' Purchase Order No. C-46250-C Work performed under Shop Order No. NEXP-450 j l 1 l 1 l WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Business Unit ' P.O. Box 355 l Pittsburgh, Pennsylvania 15230 1 8 1995 Westinghouse Electric Corporation. All rights reserved. l l
1 EXECUTIVE
SUMMARY
1 At the conclusion of Fuel Cycle 14, a reactor cavity measurement i program was instituted at Point Beach Unit 2 to provide a l continuous monitoring of the reactor pressure vessel and reactor i vessel support structure. The use of the cavity measurement program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that l exposure over the service life of the unit. During the first cycle of irradiation with cavity dosimetry installed (Cycle 15), the reactor was operating with a conventional low leakage fuel management strategy. At the onset of Cycle 16 additional neutron flux reduction at the beltline circumferential weld was achieved by the introduction of part length hafnium absorbers in selected peripheral fuel assemblies. A direct comparison of the Cycles 15, 16, 17, and 18 through 20 cavity measurements demonstrated the following incremental flux reduction for the circumferential weld: MEASURED CAVITY FLUX (E > 1.0 MeV) (n/cm2 -sec) CYCLE 15 CYCLE 16 CYCLE 17 CYCLES 18/20 4 0' 1.87E+09 1.35E+09 1.40E+09 1.32E+09 15 1.80E+09 1.28E+09 1.21E+09 1.26E+09 30 1.32E+09 1.02E+09 1.05E+09 1.18E+09 45' 1.12E+09 1.03E+09 9.76E+08 9.87E+08 MEASURED REDUCTION FACTOR CYCLE 15 CYCLE 16 CYCLE 17 CYCLES 18/20 0' 1.00 0.722 0.749 0.706 15* 1.00 0.711 0.672 0.700 30' 1.00 0.773 0.795 0.894 45" 1.00 0.920 0.871 0.881 Due to the relatively short axial extent of the hafnium inserts, the flux reduction impact on the intermediate and lower shell forgings is less dramatic than in the case of the circumferential weld. Based on the continued use of the current (average of Cycles 16 through 20) fuel loading pattern with the part length hafnium absorbers, the projected maximum fast neutron exposure of the
vessel beltline materials at 32 and 48 effective full power years of operation is summarized as follows: NEUTRON FLUENCE (E > 1.0 MeV) [n/cd] 32 EFPY 48 EFPY Beltline Circumferential Weld 2.49E+19 3.42E+19 Intermediate Shell Forging 3.01E+19 4.32E+19 Lower Shell Forging 2.52E+19 3.59E+19 Upper / Int. Shell Circ. Weld 5.48E+18 7.84E+18 Lower Shell/ Head Circ. Weld <1.00E+17 <1.00E+15 As further data are accumulated from subsecuent irradiations, the neutron environment in the vicinity of the Unit 2 pressure vessel will become better characterized and the uncertainties in the vessel exposure projections will be reduced. Thus, the measurement program will permit the assessment of vessel condition to be based on realistic exposure levels with known uncertainties and will eliminate the need for any unnecessary conservatism in the determination of vessel operating parameters. In addition, the excellent three-dimensional fluence profiles established by the measurements, enables the true effects of three-dimensional and potentially non-symmetric flux reduction measures to be accurately accounted for in a manner that would be difficult using analysis alone. All of the calculations and dosimetry evaluations presented in this report have been based on the latest available nuclear cross-section data derived from ENDF/B-VI and are intended to be consistent with the requirements of Draft Regulatory Guide DG-1025, " Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence". As such, the data provided here are intended to supersede prior evaluations documented in References 3, 4, and 5. l l l l l l
TABLE OF CONTENTS page TABLE OF-CONTENTS i LIST'OF FIGURES iii LIST OF TABLES vi 1.0 OVERVIEW OF THE PROGRAM 1-1 '2. 0 DESCRIPTION OF THE MEASUREMENT PROGRAM 2-1 2.1. Description of Reactor Cavity Dosimetry 2-1 2.2 Description of Surveillance Capsule Dosimetry 2-8 3.0 NEUTRON TRANSPORT AND DOSIMETRY EVALUATION 3-1 METHODOLOGIES 3.1 Neutron Transport Analysis Methods 3-1 3.2 Neutron Dosimetry Evaluation Methodology 3-9 3.3 Determination of Best Estimate' Pressure 3-18 Vessel Exposure 4.O RESULTS OF NEUTRON TRANSPORT CALCULATIONS 4-1 4.1 Reference Forward. Calculat-i pn 4-1 4.2 Fuel Cycle Specific Adjoint Calculations 4-15 5.0 EVALUATIONS OF SURVEILLANCE CAPSULE DOSIMETRY 5-1 5.1 Measured Reaction Rates 5-1 ) 5.2 Results of.the Least. Squares Adjustment 5-2 ! Procedure 6.0 EVALUATIONS OF REACTOR CAVITY DOSIMETRY 6-1 6.1 Cycle 15 Results 6-1 6.2 Cycle 16 Results 6-22 6.3 Cycles 17 Results 6-42 6.4-Cycles 18/20 Results 6-62 i 7.0 COMPARISON OF CALCULATIONS WITH MEASUREMENTS 7-1 7.1 Comparison of Least Squares Adjustment 7-1 Results with Calculation 7.2 Comparisons of Measured and Calculated 7-2 Sensor Reaction Rates I l 1 . I l
i 4 1 TABLE OF CONTENTS DASLR 8.0 BEST-ESTIMATE NEUTRON EXPOSURE OF PRESSURE 8-1
- VESSEL MATERIALS 8.1 Exposure Distributions Within the Beltline '8-1 l Region
. 8.2. Exposure of Specific Beltline' Materials 8-16
- 8.3 Uncertainties in Exposure Projections 8-25 I
9.0 REFERENCES
9-1
- APPENDIX A MEASURED SPECIFIC ACTIVITY AND IRRADIATION A-1
! HISTORY-OF SURVEILLANCE CAPSULE SENSOR SETS { ~ APPENDIX B MEASURED SPECIFIC ACTIVITY AND IRRADIATION B-1 i HISTORY OF CYCLE 15 REACTOR CAVITY SENSORS { APPENDIX C MEASURED SPECIFIC ACTIVITY AND IRRADIATION C-1 HISTORY OF CYCLE 16 REACTOR CAVITY SENSORS APPENDIX D MEASURED SPECIFIC ACTIVITY AND IRRADIATION D-1 ! HISTORY OF CYCLES 17 REACTOR CAVITY SENSORS 4 APPENDIX E MEASURED SPECIFIC ACTIVITY AND IRRADIATION E-1 HISTORY OF CYCLES 18/20 REACTOR CAVITY SENSORS 11
LIST OF FIGURES Ficure Title Pace 1.0-1 Description of Pressure Vessel Beltline 1-5 Materials 2.1-1 Azimuthal Location of Sensor Strings 2-5 2.1-2 Azimuthal Location of Multiple Foil 26 Sensor Sets 2.1-3 Irradiation Capsule for-Cavity Sensor 2-7
' Sets 2.2-1 Neutron Sensor Locations Within Internal 2-9 Surveillance Capsules 3.1-1 Reactor Geometry Showing a 45* r,6 Sector 3-7 3.1-2 Internal Surveillance Capsule Geometry 3-8 6.1-1 Fast Neutron Flux (E > 1.0 MeV) as a 6-18 Function of Axial Position Along the 0.0 Degree Traverse in the Reactor Cavity -
Cycle 15 Irradiation 6.1-2 Fast Neutron Flux (E > 1.0 MeV) as a 6-19 Function of Axial Position Along the 15.0 Degree Traverse in the Reactor Cavity - Cycle 15 Irradiation j l 6.1-3 Fast Neutron Flux (E > 1.0 MeV) as a 6-20 { Function of Axial Position Along the 30.0 l Degree Traverse in the Reactor Cavity.- l Cycle 15 Irradiation j 6.1-4 Fast Neutron Flux (E > 1.0 MeV) as a 6-21 l Function of Axial Position Along the 45.0 Degree Traverse in the Reactor Cavity - j Cycle 15 Irradiation l l I 111 ! 1 1 I
LIST OF FIGURES Fiqure Title Page 6.2-1 Fast Neutron Flux (E > 1.0 MeV) as a 6-38 Function of Axial Position Along the 0.0 Degree Traverse in the Reactor Cavity - Cycle 16 Irradiation 6.2-2 Fast Neutron Flux (E > 1.0 MeV) as a 6-39 Function of Axial Position Along the 15.0 Degree Traverse in the Reactor Cavity - Cycle 16 Irradiation 6.2-3 Fast Neutre,n Flux (E > 1.0 MeV) as a 6-40 Function of Axial Position Along the 30.0 Degree Traverse in the Reactor Cavity - Cycle 16 Irradiation 6.2-4 Fast Neutron Flux (E > 1.0 MeV) as a 6-41 Function of Axial Position Along the 45.0 Degree Traverse in the Reactor Cavity - Cycle 16 Irradiation 6.3-1 Fast Neutron Flux (E > 1.0 MeV) as a 6-58 Function of Axial Position Along the 0.0 Degree Traverse in the Reactor Cavity - Cycle 17 Irradiation 6.3-2 Fast Neutron Flux (E > 1.0 MeV) as a 6-59 Function of Axial Position Along the 15.0 Degree Traverse in the Reactor Cavity - Cycle 17 Irradiation 6.3-3 Fast Neutron Flux (E > 1.0 MeV) as a 6-60 Function of Axial Position Along the 30.0 Degree Traverse in the Reactor Cavity - Cycle 17 Irradiation 6.3-4 Fast Neutron Flux (E > 1.0 MeV) as a 6-61 Function of Axial Position Along the 45.0 Degree Traverse in the Reactor Cavity - Cycle 17 Irradiation iv
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LIST OF FIGURES Ficure Title Eagg 6.4-1 Fast Neucron Flux (E > 1.0 MeV) as a 6-78 ) Function of Axial Position Along the 0.0 Degree Traverse in the Reactor Cavity - ! Cycles 18/20 Irradiation i 6.4-2 Fast Neutron Flux (E > 1.0 MeV) as a 6-79 Function of Axial Position Along the 15.0 Degree Traverse in the Reactor Cavity - Cycles 18/20 Irradiation 6.4-3 Fast Neutron Flux (E > 1.0 MeV) as a 6-80 Function of Axial Position Along the 30.0 Degree Traverse in the Reactor Cavity - Cycles 18/20 Irradiation 6.4-4 Fast Neutron Flux (E > 1.0 MeV) as a 6-81 Function of Axial Position Along the 45.0 Degree Traverse in the Practor Cavity - Cycles 18/20 Irradiation 8.2-1 Fast Neutron Fluence (E > 1.0 MeV) as a 8-22 Function of Azimuthal Angle at the Inner Radius of the Beltline Circumferential Weld I '8.2-2 Fast Neutron Fluence (E > 0.1 MeV) as a 8-23 l Function of Azimuthal Angle at the Inner l Radius of the Beltline Circumferential Weld 8.2-3 Iron Atom Displacements [dpa) as a 8-24 l Function of Azimuthal Angle at the Inner l Radius of the Beltline Circumferential Weld i l l V
1 I LIST OF TABLES d Table Title Page 4.1-1 Calculated Reference Neutron Energy Spectra 4-4 at Cavity Sensor Set Locations 4.1-2 Reference Neutron Sensor Reaction Rates and 4-6 Exposure Parameters at the Cavity Sensor Set Locations
. 4.1-3 Calculated Reference Neutron Energy Spectra 4-7
! at Surveillance Capsule Center I 4.1-4 Reference Neutron Sensor Reaction Rates and 4-9 Exposure Parameters at the Center of l Surveillance Capsules t
- 4.1-5 Radial Gradient Corrections for Sensors 4-10 l Contained in Internal Surveillance Capsules 4
4.1-6 Summary of Exposure Rates at the Pressure 4-11 l Vessel Clad / Base Metal Interface l-4.1-7 Relative Radial Distribution of Neutron 4-12
- Flux (E > 1.0 MeV) Within the Pressure
! Vessel Wall . 1 1 4.1-8 Relative Radial Distribution of Neutron 4-13 I
- Flux (E > 0.1 MeV) Within the Pressure Vessel Wall 4.1-9 Relative Radial Distribution of Iron 4-14 i Displacement Rate (dpa) Within the Pressure i
j Vessel Wall l I 4.2-1 Calculated Fast Neutron Flux (E > 1.0 MeV) 4-16 ] at the Center of Reactor Vessel Surveillance Capsules 2-4.2-2 Calculated Fast Neutron Fluence 4-17 } (E > 1.0 MeV) at the Center of Reactor i Vessel Surveillance Capsules 4 i vi
. ~ _ , -_ -
LIST OF TABLES Table Title Page 4.2-3 Calculated Fast Neutron Flux (E > 1.0 MeV) 4-18 at the Pressure Vessel. Clad / Base Metal Interface 4.2-4 Calculated Fast Neutron Fluence 4-19 (E > 1.0 MeV) at the Pressure Vessel Clad / Base Metal Interface 4.2-5 Calculated Fast Neutron Flux (E > 1.0 MeV) 4-20 at the Cavity Sensor Set Locations 4.2-6 Calculated Fast Neutron Fluence 4-21 (E > 1.0 MeV) at the Cavity Sensor Set Locations 4.2-7 Calculated Fast Neutron Flux (E > 0.1 MeV) 4-22 at the Center of Reactor Vessel Surveillance Capsules 4.2-8' Calculated Fast Neutron Fluence 4-23 (E > 0.1 MeV) at the Center of Reactor Vessel Surveillance Capsules 4.2-9 Calculated Fast Neutron Flux (E > 0.1 MeV) 4-24 at the Pressure Vessel Clad / Base Metal Interface 4.2-10 Calculated Fast Neutron Fluence 4-25 (E > 0.1 MeV) at the Pressure Vessel ; Clad / Base Metal Interface 1 4.2-11 Calculated Fast Neutron Flux (E > 0.1 MeV) 4-26 at the Cavity Sensor Set Locations 4.2-12 Calculated Fast Neutron Fluence 4-27 (E > 0.1 MeV) at the Cavity Sensor Set Locations 1 i vil I i
1 l 1 LIST OF TABLES l Table Title Page l 4.2-13 Calculated Iron Displacement Rate 4-28 at the Center of Reactor Vessel Surveillance Capsules 4.2-14 Calculated Iron Displacements at the 4-29 Center of Reactor Vessel Surveillance Capsules l 4.2-15 Calculated Iron Displacement Rate 4-30 l at the Pressure Vessel Clad / Base Metal ' Interface 4.2-16 Calculated Iron Displacements at the 4-31 i Pressure Vessel Clad / Base Metal Interface 4.2-17 Calculated Iron Displacement Rate 4-32 at the Cavity Sensor Set Locations I 4.2-18 Calculated Iron Displacements at the 4-33 Cavity Sensor Set Locations 5.1-1 Summary of Reaction Rates Derived from 5-4 Multiple Foil Sensor Sets Withdrawn from Internal Surveillance Capsules 5.2-1 Derived Exposure Rates from Surveillance 5-5 Capsule V Withdrawn at the End of Fuel Cycle 1 5.2-2 Derived Exposure Rates from Surveillance 5-6 l l Capsule T Withdrawn at the End of Fuel Cycle 3 l 5.2-3 Derived Exposure Rates from Surveillance 5-7 Capsule R Withdrawn at the End of Fuel Cycle 5 5.2-4 Derived Exposure Rates from Surveillance 5-8 Capsule S Withdrawn at the End of Fuel Cycle 16 viii
k LIST OF TABLES Table Title Page 6.1-1 Summary of Reaction Rates Derived from 6-4 Multiple Foil Sensor Sets - Cycle 15 Irradiation , , 6.1-2 Fe-54(n,p) Reaction Rates Derived from 6-5 the Stainless Steel Gradient Chains -
- Cycle 15 Irradiation 6.1-3 Ni-58(n,p) Reaction Rates Derived from 6-6 3
the Stainless Steel Gradient Chains - Cycle 15 Irradiation 6.1-4 Co- 59 (n, y) Reaction Rates Derived from 6-7 t",e Stainless Steel Gradient Chains -
.e 15 Irradiation 6.1-5 Derived Exposure Rates from the Capsule H 6-8 Dosimetry Evaluation - 0.0 Degree Azimuth Core Midplane 6.1-6 Derived Exposure Rates from the Capsule J 6-9 l Dosimetry Evaluation - 15.0 Degree Azimuth l Core Midplane 6.1-7 Derived Exposure Rates from the Capsule K 6-10 Dosimetry Evaluation - 30.0 Degree Azimuth Core Midplane f
6.1-8 Derived Exposure Rates from the Capsule L 6-11 Dosimetry Evaluaticn - 45.0 Degree Azimuth Core Midplane l 1 6.1-9 Derived Exposure Rates from the Capsule G 6-12 Dosimetry Evaluation - 0.0 Degree Azimuth Core Top l l 6.1-10 Derived Exposure Rates from the Capsule I 6-13 l Dosimetry Evaluation - 0.0 Degree Azimuth ' Core Bottom ix 1 i I
)
LIST OF TABLES Table Title Page 6.1-11 Derived Exposure Rates from the Capsule XX 6-14 Dosimetry Evaluation - 0.0 Degree Azimuth Vessel Support l 6.1-12 Fast Neutron Flux (E > 1.0 MeV) as a 6-15 Function of Axial Position Within the Reactor Cavity - Cycle 15 Irradiation 6.1-13 Fast Neutron Flux (E > 0.1 MeV) as a 6-16
, Function of Axial Position Within the Reactor Cavity - Cycle 15 Irradiation 6.1-14 Iron Displacement Rate as a Function 6-17 of Axial Position Within the Reactor Cavity - Cycle 15 Irradiation 6.2-1 Summary of Reaction Rates Derived from 6-25 i Multiple Foil Sensor Sets - Cycle 16 !
Irradiation 1 6.2-2 Fe-54(n,p) Reaction Rates Derived from 6-26 the Stainless Steel Gradient Chains - Cycle 16 Irradiation 6.2-3 Ni-58(n,p) Reaction Rates Derived from 6-27 the Stainless Steel Gradient Chains - Cycle 16 Irradiation 6.2-4 Co-59(n,y) Reaction Rates Derived from 6-28 the Stainless Steel Gradient Chains - Cycle 16 Irradiation 1 6.2-5 Derived Exposure Rates from the Capsule N 6-29 l Dosimetry Evaluation - 0.0 Degree Azimuth Core Midplane 6.2-6 Derived Exposure Rates from the Capsule P 6-30 l Dosimetry Evaluation - 15.0 Degree Azimuth Core Midplane X
LIST OF TABLES Table Title Page 6.2-7 Derived Exposure Rates from the Capsule Q 6-31 Dosimetry Evaluation - 30.0 Degree Azimuth Core Midplane 6.2-8 Derived Exposure Rates from the Capsule R 6-32 l Dosimetry Evaluation - 45.0 Degree Azimuth Core Midplane 6.2-9 Derived Exposure Rates from the Capsule M 6-33 Dosimetry Evaluation - 0.0 Degree Azimuth Core Top 6.2-10 Derived Exposure Rates from the Capsule O 6-34 ) Dosimetry Evaluation - 0.0 Degree Azimuth Core Bottom 6.2-11 Fast Neutron Flux (E > 1.0 MeV) as a 6-35 Function of Axial Position Within the Reactor Cavity - Cycle 16 Irradiation 6.2-12 Fast Neutron Flux (E > 0.1 MeV) as a 6-36 Function of Axial Position Within the Reactor Cavity - Cycle 16 Irradiation 6.2-13 Iron Displacement Rate as a Function 6-37 of Axial Position Within the Reactor Cavity - Cycle 16 Irradiation 6.3-1 Summary of Reaction Rates Derived from 6-45 Multiple Foil Sensor Sets - Cycle 17 Irradiation 6.3-2 Fe-54(n,p) Reaction Rates Derived from 6-46 , the Stainless Steel Gradient Chains - I Cycle 17 Irradiation 6.3-3 Ni-58(n,p) Reaction Rates Derived from 6-47 the Stainless Steel Gradient Chains - Cycle 17 Irradiation
)
xi
LIST OF TABLES Table Title Page 6.3-4 Co-59 (n, y) Reaction Rates Derived from 6-48 the Stainless Steel Gradient Chains - Cycle 17 Irradiation 6.3-5 Derived Exposure Rates from the Capsule BB 6-49 Dosimetry Evaluation - 0.0 Degree Azimuth Core Midplane 6.3-6 Derived Exposure Rates from the Capsule DD 6-50 Dosimetry Evaluation - 15.0 Degree Azimuth Core Midplane 6.3-7 Derived Exposure Rates from the Capsule EE 6-51 Dosimetry Evaluation - 30.0 Degree Azimuth Core Midplane 6.3-8 Derived Exposure Rates from the Capsule FF 6-52 Dosimetry Evaluation - 45.0 Degree Azimuth ) Core Midplane l l 6.3-9 Derived Exposure Rates from the Capsule AA 6-53 Dosimetry Evaluation - 0.0 Degree Azimuth Core Top 6.3-10 Derived Exposure Rates from the Capsule CC 6-54 Dosimetry Evaluation - 0.0 Degree Azimuth Core Bottom 6.3-11 Fast Neutron Flux (E > 1.0 MeV) as a 6-55 Function of Axial Position Within the Reactor Cavity - Cycle 17 Irradiation 6.3-12 Fast Neutron Flux (E > 0.1 MeV) as a 6-56 Function of Axial Position Within the Reactor Cavity - Cycle 17 Irradiation 6.3-13 Iron Displacement Rate as a Function 6-57 of Axial Position Within the Reactor Cavity - Cycle 17 Irradiation xii
l LIST OF TABLES Table Title Page 6.4-1 -Summary of Reaction Rates Derived from 6-65 , Multiple Foil Sensor Sets - Cycles 18/20 l Irradiation 6.4-2 Fe- 54 (n,'p) Reaction Rates Derived from 6-66 the Stainless Steel Gradient Chains - Cycles 18/20 Irradiation 6.4-3 Ni-58 (n,p) . Reaction Rates Derived from 6-67', the Stainless Steel Gradient Chains - Cycles 18/20 Irradiation 6.4-4 Co- 59 (n, y) Reaction Rates Derived from 6-68 the Stainless Steel Gradient Chains - Cycles 18/20 Irradiation 6.4-5 Derived Exposure Rates from the Capsule NN 6-69 Dosimetry Evaluation - 0.0 Degree Azimuth Core Midplane
~
i 6.4-6 Derived Exposure R.ates from the Capsule PP 6-70 l Dosimetry Evaluation - 15.0 Degree Azimuth Core Midplane 6.4-7 Derived Exposure Rates from the Capsule QQ 6-71 l Dosimetry Evaluation - 30.0 Degree Azimuth Core Midplane 6.4-8 Derived Exposure Rates from the Capsule RR 6-72 Dosimetry Evaluation - 45.0 Degree Azimuth l Core Midplane l 6.4-9 Derived Exposure Rates from the Capsule MM 6-73 Dosimetry Evaluation - 0.0 Degree Azimuth Core' Top 6.4-10 Derived Exposure Rates from the Capsule 00 6-74 Dosimetry Evaluation 0.0 Degree Azimuth Core Bottom xiii
1 LIST OF TABLES Table Title Page 6.4-11 Fast Neutron Flux (E > 1.0 MeV) as a 6-75 Function of Axial Position Within the
. Reactor Cavity - Cycles 18/20 Irradiation 6.4-12 Fast Neutron Flux (E > 0.1 MeV) as a 6-76 Function of Axial Position Within the Reactor Cavity - Cycles 18/20 Irradiation 6.4-13 Iron Displacement Rate as a Function 6-77 , of Axial Position Within the Reactor Cavity - Cycles 18/20 Irradiation 7.1-1 Comparison of Measured and Calculated 7-3 Exposure Rates from Surveillance Capsule
. and Cavity Dosimetry Irradiations i 7.2-1 Comparison of Measured and Calculated 7-6 Neutron Sensor Reaction Rates from
- Surveillance Capsule and Cavity Dosimetry Irradiations 8.1-1 Summary of Best Estimate Fast Neutron 8-4 l (E > 1.0 MeV) Exposure Projections for the l Beltline Region of the Point Beach Unit 2 4 Reactor Pressure Vessel - 0 Degree Azimuth 8.1-2 Summary of Best Estimate Fast Neutron 8-5
. (E > 1.0 MeV) Exposure Projections for the
- Beltline Region of the Point Beach Unit 2 2 Reactor Pressure Vessel - 15 Degree Azimuth l 8.1-3 Summary of Best Estimate Fast Neutron 8-6
- (E > 1.0 MeV) Exposure Projections for the
- Beltline Region of the Point Beach Unit 2 Reactor Pressure Vessel - 30 Degree Azimuth G.1-4 Summary of Best Estimate Fast Neutron 8-7 (E > 1.0 MeV) Exposure Projections for the Beltline Region of the Poinc Beach Unit 2 Reactor Pressure Vessel - 45 Degree Azimuth 1
xiv i
LIST OF TABLES Table Title Page 8.1-5 Summary of Best Estimate Fast Neutron 8-8 (E > 0.1 MeV) Exposure Projections for the i Beltline Region of the Point Beach Unit 2 l Reactor Pressure Vessel - 0 Degree Azimuth 8.1-6 Summary of Best Estimate Fast Neutron 8-9 (E > 0.1 MeV) Exposure Projections for the l I Beltline Region of the Point Beach Unit 2 Reactor Pressure Vessel - 15 Degree Azimuth 8.1-7 Summary of Best Estimate Fast Neutron 8-10 (E > 0.1 MeV) Exposure Projections for the Beltline Region of the Point Beach Unit 2 Reactor Pressure Vessel - 30 Degree Azimuth 8.1-8 Summary of Best Estimate Fast Neutron 8-11 (E > 0.1 MeV) Exposure Projections for the Beltline Region of the Point Beach Unit 2 Reactor Pressure Vessel - 45 Degree Azimuth 8.1-9 Summary of Best Estimate Iron Atom 8-12 Displacement Projections for the Beltline Region of the Point Beach Unit 2 Reactor Pressure Vessel - 0 Degree Azimuth 8.1-10 Summary of Best Estimate Iron Atom 8-13 Displacement Projections for the Beltline Region of the Point Beach Unit 2 Reactor Pressure Vessel - 15 Degree Azimuth 8.1-11 Summary of Best Estimate Iron Atom 8-14 Displacement Projections for the Beltline Region of the Point beach Unit 2 Reactor Pressure Vessel - 30 Degree Azimuth ) 8.1-12 Summary of Best Estimate Iron Atom 8-15 Displacement Projections for the Beltline Region of the Point Beach Unit 2 Reactor Pressure Vessel - 45 Degree Azimuth XV i l
LIST OF TABLES Table Title Page 8.2 Maximum Fast Neutron' Exposure of the 8-18 Point Beach Unit 2 Beltline'Circumferential Weld (SA-1484) 8.2-2 Maximum Fast Neutron Exposure of the 8-19 Point Beach Unit 2 Intermediate Shell Forging (123V500) 8.2-3 Maximum Fast' Neutron Exposure of the 8-20 Point Beach Unit 2 Lower Shell Forging 4 (122W195) 8.2-4 Maximum Fast Neutron Exposure of the 8-21 Point Beach Unit 2 Upper / Intermediate l Shell Circumferential Weld (Heat No. 21935) l and the Upper Shell Forging 1 1 xvi
l SECTION 1.0 OVERVIEW OF THE PROGRAM l The Reactor Cavity Neutron Measurement ProgramN initiated at Point Beach Unit 2 at the start of Fuel Cycle 15 was designed to provide a mechanism for the long term continuous monitoring of l the neutron exposure of those portions of the reactor vessel and vessel support structure which may experience radiation induced increases in reference nil ductility transition temperature (RTum) over the nuclear power plant lifetime. When used in conjunction with dosimetry from internal surveillance capsules A and with the results of neutron transport calculations, the reactor cavit'/ neutron dosimetry provides an extensive plant specific measurement data base that can be used to determine the best estimate neutron exposure of the pressure vessel and to project embrittlement gradients through the vessel wall with a minimum uncertainty. Minimizing the uncertainty in the neutron , exposure projections will, in turn, help to assure that the reactor can be operated in the least restrictive mode possible l with respect to 1 - 10CFR50 Appendix G pressure / temperature limit curves for normal heatup and cooldown of the reactor coolant system. 2 - Emergency Response Guideline (ERG) pressure / temperature limit curves . 3 - Pressurized Thermal Shock (PTS) RTns screening criteria. In addition, an accurate measure of the neutron exposure of the reactor vessel and support structure can provide a sound basis for requalification should operation of the plant beyond the current design and/or licensed lifetime prove to be desirable. In the assessment of the state of embrittlement of light water reactor pressure vessels, an accurate evaluation of the neutron exposure of the materials comprising the beltline region of the vessel is required. This exposure evaluation must, in general, include assessments not only at locations of maximum exposure at 1-1
the inner diameter of the vessel, but, also, as a function of axial, azimuthal, and radial location throughout the vessel wall. A schematic of the beltline region of the Point Beach Unit 2 reactor pressure vessel is provided in Figure 1-1. In this case, the beltline region is constructed of two (2) shell forgings and two (2) circumferential welds; one joining the intermediate and lower shell forgings and one joining the intermediate shell forging to the nozzle shell course. Each of these four materials must be considered in the overall embrittlement assessments of the pressure vessel. In order to satisfy the requirements of 10CFR50 Appendix G for the calculation of pressure / temperature limit curves for normal heatup and cooldown of the reactor coolant system, fast neutron exposure levels must be defined at depths within the vessel wall equal to 25 and 75 percent of the wall thickness for each of the materials comprising the beltline region. These locations are commonly referred to as the 1/4T and 3/4T positions in the vessel wall. The 1/4T exposure levels are also used in the determination of upper shelf fracture toughness as specified in 10CFR50 Appendix G. In the determination of values of RTn3 for comparison with applicable pressurized thermal shock screening criteria for plates and circumferential welds, maximum neutron exposure levels experienced by each of the beltline materials are required. These maximum levels will, of course, occur at the vessel inner radius. In the event that a probabalistic fracture mechanics evaluation of the pressure vessel is performed, or if an evaluation of thermal annealing and subsequent material re-embrittlement is undertaken, a complete embrittlement profile is required for the entire volume of the pressure vessel beltline. The determination of this embrittlement profile would, in turn, necessitate the evaluation of neutron exposure gradients throughout the entire beltline. The methodology used to provide these required best estimate neutron expor,ure evaluations for the Point Beach Unit 2 pressure 1-2 _ j
vessel-is based on the underlying philosophy that, in order to minimize the uncertainties associated with vessel exposure projections, plant specific neutron transport calculations must be supported by benchmarking of the analytical approach,
. comparison with industry wide power reactor daca bases of I surveillance capsule and reactor cavity dosimetry, and, ultimately, by validation with plant specific surveillance capsule and reactor cavity dosimetry data bases.
That is, as a progression is made from the use of a purely analytical approach tied to experimental benchmarks to an' approach that makes use of industry and plant specific power reactor measurements to remove potential biases in the analytical method, knowledge regarding the neutron environment applicable to aLspecific reactor vessel is increased and the uncertainty associated with vessel exposure projections is minimized. With this overall methodology in mind, the Reactor Cavity Measurement Program was established to meet the following objectiven: 1 - Provide a measurement data base sufficient to: I a) remove biases that may be present in analytical predictions of neutron exposure; and ; b) support the methodology for the projection of exposure gradients through the thickness of the pressure vessel wall. 2- Establish uncertainties in the best estimate fluence projections for the pressure vessel wall. 3 - Provide a long term continuous monitoring capability for the beltline region of the pressure vessel. This report provides the results of neutron dosimetry evaluations performed subsequent to the completion of Fuel Cycle 20. Fast neutron exposure in terms of fast neutron fluence (E > 1.0 MeV) and dpa is established for all measurement locations in the reactor cavity. The analytical formalism describing the relationship mmong the measurement pointe and locations within 1-3 i
the pressure vessel wall is described and used to project the k' exposure of the vessel itself. Results of exposure evaluations from surveillance capsule' dosimetry withdrawn'at the end of Fuel Cycles 1, 3, 5, and 16 as well as cavity dosimetry results from-Cycles 15, 16, and 17 are
' incorporated to provide the integrated exposure of the pressure '
. vessel from plant startup through the end of Cycle 20. Also, uncertainties associated with the derived exposure parameters at the measurement' locations and with the projected exposure of the pressure vessel are provided.
In addition to the evaluation of the current exposure of the reactor vessel beltline materials, projections of the future exposure of the vessel are also provided. Current evaluations
- and future projections are provided fc each of the beltline weldments as well as for the forgings comprising both the intermediate and lower shells.
All of the calculations and dosimetry evaluations presented in this report are intended to meet the requirements of Draft Regulatory Guide DG-1025, " Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence"; and, have been based on the latest available nuclear cross-section data derived from ENDF/B-VI. As such, the data provided here are intended to supersede prior evaluations documented in references 3, 4, and 5. 1-4
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2.0 DESCRIPTION
OF THE MEASUREMENT PROGRAM 2.1 - Description of Reactor Cavity Dosimetry To achieve the goals of the Reactor Cavity Neutron Measurement
-Program, comprehensive multiple foil sensor sets including radiometric monitors (RM) and solid state track recorders (SSTR) were installed at several locations in the reactor cavity to characterize the neutron energy spectra within the beltline region of the reactor vessel. In addition, gradient chains were used in conjunction with the encapsulated sensors to complete the azimuthal and axial mapping of the neutron environment over the regions of interest.
Placement of the multiple foil sensor sets'was such that spectra evaluations could be made at four azimuthal locations at an axial elevation representative of the midplane of the reactor core. ! 5 The intent here was to determine changes in spectra caused by l I varying amounts of water located between the core and the
- pressure vessel. Due to the irregular shape of the reactor core, water thickness varies significantly as a function of azimuthal angle. In addition to the four midplane sensor sets, two l multiple foil packages were positioned opposite the top and bottom of the active core at the azimuthal angle corresponding to j the maximum neutron flux. Here the intent was to measure variations in neutron spectra over the the core height; particularly near the top of the fuel where backscattering of neutrons from primary loop nozzles and vessel support structures !
could produce significant perturbations. At each of the four azimuthal locations selected for core midplane spectra measurements, gradient chains extended over a fourteen foot height centered on the core midplane. The sensor set deployment described in the preceding paragraphs is ' characteristic of the basic long term monitoring program designed to provide fast neutron exposure assessments for materials comprising the beltline region of the reactor pressure vessel. During the Cycle 15 irradiacion, an additional multiple 2-1
foil sensor set was included in the vicinity of the reactor i vessel supports in order to determine the exposure rate and neutron spectrum at this location well above the beltline region of the reactor vessel. This capsule placement represented a one time measurement that was not repeated as a part of the long term
- monitoring efforts.
4 1 2.1.1 Sensor Placement in the Reactor Cavity A detailed description of the cavity dosimetry hardware and plant specific installation can be found in Reference 1. However, the following information is provided in this report to orient the reader to the plant geometry and the specifics of the sensor
- sets.
The placement of the individual multiple foil sensor sets and
~
gradient chains within the reactor cavity is illustrated in Figures 2.1-1 and 2.1-2. In Figure 2.1-1 a plan view of the azimuthal locations of the four strings of sensor sets is depicted. The strings were located at azimuthal positions of 0, 15, 30, and 45 degrees relative to the core cardinal axis. The sensor strings were hung in the annular gap between the pressure vessel insulation and the primary biological shield at a nominal radius of 79 inches relative to the core centerline. 1 In Figure 2.1-2, the axial extent of each of the sensor set strings is illustrated along with the locations of the multiple foil holders. At the O degree azimuth, multiple foil sets were positioned at the core midplane, opposite the top and bottom of the active fuel, and, during Cycle 15 only, at the elevation of the reactor vessel support. At the 15, 30, and 45 degree azimuthal locations, multiple foil sets were positioned only opposite the core midplane. In all cases, stainless steel gradient chains extended 7 feet relative to the midplane of the active core. 1 The sensor sets and gradient chains were suspended from two ' support bars' mounted on a support frame assembled around the outlet nozzle support shoe of primary loop A. The bottom edges j of the support bars were positioned 26.625 inches above the top j 2-2 l l
- _ _ _ . = _ _ - ..
of the active fuel. The sensor sets and gradient chains were retained and supported at the bottom by chain clamps attached to stainless steel eye nuts with stainless steel threaded chain connectors. The eye nuts were, in turn, atteched to threaded studs embedded in the sump wall. The design of the dosimetry support bars and frames along with the gradient chains and stops ensured correct axial and azimuthal positioning of the dosimetry relative to well known reactor support features. 2.1.2 Description of Irradiation Capsules The sensor sets used to characterize the neutron spectra within the reactor cavity were retained in 3.87 inch x 1.00 inch x 0.50 inch rectangular aluminum 6061 capsules such as that shown in Figure 2.1-3. Each capsule included three compartments to hold the neutron sensors. The top compartment (position 1) was intended to accomodate bare radiometric monitors and SSTR packages, whereas, the two remaining compartments (positions 2 and 3) were meant to house cadmium shielded packages. The separation between positions 1 and 2 was such that cadmium shields inserted into position 2 did not introduce perturbations in the thermal flux in position 1. Aluminum 6061 was selected for the dosimeter capsules in order to minimize neutron flux perturbations at the sensor set locations as well as to limit the radiation levels associated with post-irradiation shipping and handling of the capsules. A summary of the contents of the multiple foil capsules used during each cycle of irradiation is provided in the appendices to this report. 2.1.3 - Description of Gradient Chains i Along with the multiple foil sensor sets placed at discrete locations within the reactor cavity, gradient chains were employed to obtain axial variations of fast neutron exposure along each of the four traverses. Subsequent to irradiation these gradient chains were removed from the cavity and segmented to provide neutron reaction rate measurements at one foot intervals over the height of the axial traverses. These gradient 2-3
chains consisted'of Type 304 stainless steel bead chain of 0.188 inch diameter. 4 When coupled with a chemical analysis, the stainless steel gradient chains yielded activation results for the Fe-54 (n,p) , Ni-58(n,p), and Co-59 (n, y) reactions. The high purity iron, nickel, and cobalt-aluminum foils contained in the multiple foil sensor sets established a direct correlation with the measured reaction rates from the stainless steel chain; and.provided an overcheck on the chemical analysis of the type 304 steel'. , 2-4
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2.2 - Description of Surveillance Capsule Dosimetry Over the course of the first 16 fuel cycles at Point Beach Unit 2, four materials surveillance capsules were withdrawn from their positions between the thermal shield and the reactor vessel. The neutron dosimetry contained within these capsules provided a measure of the integral exposure received by each of the capsules during its respective irradiation period; and established a measurement continuity between the initial startup of the reactor and the initiation of the Reactor Cavity Measurement Program. The specific withdrawal dates of these four capsules were as follows: Capsule V End of Cycle 1 10/74 Capsule T End of Cycle 3 03/77 Capsule R End of Cycle 5 03/79 Capsule S End of Cycle 16 10/90 The type and location of the neutron sensors included in the materials surveillance program are described in some detail in Reference 2; and, are illustrated schematically in Figure 2.2-1 of this report. Relative to Figure 2.2-1, copper, nickel, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers , at several axial levels within each capsule. The l cadmium-shielded uranium and neptunium fission monitors were accomodated within a dosimeter block located near the center of the capsule. In addition to these high purity sensors, iron dosimeters were also obtained by removing samples from several charpy test specimens from various locations within the capsule. Specific information pertinent to the individual sensor sets l included in Capsules V, T, R and S is provided in the appendices to this report. 2-8 I
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4 -_ ~. . . _ . _ .. . _ ._ _ _ . _ . _ . _ . _ _ . _ _ _ _ _ . . .._ _ _ . _ _ _ _ _._ .~ L l' l SECTION'3.0 NEUTRON TRANSPORT'AND DOSIMETRY EVALUATION METHODOLOGIES r
.- As noted in Section 1.0 of this report,-the best. estimate.
- exposure of the reactor pressure vessel was developed using a combination of absolute plant specific neutron transport calculations and plant specific measurements from the reactor-
~
cavity and: internal surveillance capsules. In this section, the
- neutron. transport and~ dosimetry evaluation methodologies are discussed in:some detail; and the approach used to combine the calculations and measurements to produce-the best estimate vessel exposure is presented.
3.1 - Neutron-Transport Analysis Methods Fast neutron exposure-calculations for the reactor and cavity. : geometry were carried out using both forward and adjoint discrete ordinates transport techniques. A single forward calculation
.provided'the relative energy distribution of neutrons for use as input to neutron dosimetry evaluations as well as for use in !
relating measurement results to the actual exposure at key locations in the pressure vessel wall. A series of adjoint calculations, on the other hand, established the means to compute
, absolute exposure rate values using fuel cycle specific core power distributions; thus, providing a direct comparison with all dosimetry.results obtained over the operating history of the reactor.
In combination, the absolute cycle specific data from the adjoint evaluations together with relative neutron energy spectra distributions 1from the forward calculation provided the means to: 1 - Evaluate neutron dosimetry from reactor cavity and surveillance capsule locations.
~
2- Enable a direct comparison of analytical p'ediction
' with' measurement.:
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3- Determine plant specific bias factors to be used in the evaluation of the best estimate exposure of the reactor pressure vessel. 4 - Establish a mechanism for orojection of pressure vessel exposure as the design of each new fuel cycle evolves. 3.1.1 - Reference Forward Calculation A plan view of the reactor geometry at the core midplane elevation is shown in Figure.3.1-l'. Since the reactor exhibits 1/8 core symmetry only a 0-45 degree sector is depicted. In addition to the core, reactor internals, pressure vessel, and the primary biological shield, the model also included explicit representations of the surveillance capsules, the pressure vessel cladding, and the mirror insulation located external to the vessel. The model depicted in Figure 3.1-1 was developed using nominal design dimensions for all components. Specified tolerances in the design dimensions are reflected in the overall uncertainty assessments associated with projected neutron exposures. This modeling approach is consistent with the guidelines of DG-1025. A description of a single surveillance capsule attached to the thermal shield is shown in Figure 3.1-2. From a neutronic standpoint, the inclusion of the surveillance capsules and associated support structures in the analytical model is significant. Since the presence of the capsules and structure has a marked impact on the magnitude of the neutron flux as well as on the relative neutron energy spectra at dosimetry locations within the capsules, a meaningful comparioon of measurement and calculation can be made only if these perturbation effects are properly accounted for in the analysis. In contrast to the relatively massive stainless steel and carbon steel structures associated with the internal surveillance capsules, the small aluminum capsules used in the reactor cavity measurement program were designed to minimize perturbations in the neutron flux and, thus, to provide free field data at the 3-2
)
i
O measurement locations. Therefore, explicit'modeling of'these small capsules in the forward transport model was not required. The forward transport calculation for the reactor model depicted-in Figures 3.1-1.and 3.1-2 was carried out in r,6 geometry.using the DORT two-dimensional discrete ordinates code l* land the BUGLE-93 cross-section library M. The BUGLE-93 library is a.47
' neutron group, ENDFB-VI based, data set produced specifically'for light water. reactor applications. In these analyses, anisotropic
. scattering.was treated with a P3 expansion of the scattering cross-sections.and the angular discretization was modeled with an Se order of angular quadrature. The reference forward
. calculation.was normalized to a core midplane power density
; characteristic of operation at a thermal power level of.1518 MWt.
The spatial core' power distribution utilized in the reference forward calculation was derived.from statistical studies of long-term operation of Westinghouse 2-loop plants. Inherent in the development of this reference core power distribution was the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Due to the use of this bounding spatial power distribution, the results from the reference forward calculation establish conservative exposure projections for reactors of this design operating at the stretch rating of 1518 MWt. Since it is unlikely that actual reactor operation would result in the implementation of a power distribution at the nominal +2a level for a large number of fuel cycles and, further, because of the widespread implementation of low leakage fuel management strategies, the fuel cycle specific calculations for this reactor L result in exposure rates well below these conservative
. predictions. This difference between the conservative forward calculation and the fuel cycle specific computations is
- ' illustrated-by a comparison of the analytical results given in Section 4.0 of this report.
3-3 1 i 4
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3.1.2 - Cycle Specific Adjoint Calculations All adjoint analyses were also carried out using an Se order of angular quadrature and the P3 cross-section approximation from the BUGLE-93 library. Adjoint source locations were chosen at. each of the azimuthal locations containing cavity dosimetry as well as at several key azimuths on the pressure vessel inner radius. In addition, adjoint calculations were carried out for sources positioned at the geometric center of capsules located at 13, 23, and 33 degrees relative to the core cardinal axes. Again, these calculations were run in r,6 geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, $(E > 1.0 MeV). The importance functions generated from these individual adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for each of the fuel cycles to date; and, established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles. Having the importance functions and appropriate core source distributions, the response of interest can be calculated as:
$ ( Ro, 6o) = j, j, ja I(r,6,E) S(r,6,E) r dr d6 dE where: $ (Ro, 6o) =
Neutron flux (E > 1.0 MeV) at radius Ro and azimuthal angle 6o. I(r,6,E) = Adjoint importance function at radius r, azit.nthal angle 6, and neutron source energy E. S (r,6, E) = Neutron source strength at core location r,6 and energy E. 3-4
i i It is important to note that the cycle specific neutron sou rce : distributions, S (r,6, E) , utilized with the adjoint importante functions, I(r,6,E), permited the use not only of fuel cycle specific spatial variations of fission rates within the reactor ! core; but, also allowed for the inclusion of the effects of the differing neutron yield per fission and the variation in fission ; spectrum introduced by the build-in of plutonium isotopes as the l burnup of individual fuel assemblies increased. Although the adjoint importance functions used in these r_nalyses were based on a response function defined by the throchold neutron flux (E > 1.0 MeV), prior calculationsm have shown that, .while the' implementation of low leakage loading patterns significantly impact the magnitude and the spatial distribution of the neutron field, changes in the relative neutron energy l spectrum are of second order. Thus, for a given location the exposure parameter ratios such as [dpa/sec]/[$(E > 1.0 MeV)] are insensitive to changing core source distributions. In the application of these adjoint importance functions to the current evaluations, therefore, calculation of the iron displacement , rates (dpa/sec) and the neutron flux (E > 0.1 MeV) were computed I on a cycle specific basis by using the appropriate [dpa/sec] / [4(E > 1.0 MeV)] and [$ (E > 0.1 MeV) ] / [$ (E > 1. 0 MeV) ] ratios from the reference forward analysis in conjunction with the cycle specific &(E > 1.0 MeV) solutions from the individual adjoint evaluations. In particular, after defining the following exposure rate ratios, (dpaisec} g, ,
$(E> 1.0 MeV)
R, , &(E > 0.1 Men 4(E > 1.0 MeP) 3-5
the corresponding fuel cycle specific exposure rates at the adjoint source locations were computed from the following relations: dpa/sec = ($(E > 1.0 MeV)] R3
$(E > 0.1 MeV) = [$(E > 1.0 MeV)] R2 All fuel cycle specific absolute calulations were also normalized to the current rated power level for Point Beach Unit 2, 1518 MWt..
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4 3.2 - Neutron Dosimetry Evaluation Methodology T The use of passive neutron sensors such as those included in the j internal surveillance capsule and reactor cavity dosimetry sets l does not yield a direct measure of the energy dependent neutron I flux level at the measurement location. 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 flux level and, hence, time integrated exposure (fluence) experienced by the sensors may be developed ] from the measurements only if the sensor characteristics and the parameters of the irradiation are well known. In particular, the l following variables are of interest: i i 1 1 - The measured specific activity of each sensor 2 - The physical characteristics of each sensor i 3 - The operating history of the reactor 4 - The energy response of each sensor j 5 - The neutron energy spectrum at the sensor location ! In this section the procedures used by Westinghouse to determine j sensor specific activities, to develop reaction rates for l individual sensors from the measured specific activities and the operating history of the reactor, and to derive key fast neutron exposure parameters from the measured reaction rates are described. f For the most part, these procedures apply to all of the , evaluations provided in this report. However, in the case of j internal surveillance capsule V, the specific activities of the multiple foil sensor set were determined from prior analysis j performed by Battelle Columbus Laboratory. In this case, the source of the measured specific activity data was referenced and
- the remainder of the data evaluation proceeded using the methodology described in this section.
l 3-9
1 ~3.2.1 - Determination of Sensor Reaction Rates Following irradiation, the multiple foil sensor sets'from surveillance capsule and cavity irradiations along with reactor cavity, gradient chains.were recovered and transported to Pittsburgh for evaluation. . Analysis of all radiometric foils;and gradient' chains was performed at the Westinghouse Analytical Services: Laboratory; while the evaluation of the SSTR sensors from the. cavity irradiations was carried out at the Westinghouse Science and Technology Center Track Recorder Laboratory. 3.2.1.1 - Radiometric Sensors The specific activity of each of the radiometric sensors and gradient chain segments was determined using established ASTM procedures" *"*H9 Following sample preparation and weighing, the specific activity of.each sensor was determined by means of a lithium drifted germanium, Ge(Li), gamma spectrometer. In the case of the surveillance capsule and cavity multiple foil sensor sets, these analyses were performed by direct counting of each of the individual foils or wires; or, as in the case ot'U-238 and Np-237 fission monitors from internal surveillance capsules, by direct counting preceded by dissolution and chemical separation , of' cesium from the' sensor. For the stainless steel gradient chains used in the cavity irradiations, individual sensors were obtained by cutting the chains or wires into a series of segments to provide data points at one foot intervals over an axial' span encompassing 7 feet relative to the reactor core midplane. The irradiation history of the reactor over its operating lifetime was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report". In particular, operating data were extracted from that report on a monthly basis from reactor startup to the end of the current evaluation period. For the sensor sets utilized in surveillance capsule and reactor cavity ' irradiations, 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. 3-10
Having the measured specific activities, the operating history of the reactor, and the physical characteristics of the sensors, reaction rates referenced to full power operation at 1518 MWt were determined from the following equation: A R. No F Y[ C,\1 - c'"1 e~"'
) <
where: A = measured specific activity (dps/gm) R = reaction rate averaged over the irradiation period and referenced to operation at a core power level of Pra (rps/ nucleus). No = number of target element atoms per gram of sensor. F = weight fraction of the target isotope in the sensor material. Y = number of product atoms produced per reaction. Pj = average core power level during irradiation period j (MW). P,.r - maximum or reference core power level of the reactor (B9f) . Cj = calculated ratio of $(E > 1.0 MeV) during i irradiation period j to the time weighted average ;
$(E > 1.0 MeV) over the entire irradiation period.
A = decay constant of the product isotope (sec) . tj = length of irradiation period j (sec). to = decay time following irradiation period j (sec). and the summation is carried out over the total number of monthly intervals comprising the total irradiation period. In the above equation, the ratio P j /Pa r accounts for month by month variation of power level within a given fuel cycle. The ratio C3 is calculated for each fuel cycle using the adjoint transport methodology and accounts for the change in sensor reaction rates caused by variations in flux level due to changes in core power spatial distributions from fuel cycle to fuel 3-11
For a single cycle irradiation C = 1.0. However, for
~
cycle. i multiple cycle irradiations, particularly those employing low leakage fuel management the additional C correction must be utilized.
~
3.2.1.2 - Solid State Track Recorders-Following preparation of the mica discs,-all of'the solid state track < recorders were scanned either manually or with the-Westinghouse STC Automated Track Scanner to determine the number of fissions that occured during the course of the irradiation of
.the sensor' sets. Since the SSTR sensors are. integrating devices not susceptible to radioactive decay of a product isotope, the measurements of total fissions per atom, A, were converted
'directly to reaction rates using the following relationship:
R, A P E) ') where the denominator in the above equation represents the total effective full power seconds of reactor operation during the
- irradiation =of the solid state track recorders.
1 l- The SSTR fissionable deposits were designed for reuse in the long i term monitoring program. Therefore, following processing each sensor was carefully examined to assure that the deposits were neither damaged nor contaminated during irradiation, handling, j and: post-irradiation processing.
- In particular, . these examinations were designed to assure that, E in all-cases, the fission tracks were confined to an area corresponding to'the active portion of the fissionable deposit
- and that the edges of the active area were sharply. defined with a
!. sufficient drop-off in track density to indicate acceptable t- - signal to background ratios for the measurements. Each mica SSTR 3-12 i
._ ~ _ _ ~ . _ _ _ _ _. - . .
and fissionable deposit was also closely inspected under a microscope to verify that no physical damage had occured during exposure or shipment. Selected deposits were also subjected to mass recalibration to verify that no deposit mass had been lost during shipping or exposure. 3.2.1.3 - Corrections to Reaction Rate Data Prior to using the measured reaction rates in the least squares adjustment procedure discussed in Section 3.2.2 of this report, additional corrections were made to the U-238 foil and SSTR measurements to account for the presence of U-235 impurities in the sensors as well as to adjust for the build-in of plutonium e isotopes over the course of the irradiation. These corrections were location and fluence dependent and were derived from a combination of data from the reference forward transport calculation and the cycle specific adjoint analyses as well as
- from measurements made with the U-235 solid state track recorders.
In addition to the corrections made for the presence of U-235 in i the U-238 fission sensors, corrections were also made to both the U-238 and Np-237 sensor reaction rates to account for gamma ray induced fission reactions occuring over the course of the irradiation. These photo-fission corrections were, likewise, ; , location dependent and were based on the reference transport calculations described in Section 3.1.1. In performing the dosimetry evaluations for the internal : surveillance capsules, the sensor reaction rates measured at the l locations shown in Figure 2.2-1 were indexed to the geometric center of the capsules prior to use in the spectrum adjustment procedure. This indexing procedure required correcting the measured reaction rates by the application of analytically determined spatial gradients. For the Point Beach Unit 2 surveillance capsules, the gradient correction factors for each sensor reaction were obtained from the reference forward transport calculation and were used in a multiplicative fashion to relate individual measured reaction rates to the corresponding 4 value at the geometric center of the surveillance capsule. In ; 3-13
the case of the reactor cavity sensors, all of the monitors were located at the same radial and azimuthal location. Thus, gradient corrections were not required in the evaluation of these dosimetry sets. 3.2.2 - Least Squares Adjustment Procedure Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code @l. The FERRET approach used the measured reaction rate data, sensor reaction cross-sections, and a calculated trial spectrum as input and proceeded to adjust the group fluxes from the trial spectrum to produce a best fit (in a least squares sense) to the measured reaction rate data. The
" measured" exposure parameters along with the associated uncertainties were then obtained from the adjusted spectrum.
In the FERRET evaluations, a log-normal least squares algorithm weights both the trial values and the measured data in accordance with the assigned uncertaintiee and correlations. In general, the measured values f are linearly related to the flux & by some response matrix A: I"' " )[ Ag &f' s where i indexes the measured values belonging to a single data set s, g designates the energy group, and a delineates spectra i that may be simultaneously adjusted. For example, R, = [ ag h, s l l relates a set of measured reaction rates Ri to a single spectrum ) 4,by.the multigroup reaction cross-section ag. The log-normal approach automatically accounts for the physical constraint of i positive fluxes, even with large assigned uncertainties. l l 3-14
In the least squares adjustment, the continuous quantities (i.e., neutron spectra and cross-sections) were approximated in a multi-group format consisting of 53 energy groups. The trial input spectrum was converted to the FERRET 53 group structure using the SAND-II codeU!I. This procedure was carried out by first expanding the 47 group calculated spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure in regions where group boundaries do not coincide. The 620 point spectrum was then re-collapsed into the group structure used in FERRET. The sensor set reaction cross-sections, obtained from the ENDF/B-VI dosimetry filep21, were also collapsed into the 53 energy group structure using the SAND-II code. In this instance, the trial spectrum, as expanded to 620 groups, was employed as a weighting function in the cross-section collapsing procedure. Reaction cross-section uncertainties in the form of a 53 x 53 covariance matrix for each sensor reaction were also constructed from the information contained on the ENDF/B-VI data files. These matrices included energy group to energy group uncertainty correlations for each of the individual reactions. However, correlations between cross-sections for different sensor reactions were not included. The omission of this additional uncertainty information does not significantly impact the results of the adjustment. I Due to the importance of providing a trial spectrum that exhibits a relative energy distribution close to the actual spectrum at the sensor set locations, the neutron spectrum input to the l FERRET evaluation was obtained from the plant specific l calculation for each dosimetry location. While the 53 x 53 group l covariance matrices applicable to the sensor reaction cross- ! sections were developed from the cross-section data files, the l covariance matrix for the input trial spectrum was constructed ! I from the following relation: l M,, = R,' + R, R,r P,, i where R, specifies an overall fractional normalization l uncertainty (i .e. , complete correlation) for the set of values. l 3-15 l l l
The fractional uncertainties R, specify additional random uncertainties for group g that are correlated with a correlation matrix given by: P,, = [1 -0] 6,, + 0 c -" where: H W8'Y 2y 2 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes short range correlations over a group range y (6 specifies the strength of the latter term). The value of 6 is 1 when g - g' and 0 otherwise. For the trial spectrum used in the current evaluations, a short range correlation of y = 6 groups was used. This choice implies that neighboring groups are strongly correlated when 6 is close to 1. Strong long range correlations (or anti-correlations) were justified based on information presented by R. E. Maerker t23 ). Maerker's results are closely duplicated when y = 6. For the integral reaction rate covariances, simple normalization and random uncertainties were combined as deduced from experimental uncertainties. In performing the least squares adjustment with the FERRET code, the input spectra from the reference forward transport calculations were normalized to the absolute calculations from the cycle specific adjoint analyses. The specific normalization factors for individual evaluations depended on the location of the sensor set as well as on the neutron flux level at that location. 3-16
l The specific assignment of uncertainties in the measured reaction rates and the: input (trial) spectra used in the FERRET evaluations was as:follows: REACTION RATE UNCERTAINTY 5% FLUX NORMALIZATION UNCERTAINTY 30% FLUX GROUP UNCERTAINTIES I
' (E > 0. 0055 MeV) . 30%
(0.68 ev'< E < 0.0055 MeV) 58% I (E <10.68 ev) 104% 1 SHORT RANGE CORRELATION (E >'O.0055 MeV) 0.9 ,
-(0.68 ev < E < 0.0055 MeV) 0.5 _
(E < 0.68 ev) 0.5 FLUX GROUP CORRELATION RANGE (E > 0.0055 MeV) 6 (0.68 ev < E < 0.0055 MeV) 3
-(E < 0.68 ev) 2 It should be noted that the uncertainties listed for the upper energy ranges extend down to the lower range. Thus, the 58%
group uncertainty in the second range is made up of a 30% uncertainty with a 0.9 short range correlation and a range of 6, and a second part of magnitude 50% with a 0.5 correlation and a range of 3. These input uncertainty assignments were based on prior ' experience in using the FERRET least squares adjustment approach in the analysis of neutron dosimetry from surveillance capsule, reactor _ cavity, and benchmark irradiations. The values are-liberal enough to permit adjustment of the input spectrum to fit the. measured data for all practical applications. 3-17
3.3 - Determination of Best Estimate Pressure Vessel Exposure As noted earlier in this report, the best estimate exposure of the reactor pressure vessel was developed using a combination.of absolute plant specific transport calculations based on the methodology discussed in Section 3.1 and plant specific measurement data determined using the measurement evaluation techniques described in Section 3.2. In particular, the best estimate vessel exposure is obtained from the following relationship:
*aw ze " K #c*.
where: e ue s. = The best estimate fast neutron exposure at the location of interest. K = The plant specific measurement / calculation (M/C) bias factor derived from all available surveillance capsule and reactor cavity dosimetry data. 4au. - The absolute calculated fast neutron exposure at the location of interest. The approach defined in the above equation is based on the l premise that the measurement data represent the most accurate l plant specific information available at the locations of the dosimetry; and, further that the use of the measurement data on a I plant specific basis essentially removes biases present in the I analytical approach and mitigates the uncertainties that would result from the use of analysis alone. That is, at the measurement points the uncertainty in the best estimate exposure is dominated by the uncertainties in the measurement process. At locations within the pressure vessel wall, additional uncertainty is incurred due to the analytically determined relative ratios i l among the various' measurement points and locations within the pressure vessel wall. 3-18 l l l
The implementation of this approach acts to remove plant specific biases associated with the definition of the core source, actual vs. assumed reactor dimensions, and operational variations in water density within the reactor. As a result, the overall uncertainty in the best estimate exposure projections within the vessel wall depend on the individual uncertainties in the measurement process, the uncertainty in the dosimetry location, and in the uncertainty in the calculated ratio of the neutron exposure at the point of interest to that at the measurement location. The uncertainties in the measured flux were derived directly from the results of the least squares evaluation of dosimetry data. The positioning uncertainties were taken from parametric studies of sensor position performed as part of an analytical sensitivity evaluation of the Point Beach Unit 2 reactor. The uncertainties in the exposure ratios relating dosimetry results to positions within the vessel wall were based on analytical sensitivity studies of the vessel thickness tolerance for the cavity data and on downcomer water density variations and vessel inner radius tolerance for the surveillance capsule measurements. l l l 3-19 l l i
SECTION 4.0 RESULTS OF NEUTRON TRANSPORT CALCULATIONS 4.1 Reference Forward Calculation As noted in Section 3.0 of this report, data from the reference forward transport calculation were used in evaluating dosimetry from both reactor cavity and surveillance capsule irradiations as well as in relating the results of these evaluations to the neutron exposure of the pressure vessel wall. In this section,
- the key data extracted from the reference forward calculation is presented and its relevance to the dosimetry evaluations and vessel exposure projections is discussed. The reader should - recall . hat the results of the reference forward transport calculation were intended for use on a relative basis and, therefore, should not be used for absolute comparison with measurement. All absolute comparisons were based on the results of the fuel cycle specific adjoint calculations discussed in Section 4.2.
4.1.1 - Cavity Sensor Set Locations Data from the reference forward calculation pertinent to cavity l sensor evaluations are provided in Tables 4.1-1 and 4.1-2. In Table 4.1-1, the calculated neutron energy spectra applicable to the sensor locations at 0.0, 15.0, 30.0 and 45.0 degrees relative to the core cardinal axes are listed. These data represent the trial spectra used as the starting guess in the
- FERRET least squares adjustment evaluations of the cavity sensor ,
- sets. On a relative basis these calculated energy distributions i 4 establish a baseline against which adjusted spectra may be 1
compared; and, when coupled with the adjoint results of Section 4.2, provide an analytical prediction of absolute neutron spectra at the sensor set locations for each irradiation period. 4-1 I
In Table 4.1-2, the calculated neutron sensor reaction rates associated with the spectra from Table 4.1-1 are provided along with the reference exposure rates in terms of $(E > 1.0 MeV),
$(E > 0.1 MeV) and dpa/sec. Also listed are the associated exposure rate ratios calculated for each of the cavity sensor set locations.
The reference reaction rates, exposure rates, and exposure rate ratios were used in conjunction with fuel cycle specific adjoint transport calculations from Section 4.2 to provide calculated sensor set reaction rates and to project sensor set exposures in terms of 4(E > 0.1 MeV) and dpa/sec for each irradiation period. 4.1.2 - Surveillance Capsule Locations Data from the reference forward calculation pertinent to surveillance capsule evaluations are provided in Tables 4.1-3 through 4.1-5. In Table 4.1-3, the calculated neutron energy spectra at the geometric center of surveillance capsules located at 13, 23, and 33 degrees relative to the core cardinal axes are listed. In Table 4.1-4, the calculated neutron sensor reaction rates and exposure rate ratios associated with the spectra from Table 4.1-3 are provided along with the calculated exposure rates in terms of p(E > 1.0 MeV), $(E > 0.1 MeV) and dpa/sec. Again, these data are applicable to the geometric center of each surveillance capsule. These tabulated data were used in the surveillance capsule dosimetry evaluations and exposure calculations in the same fashion as was the case for the cavity sensor sets. As noted earlier in this report, surveillance capsule dosimetry evaluations also require spatial gradient corrections to be applied to measured reaction rates in sensors dispersed throughout the capsule. In the case of the Point Beach Unit 2 surveillance capsules, neutron sensors were positioned within the specimen array as shown in Figure 2.2-1. In Table 4.1-5, gradient correction factors applicable to the various dosimetry locations are provided for each sensor reaction. 4-2 l l
4.1.3 - Pressure Vessel Wall Data from the reference forward calculation pertinent to the pressure' vessel wall are provided in Tables 4.1-6 through 4.1-9. In Table 4.1-6, the calculated azimuthal distribution of exposure rates in terms of $(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec are listed at approximately 5 degree intervals over the reactor geometry. These data are applicable to the clad / base metal interface. Also given in Table 4.1-6 are the exposure rate ratios [$ (E > 0.1 MeV) ] / [$ (E > 1. 0 MeV) ] and [dpa/sec] / [$ (E > 1.0 MeV)] that provide an indication of the variation in neutron spectrum as a function of azimuthal angle at the pressure vessel inner radius.
. Radial gradient information for $(E > 1.0 MeV), $(E > 0.1 Mev),
and dpa/sec is given in Tables 4.1-7, 4.1-8, and 4.1-9, respectively. These data are presented on a relative basis for each exposure parameter at the 0, 15, 30, and 45 degree azimuthal' locations. Exposure rate distributions within the vessel wall ; are obtained by normalizing the calculated or best estimate exposure at the vessel inner radius to the gradient data given in Tables 4.1-7 through 4.1-9. l 4-3 l 1 f _ __ __
TABLE 4.1-1 CALCULATED REFERENCE NEUTRON ENERGY SPECTRA AT CAVITY SENSOR SET LOCATIONS [n/cm2 -sec) 1518 MWt; F, - 1.2 LOWER ENERGY AZIMUTHAL ANGLE (MeV) 0 . 0' 15.0* 30.0' 45.0* 1.42E+01- 6.55E+05 5.19E+05 3.99E+05 3.83E+05~ 1.22E+01 1.94E+06 1.52E+06 1.15E+06 1.10E+06 1.00E+01 8.02E+06 6.14E+06 4.59E+06 4.31E+06 8.61E+00 1.50E+07 1.14E+07 8.44E+06 7.86E+06 7.41E+00 2.39E+07 1.79E+07 1.31E+07 1.21E+07~ 6.07E+00 5.15E+07 3.83E+07 2.78E+07 2.54E+07 4.97E+00 7.54E+07 5.62E+07 4.03E+07 3.61E+07 3.68E+00 1.52E+08 1.14E+08 8.00E+07 6.98E+07 3.01E+00- 1.28E+08 9.65E+07 6.70E+07 5.76E+07 2.73E+00 1.06E+08 7.96E+07 5.47E+07 4.69E+07 2.47E+00 1.34E+08 1.02E+08 7.02E+07 5.95E+07 2.37E+00 8.19E+07. 6.26E+07 4.30E+07 3.63E+07 2.35E+00 2.52E+07 1.89E+07 1.28E+07 1.09E+07 2.23E+00 1.09E+08 8.25E+07 5.61E+07 4.77E+07 1.92E+00 3.06E+08 2.32E+08 1.58E+08 1.34E+08 1.65E+00 4.28E+08 3.26E+08 2.22E+08 1.86E+08 1.35E+00 7.32E+08 5.69E+08 3.86E+08 3.20E+08 1.00E+00 1.69E+09 1.34E+09 9.09E+08 7.44E+08 8.21E 1.58E+09 1.27E+09 8.60E+08 6.98E+08 7.43E-01 7.57E+08 6.42E+08 4.40E+08 3.44E+08 6.08E-01 3.27E+09 2.68E+09 1.82E+09 1.47E+09 4.98E-01 3.08E+09 2.60E+09 1.78E+09 1.40E+09 3.69E-01 3.20E+09 2.73E+09 1.89E+09 1.48E+09 ) NOTE: The upper energy of group 1 is 17.33 Mev. 4-4 i 1
TABLE 4.1-1 (continued) CALCULATED REFERENCE NEUTRON ENERGY SPECTRA AT CAVITY SENSOR SET LOCATIONS [n/cm2 -sec) 1518 MWt; F, = 1. 2 LOWER ENERGY AZIMUTHAL ANGLE (MeV) 0 . 0" _15 . 0' 30.0' 45.0' 2.97E-01 4.34E+09 3.67E+09 2.53E+09 2.00E+09 1.83E-01 5.44E+09 4.84E+09 3.37E+09 2.58E+09 1.11E-01 5.93E+09 5.26E+09 3.68E+09 2.84E+09 6.74E-02 3.96E+09 3.57E+09 2.51E+09 1.93E+09 4.09E-02 3.02E+09 2.74E+09 1.94E+09 1.49E+09 3.18E-02 9.38E+08 8.69E+08 6.21E+08 4.72E+08 2.61E-02 4.40E+08 4.16E+08 2.96E+08 2.20E+08 2.42E-02 1.61E+09 1.37E+09 9.63E+08 7.79E+08 2.19E-02 9.99E+08 8.88E+08 6.23E+08 4.84E+08 1.50E-02 1.89E+09 1.77E+09 1.25E+09 9.29E+08 7.10E-03 2.52E+09 2.36E+09 1.69E+09 1.28E+09 3.36E-03 2.87E+09 2.65E+09 1.90E+09 1.46E+09 1.59E-03 2.42E+09 2.24E+09 1.62E+09 1.25E+09 4.54E-04 3.73E+09 3.45E+09 2.50E+09 1.95E+09 2.14E-04 1.87E+09 1.74E+09 1.27E+09 9.87E+08 1.01E-04 2.00E+09 1.85E+09 1.35E+09 1.06E+09 I 3.73E-05 2.57E+09 2.37E+09 1.73E+09 1.37E+09 I 1.07E-05 3.00E+09 2.76E+09 2.02E+09 1.61E+09 5.04E-06 1.63E+09 1.50E+09 1.10E+09 8.76E+08 1.86E-06 2.02E+09 1.86E+09 1.36E+09 1.09E+09 8.76E-07 1.42E+09 1.30E+09 9.59E+08 7.71E+08 4.14E-07 1.07E+09 9.87E+08 7.26E+08 5.86E+08 1.00E-07 2.52E+09 2.26E+09 1.67E+09 1.40E+09 0.00 6.55E+09 5.21E+09 3.79E+09 3.64E+09 NOTE: The upper energy of group 1 is 17.33 Mev. 4 4 4-5 J
TABLE 4.1-2 REFERENCE NEUTRON SENSOR REACTION RATES AND EXPOSURE PARAMETERS AT THE CAVITY SENSOR SET LOCATIONS 1518 MWt; F, = 1.20 AZIMUTHAL ANGLE 0 . 0' 15 . 0* - 30.0* 45.0' Reaction Rate (ros/ nucleus) Cu-63(n,a) (Cd) 2.05E-18 1.54E-18 1.13E-18 1.04E-18 Ti-46(n,p) (Cd) 3.06E-17 2.29E-17 1.66E-17 1.51E-17 Fe-54(n,p) (Cd) 1.99E-16 1.50E-16 1.06E-16 9.31E-17 Ni-58(n,p) (Cd) 2.87E-16 2.16E-16 1.52E-16 1.33E-16 U-238(n,f) (Cd) 1.18E-15 9.03E-16 6.22E-16 5.29E-16 Np-237(n,f) (Cd) 1.61E-14 1.30E-14 8.94E-15 7.24E-15 i Co- 59 (n, y) 3.88E-13 3.34E-13 2.44E-13 2.10E-13 Co- 59 (n, y) (Cd) 1.82E-13 1.68E-13 1.23E-13 9.68E-14 U-235(n,f) 3.64E-12 3.01E-12 2.18E-12 2.00E-12 U-235(n,f) (Cd) 6.47E-13 5.92E-13 4.31E-13 3.42E-13 U-238(y,f) 4.25E-17 3.43E-17 2.49E-17 2.27E-17 Np-237(y,f) 1.29E-16 1.04E-16 7.15E-17 6.52E-17 Neutron Flux (n/cm2 -sec) 4(E > 1.0 MeV) 4.14E+09 3.20E+09 2.19E+09 1.83E+09
$ (E > 0.1 MeV) 3.25E+10 2.76E+10 1.90E+10 1.50E+10 doa/sec Displacement Rate 1.17E-11 9.70E-12 6.70E-12 5.36E-12 $ (E > 0.1) /4 (E > 1. 0) 7.76 8.63 8.72 8.20
[dpa/sec] /$ (E > 1.0) 2.83E-21 3.03E-21 3.06E-21 2.93E-21 U238(y,f)/U238(n,f) 0.036 0.038 0.040 0.043 Np237(y,f)/Np237(n,f) 0.008 0.008 0.008 0.009 4-6
TABLE 4.1-3 CALCULATED REFERENCE NEUTRON ENERGY SPECTRA SURVEILLANCE CAPSULE CENTER [n/cm2 -sec] 1518 MWt; F = 1. 2
' LOWER ENERGY AZIMUTHAL ANGLE IMeV) 13.0' 23.0' 33.0 >
1.42E+01 1.63E+07 1.34E+07 1.16E+07 1.22E+01 5.40E+07 4.34E+07 3.75E+07 1.00E+01 2.47E+08 1.94E+08 1.67E+08 8.61E+00 4.97E+08 3.83E+08 3.32E+08 7.41E+00 9.11E+08 6.82E+08 5.94E+08 6.07E+00 2.28E+09 1.69E+09 1.47E+09 4.97E+00 3.71E+09 2.64E+09 2.33E+09 3.68E+00 8.25E+09 5.50E+09 4.98E+09 3.01E+00 7.23E+09 4.53E+09 4.19E+09 2.73E+00 5.72E+09 3.55E+09 3.29E+09 2.47E+00 6.89E+09 4.22E+09 3.94E+09 2.37E+00 4.02E+09 2.47E+09 2.30E+09 2.35E+00 1.12E+09 6.91E+08 6.44E+08 2.23E+00 4.91E+09 2.99E+09 2.80E+09 1.92E+00 1.39E+10 8.32E+09 7.84E+09 . 1.65E+00 1.72E+10 1.00E+10 9.54E+09 1.35E+00 2.71E+10 1.55E+10 1.48E+10 1.00E+00 5.51E+10 3.04E+10 2.93E+10 8.21E-01 3.95E+10 2.13E+10 2.06E+10 7.43E-01 2.01E+10 1.08E+10 1.05E+10
- 6.08E-01 6.22E+10 3.24E+10 3.17E+10 4.98E-01 5.22E+10 2.69E+10 2.63E+10 3.69E-01 5.72E+10 2.95E+10 2.88E+10 NOTE
- The upper energy of group 1 is 17.33 Mev.
i 4-7
TABLE 4.1-3 (continued) CALCULATED REFERENCE NEUTRON ENERGY SPECTRA 2 SURVEILLANCE CAPSULE CENTER- (n/cm -sec) 1518 MWt; F, - 1.2 LOWER ENERGY AZIMUTHAL ANGLE (MeV) 13.0' 23.0' 33.0' 2.978-01 5.58E+10 2.84E+10 2.78E+10 1.83E-01 7.43E+10 3.79E+10 3.71E+10 1.11E-01 7.06E+10 3.56E+10 3.50E+10 6.74E-02 5.41E+10 2.72E+10 2.67E+10 4.09E-02 4.48E+10 2.25E+10 2.21E+10 3.18E-02 1.64E+10 8.24E+09 8.10E+09 2.61E-02 7.35E+09 3.59E+09 3.51B+09 2.42E-02 1.49E+10 7.43E+09 7.33E+09 2.19E-02 9.31E+09 4.63E+09 4.56E+09 1.50E-02 2.39E+10 1.20E+10 1.18E+10 7.10E-03 4.41E+10 2.22E+10 2.18E+10 3.36E-03 5.02E+10 2.52E+10 2.47E+10 1.59E-03 4.68E+10 2.33E+10 2.30E+10 4.54E-04 7.73E+10 3.85E+10 3.79E+10 2.14E-04 4.26E+10 2.12E+10 2.08E+10 1.01E-04 4.70E+10 2.33E+10 2.30E+10 3.73E-05 6.16E+10 3.04E+10 3.00E+10 1.07E-05 7.50E+10 3.69E<10 3.65E+10 5.04E-06 4.29E+10 2.11E+10 2.08E+10 1.86E-06 5.63E+10 2.78E+10 2.74E+10 8.76E-07 4.16E+10 2.06E+10 2.02E+10 4.14E-07 3.52E+10 1.76E+10 1.72E+10 1.00E-07 8.60E+10 4.39E+10 4.25E+10 0.00 2.45E+11 1.31E+11 1.23E+11 NOTE: The upper energy of group 1 is 17.33 Mev. 4-8
TABLE 4.1-4 REFERENCE NEUTRON SENSOR REACTION RATES AND EXPOSURE PARAMETERS AT THE CENTER OF SURVEILLANCE CAPSULES 1518 MWt; F, = 1.20 13.0' 23.0' 33.0" Reaction Rate (ros/ nucleus) Cu- 63 (n, a) 8.11E-17 6.06E-17 5.29E-17 Fe- 54 (n, p) 9.91E-15 6.64E-15 6.03E-15 Ni-58(n,p) 1.38E-14 9.14E-15 8.29E-15 U-238(n,f) (Cd) 5.14E-14 3.16E-14 2.94E-14 Np- 2 3 7 (n, f ) (Cd) 4.43E-13 2.48E-13 2.38E-13 Co-59 (n, y) 1.13E-11 5.87E-12 5.62E-12 Co- 59 (n, y) (Cd) 4.40e-12 2.17E-12 2.15E-12 U-238 (y, f) 2.58E-15 1.45E-15 1.40E-15 Np- 2 3 7 (7, f ) 7.21E-15 4.06E-15 3.91E-15 Neutron Flux (n/cm2 -sec)
$(E > 1.0 MeV) 1.59E+11 9.35E+10 8.83E+10
$(E > 0.1 MeV) 6.02E+11 3.22E+11 3.11E+11 doa/sec Displacement Rate 2.83E-10 1.59E-10 1.52E-10
$ (E > 0.1) /$ (E > 1. 0) 3.79 3.44 3.52 [dpa/sec]/$(E > 1.0) 1.78E-21 1.70E-21 1.72E-21 U23 8 (7, f) /U23 8 (n, f) 0.050 0.046 0.048 Np237 (y, f) /Np237 (n, f) 0.016 0.016 0.016 4-9
t TABLE 4.1-5 RADIAL GRADIENT CORRECTIONS FOR SENSORS CONTAINED IN INTERNAL SURVEILLANCE CAPSULES 13' CAPSULE Radial Location (cm) 157.59 158.35 158.59 Cu- 63 (n, a) 0.88 1.00 1.04 Fe- 54 (n, p) 0.87 .1. '00 1.05 .Ni-58(n,p) 0.87 1.00 1.05 U-238(n,f) Cd 0.87 1.00 1.05 Np-237(n,f) Cd 0.87 1.00- 1.05 Co-59 (n, y) 0.95 1.00 0.99 Co-59 (n, y) Cd 0.83 1.00. 1.07 23' CAPSULE
' Radial Location (cm) 157.59 158.35 158.59 Cu- 63 (n, a) 0.87 1.00 1.04 Fe-54 (n, p) 0.86 1.00 1.04 Ni-58(n,p) 0.86 1.00 1.04 U-238(n,f) Cd 0.87 1.00 1.05 Np-237(n,f) Cd 0.87 1.00 1.05 Co-59 (n, y) 0.95 1.00 0.99 Co-59 (n, y) Cd 0.84 1.00 1.08 33 CAPSULE Radial Location (cm) 157.59 158.35 158.59 Cu- 63 (n, a) 0.87 1.00 1.04 Fe- 54 (n, p) 0.86 1.00 1.05 Ni-58(n,p) 0.86 1.00 1.05 U-238 (n, f) Cd 0.86 1.00 1.05 Np- 237 (n, f ) Cd 0.87 1.00 1.05 i Co- 59 (n, y) 0.95 1.00 0.99 l Co-59 (n, y) Cd 0.83 1.00 1.08 l 1
l l l 4-10 l 1 l
TABLE 4.1-6
SUMMARY
OF EXPOSURE RATES AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE FLUX (n/cm2 -sec) THETA [E > 0.11 doa/sec (deo) (E > 1.0) (E > 0.1) doa/sec IE > 1.01 IE > 1.01 0.25 5.32E+10 1.46E+11 8.68E-11 2.74 1.63E-21 4.75 5.11E+10 1.40E+11' 8.35E-11 2.75 1.64E-21 9.75 4.37E+10 1.23E+11 7.22E-11 2.81 1.65E-21 15.00 3.25E+10 9.45E+10 5.46E-11 2.91 1.68E-21 19.75 2.58E+10 7.13E+10 4.27E-11 2.76 1.65E-21 24.75 2.29E+10 6.36E+10 3.79E-11 2.78 1.66E-21 30.00 2.22E+10 6.05E+10 3.65E-11 2.73 1.64E-21 34.75 2.05E+10 5.72E+10 3.40E-11 2.79 1.66E-21 39.75 1.96E+10 5.17E+10 3.19E-11 2.63 1.62E-21 44.75 1.87E+10 4.91E+10 3.03E-11 2.63 1.62E-21 l 4-11
TABLE 4.1-7 RELATIVE RADIAL DISTRIBUTION OF NEUTRON FLUX (E > 1.0 MeV) WITHIN THE PRESSURE VESSEL WALL RADIUS AZIMUTHAL ANGLE (cm) O' .15' . 30' 45' 168.04 1.000 1.000 1.000 1.000 168.27 0.987 0.987 0.985 0.987 168.88 0.940 0.942 0.937 0.942 169.75 0.862 0.865 0.857 0.866 170.93 0.754 0.757 0.749 0.760 172.25 0.639 0.644 0.636 0.647 173.53 0.540 0.546 0.539 0.550 174.98 0.444 0.451 0.444 0.454 176.46 0.362 0.370 0.363 0.372 177.58 0.308 0.317 0.311 0.318 179.03 0.250 0.259 0.253 0.260 180.66 0.196 0.206 0.201 0.206 181.63 0.169 0.179 0.175 0.178 182.60 0.144 0.154 0.151 0.154 184.06 0.110 0.122 0.120 0.122 184.87 0.101 0.113 0.112 0.113 Note: Base Metal Inner Radius = 168.04 cm. Base Metal 1/4T = 172.25 cm. Base Metal 1/2T = 176.46 cm. Base Metal 3/4T = 180.66 cm. Base Metal Outer Radius = 184.87 cm. i i a 4-12 __m. .z__1 , - - .__ .__ -
. . . .-. . .~. . - - . ..
l TABLE 4.1-8 RELATIVE RADIAL DISTRIBUTION OF NEUTRON FLUX (E > 0.1 MeV) WITHIN THE PRESSURE VESSEL WALL RADIUS AZIMUTHAL ANGLE (cm) 0* 15' 30* 45 168.04 1.000 1.000 1.000 1.000 168.27 1.005 1.007 1.005 1.007 168.88 1.002 1.007 1.004 1.008 169.75 0.980 0.990 0.985 0.992 170.93 0.934 0.948 0.945 0.953 172.25 0.873 0.891 0.889 0.899 173.53 0.809 0.831 0.831 0.841 174.98 0.736 0.763 0.763 0.773 176.46 0.662 0.693 0.694 0.703 177.58 0.606 0.640 0.642 0.650 179.03 0.536 0.573 0.577 0.582 180.66 0.461 0.502 0.507 0.509 181.63 0.416 0.458 0.466 0.465 182.60 0.369 0.415 0.423 0.421 184.06 0.298 0.348 0.361 0.357 184.87 0.276 0.327 0.343 0.339 Note: Base Metal Inner Radius = 168.04 cm. Base Metal 1/4T = 172.25 cm. Base Metal 1/2T = 176.46 cm. Base Metal 3/4T = 180.66 cm. Base Metal Outer Radius = 184.87 cm. l 4-13
TABLE 4.1-9 RELATIVE RADIAL DISTRIBUTION OF IRON DISPLACEMENT RATE (dpa) ' WITHIN THE PRESSURE VESSEL WALL RADIUS AZIMUTHAL ANGLE (cm) 0* 15* __.3.D* _ 45* 168.04 1.000 1.000 1.000 1.000 168.27 0.988 0.990 0.988 0.989 168.88 0.951 0.955 0.950 0.954 169.75 0.889 0.896 0.889 0.857 170.93 0.804 0.814 0.805 0.812 172.25 0.712 0.726 0.716 0.723 173.53 0.630 0.648 0.638 0.644 174.98 0.547 0.568 0.558 0.563 176.46 0.472 0.495 0.486 0.490 177.58 0.420 0.445 0.436 0.439 179.03 0.360 0.386 0.379 0.380 180.66 0.301 0.328 0.322 0.322 181.63 0.267 0.296 0.291 0.289 182.60 0.234 0.264 0.261 0.258 184.06 0.187 0.219 0.220 0.216 184.87 0.173 0.206 0.208 0.205 Note: Base Metal Inner Radius = 168.04 cm. Base Metal 1/4T = 172.25 cm. Base Metal 1/2T = 176.46 cm. Base Metal 3/4T = 180.66 cm. Base Metal Outer Radius = 184.87 cm. 4-14
. ~_ .- .. . .. . .. -. - _ - . - -
i i 4.2 - Fuel Cycle Specific Adjoint Calculations l' . l Results of the fuel cycle specific adjoint transport calculations for the first 20 cycles of operation at Point Beach Unit 2 are
= summarized in Tables 4.2-1 through 4.2-18. The data listed in these tables establish the means for absolute comparison of analysis and measurement for the Cycles 15, 16, 17, and 18/20 4
cavity dosimetry irradiations as well as for the four sets of surveillance capsule dosimetry withdrawn to date. These results ! l also provide the fuel cycle specific relationship among the I
' I
, surveillance capsule and reactor cavity measurement locations and key positions at the inner radius of the pressure vessel wall. I The core power distributions used in the cycle specific fast neutron exposure calculations for Fuel Cycles 1 through 20 were taken from the fuel cycle design reports applicable to Point ) Beach Unit 2* *"8"33 The data extracted from the fuel cycle ; design reports represented cycle averaged relative fuel assembly l powers and burnups as well as cycle averaged relative axial l distributions. Therefore, the results of the adjoint evaluation i provided data in terms of fuel cycle averaged neutron flux which, when multiplied by the appropriate fuel cycle length, produced the incremental fast neutron exposure for the fuel cycle. The calculated fast neutron flux (E > 1.0 MeV) and cumulative fast neutron fluence at the center of surveillance capsules located at 13, 23, and 33 degrees are provided for each of the 20 operating fuel cycles in Tables 4.2-1 and 4.2-2, respectively. The data as tabulated are applicable to the axial core midplane. Similar data applicable to the pressure vessel inner radius are given in Tables 4.2-3 and 4.2-4 and data pertinent to the cavity dosimetry sensor locations are listed in Tables 4.2-5 and 4.2-6. Exposure parameter ratios necessary to convert the cycle specific data listed in Tables 4.2-1 through 4.2-6 to other key fast neutron exposure units are given in Section 4.1 of this report. Application of these ratios to the data from Tables 4.2-1 through 4.2-6 yielded corresponding exposure data in terms of flux / fluence (E > 0.1 MeV) (Tables 4.2.7 through 4.2.12) and iron atom displacements (Tables 4.2.13 through 4.2.18). 4-15
TABLE 4.2-1 CALCULATED FAST NEUTRON FLUX (E > 1.O MeV) AT THE CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES NEUTRON FLUX (n/cm2 -sec) CYCLE No. 13' 23' 33' 1 1.35E+11 7.71E+10 7.18E+10 2 1.37E+11 8.14E+10 7.84E+10 3 1.35E+11 7.99E+10 7.77E+10 4 1.28E+11 7.72E+10 7.42E+10 5 1.35E+11 8.09E+10 7.44E+10 6 1.06E+11 7.74E+10 7.68E+10 7 1.04E+11 6.73E+10 6.46E+10 8 1.05E+11 6.60E+10 6.16E+10
- 9. 1.09E+11 6.71E+10 6.24E+10 10 9.88E+10 6.60E+10 6.39E+10 11 9.45E+10 6.86E+10 6.31E+10 12 9.90E+10 6.81E+10 6.02E+10 13 9.16E+10 6.44E+10 5.71E+10 14 9.43E+10 6.69E+10 6.21E+10 15 9.11E+10 6.47E+10 5.69E+10 16 7.32E+10 5.23E+10 4.98E+10 17 7.30E+10 5.32E+10 5.12E+10 18 7.14E+10 5.41E+10 5.28E+10 19 7.30E+10 5.35E+10 5.13E+10 20 7.24E+10 5.41E+10 5.42E+10 4-16 l
)
TABLE 4.2-2 CALCULATED FAST NEUTRON FLUENCE- (E > 1.0 MeV) AT THE CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES IRRADIATION CUMULATIVE FLUENCE END OF TIME (n/cm) 2 CYCLE (EFPS) 13* 23 33 1 4.81E+07 6.50E+18 3.71E+18 3.45E+18 2 8.13E+07 1.11E+19 6.41E+18 6.05E+18 3 1.09E+08 1.48E+19 8.61E+18 8.19E+18 4 1.36E+08 1.83E+19 1.07E+19 1.02E+19 5 1.64E+08 2.20E+19 1.30E+19 1.23E+19 6 1.91E+08 2.49E+19 1.51E+19 1.44E+19 7 2.20E+08 2.79E+19 1.70E+19 1.62E+19 8 2.47E+08 3.07E+19 1.88E+19 1.79E+19 9 2.72E+08 3.35E+19 2.05E+19 1.94E+19 10 3.09E+08 3.72E+19 2.29E+19 2.18E+19 11 3.36E+08 3.97E+19 2.48E+19 2.35E+19 12 3.61E+08 4.22E+19 2.65E+19 2.51E+19 13 3.87E+08 4.45E+19 2.81E+19 2.65E+19 14 4.14E+08 4.71E+19 3.00E+19 2.82E+19 15 4.39E+08 4.94E+19 3.16E+19 2.96E+19 16 4.66E+08 5.14E+19 3.30E+19 3.10E+19 17 4.93E+08 5.33E+19 3.44E+19 3.23E+19 18 5.20E+08 5.52E+19 3.59E+19 3.38E+19 19 5.46E+08 5.71E+19 3.73E+19 3.51E+19 20 5.74E+08 5.92E+3 9 3.88E+19 3.66E+19 l l 1 4-17
TABLE 4.2-3 CALCULATED FAST NEUTRON FLUX (E > 1.0-MeV) AT THE PRESSURE-VESSEL CLAD / BASE METAL INTERFACE NEUTRON FLUX (n/cm2- sec) CYCLE No. O' 15 30' 45* 1 4.53E+10 2.75E+10 1.81E+10 1.54E+10 2 4.60E+10 2.80E+10 1.96E+10 1.70E+10 3 4.59E+10 2.76E+10 1,94E+10 1.69E+10 4 4.31E+10 2.63E+10 1.86E+10 1.59E+10 5 4.39E+10 2.76E+10 1.90E+10 1.50E+10 6 3.41E+10 2.28E+10 1.93E+10 1.57E+10 7 3.60E+10 2.20E+10 1.63E+10 1.52E+10 8 3.74E+10 2.21E+10 1.57E+10 1.45E+10 9 3.84E+10 2.29E+10 1.59E+10 1.47E+10 10 3.40E+10 2.10E+10 1.61E+10 1.47E+10 11 2.83E+10 2.06E+10 1.63E+10 1.37E+10 12 3.07E+10 2.14E+10 1.58E+10 1.27E+10 13 2.84E+10 1.99E+10 1.50E+10 1.22E+10 14 2.86E+10 2.05E+10 1.60E+10 1.42E+10 15 2.77E+10 1.98E+10 1.50E+10 1.19E+10 16 2.26E+10 1.60E+10 1.27E+10 1.13E+10 17 2.27E+10 1.60E+10 1.30E+10 1.19E+10 18 2.17E+10 1.58E+10 1.34E+10 1.22E+10 19 2.28E+10 1.61E+10 1.31E+10 1.17E+10 20 2.24E+10 1.60E+10 1.36E+10 1.27E+10 i 4-18 1
i TABLE 4.2-4 CALCULATED FAST NEUTRON FLUENCE (E > 1.0 MeV) AT THE ; PRESSURE VESSEL CLAD / BASE METAL INTERFACE l IRRADIATION CUMULATIVE FLUENCE 2 END OF TIME (n/cm) ; CYCLE (EFPS) O' 15* 30' 45" j 1 4.81E+07 2.18E+18 1.32E+18 8.73E+17 7.40E+17
]
2 8.13E+07 3.71E+18 2.25E+18 1.52E+18 1.30E+18 I 3 1.09E+08 4.97E+18 3.01E+18 2.06E+18 1.77E+18 4 1.36E+08 6.1SE+18 3.73E+18 2.57E+18 2.20E+18 5 1.64E+08 7.38E+18 4.50E+18 3.10E+18 2.62E+18 6 1.91E+08 8.31E+18 5.13E+18 3.62E+18 3.05E+18 7 2.20E+08 9.32E+18 5.75E+18 4.08E+18 3.48E+18 8 2.47E+08 1.03E+19 6.34E+18 4.51E+18 3.87E+18 9 2.72E+08 1.13E+19 6.92E+18 4.90E+18 4.24E+18 10 3.09E+08 1.26E+19 7.71E+18 5.51E+18 4.79E+18 ) 11 3.36E+08 1.33E+19 8.26E+18 5.95E+18 5.16E+18 j 12 3.61E+08 1.41E+19 8.80E+18 6.35E+18 5.48E+18 13 3.87E+08 1.48E+19 9.31E+18 6.73E+18 5.79E+18 j 14 4.14E+08 1.56E+19 9.86E+18 7.16E+18 6.17E+18 ; 15 4.39E+08 1.63E+19 1.04E+19 7.54E+18 6.48E+18 16 4.66E+08 1.69E+19 1.08E+19 7.88E+18 6.78E+18 17 4.93E+08 1.75E+19 1.12E+19 8.23E+18 7.10E+18 18 5.20E+08 1.81E+19 1.16E+19 8.59E+18 7.42E+18 5.46E+08 1.87E+19 8.93E+18 19 1.21E+19 7.73E+18 20 5.74E+08 1.93E+19 1.25E+19 9.31E+18 8.09E+18 4-19
TABLE'4.2-5 CALCULATED FAST NEUTRON FLUX (E > 1.O MeV) AT THE
' CAVITY SENSOR SET LOCATIONS NEUTRON FLUX (n/cm2 -sec)
CYCLE No. O' 15' _39'_ 45' 1 3.46E+09 2.67E+09 1.78E+09 1.48E+09 2 3.52E+09 2.73E+09 1.90E+09 1.62E+09 3 3.50E+09 2.70E+09 1.88E+09 1.60E+09 4 3.29E+09 2.57E+09 1.80E+09 1.52E+09 5 3.38E+09 2.65E+09 1.82E+09 1.48E+09 6 2.67E+09 2.23E+09 1.78E+09 1.51E+09 7 2.75E+09 2.16E+09 1.58E+09 1.40E+09 8 2.82E+09 2.18E+09 1.54E+09 1.34E+09 9 2.91E+09 2.25E+09 1.57E+09 1.36E+09 10 2.60E+09 2.07E+09 1.55E+09 1.37E+09 11 2.27E+09 1.96E+09 1.53E+09 1.30E+09 12 2.433+09 2.04E+09 1.50E+09 1.23E+09 13 2.25E+09 1.90E+09 1.42E+09 1.18E+09 14 2.28E+09 1.95E+09 1.52E+09 1.32E+09 15 2.21E+09 1.88E+09 1.42E+09 1.16E+09 16 1.80E+09 1.54E+09 1.21E+09 1.06E+09 17 1.81E+09 1.55E+09 1.24E+09 1.10E+09 18 1.74E+09 1.52E+09 1.26E+09 1.13E+09 19 1.81E+09 1.55E+09 1.24E+09 1.09E+09 20 1.79E+09 1.55E+09 1.29E+09 1.17E+09 i l 4-20
TABLE 4.2-6 s CALCULATED FAST NEUTRON FLUENCE (E > 1.0 MeV) AT THE CAVITY SENSOR SET LOCATIONS IRRADIATION CUMULATIVE FLUENCE 2 END OF TIME (n/cm) CYCLE (EFPS) 0* 15' 30 45 1 4.81E+07 1.67E+17 1.28E+17 8.55E+16 7.12E+16 1 2 8.13E+07 2.83E+17 2.19E+17 1.49E+17 1.25E+17 3 1.09E+08 3.80E+17 2.93E+17 2.00E+17 1.69E+17 4 1.36E+08 4.70E+17 3.64E+17 2.49E+17 2.11E+17 5 1.64E+08 5.64E+17 4.38E+17 3.00E+17 2.52E+17 4 6 1.91E+08 6.37E+17 4.99E+17 3.49E+17 2.93E+17
- 7 2.20E+08 7.15E+17 5.60E+17 3.94E+17 3.33E+17 8 2.47E+08 7.91E+17 6.19E+17 4.35E+17 3.69E+17 9 2.72E+08 8.63E+17 6.75E+17 4.74E+17 4.03E+17 10 3.09E+08 9.61E+17 7.53E+17 5.33E+17 4.54E+17 11 3.36E+08 1.02E+18 8.05E+17 5.74E+17 4.89E+17 12 3.61E+08 1.08E+18 8.57E+17 6.12E+17 5.20E+17 i 13 3.87E+08 1.14E+18 9.05E+17 6.48E+17 5.51E+17 1
14 4.14E+08 1.20E+18 9.58E+17 6.89E+17 5.87E+17 . 15 4.39E+08 1.26E+18 1.01E+18 7.25E+17 6.16E+17 1 16 4.66E+08 1.31E+18 1.05E+18 7.58E+17 6.45E+17 17 4.93E+08 1.36E+18 1.09E+18 7.91E+17 6.74E+17 18 5.20E+08 1.40E+18 1.13E+18 8.25E+17 7.04E+17 5.46E+08 1.17E+18 19 1.45E+18 8.57E+17 7.33E+17 20 5.74E+08 1.50E+18 1.21E+18 8.93E+17 7.65E+17 4 4 i 4-21 4
TABLE 4.2-7 l CALCULATED FAST NEUTRON FLUX (E > 0.1 MeV) AT THE CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES NEUTRON FLUX (n/cm2 -sec) CYCLE No. 13' 23' 33' 1 5.12E+11 2.65E+11 2.53E+11 2 5.20E+11 2.80E+11 2.77E+11 3 5.13E+11 2.75E+11 2.74E+11 4 4.85E+11 2.66E+11 2.62E+11 5 5.10E+11 2.79E+11 2.62E+11 6 4.01E+11 2.67E+11 2.71E+11 7 3.95E+11 2.32E+11 2.28E+11 8 3.98E+11 2.27E+11 2.17E+11 9 4.15E+11 2.31E+11 2.20E+11 10 3.74E+11 2.27E+11 2.26E+11 11 3.58E+11 2.36E+11 2.23E+11 12 3.75E+11 2.35E+11 2.12E+11 13 3.47E+11 2.22E+11 2.01E+11 14 3.57E+11 2.30E+11 2.19E+11 15 3.45E+11 2.23E+11 2.01E+11 16 2.77E+11 1.80E+11 1.76E+11 17 2.76E+11 1.83E+11 1.81E+11 18 2.70E+11 1.86E+11 1.86E+11 19 2.77E+11 1.84E+11 1.81E+11 . 20 2.74E+11 1.86E+11 1.91E+11 4-22
TABLE 4.2-8 CALCULATED FAST NEUTRON FLUENCE (E > 0.1 MeV) AT THE CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES IRRADIATION CUMULATIVE FLUENCE END OF TIME (n/cm) 2 CYCLE (EFPS) 13" 23' 33 1 4.81E+07 2.46E+19 1.28E+19 1.22E+19 2 8.13E+07 4.19E+19 2.21E+19 2.14E+19 3 1.09E+08 5.60E+19 2.96E+19 2.89E+19 4 1.36E+08 6.93E+19 3.69E+19 3.61E+19 5 1.64E+08 8.35E+19 4.47E+19 4.34E+19 6 1.91E+08 9.45E+19 5.19E+19 5.08E+19 7 2.20E+08 1.06E+20 5.85E+19 5.72E+19 8 2.47E+08 1.16E+20 6.46E+19 6.31E+19 9 2.72E+08 1.27E+20 7.04E+19 6.86E+19 10 3.09E+08 1.41E+20 7.89E+19 7.71E+19 11 3.36E+08 1.50E+20 8.53E+19 8.30E+19 12 3.61E+08 1.60E+20 9.12E+19 8.84E+19 13 3.87E+08 1.69E+20 9.68E+19 9.35E+19 14 4.14E+08 1.78E+20 1.03E+20 9.95E+19 15 4.39E+08 1.87E+20 1.09E+20 1.05E+20 16 4.66E+08 1.95E+20 1.14E+20 1.09E+20 17 4.93E+08 2.02E+20 1.18E+20 1.14E+20 18 5.20E+08 2.09E+20 1.23E+20 1.19E+20 19 5.46E+08 2.16E+20 1.28E+20 1.24E+20 20 5.74E+08 2.24E+20 1.33E+20 1.29E+20 I 4-23
r TABLE 4.2 CALCULATED FAST NEUTRON FLUX (E > 0.1 MeV) AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE NEUTRON FLUX (n/cm2- sec) CYCLE No. 0* 15* 30' 45* 1 1.24E+11 8.00E+10 4.94E+10 4.05E+10 2 1.26E+11 8.17E+10 5.35E+10 4.47E+10 3 1.26E+11 8.05E+10 5.28E+10 4.44E+10 4 1.18E+11 7.66E+10 5.07E+10 4.20E+10 5 1.20E+11 8.05E+10 5.17E+10 3.95E+10 6 9.34E+10 6.65E+10 '5.26E+10 4.12E+10 7 9.85E+10 6.42E+10 4.44E+10 4.00E+10 8 1.02E+10 6.45E+10 4.28E+10 3.81E+10 9 1.05E+10 6.68E+10 4.33E+10 3.86E+10 10 9.31E+10 6.13E+10 4.39E+10 3.88E+10 11 7.74E+10 6.00E+10 4.44E+10 3.60E+10 12 8.42E+10 6.22E+10 4.30E+10 3.34E+10 13 7.76E+10 5.80E+10 4.08E+10 3.22E+10 14 7.82E+10 5.97E+10 4.35E+10 3.74E+10 15 7.59E+10 5.78E+10 4.08E+10 3.13E+10 16 6.20E+10 4.66E+10 3.47E+10 2.98E+10 17 6.22E+10 4.67E+10 3.55E+10 3.13E+10 18 5.94E+10 4.61E+10 3.65E+10 3.22E+10 19 6.23E+10 4.68E+10 3.57E+10 3.09E+10 20 6.13E+10 4.65E+10 3.71E+10 3.35E+10 4-24
l TABLE 4.2-10 l CALCULATED FAST NEUTRON FLUENCE (E > 0.1 MeV) AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE IRRADIATION CUMULATIVE FLUENCE END OF TIME (n/cm) 2 j CYCLE (EFPS) 0* 15 30* 45 1 4.81E+07 5.97E+18 3.85E+18 2.38E+18 1.95E+18 2 8.13E+07 1.02E+19 6.56E+18 4.15E+18 3.43E+18 3 1.09E+08 1.36E+19 8.78E+18 5.61E+18 4.65E+18 4 1.36E+08 1.68E+19 1.09E+19 7.00E+18 5.80E+18
-5 1.64E+08 2.02E+19 1.31E+19 8.44E+18 6.90E+18 6 1.91E+08 2.27E+19 1.49E+19 9.88E+18 8.03E+18 7 2.20E+08 2.55E+19 1.67E+19 1.11E+19 9.16E+18 8 2.47E+08 2.83E+19 1.85E+19 1.23E+19 1.02E+19 9 2.72E+08 3.09E+19 2.02E+19 1.34E+19 1.12E+19 10 3.09E+08 3.44E+19 2.25E+19 1.50E+19 1.26E+19 11 3.36E+08 3.65E+19 2.41E+19 1.62E+19 1.36E+19 12 3.61E+08 3.86E+19 2.56E+19 1.73E+19 1.44E+19 i 13 3.87E+08 4.06E+19 2.71E+19 1.83E+19 1.52E+19 l 14 4.14E+08 4.27E+19 2.87E+19 1.95E+19 1.63E+19 15 4.39E+08 4.47E+19 3.02E+19 2.06E+19 1.71E+19 16 4.66E+08 4.63E+19 3.15E+19 2.15E+19 1.79E+19 17 4.93E+08 4.80E+19 3.27E+19 2.24E+19 1.87E+19 18 5.20E+08 4.96E+19 3.39E+19 2.34E+19 1.95E+19 19 5.46E+08 5.12E+19 3.52E+19 2.43E+19 2.04E+19 20 5.74E+08 5.29E+19 3.65E+19 2.54E+19 2.13E+19 1
l S 4-25 4 e l
I ! l ! l- l l
. TABLE.4.2-11 i I
l l CALCULATED FAST NEUTRON FLUX (E'>'O.1 MeV) AT THE l CAVITY SENSOR SET LOCATIONS NEUTRON FLUX (n/cm?-sec) . CYCLE No. O' 15' 30* 45' 1 2.72E+10 2.30E+10 1.55E+10 1.22E+10 2 2.76E+10 2.35E+10 l'65E+10
. 1.33E+10 l 3 2.75E+10 2.33E+10 1.63E+10 1.32E+10 4 2.59E+10 2.21E+10 1.56E+10 1.25E+10 l 5 2.66E+10 2.29E+10 1.59E+10 1.21E+10 l 6 2.10E+10 1.92E+10 1.55E+10 1.24E+10 >
7 2.16E+10 1.86E+10 1.38E+10 1.15E+10 8 2.22E+10 1.88E+10 1.34E+10 1.10E+10 9 2.29E+10 1.94E+10 1.36E+10 1.12E+10 10 2.04E+10 1.78E+10 1.35E+10 1.12E+10 11 1.78E+10 1.69E+10 1.33E+10 1.07E+10 12 1.91E+10 1.75E+10 1.30E+10 1.01E+10 ; 13 1.76E+10 1.63E+10 1.24E+10 9.69E+09 14 1.79E+10 1.68E+10 1.32E+10 1.09E+10 15 1.74E+10 1.62E+10 ~1.23E+10 9.53E+09 16 1.42E+10 1.32E+10 1.05E+10 8.69E+09 ) 17 1.42E+10 1.33E+10 1.08E+10 9.04E+09
]
18 1.37E+10 1.31E+10 1.10E+10 9.27E+09 19 1.42E+10 1.33E+10 1.08E+10 8.97E+09 20 1.41E+10 1.33E+10 1.12E+10 9.57E+09 4-26 I I
~ TABLE 4.2-12 CALCULATED FAST NEUTRON FLUENCE (E > 0.1 MeV) AT THE CAVITY SENSOR SET LOCATIONS I IRRADIATION CUMULATIVE FLUENCE
, END OF TIME (n/cm) 2 CYCLE (EFPS) 0' 15' 30' 45 1 4.81E+07 1.31E+18 1.10E+18 7.43E+17 5.84E+17 2 8.13E+07 2.23E+18 1.89E+18 1.29E+18 1.02E+18 l
3 1.09E+08 2.98E+18 2.53E+18 1.74E+18 1.39E+18 4 1.36E+08 3.69E+18 3.13E+18 2.17E+18 1.73E+18 5 1.64E+08 4.43E+18 3.77E+18 2.61E+18 2.07E+18 6 1.91E+08 5.01E+18 4.30E+18 3.03E+18 2.41E+18 7 2.20E+08 5.61E+18 4.82E+18 3.42E+18 2.73E+18 ! 8 2.47E+08 6.21E+18 5.33E+18 3.78E+18 3.03E+18 9 2.72E+08 6.78E+18 5.81E+18 4.13E+18 3.31E+18 10 3.09E+08 7.55E+18 6.48E+18 4.63E+18 3.73E+18 11 3.36E+08 8.03E+18 6.94E+18 4.99E+18 4.02E+18 ~ 12 3.61E+08 8.51E+18 7.38E+18 5.32E+18 4.27E+18 13 3.87E+08 8.96E+18 7.80E+18 5.64E+18 4.52E+18 14 4.14E+08 9.45E+18 8.25E+18 6.00E+18 4.82E+18 3 15 4.39E+08 9.89E+18 8.67E+18 6.31E+18 5.06E+18 i 16 4.66E+08 1.03E+19 9.02E+18 6.59E+18 5.29E+18 17 4.93E+08 1.06E+19 9.38E+18 6.88E+18 5.53E+18 18 5.20E+08 1.10E+19 9.73E+18 7.17E+18 5.78E+18 19 5.46E+08 1.14E+19 1.01E+19 7.45E+18 6.01E+18 20 5.74E+08 1.18E+19 1.04E+19 7.77E+18 6.28E+18 4 e 4-27
TABLE 4.2-13 CALCULATED IRON ATOM DISPLACEMENT RATE AT THE CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES DISPLACEMENT RATE (dpa/sec) CYCLE No. 13' 23" _23'_ 1 2.41E-10 1.31E-10 1.23E-10 2 2.44E-10 1.39E-10 1.35E-10 3 2.41E-10 1.36E-10 1.34E-10 4 2.28E-10 1.32E-10 1.28E-10 5 2.40E-10 1.38E-10 1.28E-10 6 1.88E-10 1.32E-10 1.32E-10 7 1.86E-10 1.15E-10 1.11E-10 8 1.87E-10 1.12E-10 1.06E-10 9 1.99E-10 1.14E-10 1.07E-10 10 1.76E-10 1.12E-10 1.10E-10 11 1.68E-10 1.17E-10 1.09E-10 12 1.76E-10 1.16E-10 1.03E-10 13 1.63E-10 1.10E-10 9.81E-11 14 1.68E-10 1.14E-10 1.07E-10 15 1.62E-10 1.10E-10 9.79E-11 16 1.30E-10 8.92E-11 8.57E-11 17 1.30E-10 9.06E-11 8.81E-11 18 1.27E-10 9.22E-11 9.08E-11 19 1.30E-10 9.11E-11 8.82E-11 20 1.29E-10 9.21E-11 9.32E-11 4-28
t.. TABLE 4.2-14 CALCULATED IRON ATOM DISPLACEMENTS AT THE CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES IRRADIATION CUMULATIVE DISPLACEMENTS END OF TIME (dpa) CYCLE (EFPS) 13* 23* 33* 1 4.81E+07 1.16E-02 6.32E-03 5.94E-03 2 8.13E+07 1.97E-02 1.09E-02 1.04E-02 3 1.09E+08 2.63E-02 1.47E-02 1.41E-02 4 1.36E+0B 3.26E-02 1.83E-02 1.76E-02 5 1.64E+08 3.92E-02 2.21E-02 2.11E-02 6 1.91E+08 4.44E-02 2.57E-02 2.48E-02 7 2.20E+08 4.96E-02 2.90E-02 2.79E-02 8 2.47E+08 5.47E-02 3.20E-02 3.08E-02 9 2.72E+08 5.95E-02 3.48E-02 3.340-02 10 3.09E+08 6.62E-02 3.91E-02 3.76E-02 11 3.36E+08 7.07E-02 4.22E-02 4.05E-02 12 3.61E+08 7.51E-02 4.51E-02 4.31E-02 13 3.87E+08 7.93E-02 4.79E-02 4.56E-02 14 4.14E+08 8.38E-02 5.10E-02 4.85E-02 15 4.39E+08 8.80E-02 5.38E-02 5.10E-02 16 4.66E+08 9.15E-02 5.62E-02 5.33E-02 17 4.93E+08 9.49E-02 5.87E-02 5.56E-02 18 5.20E+08 9.83E-02 6.11E-02 5.81E-02 19 5.46E+08 1.02E-01 6.35E-02 6.04E-02 20 5.74E+08 1.05E-01 6.61E-02 6.30E-02 k t 4-29
f TABLE 4.2-15 CALCULATED IRON ATOM DISPLACEMENT RATE AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE DISPLACEMENT RATE (dpa/sec) CYCLE No. O' 15' 30' 45* 1 7.40E-11 4.62E-11 2.98E-11 2.50E-11 2 7.51E-11 4.71E-11 3.22E-11 2.75E-11
- 3. 7.50E-11 4.64E-11 3.18E-11 2.74E-11 4 7.03E-11 4.42E-11 3.06E-11 2.59E-11 5 7.16E 4.64E-11 3.12E-11 2.44E-11 6 5.57E-11 3.83E-11 3.17E-11 2.54E-11 7 5.87E-11 3.70E-11 2.68E-11 2.46E-11 8 6.10E-11 3.72E-11 2.58E-11 2.35E-11 9 6.26E-11 3.85E-11 2.61E-11 2.38E-11 10 5.55E-11 3.535-11 2.65E-11 2.39E-11 11 4.61E-11 3.46E-11 2.68E-11 2.22E-11 12 5.02E-11 3.59E-11 2.59E-11 2.06E-11 13 4.63E-11 3.34E-11 2.45E-11 1.98E-11 14 4.66E-11 3.44E-11 2.62E-11 2.30E-11 15 4.53E-11 3.33E-11 2.46E-11 1.93E-11 16 3.70E-11 2.69E-11 2.09E-11 1.84E-11 17 3.71E-11 2.69E-11 2.14E-11 1.93E-11 18 3.54E-11 2.66E-11 2.20E-11 1.98E-11 19 3.71E-11 2.70E-11 2.15E-11 1.90E-11 20 3.65E-11 2.68E-11 2.23E-11 2.06E-11 4-30
TABLE 4.2-16 CALCULATED IRON ATOM DISPLACEMENTS AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE IRRADIATION CUMULATIVE DISPLACEMENTS END OF TIME (dpa) CYCLE (EFPS) 0 15' 30' 45' 1 4.81E+07 3.56E-03 2.22E-03 1.43E-03 1.20E-03 2 8.13E+07 6.05E-03 3.79E-03 2.50E-03 2.12E-03 3 1.09E+08 8.11E-03 5.06E-03 3.38E-03 2.87E-03 4 1.36E+08 1.00E-02 6.27E-03 4.21E-03 3.58E-03 5 1.64E+08 1.20E-02 7.57E-03 5.08E-03 4.26E-03 6 1.91E+08 1.36E-02 8.62E-03 5.95E-03 4.95E-03 7 2.20E+08 1.52E-02 9.66E-03 6.70E-03 5.65E-03 8 2.47E+08 1.69E-02 1.07E-02 7.40E-03 6.28E-03 9 2.72E+08 1.84E-02 1.16E-02 8.05E-03 6.88E-03 10 3.09E+08 2.05E-02 1.30E-02 9.05E-03 7.78E-03 11 3.36E+08 2.18E-02 1.39E-02 9.77E-03 8.38E-03 12 3.61E+08 2.30E-02 1.48E-02 1.04E-02 8.89E-03 13 3.87E+08 2.42E-02 1.56E-02 1.11E-02 9.40E-03 14 4.14E+08 2.55E-02 1.66E-02 1.18E-02 1.00E-02 15 4.39E+08 2.66E-02 1.74E-02 1.24E-02 1.05E-02 ) 16 4.66E+08 2.76E-02 1.82E-02 1.29E-02 1.10E-02 l 17 4.93E+08 2.86E-02 1.89E-02 1.35E-02 1.15E-02 18 5.20E+08 2.95E-02 1.96E-02 1.41E-02 1.21E-02 19 5.46E+08 3.05E-02 2.03E-02 1.47E-02 1.26E-02 20 5.74E+08 3.15E-02 2.10E-02 1.53E-02 1.31E-02 l 4-31 i
TABLE 4.2-17 I CALCULATED IRON ATOM DISPLACEMENT RATE AT THE I CAVITY SENSOR SET LOCATIONS l DISPLACEMENT RATE l (dpa/sec) , CYCLE No. 0* 15' 30* .. 4 5'_ . 1 9.76E-12 8.08E-12 5.44E-12 4.34E-12 2 9.92E-12 8.29E-12 5.82E-12 4.74E-12 3 9.87E-12 8.19E-12 5.75E-12 4.70E-12 4 9.29E-12 7.78E-12 5.50E-12 4.47E-12 5 9.53E-12 8.07E-12 5.58E-12 4.34E-12 - 6 7.54E-12 6.76E-12 5.45E-12 4.43E-12 4'.11E-12 7 .7.74E-12 6.56E-12 4.85E-12 , 8 7.95E-12 6.61E-12 4.72E-12 3.94E-12 9 8.20E-12 6.81E-12 4.80E-12 4.00E-12 10 7.33E-12 6.27E-12 4.75E-12 4.01E-12 11 6.39E-12 5.94E-12 4.70E-12 3.83E-12 12 6.85E-12 6.17E-12 4.59E-12 3.62E-12 13 6.33E-12 5.75E-12 4.35E-12 3.46E-12 14 6.43E-12 5.92E-12 4.66E-12 3.89E-12 15 6.23E-12 5.71E-12 4.33E-12 3.41E-12 16 5.08E-12 4.66E-12 3.71E-12 3.11E-12 ] 4 17 5.10E-12 4.68E-12 3.80E-12 3.23E-12 l 4 18 4.92E-12 4.62E-12 3.87E-12 3.31E-12 19 5.11E-12 4.70E-12 3.80E-12 3.21E-12 20 5.05E-12 4.69E-12 3.94E-12 3.42E-12 4 A 4-32 I 1
l i I TABLE 4.2-18 CALCULATED IRON ATOM DISPLACEMENTS AT THE CAVITY SENSOR SET LOCATIONS : IRRADIATION CUMULATIVE FLUENCE END OF TIME (n/cm) 2 CYCLE (EFPS) 0' 15' 30' 45' I 1 4.81E+07 4.70E-04 3.89E-04 2.62E-04 2.09E-04 2 8.13E+07 7.99E-04 6.64E-04 4.55E-04 3.66E-04 3 1.09E+08 1.07E-03 .8.89E-04 6.13E-04 4.96E-04 4 1.36E+08 1.33E-03 1.10E-03 7.64E-04 6.18E-04 5 1.64E+08 1.59E-03 1.33E-03 9.19E-04 7.39E-04 6 1.91E+08 1.80E-03 1.51E-03 1.07E-03 8.60E-04 1 7 2.20E+08 2.02E-03 1.70E-03 1.21E-03 9.76E-04 l 8 2.47E+08 2.23E-03 1.87E-03 1.33E-03 1.08E-03 l 9 2.72E+08 2.43E-03 2.05E-03 1.45E-03 1.18E-03 10 3.09E+08 2.71E-03 2.28E-03 1.63E-03 1.33E-03 l 11 3.36E+08 2.88E-03 2.44E-03 1.76E-03 1.44E-03 12 3.61E+08 3.06E-03 2.60E-03 1.87E-03 1.53E-03 13 3.87E+08 3.22E-03 2.74E-03 1.98E-03 1.62E-03 14 4.14E+08 3.39E-03 2.90E-03 2.11E-03 1.72E-03 15 4.39E+08 3.55E-03 3.05E-03 2.22E-03 1.81E-03 16 4.66E+08 3.69E-03 3.17E-03 2.32E-03 1.89E-03 17 4.93E+08 3.82E-03 3.30E-03 2.42E-03 1.98E-03 18 5.20E+08 3.95E-03 3.42E-03 2.52E-03 2.07E-03 19 5.46E+08 4.09E-03 3.55E-03 2.62E-03 2.15E-03 20 5.74E+08 4.23E-03 3.68E-03 2.73E-03 2.25E-03 1 4-33
SECTION 5.0 EVALUATIONS OF SURVSILLANCE CAPSULE DOSIMETRY In'this'section, the results of the evaluations of the four neutron sensor sets withdrawn as a part of the Point Beach 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 (EFPS)
.V 13' END OF CYCLE 1 4.81E+07 T 23* END OF CYCLE 3 1.09E+08 R 13' END OF CYCLE 5 1.64E+08 S 33' END OF CYCLE 16 4.66E+08 5.1 - Measured Reaction Rates With the exception of Capsule V, radiometric counting of each of these capsule dosimetry data sets w&s accomplished by Westinghouse using the procedures discussed in Section 3.0 of this report H 'd"1 The measured specific activities are included in Appendix A to this report. Radiometric counting of the sensors from Capsule V, on the other hand, was carried out by the l Battelle Memorial Institute H'I. However, in this case, the measured specific activities were not reported.
The irradiation history of the Point Beach Unit 2 reactor during the first 16 fuel cycles is also listed in Appendix A. The irradiation history was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report" for the applicable operating periods. In addition to the reactor power history, for the multiple cycle irradiations Capsules T,R, and S, the flux level adjustment factors for each cycle are also tabulated in Appendix A. These adjustment factors were determined from the fuel cycle specific adjoint calculations described in Section 4.2 of this report. 5-1
, Based on the irradiation history, the individual sensor characteristics, capsule gradient corrections, and the measured specific activities, reaction rates averaged over the appropriate j irradiation periods and referenced to a core power level of 1518 j MWt were computed for the sensor sets removed from Capsules T, R, I and S. In the cast of Capsule V, reaction rates were developed directly from the derived neutron flux and spectrum averaged . reaction cross-sections reported in reference 46. The computed reaction rates for the multiple foil sensor sets from each of the four internal surveillance capsules are provided in Table 5.1-1. In regard to the data listed in Table 5.1-1, the fission rate measurements for the U-238 sensors include corrections for U-235 impurities, the build-in of Plutonium isotopes during the long irradiations, and for the effects of 7,f reactions. Likewise, the fission rate measurements for the Np-237 sensors include adjustments for 7,f reactions occuring over the course of the respective irradiation periods. i 5.2 - Results of the Least Squares Adjustment Procedure The results of the application of the least squares adjustment procedure to the four sets of surveillance capsule dosimetry are provided in Table 5.2-1 through 5.2-4. In these tables, the derived exposure experienced by the capsule along with data illustrating the fit of both the trial and adjusted spectra to the measurements are given. Also included in the tabulations are the la uncertainties associated with each of the derived exposure rates. In regard to the comparisons listed in Table 5.2-1 and 5.2-2, it should be nited that the columns labeled " trial calc" were obtained by normalizing the neutron spectral data from Table 4.1-3 to the absolute calculated neutron flux (E > 1.0 MeV) averaged over the applicable irradiation periods (Cycle 1 for Capsule V, Cycles 1 through 3 for Capsule T, Cycles 1 through 5 for Capsule R, and Cycles 1 through 16 for Capsule S) as discussed in Section 3.0. Thus, the comparisons illustrated in Tables 5.2-1 through 5.2-4 indicate the degree to which the calculated neutron energy specta matched the measured sensor data 5-2 1
)
i-before and.after adjustment. Absolute comparisons are discussed further in Sectionl7.0 of this report. l l 5-3 h -.-,.,_q_ .._-. .-.,., -
-- =__.-.- -
TABLE 5.1-1
SUMMARY
OF REACTION RATES DERIVED FROM MULTIPLE FOIL SENSOR SETS WITHDRAWN FROM INTERNAL SURVEILLANCE CAPSULES REACTION RATE (ros/ nucleus) CAPSULE CAPSULE CAPSULE CAPSULE REACTION V T R S Cu63 (n, of) Co60 6.28E-17 4.86E-17 6.77E-17 4.26E-17 Fe54(n,p)Mn54 7.79E-15 5.53E-15 7.86E-15 NiS 8 (n, p) Co58 7.34E-15 1.12E-14 5.51E-15 U238(n,f)Cs137 Cd 4.84E-14 2.69E-14 4.54E-14 2.46E-14 Np237 (n, f) Cs137 Cd 4.24E-13 2.37E-13 4.19E-14 1.95E-13 CoS9 (n, y) Co60 7.60E-12 5.13E-12 9.34E-12 3.72E-12 CoS9 (n, y) Co60 Cd 3.10E-12 1.97E-12 3.79E-12 1.60E-12 P O l I i I 5-4 l l
)
_- .-. - - - . . _ . .- . ... . _ ~ . . =. -- .. . - l
- TABLE 5.2.1 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE V DOSIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 1 TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 1.35E+11 1.44E+11 8%
4(E > 0.1 MeV) 5.12E+11 5.71E+11 15% 4(E < 0.414 eV) 2.46E+11 1.92E+11 20% dpa/sec 2.41E-10 2.60E-10 11%' COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE V REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu - 63 (n, at) 6.28E-17 6.90E-17 6.26E-17 0.91 1.00 Fe-54(n,p) 7.79E-15 8.44E-15 7.96E-15 0.92 0.98 U-230(n,f) Cd 4.84E-14 4.37E-14 4.54E-14 1.11 1.07 Np-237(n,f) Cd 4.24E-13 3.77E-13 4.18E-13 1.13 1.01 Co- 59 (n, y) 7.60E-12 9.62E-12 7.61E-12 0.79 1.00 Co- 59 (n, y) Cd 3.10E-12 3.74E-12 3.10E-12 0.83 1.00 i 5-5
1 l I l I TABLE 5.2.2 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE T DOSIMETRY WITHDRAWN AT.THE END OF FUEL CYCLE 3 TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY p(E > 1.0 MeV) 7.91E+10 8.21E+10 8% p(E > 0.1 MeV) 2.72E+11 2.97E+11 15% p(E < 0.414 eV) 1.28E+11 1.29E+11 18% dpa/sec 1.35E-10 1.43E-10 10% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE T REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, cr) 4.86E-17 5.13E-17 4.86E-17 0.95 1.00 Fe- 54 (n, p) 5.53E-15 5.62E-15 5.49E-15 0.98 1.01 Ni-58(n,p) 7.34E-15 7.73E-15 7.43E-15 0.95 0.99 U-238(n,f) Cd 2.69E-14 2.67E-14 2.70E-14 1.01 1.00 Np-237(n,f) Cd 2.37E-13 2.10E-13 2.30E-13 1.13 1.03 Co- 59 (n, y) 5.13E-12 4.97E-12 5.13E-12 1.03 1.00 Cc-59 (n,y) Cd 1.97E-12 1.84E-12 1.97E-12 1.07 1.00 5-6 _. .. . . _ - . ~ . _ - _ . . - _ .
l l 1 TABLE 5.2.3 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE R DOSIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 5 ' TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$ (E > 1. 0 MeV) 1.34E+11 1.39E+11 8%
$(E > 0.1 MeV) 5.09E+11 5.61E+11 15%
$(E < 0.414 eV) 2.45E+11 2.34E+11 19%
dpa/sec 2.39E-10 2.55E-10 11% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE R REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, a) 6.77E-17 6.86E-17 6.71E-17 0.99 1.01 Fe-54(n,p) 7.86E-15 8.39E-15 8.00E-15 0.94 0.98
)
Ni- 58 (n, p) 1.12E-14 1.17E-14 1.13E-14 0.96 0.99 : U-238(n,f) Cd 4.54E-14 4.35E-14 4.38E-14 1.04 1.04 I Np-237(n,f) Cd 4.19E-13 3.75E-13 4.08E-13 1.12 1.03 Co-59 (n, y) 9.34E-12 9.56E-12 9.34E-12 0.98 1.00 Co- 59 (n, y) Cd 3.79E-12 3.72E-12 3.79E-12 1.02 1.00
)
i 1 5-7 )' s p g. ,. a.- - , ,. -
TABLE 5.2.4 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE S DOSIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 16 TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 6.65E+10 6.81E+10 8%
$(E > 0.1 MeV) 2.35E+11 2.54E+11 15%
$(E < 0.414 eV) 1.08E+11 8.91E+10 20%
dpa/sec 1.14E-10 1.20E-10 10% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE S REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, or) 4.26E-17 3.98E-17 4.18E-17 1.07 1.02 Ni- 5 8 (n, p) 5.51E-15 6.24E-17 5.77E-15 0.88 0.96 U-238(n,f) Cd 2.46E-14 2.21E-14 2.25E-14 1.11 1.09 Np-237(n,f) Cd 1.95E-13 1.79E-13 1.91E-13 1.09 1.02 Co-59(n,y) 3.72E-12 4.23E-12 3.73E-12 0.88 1.00 Co-59(n,y) Cd 1.60E-12 1.62E-12 1.60E-12 0.99 1.00 4 5-8
l l l l SECTION 6.0 l l EVALUATIONS OF REACTOR CAVITY DOSIMETRY In this section, the results of the evaluations of all neutron sensor sets irradiated since the inception of the Reactor Cavity Measurement Program are presented. At Point Beach Unit 2 the program was initiated prior to the startup of Cycle 15; and, to date, has included measurement evaluations at the conclusion of Cycles 15, 16, 17 and 20. The evaluation of each of these sets of measured data was accomplished using a consistent approach based on the methodology discussed in Section 3.0, resulting in an accurate data base to be used in defining the best estimate neutron exposure of the reactor vessel wall. 6.1 - Cycle 15 Results 6.1.1 - Measured Reaction rates l l During the Cycle 15 irradiation, seven multiple foil sensor sets, and four stainless steel gradient chains were deployed in the reactor cavity as depicted in Figures 2.1-1 and 2.1-2. The capsule identifications associated with each of the multiple foil sensor sets were as follows: CAPSULE IDENTIFICATION AZIMUTH VESSEL CORE CORE CORE (degrees) SUPPORT TOP MIDPLANE BOTTOM 0.0 XX G H I 15.0 J 30.0 K
)
45.0 L The contents of each of these irradiation capsules is specified in Appendix B to this report. The irradiation history of the Point Beach Unit 2 reactor during Cycle 15 is also listed in Appendix B. The irradiation history was obtained from NUREG-0020, " Licensed Operating Reactors Status l 6-1 1 l l
I Summary Report" for the applicable operating period. Based on this reactor operating history, the individual sensor characteristics, and the measured specific activities given in Appendix B, cycle average reaction rates referenced to a core power level of 1518 MWt were computed for each multiple foil sensor and gradient chain segment. The computed reaction rates for the multiple foil sensor sets, including radiometric foils and solid state track recorders, irradiated during Cycle 15 are provided in Table 6.1-1. Corresponding reaction rate data from the the four stainless steel gradient chains are recorded in Tables 6.1-2 through 6.1-4 for the Fe-54(n,p), Ni-58(n,p), and Co-59(n,7) reactions, respectively. In regard to the data listed in Table 6.1-1, the Fe-54(n,p) reaction rates represent an average of the bare and cadmium covered measurements for each capsule. Likewise, the U-238(n,f) reaction rates were obtained by averaging the results of the radiometric foil and solid state track recorder data. In addition, the fission rate measurements include corrections for U-235 impurities in the U-238 sensors as well as corrections for photo-fission reactions in both the U-238 and Np-237 sensors. 6.1.2 - Results of the Least Squares Adjustment Procedure The results of the application of the least squares adjustment procedure to the seven sets of multiple foil measuremencs ; obtained from the Cycle 15 irradiation are provided in Tables 6.1-5 through 6.1-11. In these tables, the derived exposure experienced at each sensor set location along with data illustrating the fit of both the trial and adjusted spectra to the measurements are given. Also included in the tabulations are the la uncertainties associated with each of the derived exposure rates. i l In regard to the comparisons listed in Tables 6.1-5 through 6.1-11, it should be noted that the columns labeled " trial cale" were obtained by normalizing the neutron spectral data from Table 4.1-2 to the absolute calculated neutron flux (E > 1.0 MeV) 1
)
6-2 l l I
averaged over the Cycle 15 irradiation period as discussed in l Section 3.0. Thus, the comparisons illustrated in Tables 6.1-5 through 6.1-11 indicate the degree to which the calculated neutron energy spectra matched the measured data before and after adjustment. Absolute comparisons of calculation and measurement are discussed further in Section 7.0 of this report. Complete traverses of fast neutron exposure rates in the reactor cavity were developed by combining the results of the least squares adjustment of the multiple foil data with the Fe-54 (n,p) reaction rate measurements from the gradient chains. The gradient data were employed to establish relative axial distributions over the measurement range and these relative distributions were then normalized to the FERRET results from the midplane sensor sets to produce axial distributions of exposure rates in terms of $ (E > 1.0 MeV) , 4(E > 0.1 MeV), and dpa/sec in the reactor cavity. The resultant axial distributions of 4(E > 1.0 MeV),
$(E > 0.1 MeV), and dpa/sec from the gradient chain measurements are given in Tables 6.1-12, 6.1-13, and 6-14, respectively. The !
distributions of $(E > 1.0 MeV) are depicted graphically in ! l Figures 6.1-1 through 6.1-4. In these graphical presentations, results for axial locations of -6.0, 0.0, and +6.0 feet relative to the core midplane represent the explicit results of the FERRET l l evaluations summarized in Tables 6.1-5 through 6.1-10, while results at the remaining axial locations depict the normalized data from the gradient chains. 6-3
TABLE 6.1-1 SUM 4ARY OF REACTION RATES DERIVED FROM MULTIPLE FOIL SENSOR SETS CYCLE 15~ IRRADIATION REACTION RATE (ros/nuc1eus) CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE H J K L G I XX Cu- 63 (n, a) Cd 1.07E-18 9.85E-19 7.45E-19 7.14E-19 3.86E-19 3.83E-19 3.72E-20 Ti-46(n,p) Cd 1.66E-17 1.52E-17 1.14E-17 1.08E-17 6.55E-18 6.49E-18 6.23B-19 Fe-54(n,p) Cd 9.49E-17 8.52E-17 6.58E-17 5.52E-17 3.39E-17 3.64E-17 3.76E Ni-58(n,p) Cd 1.38E-16 1.21E-16 9.02E-17 8.06E-17 5.35E-17 5.24E-17 5.92E-18 U-238(n,f) Cd 5.49E-16 5.02E-16 3.73E-16 3.31E-16 2.19E-16 1.97E-16 3.27E-17 Np-237 (n, f) Cd 8.02E-15 7.82E-15 5.61E-15 4.89E-15 3.39E-15 3.23E-15 7.58E-16 Co-59 (n, y) 1.14E-13 1.42E-13 1.16E-13 7.34E-14 3.86E 4.53E-14 1.41E-14 Co-59 (n, y) Cd 6.71E-14 8.11E-14 6.60E-14 4.81e-14 2.85E-14 2.94E-14 1.01E-14 U-23 5 (n, f) 9.60E-13 1.18E-12 1.19E-12 5.47E-13 3.09E-13 3.52E-13 1.16E-13 U-235(n,f) Cd 3.12E-13 4.05E-13 3.12B-13 2.05E-13 8.26E-14 7.52E-14 5.16E-14'- i Note: Cd indicates that the sensor was cadmium covered. l 6-4 l l l
TABLE 6.1-2 Fe-54(n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS - CYCLE 15 IRRADIATION FEET REACTION RATE (rps/ nucleus) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 1.99E-17 1.54E-17 1.30E-17 1.15E-17
+5.5 4.85E-17 4.25E-17 3.17E-17 2.65E-17
+4.5 7.10E-17 6.46E-17 4.86E-17 4.25E-17
+3.5 8.31E-17 7.16E-17 5.35E-17 4.73E-17
+2.5 9.27E-17 7.68E-17 5.71E-17 5.22E-17
+1.5 9.15E-17 7.61E-17 5.44E-17 4.98E-17
+0.5 8.89E-17 7.68E-17 5.50E-17 5.01E-17
-0.5 9.27E-17 8.19E-17 5.95E-17 5.15E-17
-1.5 8.51E-17 7.93E-17 5.68E-17 5.05E-17
-2.5 8.12E-17 7.87E-17 6.03E-17 4.96E-17
-3.5 7.99E-17 7.36E-17 5.16E-17 4.73E-17
-4.5 7.16E-17 6.72E-17 4.51E-17 4.53E-17 1 I
-5.5 4.83E-17 4.31E-17 2.85E-17 2.84E-17
-6.5 1.78E-17 1.65E-17 1.24E-17 1.17E-17 2
N 5 I 1 6 6-5 J
I l TABLE 6.1-3 i Ni-58(n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS - CYCLE 15 IRRADIATION FEET REACTION RATE (rps/nucleas) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 2.96E-17 2.34E-17 1.91E-17 1.76E-17
+5.5 7.35E-17 5.81E-17 4.89E-17 4.07E-17
+4.5 1.08E-16 9.40E-17 6.95E-17 6.15E-17
+3.5 1.21E-16 1.01E-16 7.80E-17 6.83E-17
+2.5 1.32E-16 1.14E-16 8.43E-17 7.35E-17
+1.5 1.30E-16 1.12E-16 8.54E-17 7.17E-17
+0.5 1.22E-16 1.11E-16 8.09E-17 7.12E-17
-0.5 1.28E-16 1.16E-16 8.37E-17 7.52E-17
-1.5 1.18E-16 1.12E-16 8.09E-17 7.23E-17
-2.5 1.21E-16 1.06E-16 8.03E-17 7.12E-17 ,
-3.5 1.16E-16 1.06E-16 7.63E-17 6.61E-17
-4.5 1.06E-16 9.74E-17 6.83E-17 6.21E-17
-5.5 7.12E-17 6.43E-17 4.52E-17 4.30E-17
-6.5 2.79E-17 2.56E-17 1.89E-17 1.72E-17 l
1 l l 6-6 , 1 l
TABLE 6.1-4 Co-59(n,y)' REACTION RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS,- CYCLE 15 IRRADIATION FEET REACTION RATE (rps/ nucleus) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 2.40E-14 2.47E-14 2.18E-14 1.83E-14
+5.5 5.40E-14 7.97E-14 6.10E-14 3.75E-14
+4.5 6.95E-14 1.15E-13 8.76E-14 5.21E-14
+3.5 8.53E-14 1.33E-13 1.03E-13 5.99E-14
+2.5 9.49E-14 1.44E-13 1.10E-13 6.73E-14
+1.5 9.78E-14 1.44E-13 1.15E-13 7.06E-14
+0.5 1.01E-13 1.40E-13 1.19E-13 7.06E-14
-0.5 1.14E-13 1.44E-13 1.17E-13 7.18E-14
-1.5 1.06E-13 1.37E-13 1.14E-13 7.01E-14
-2.5 1.01E-13 1.29E-13 1.09E-13 6.73E-14
-3.5 9.27E-14 1.20E-13 9.89E-14 6.16E-14
-4.5 7.46E-14 1.00E-13 7.86E-14 5.11E-14 1
-5.5 4.94E-14 7.12E-14 4.21E-14 3.63E-14 !
-6.5 3.81E-14 3.85E-14 2.93E-14 2.76E-14 6-7 s
TABLE 6.1-5 DERIVED EXPOSURE RATES FROM THE CAPSULE H DOSIMETRY EVALUATION 0.0 DEGREE AZIMUTH ; CORE MIDPLANE TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 2.22E+09 1.95E+09 8%
$(E > 0.1'MeV) 1.74E+10 1.62E+10 17%
$(E < 0.414 eV) 4.17E+09 1.90E+09 25%
dpa/sec 1.24E-12 5.72E-12 13% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0.0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (rps/ nucleus)
- TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED j Cu- 63 (n, a) Cd 1.07E-18 1.10E-18 1.07E-18 0.97 1.00 i Ti-46(n,p) Cd 1.66E-17 1.64E-17 1.63E-17 1.01 1.02 Fe-54(n,p) Cd 9.49E-17 1.07E-16 9.65E-17 0.89 0.98 1 Ni- 58 (n, p) Cd 1.38E-16 1.54E-16 1.38E-16 0.90 1.00 U-238(n,f) Cd 5.49E-16 6.35E-16 5.58E-16 0.87 0.98 i Np-237(n,f) Cd 8.02E-15 8.62E-15 7.86E-15 0.93 1.02
! Co- 59 (n, y) 1.14E-13 2.08E-13 1.14E-13 0.55 1.00 Co- 59 (n,y) Cd 6.71E-14 9.78E-14 6.73E-14 0.69 1.00 i, U-235(n,f) 9.60E-13 1.95E-12 9.71E-13 0.49 0.99
, U-235 (n, f) Cd 3.12E-13 3.47E-13 3.05E-13 0.90 1.02 I
i a 4 6-8 l l r
TABLE 6.1-6 I DERIVED EXPOSURE RATES FROM THE CAPSULE J DOSIMETRY EVALUATION 15.0 DEGREE AZIMUTH - CORE MIDPLANE ( TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY l
$(E > 1.0 MeV) 1.89E+09 1.78E+09 8% ;
$(E > 0.1 MeV) 1.63E+10 1.66E+10 17%
$(E < 0.414 eV) 3.76E+09 2.40E+09 25%
dpa/sec 5.73E-12 5.74E-12 13% 1 l 4 COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES I 15.0 DEGREE AZIMUTH - CORE MIDPLANE I
, REACTION RATE (rps/ nucleus)
TRIAL ADJUSTED M/C M/C 4 MEASURED CALC. CALC. TRIAL ADJUSTED l Cu- 63 (n, cr) Cd 9.85E-19 9.11E-19 9.83E-19 1.08 1.00 I
- ' Ti-46(n,p) Cd 1.52E-17 1.36E-17 1.49E-17 1.12 1.02 Fe-54(n,p) Cd 8.52E-17 8.83E-17 8.65E-17 0.97 0.99 4
Ni-58(n,p) Cd 1.21E-16 1.27E-16 1.23E-16 0.95 0.98 J U-238(n,f) Cd 5.02E-16 5.33E-16 5.04E-16 0.94 1.00 Np-237(n,f) Cd 7.82E-15 7.70E-15 7.65E-15 1.02 1.02
) Co- 59 (n, y) 1.42E-13 1.98E-13 1.41E-13 0.72 1.01 Co-59(n,y) Cd 8.11E-14 9.93E-14 8.20E-14 0.82 0.99
- U-235(n,f) 1.18E-12 1.78E-12 1.22E-12 0.66 0.97 i U-235(n,f) Cd 4.05E-13 3.49E-13 3.87E-13 1.16 1.05 6-9 a-
l TABLE 6.1-7 DERIVED EXPOSURE RATES FROM THE CAPSULE K DOSIMETRY EVALUATION 30.0 DEGREE AZIMUTH - CORE MIDPLANE TRIAL ADJUSTED la VALUE VALUE IIRCFlRTAINTY
$(E > 1.0 MeV) 1.42E+09 1.31E+09 8%
$(E > 0.1 MeV) 1.24E+10 1.21E+10 17%
$(E < 0.414 eV) 3.01E+09 2.21E+09 24%
dpa/sec 4.35E-12 4.20E-12 13% 1 COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 30.0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (rps/ nucleus) TRIAL ADJUSTED h C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, or) Cd 7.45E-19 7.34E-19 7.46E-19 1.02 1.00 Ti-46(n,p) Cd 1.14E-17 1.08E-17 1.12E-17 1.06 1.02 Fe- 54 (n, p) Cd 6.58E-17 6.85E-17 6.59E-17 0.96 1.00 Ni-58(n,p) Cd 9.02E-17 9.84E-17 9.18E-17 0.92 0.98 U-238(n,f) Cd 3.73E-16 4.03E-16 3.75E-16 0.93 1.00 Np-237(n,f) Cd 5.60E-15 5.80E-15 5.54E-15 0.97 1.01 Co-59(n,7) 1.16E-13 1.58E-13 1.20E-13 0.73 0.97 Co-59(n,y) Cd 6.60E-14 7.97E-14 6.51E-14 0.83 1.01 ; l U-235(n,f) 1.19E-12 1.42E-12 1.10E-12 0.84 1.08 U-235(n,f) Cd 3.12E-13 2.80E-13 3.08E-13 1.11 1.01 1 1 6-10 l
TABLE 6.1-8 DERIVED EXPOSURE RATES FROM THE CAPSULE L DOSIMETRY EVALUATION 45.0 DEGREE AZIMUTH - CORE MIDPLANE TRIAL ADJUSTED la VALUE. VALUE UNCERTAINTY
$(E > 1.0 MeV) 1.17E+09 1.12E+09 8%
$(E > 0.1 MeV) 9.60E+09 9.99E+09 17%
$(E < 0.414 eV) 2.75E+09 1.03E+09 26%
dpa/sec 3.43E-12 3.49E-12 13% i COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES l 45.0 DEGREE AZIMUTH - CORE MIDPLANE 1 l REACTION RATE (rps/ nucleus) l TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 6 3 (n, or) Cd 7.14E-19 6.68E-19 7.10E-19 1.07 1.01 Ti-46(n,p) Cd 1.08E-17 9.68E-18 1.05E-17 1.12 1.03 Fe-54(n,p) Cd 5.52E-17 5.96E-17 5.71E-17 0.93 0.97 Ni-58(n,p) Cd 8.06E-17 8.49E-17 8.14E-17 0.95 0.99 U-238(n,f) Cd 3.31E-16 3.39E-16 3.26E-16 0.98 1.02 Np-237(n,f) Cd 4.88E-15 4.64E-15 4.73E-15 1.05 1.03 Co- 59 (n, y) 7.34E-14 1.35E-13 7.37E-14 0.54 1.00 Co-59(n,y) Cd 4.81E-14 6.20E-14 4.81E-14 0.78 1.00 U-235 (n, f) 5.47E-13 1.28E-12 5.56E-13 0.43 0.98 U-235 (n, f) Cd 2.05E-13 2.19E-13 2.01E-13 0.94 1.02 a 5 1 4 2 6-11 J 4
TABLE 6.1-9 DERIVED EXPOSURE RATES FROM THE CAPSULE G DOSIMETRY EVALUATION 0.0 DEGREE AZIMUTH - CORE TOP TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 7.84E+08 7.73E+08 8%
$(E > 0.1 MeV) 6.16E+09 6.41E+09 16%
$(E < 0.414 eV) 1.47E+09 5.37E+08 25%
dpa/sec 2.21E-12 2.26E-12 13% i COMPARISON OF. MEASURED AND CALCULATED SENSOR REACTION RATES i 0.0 DEGREE AZIMUTH - CORE TOP i REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C
- MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, ot) Cd 3.86E-19 3.88E-19 3.88E-19 1.00 1.00 Ti-46(n,p) Cd 6.55E-18 5.81E-18 6.30E-18 1.13 1.04 i Fe- 54 (n, p) Cd 6.39E-17 3.78E-17 3.57E-17 0.90 0.95 Ni-58(n,p) Cd 5.35E-17 5.42E-17 5.31E-17 0.99 1.01
, U-238 (n, f) Cd 2.19E-16 2.24E-16 2.17E-16 0.98 1.01 4 Np-237(n,f) Cd 3.38E-15 3.05E-15 3.22E-15 1.11 1.05 Co- 59 (n, y) 3.86E-14 7.35E-14 4.12E-14 0.53 0.94 Co-59 (n, y) Cd 2.85E-14 3.46E-14 2.73E-14 0.82 1.04 U-235(n,f) 3.09E-13 6.90E-13 2.82E-13 0.45 1.10 U-235 (n, f) Cd 8.26E-14 1.23E-13 8.84E-14 0.67 0.93 4 1 6-12 I n
,= - - - . a--
TABLE 6.1-10 i DERIVED EXPOSURE RATES FROM THE CAPSULE I DOSIMETRY EVALUATION l 0.0 DEGREE AZIMUTH - CORE BOTTOM l I l TRIAL ADJUSTED la l VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 8.42E+08 7.42E+08 8% l
$(E > 0.1 MeV) 6.62E+09 5.93E+09 16% i p(E < 0.414 eV) 1.58E+09 6.82E+08 23%
dpa/sec 2.38E-12 2.12E-12 13% 1 i l l COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0.0 DEGREE AZIMUTH - CORE BOTTOM REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, at) Cd 3.83E-19 4.17E-19 3.87E-19 0.92 0.99 Ti-46(n,p) Cd 6.49E-18 6.24E-18 6.29E-18 1.04 1.03 Fe-54 (n, p) Cd 3.64E-17 4.05E-17 3.68E-17 0.90 0.99 Ni-58(n,p) Cd 5.24E-17 5.83E-17 5.25E-17 0.90 1.00 U-238(n,f) Cd 1.97E-16 2.41E-16 2.10E-16 0.82 0.94 Np-237(n,f) Cd 3.22E-15 3.27E-15 3.05E-15 0.99 1.06 Co-59 (n,7) 4.53E-14 7.89E-14 4.68E-14 0.57 0.97 Co-59 (n, y) Cd 2.94E-14 3.71E-14 2.87E-14 0.79 1.02 U-235(n,f) 3.52E-13 7.41E-13 3.36E-13 0.48 1.05 U-235 (n, f) Cd 7.52E-14 1.32E-13 8.09E-14 0.57 0.93 6-13
TABLE 6.1-11 DERIVED EXPOSURE RATES FROM THE CAPSULE XX DOSIMETRY EVALUATION 0.0 DEGREE AZIMUTH - VESSEL SUPPORT TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 8.70E+07 1.25E+08 8%
$(E > 0.1 MeV) 6.83E+08 1.46E+09 17%
$(E < 0.414 eV) 1.63E+08 1.78E+08 27%
dpa/sec 2.45E-13 4.74E-13 12% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0.0 DEGREE AZIMUTH - VESSEL SUPPORT REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, a) Cd 3.72E-20 4.31E-20 3.75E-20 0.86 0.99 Ti-46(n,p) Cd 6.23E-19 6.44E-19 6.06E-19 0.97 1.03 Fe-54 (n, p) Cd 3.76E-18 4.19E-18 3.94E-18 0.90 0.95 Ni-58(n,p) Cd -5.92E-18 6.02E-18 6.03E-18 0.98 0.98 U-238(n,f) Cd 3.28E-17 2.49E-17 3.09E-17 1,32 1.06 Np-237(n,f) Cd 7.58E-16 3.38E-16 6.86E-16 2.24 1.11 Co-59 (n, y) 1.41E-14 8.15E-15 1.45E-14 1.73 0.97 Co- 59 (n, y) Cd 1.01E-14 3.83E-15 9.95E-15 2.64 1.02 U-235(n,f) 1.16E-13 7.65E-14 1.11E-13 1.52 1.05 U-235(n,f) Cd 5.16E-14 1.36E-14 4.83E-14 3.79 1.07 I 6-14
TABLE 6.1-12 FAST' NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY. CYCLE 15 IRRADIATION I FEET NEUTRON FLUX (n/cnf-sec) , FROM l MIDPLANE 0.0 DEG 14.0 DEG 30.0 DEG 45.0 DEG
+6.5 4.27E+08 3.45E+08 2.97E+08 2.53E+08
+5.5 1.04E+09 9.53E+08 7.25E+08 5.84E+08
+4.5 1.52E+09 1.45E+09 1.11E+09 9.37E+08
+3.5 1.79E+09 1.61E+09 1.23E+09 1.04E+09
+2.5 1.99E+09 1.72E+09 1.31E+09 1.15E+09 l
+1.5 1.96E+09 1.71E+09 1.25E+09 1.10E+09
+0.5 1.91E+09 1.72E+09 1.26E+09 1.10E+09
-0.5 1.99E+09 1.84E+09 1.36E+09 1.14E+09
-1.5 1.83E+09 1.78E+09 1.30E+09 1.11E+09
-2.5 1.74E+09 1.77E+09 1.38E+09 1.09E+09
-3.5 1.72E+09 1.65E+09 1.18E+09 1.04E+09
-4.5 1.54E+09 1.51E+09 1.03E+09 1.00E+09
-5.5 1.04E+09 9.68E+08 6.51E+08 6.26E+08
-6.5 3.82E+08 3.70E+08 2.84E+08 2.58E+08 6-15
l TABLE 6.1-13 l 1 FAST-NEUTRON FLUX (E > 0.1 MeV) - AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLE 15 IRRADIATION e i 2 FEET NEUTRON FLUX (n/cm -sec) FROM ' MIDPLANE 0.0 DEG 15.0'DEG 30.0 0];i 45.0 DEQ
+6.5' 3.55E+09- 3.21E+09- 2.74E+09 2.25E+09
+5.5 8.65E+09 8.89E+09 '6.69E+09 5.21E+09 '
+4.5' 1.27E+10- 1.35E+10 1.03E+10 8.36E+09
+3.5 1.48E+10 1.50E+10 1.13E+10 9.30E+09 ,
+2.5 1.65E+10 1.61E+10 1.21E+10 1.03E+10
+1.5 1.63E+10 1.59E+10 1.15E+10 9.80E+09 '
+0.5 1.59E+10 1.61E+10 1.16E+10 9.85E+09
-0.5 1.65E+10 1.71E+10 1.26E+10 1.01E+10
-1.5 1.52E+10 .1.66E+10 1.20E+10 9.93E+09
-2.5 1.45E+10 1.65E+10 1.27E+10 9.76E+09 e
-3.5 1.43E+10 1.54E+10 1.09E+10 9.31E+09
-4.5 1.28E+10 1.41E+10 9.53E+09 8.92E+09 .f
-5.5 8.61E+09 9.02E+09 6.02E+09 5.59E+09
-6.5 3.17E+09 3.45E+09 2.62E+09 2.30E+09 4
. 1 I
[. j i 6-16 I r t [
TABLE 6.1-14 IRON DISPLACEMENT RATE AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLE 15 IRRADIATION I FEET DISPLACEMENT RATE (dpa/sec) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 1.25E-12 1.11E-12 9.52E-13 7.87E-13 +5.5 3.05E-12 3.07E-12 2.32E-12 1.82E-12 +4.5 4.47E-12 4.67E-12 3.57E-12 2.92E-12 +3.5 5.24E-12 5.18E-12 3.93E-12 3.25E-12 +2.5 5.84E-12 5.56E-12 4.19E-12 3.59E-12 +1.5- 5.76E-12 5.51E-12 3.99E-12 3.42E-12 +0.5 5.60E-12 5.56E-12 4.04E-12 3.44E-12 -0.5 5.84E-12 5.92E-12 4.36E-12 3.54E-12 -1.5 5.36E-12 5.74E-12 4.17E-12 3.47E-12 -2.5 5.12E-12 5.69E-12 4.42E-12 3.41E-12 -3.5 5.04E-12 5.32E-12 3.78E-12 3.25E-12 -4.5 4.51E-12 4.86E-12 3.31E-12 3.12E-12 , -5.5 3.04E-12 3.12E-12 2.09E-12 1.95E-12 l -6.5 1.12E-12 1.19E-12 9.11E-13 8.04E-13 l
l
)
6-17
1 FIGURE 6.1-1 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 0.0 DEGREF TRAVERSE IN THE REACTOR CAVITY CYCLE 15 IRRADIATION 1.0E+10
,= .- =,
81.0E+09 m - [ \ , y I i , o \ t I L 5 Ir 8
=
z81.0E+08 1.0E+07 ' ' ' ' '-
-8 -6 -4 -2 0 2 4 6 8 s Distance From Core Midplane (ft) 6-18
1 I FIGURE 6.1-2 j FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 15.0 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 15 IRRADIATION l 1.0E+10 i 1
,= --
== n m 1.0E+09
?
\ .
N / \ E j g
$ I \ l
- a s 5
E 8 z51.0E+08 4 1.0E+07 ' ' '
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) 1 6-19
l 1 FIGURE 6.1-3 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 30.O DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 15 IRRADIATION 1.0E+10
/
/* "w% \
81.0E+09 'm m
? ,/ s 5 / (
3 , x '
/ 1_ f 3
LL 8
=
z51.0E+08 j 1.0 E +07
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) 6-20
FIGURE 6.1-4 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 45.0 DEGREE TRAVERSE IN THE REACTOR CAVITY , CYCLE 15 IRRADIATION 1.0E +10
' ' " =
$ 1.0E+09 N
- ~
? / \
d \
/ T r
3 1
/ \
E 4 L u. 8
=
z51.0E+08 i l 1.0E+07 ' ' ' l
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) i 1
l l 6-21 l
r
-6.2 - Cycle 16 Results I- 612.1 --Measured Reaction. rates '
During the Cycle 16 irradiation, six multiple foil sensor sets and four stainless steel gradient chains were deployed in the reactor cavity as depicted in Figures 2.1-1 and 2.1-2. The capsuleLidentifications associated with each of the multiple foil sensor sets were as follows: CAPSULE IDENTIFICATION AZIMUTH CORE -CORE CORE (decrees) TOP MIDPLANE BOTTOM 0.0 M N O 15.0 P 30.0 Q 45.0 R The contents of each of these irradiation capsules are specified in Appendix C to.this report. The irradiation history of the Point Beach Unit 2 reactor during Cycle 16 is also listed in Appendix C. The irradiation history was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report" for the applicable operating period. Based on this reactor operating history, the individual sensor characteristics, and the measured specific activities given in Appendix C, cycle average reaction rates referenced to a core power level of 1518 MWt were computed for each multiple foil sensor and gradient chain segment. The computed reaction rates for the multiple foil sensor sets, ' including radiometric foils and solid state track recorders, irradiated during Cycle 16 are provided in Table 6.2-1. Corresponding reaction rate data from the four stainless steel gradient chains are recorded in Tables 6.2-2 through 6.2-4 for , l the Fe-54(n,p), Ni-58(n,p), and Co-59 (n,y) reactions, respectively. In regard to the data listed in Table 6.2-1, the Fe-54 (n,p) ' reaction rates represent an average of the bare and cadmium 6-22 l I
covered measurements for.each capsule. Likewise, the U-238(n,f) reaction rates were obtained by. averaging the results of the radiometric foil and solid state-track recorder data. In addition, the fission rate measurements include corrections for
~
U-235 impurities in the U-238 sensors as well as corrections for photo-fission reactions in both the U-238 and Np-237 sensors. 6.2.2 - Results of the Least Squares Adjustment Procedure The results of the application of the least squares adjustment procedure to the six sets of multiple foil measurements obtained from the Cycle 16 irradiation are provided in Tables 6.2-5 through 6.2-10. In these tables, the derived exposure experienced at each sensor set location along with data illustrating the fit of both the trial and adjusted spectra to the measurements are given. Also included in the tabulations are the la uncertainties associated with each of the derived exposure rates. 1 In regard to the comparisons listed in Tables 6.2-5 through 6.2-10, it should be noted that the columns labeled " trial calc" were obtained by normalizing the neutron spectral data from Table l 4.1-1 to the absolute calculated neutron flux (E > 1.0 MeV) averaged over the Cycle 16 irradiation period as discussed in Section 3.0. Thus, the comparisons illustrated in Tables 6.2-5 ; through 6.2-10 indicate the degree to which the calculated neutron energy spectra matched the measured data before and after adjustment. Absolute comparisons are discussed further in Section 7.0 of this report. Complete traverses of fast neutron exposure rates in the reactor cavity were developed by combining the results of the least squares adjustment of the multiple foil data with the Fe-54 (n,p) reaction rate measurements from the gradient chains. The 4 gradient data were employed to establish relative axial distributions over the measurement range and these relative distributions were then normalized to the FERRET results from the midplane sensor sets to produce axial distributions of exposure rates in terms of $(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec in the reactor cavity. 6-23
The resultant axial distributions of $(E > 1.0 MeV),
; $(E > 0.1 MeV), and dpa/sec from the gradient chain measurements j- are given in Tables 6.2-11, 6.2-12, and 6.2-13, respectively.
The distributions of 4(E > 1.0 MeV) are depicted graphically in Figures 6.2-1 through 6.2-4. In these graphical presentations,
; results for axial locations of -6.0, 0.0, and +6.0 feet relative
- to the core midplane represent the explicit results of the FERRET evaluations summarized in Tables 6.2-5 through 6.2-10, while results at the remaining axial-locations depict the normalized data from the gradient chains.
i ) 4 1 1 2 i d 1 4 + 6-24
i-l
+
l.
- l. TABLE 6.2-1 l
l l SUB9fARY OF REACTION RATES DERIVED FROM MULTIPLE FOIL SENSOR SETS ' CYCLE 16 IRRADIATION ! L , i REACTION RATE (ros/ nucleus) l CAPSULE CAPSULE CAPSULE- CAPSULE CAPSULE? CAPSULE N P O R M O Cu- 63 (n, a) Cd 8.59E-19 7.50E-19 6.71E-19 6.61E-19 4.11E-19 3.69E-19 Ti-46(n,p) Cd 1.30E-17 1.12E-17 9.81E-18 9.36E 6.53E-18 5.90E-18 ! Fe-54(n,p) Cd 7.47E-17 6.35E-17 5.53E-17 5.25E-17 3.52E-17 .3.50E-17 Ni-58(n,p) Cd
~
1.07E-16 9.25E-17 7.85E-17 7.44E-17 5.52E-17 5.08E-17 U-238(n,f) Cd 4.24E-16 3.94E-16 2.87E-16 2.90E-17 2.25E-16 1.89E-16 Np-237(n,f) Cd 6.45E-15 5.66E-15 4.59E-15 4.47E-15 3.18E-15 3.14E-15 i Co-59 (n, y) 8.74E-14 1.12E-13 1.00E-13 6.23E-14 3.68E-14 4.08E-14 i Co-59(n,7) Cd 5.26E-14 6.71E-14 5.44E-14 4.14E-14 2.63E-14 ~2.65E-14 U-235 (n, f) 8.27E-13 1.02E-12 9.98E-13 4.41E-13 3.66E-13 I U-235(n,f)- Cd 2.66E-13 3.09E-13 2.32E-13 1.95E-13 7.72E-14. 7.27E-14 r n Note: Cd indicates that the sensor was cadmium covered. I 6-25 '
f TABLE 6.2-2 Fe-54(n,p) REACTION' RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS- CYCLE 16 IRRADIATION FEET REACTION RATE (rps/ nucleus) FROM-MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 1.93E-17 1.42E-17 1.19E-17 1.08E-17
+5.5 4.68E-17 3.52E-17 2.99E-17 2.57E-17
+4.5 7.39E-17 5.65E-17 4.34E-17 3.99E-17
+3.5 8.24E-17 6.76E-17 4.54E-17 4.19E-17
+2.5 8.77E-17 6.81E-17 5.01E-17 4.71E-17
+1.5 8.34E-17 6.44E-17 4.81E-17 4.67E-17
+0.5 7.50E-17 5.86E-17 5.04E-17 4.64E-17
-0.5 6.02E-17 5.23E-17 4.85E-17 4.57E-17
-1.5 5.91E-17 5.27E-17 4.71E-17 4.60E-17
-2.5 6.65E-17 5.15E-17 4.82E-17 4.40E-17
-3.5 6.92E-17 6.13E-17 4.37E-17 4.30E-17
-4.5 6.87E-17 5.86E-17 4.07E-17 4.06E-17
-5.5 4.66E-17 3.67E-17 2.69E-17 2.70E-17
-6.5 1.77E-17 1.57E-17 1.13E-17 1.07E-17 6-26
TABLE 6.2-3 Ni-58(n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS - CYCLE 16 IRRADIATION FEET REACTION RATE (rps/ nucleus) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 2.91E-17 2.11E-17 1.76E-17 1.54E-17
+5.5 7.10E-17 5.54E-17 4.33E-17 3.65E-17
+4.5 1.11E-16 8.63E-17 6.13E-17 5.75E-17
+3.5 1.18E-16 9.14E-17 6.66E-17 6.16E-17
+2.5 1.24E-16 9.52E-17 7.34E-17 6.60E-17
+1.5 1.15E-16 8.93E-17 7.04E-17 6.34E-17
+0.5 1.04E-16 8.43E-17 7.07E-17 6.34E-17
-0.5 8.69E-17 7.34E-17 7.10E-17 6.69E-17
-1.5 8.37E-17 7.60E-17 6.69E-17 6.28E-17 .
-2.5 9.19E-17 7.66E-17 6.75E-17 6.45E-17
-3.5 1.05E-16 8.43E-17 6.45E-17 6.13E-17
-4.5 9.81E-17 8.52E-17 6.01E-17 5.98E-17
-5.5 6.98E-17 5.42E-17 3.89E-17 3.98E-17
-6.5 2.87E-17 2.17E-17 1.60E-17 1.70E-17 6-27
TABLE 6.2-4 Co-59(n,7) REACTION RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS - CYCLE 16 IRRADIATION FEET REACTION RATE (rps/ nucleus) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 2.25E-14 2.25E-14 1.95E-14 1.67E-14
+5.5 5.14E-14 7.40E-14 5.53E-14 3.38E-14
+4.5 6.88E-14 1.06E-13 7.82E-14 4.68E-14
+3.5 7.98E-14 1.19E-13 8.97E-14 5.48E-14
+2.5 8.66E-14 1.25E-13 9.70E-14 5.89E-14
+1.5 8.40E-14 1.19E-13 9.59E-14 6.26E-14
+0.5 8.50E-14 1.13E-13 9.80E-14 6.21E-14
-0.5 8.34E-14 1.04E-13 9.54E-14 6.00E-14
-1.5 8.03E-14 9.96E-14 9.02E-14 6.05E-14
-2.5 7.87E-14 9.70E-14 0.97E-14 5.58E-14
-3.5 8.08E-14 9.91E-14 8.24E-14 5.27E-14
-4.5 6.62E-14 8.66E-14 6.47E-14 4.35E-14
-5.5 4.73E-14 6.36E-14 3.66E-14 3.24E-14
-6.5 3.48E-14 3.40E-14 2.59E-14 2.41E-14 6-28
TABLE 6.2-5 DERIVED EXPOSURE RATES FROM THE LAPSULE N DOSIMETRY EVALUATION l 0.O DEGREE AZIMUTH - CORE MIDPLANE l l l TRIAL ADJUSTED 1cr VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 1.80E+09 1.53E+09 8%
$(E > 0.1 MeV) 1.41E+10 1.31E+10 17%
$(E < 0.414 eV) 3.38E+09 1.52E+09 25%
dpa/sec 5.08E-12 4.59E-12 13% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0.0 DEGREE AZIMUTH - CORE MIDPLANE , REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C i MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, cr) Cd 8.59E-19 8.91E-19 8.56E-19 0.96 1.00 Ti-46(n,p) Cd 1.30E-17 1.33E-17 1.28E-17 0.98 1.02 Fe-54(n,p) Cd 7.47E-17 8.66E-17 7.57E-17 0.86 0.99 Ni-58(n,p) Cd 1.07E-16 1.24E-16 1.08E-16 0.86 0.99 U-238(n,f) Cd 4.24E-16 5.15E-16 4.35E-16 0.82 0.98 Np-237(n,f) Cd 6.45E-15 6.99E-15 6.28E-15 0.92 1.03 Co-59 (n,7) 8.74E-14 1.69E-13 8.97E-14 0.52 0.97 I Co- 59 (n, y) Cd 5.26E-14 7.93E-14 5.22E-14 0.66 1.01 U-235(n,f) 8.27E-13 1.58E-12 7.95E-13 0.52 1.03 U-235(n,f) Cd 2.66E-13 2.81E-13 2.61E-13 0.95 1.04 i 6-29
~ __ _ - - .
1 TABLE 6.2-6 DERIVED EXPOSURE RATES FROM THE CAPSULE P DOSIMETRY EVALUATION 15.0 DEGREE AZIMUTH - CORE MIDPLANE TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 1.54E+09 1.35E+09 8%
$(E > 0.1 MeV) 1.33E+10 1.24E+10 17%
$(E < 0.414 eV) 3.07E+09 1.93E+09 25%
dpa/sec 4.67E-12 4.29E-12 13% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 15.0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, ot) Cd 7.50E-19 7.42E-19 7.45E-19 1.01 1.01 Ti-46(n,p) Cd 1.12E-17 1.10E-17 1.10E-17 1.02 1.02 Fe-54 (n, p) Cd 6.35E-17 7.20E-17 6.50E-17 0.88 0.98 , Ni-58(n,p) Cd 9.25E-17 1.04E-16 9.32E-17 0.89 0.99 l U-238(n,f) Cd 3.94E-16 4.34E-16 3.84E-16 0.91 1.03 Np-237(n,f) Cd 5.66E-15 6.28E-15 5.65E-15 0.90 1.00 Co- 59 (n, y) 1.12E-13 1.61E-13 1.14E-13 0.70 0.98 Co- 59 (n, y) Cd 6.71E-14 8.09E-14 6.66E-14 0.83 1.01 U-235(n,f) 1.02E-12 1.45E-12 9.84E-13 0.70 1.04 U-235(n,f) Cd 3.09E-13 2.85E-13 3.03E-13 1.08 1.02 l l 1 f I 6-30
- l l
TABLE 6.2-7 DERIVED EXPOSURE RATES FROM THE CAPSULE Q DOSIMETRY EVALUATION , 30.0 DEGREE AZIMUTH - CORE MIDPLANE TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 1.21E+09 1.06E+09 8%
$(E > 0.1 MeV) 1.05E+10 9.77E+09 17%
$(E < 0.414 eV) 2.57E+09 1.93E+09 23%
dpa/sec 3.71E-12 3.40E-12 13% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 30.0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, a) Cd 6.71E-19 6.25E-19 6.68E-19 1.07 1.00 , Ti-46(n,p) Cd 9.81E-18 9.20E-18 9.68E-18 1.07 1.01 Fe- 54 (n, p) Cd 5.53E-17 5.84E-17 5.56E-17 0.95 1.00 Ni-58(n,p) Cd 7.85E-17 8.39E-17 7.87E-17 0.94 1.00 1 U-238 (n, f) Cd 2.87E-16 3.44E-16 3.04E-16 0.83 0.94 Np-237(n,f) Cd 4.59E-15 4.94E-15 4.48E-15 0.93 1.03 Co- 59 (n, y) 1.00E-13 1.35E-13 1.03E-13 0.74 0.97 Co- 59 (n, y) Cd 5.44E-14 6.79E-14 5.38E-14 0.80 1.01 U-235(n,f) 9.98E-13 1.21E-12 9.41E-13 0.83 1.06 U-23 5 (n, f) Cd 2.32E-13 2.38E-13 2.32E-13 0.98 1.00 j 6-31
TABLE 6.2-8
. DERIVED-EXPOSURE RATES FROM THE CAPSULE R DOSIMETRY EVALUATION 45.0 DEGREE AZIMUTH - CORE MIDPLANE TRIAL ADJUSTED icr VALUE VALUE IFCERTAINTY 1.06E+09 1.03E+09
$(E > 1.0 MeV) 8%
$(E > 0.1 MeV) 8.70E+09 9.06E+09 17%
$(E < 0.414 eV) 2.49E+09 8.11E+08 27%
dpa/sec 3.11E-12 3.17E-12 13% i COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 45.0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 6 3 (n, cr) Cd 6.61E-19 6.05E-19 6.54E-19 1.09 1.01 Ti-46(n,p) Cd 9.36E-18 8.77E-18 9.27E-18 1.07 1.01 Fe- 54 (n, p) Cd 5.25E-17 5.40E-17 5.32E-17 0.97 0.99 Ni-58(n,p) Cd 7.44E-17 7.69E-17 7.50E-17 0.97 0.99 U-238(n,f) Cd 2.90E-16 3.07E-16 2.95E-16 0.95 0.98 Np-237(n,f) Cd 4.48E-15 4.20E-15 4.32E-15 1.07 1.04 Co-59 (n, y) 6.23E-14 1.22E-13 6.19E-14 0.51 1.01 I Co- 59 (n, y) Cd 4.14E-14 5.62E-14 4.18E-14 0.74 0.99 U-235(n,f) 4.41E-13 1.16E-12 4.62E-13 0.38 0.96 U-235(n,f) Cd 1.95E-13 1.98E-13 1.87E-13 0.99 1.04 6-32 l
)
TABLE 6.2-9 DERIVED EXPOSURE RATES FROM THE CAPSULE M DOSIMETRY EVALUATION 0.0 DEGREE AZIMUTH.- CORE TOP TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 8.15E+08 7.70E+08 8%
$(E > 0.1 MeV) 6.40E+09 6.07E+09 16%
$(E < 0.414 eV) 1.53E+09 4.41E+08 30%
dpa/sec 2.30E-12 2.17E-12 13% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0.0 DEGREE AZIMUTH - CORE TOP REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, cr) Cd 4.11E-19 4.03E-19 4.10E-19 1.02 1.00 Ti-46(n,p) Cd 6.53E-18 6.03E-18 6.36E-18 1.08 1.03 Fe-54 (n, p) Cd 3.52E-17 3.92E-17 3.69E-17 0.90 0.95 Ni-58(n,p) Cd 5.52E-17 5.63E-17 5.47E-17 0.98 1.01 U-238(n,f) Cd 2.25E-16 2.33E-16 2.21E-16 0.97 1.02 Np-237(n,f) Cd 3.18E-15 3.16E-15 3.08E-15 1.01 1.03 Co- 59 (n, y) 3.68E-14 7.63E-14 3.73E-14 0.48 0.99 Co- 59 (n, y) Cd 2.63E-14 3.59E-14 2.60E-14 0.73 1.01 U-235 (n, f) Cd 7.72E-14 1.27E-13 8.09E-14 0.61 0.95 6-33
4 TABLE 6.2-10 DERIVED EXPOSURE RATES FROM THE CAPSULE O DOSIMETRY EVALUATION 0.0 DEGREE AZIMUTH - CORE BOTTOM TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 8.10E+08 7.21E+08 8%
p(E > 0.1 MeV) 6.36E+09 5.81E+09 16% ((E < 0.414 eV) 1.52E+09 6.70E+08 23% dpa/sec 2.29E-12 2.06E-12 13% 4 1 COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0.0 DEGREE AZIMUTH - CORE BOTTOM 4 REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu-63 (n, a) Cd 3.69E-19 4.01E-19 3.71E-19 0.92 1.00 TI-46(n,p) Cd 5.90E-18 6.00E-18 5.79E-18 0.98 1.02 Fe-54 (n,p) Cd 3.50E-17 3.90E-17 3.52E-17 0.90 0.99 Ni-58(n,p) Cd 5.08E-17 5.60E-17 5.07E-17 0.91 1.00 U-238(n,f) Cd 1.89E-16 2.32E-16 2.03E-16 0.82 0.93
- Np-237(n,f) Cd 3.14E-15 3.15E-15 2.97E-15 1.00 1.06 i Co- 59 (n, y) 4.08E-14 7.59E-14 4.33E-14 0.54 0.94 Co-59 (n, y) Cd 2.65E-14 3.57E-14 2.55E-14 0.74 1.04 U-235(n,f) 3.66E-13 7.12E-13 3.32E-13 0.51 1.10 U-235(n,f) Cd 7.27E-14 1.27E-13 7.86E-14 0.57 0.93 4
6-34
TABLE 6.2-11 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLE 16 IRRADIATION FEET NEUTRON FLUX (n/cm2 -sec) FROM-MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG r
'+6. 5 4.36E+08 3.44E+08 2.55E+08 2.41E+08
+5.5 1.06E+09 8.56E+08 6.41E+08 5.74E+08
+4.5 1.67E+09 1.38E+09 9.29E+08 8.92E+08
+3.5 1.86E+09 1.65E+09 9.72E+08 9.38E+08
+2.5 1.98E+09 1.66E+09 1.07E+09 1.05E+09
+1.5 1.89E+09 1.57E+09 1.03E+09 1.04E+09
+0.5 1.70E+09 1.43E+09 1.08E+09 1.04E+09
-0.5 1.36E+09 1.27E+09 1.04E+09 1.02E+09
-1.5 1.34E+09 1.28E+09 1.01E+09 1.03E+09
-2.5 1.51E+09 1.25E+09 1.03E+09 9.86E+08
-3.5 1.57E+09 1.49E+09 9.36E+08 9.63E+08
-4.5 1.55E+09 1.43E+09 8.71E+08 9.08E+08
-5.5 1.06E+09 8.94E+08 5.77E+08 6.05E+08
-6.5 4.02E+08 3.82E+08 2.41E+08 2.40E+08 l
6-35
TABLE 6.2-12 FAST NEUTRON FLUX (E > 0.1 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLE 16 IRRADIATION FEET NEUTRON FLUX (n/cm2 -sec) FROM MIRPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 3.73E+08 3.16E+08 2.35E+09 2.12E+09 +5.5 9.08B+09 7.86E+09 5.90E+09 5.05E+09 +4.5 1.43E+10 1.26E+10 8.56E+09 7.85E+09 +3.5 1.60E+10 1.51E+10 8.96E+09 8.25E+09 +2.5 1.70E+10 1.52E+10 9.90E+09 9.26E+09 +1.5 1.62E+10 1.44E+10 9.50E+09 9.19E+09 +0.5 1.45E+10 1.31E+10 9.95E+09 9.13E+09 -0.5 1.17E+10 1.17E+10 9.59E+09 8.99E+09 -1.5 1.15E+10 1.18E+10 9.29E+09 9.05E+09 -2.5 1.29E+10 1.15E+10 9.52E+09 8.67E+09 -3.5 1.34E+10 1.37E+10 8.63E+09 8.47E+09 -4.5 1.33E+10 1.31E+10 8.03E+09 7.99E+09 -5.5 9.04E+09 8.21E+09 5.32E+09 5.32E+09 -6.5 3.44E+09 3.51E+09 2.22E+09 2.11E+09 6-36 i
TABLE 6.2-13 i IRON DISPLACEMENT RATE AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLE 16 IRRADIATION l I 1 FEET DISPLACEMENT RATE (dpa/sec) j FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 1.31E-12 1.09E-12 8.17E-13 7.42E-13 +5.5 3.18E-12 2.72E-12 2.05E-12 1.77E-12 +4.5 5.02E-12 4.37E-12 2.98E-12 2.75E-12 l +3.5 5.59E-12 5.23E-12 3.12E-12 2.89E-12 +2.5 5.95E-12 5.27E-12 3.45E-12 3.24E-12 +1.5 5.67E-12 4.98E-12 3.31E-12 3.22E-12 .+0.5 5.09E-12 4.54E-12 3.46E-12 3.19E-12 -0.5. 4.09E-12 4.04E-12 3.34E-12 3.15E-12 -1.5 4.02E-12 4.07E-12 3.23E-12 3.17E-12 -2.5 4.52n 12 3.98E-12 3.31E-12 3.03E-12 -3.5 4.70E-12 4.74E-12 3.00E-12 2.96E-12 -4.5 4.66E-12 4.54E-12 2.80E-12 2.79E-12 -5.5 3.17E-12 2.84E-12 1.85E-12 1.86E-12 -6.5 1.20E-12 1.21E-12 7.73E-13 7.38E-13 1
6-37
FIGURE 6.2-1 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 0.0 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 16 IRRADIATION 1.0E +10
~
my f
$m 1.0E+09 ,
f , ek ; , E , n
$ ! h 5
E 8
=
z51.0E+08 1.0 E +07
-8 -6 -4 -2 0 2 4 6 8 i Distance From Core Midplane (ft) l l
l l 6-38 ; 1 i I
- .~.
FIGURE 6.2-2 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 15.0 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 16 IRRADIATION l 1.0E+10 m n _ 2- d 81.0E +09 > ,
?
N r
/ i, A L g
I \ h 5 E 8 x z81.0E +08 _ 1.0E +07
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) 6-39
FIGURE 6.2-3 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 30.0 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 16 IRRADIATION 1.0E +10
=- ^-
$ 1.0E+09 m -
? ,' %
/ h P
5 T y ( ,
~
) \
.E J b u.
E i b 1 z51.0E+08 l
-s_
1.0E+07
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) l 6-40
l FIGURE 6.2-4 l FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 45.0 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 16 IRRADIATION l l l 1.0E+10
$ 1.0E+09
?
1 f- 's \ i g i s r T 3
~
/ (
E u. J \ 8
=
z51.0E+08 4 l 1.0E+07
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) 6-41
6.3 - Cycle 17 Results 6.3.1 - Measured Reaction rates During the Cycle 17 irradiation, six multiple foil sensor sets and four stainless steel gradient chains were deployed in the reactor cavity as depicted in Figures 2.1-1 and 2.1-2. The capsule identifications associated with each of the multiple foil sensor sets were as follows: 4 CAPSULE IDENTIFICATION AZIMUTH CORE CORE CORE (decrees) TOP MIDPLANE BOTTOM 0.0 AA BB CC 15.0 DD 30.0 EE 45.0 FF The contents of each of these irradiation capsules are specified in Appendix D to this report. The irradiation history of the Point Beach Unit 2 reactor during Cycle 17 is also listed in Appendix D. The irradiation history was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report" for the applicable operating period. Based on this reactor operating history, the individual sensor characteristics, and the measured specific activities given in Appendix D, cycle average reaction rates referenced to a core power level of 1518 MWt were computed for each multiple foil sensor and gradient chain segment. The computed reaction rates for the multiple foil sensor sets, including radiometric foils and solid state track recorders, irradiated during Cycle 17 are provided in Table 6.3-1. Corresponding reaction rate data from the four stainless steel gradient chains are recorded in Tables 6.3-2 through 6.3-4 for the Fe-54 (n,p) , Ni- 58 (n, p) , 'and Co-59 (n, y) reactions, respectively. In regard to the data listed in Table 6.3-1, the Fe-54 (n,p) reaction rates represent an average of the bare and cadmium 6-42
covered measurements for each capsule. Likewise, the U-238(n,f) j reaction rates were obtained by averaging the results of the radiometric foil and solid state track recorder data. In I addition, the fission rate measurements include corrections for U-235 impurities in the U-238 sensors as well as corrections for i photo-fission reactions in both the U-238 and Np-237 sensors. l 6.3.2 - Results of the Least Squares Adjustment Procedure The results of the application of the least squares adjustment procedure to the six sets of multiple foil measurements obtained from the Cycle 17 irradiation are provided in Tables 6.3-5 through 6.3-10. In these tables, the derived exposure experienced at each sensor set location along with data illustrating the fit of both the trial and adjusted spectra to ! the measurements are given. Also included in the tabulations are l the la uncertainties associated with each of the derived exposure rates. In regard to the comparisons listed in Tables 6.3-5 through 6.3-10, it should be noted that the columns labeled " trial calc" were obtained by normalizing the neutron spectral data from Table 4.1-1 to the absolute calculated neutron flux (E > 1.0 MeV) averaged over the Cycle 17 irradiation period as discussed in Section 3.0. Thus, the comparisons illustrated in Tables 6.3-5 through 6.3-10 indicate the degree to which the calculated neutron energy spectra matched the measured data before and after adjustment. Absolute comparisons are discussed further in Section 7.0 of this report. Complete traverses of fast neutron exposure rates in the reactor cavity were developed by combining the results of the least squares adjustment af the multiple foil data with the Pe-54 (n,p) reaction rate measurements from the gradient chains. The gradient data were employed to establish relative axial distributions over the measurement range and these relative distributions were then normalized to the FERRET results from the . midplane sensor sets to produce axial distributions of exposure rates in terms of $(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec in the reactor cavity. 6-43
The resultant axial distributions of $(E > 1.0 MeV),
$(E > 0.1 MeV), and dpa/sec from the gradient chain measurements are given in Tables 6.3-11, 6.3-12, and 6.3-13, respectively.
The distributions of $(E > 1.0 MeV) are depicted graphically in Figures 6.3-1 through 6.3-4. In these graphical cresentations, results for axial locations of -6.0, 0.0, and +6.0 feet relative to the core midplane represent the explicit resultu of the FERRET n evaluations summarized in Tables 6.3-5 through 6.3.10, while results at the remaining axial locations depict the normalized data from the gradient chains.
- f 6-44 i
TABLE 6.3-1
SUMMARY
OF REACTION RATES DERIVED FROM MULTIPLE FOIL SENSOR Sh'rS CYCLES 17 IRRADIATION REACTION RATE (ros/ nucleus) t CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE BB DD EE FF AA CC : Cu- 63 (n, a) Cd 9.04E-19 8.01E-19 6.79E-19 6.97E-19 4.23E-19 3.92E-19 Ti-46(n,p) Cd 1.34E-17 1.18E-17 1.02E-17 1.00E-17 6.68E-18 6.19E-18 Fe-54(n,p) Cd 7.82E-17 6.82E-17 5.66E-17 5.43E-17 3.57E-17 3.63E-17 Ni-58 (n,p) Cd 1.10E-16 9.61E-17 8.01E-17 7.65E-17 5.67E-17 5.23E-17 U-238(n,f) Cd 3.98E-16 3.93E-16 3.08E-16 2.78E-16 2.26E-16 1.85E-16 Np-237(n,f) Cd 6.67E-15 5.39E-15 4.70E-15 3.90E-15 2.51E-15 Co- 59 (n, y) 9.31E-14 1.17E-13 1.05E-13 6.50E-14 3.89E-14 4.25E-14 Co- 59 (n, y) Cd 5.52E-14 6.68E-14 5.76E-14 4.40E-14 2.79E-14 2.79E-14 U-235(n,f) 7.83E-13 1.10E-12 1.02E-12 4.52E-13 3.32E-13 3.95E-13 U-235 (n, f) Cd 2.63E-13 2.98E-13 2.66E-13 1.97E-13 9.87E-14 7.12E-14 Note: Cd indicates that the sensor was cadmium covered. i 6-45
TABLE 6.3-2 Fe-54(n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS - CYCLE 17 IRRADIATION FEET REACTION RATE (rps/ nucleus) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 2.03E-17 1.52E-17 1.17E-17 1.09E-17
+5.5 5.05E-17 3.86E-17 2.97E-17 2.53E-17 44.5 7.51E-17 5.25E-17 4.50E-17. 3.95E-17 3.5 8.39E-17 6.33E-17 5.30E-17 4.55E-17
+2.C 8.83E-17 7.02E-17 4.96E-17 5.05E-17
+1.5 8.84E-17 6.82E-17 5.05E-17 4.71E-17
+0.5 7.85E-17 6.43E-17 5.40E-17 4.59E-17
-0.5 6.77E-17 6.13E-17 5.35E-17 5.30E-17
-1.5 6.67E-17 5.54E-17 5.10E-17 4.91E-17
-2.5 6.57E-17 6.23E-17 5.20E-17 5.15E-17
-3.5 7.26E-17 6.13E-17 4.87E-17 4.74E-17
-4.5 7.06E-17 6.13E-17 4.62E-17 4.19E-17
-5.5 5.00E-17 4.07E-17 2.80E-17 2.91E-17
-6.5 2.08E-17 1.49E-17 1.13E-17 1.28E-17 6-46
TABLE 6.3-3 Ni-58(n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS - CYCLES 17 IRRADIATION FEET REACTION RATE (rps/ nucleus) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 3.14E-17 2.33E-17 1.87E-17 1.64E-17
+5.5 7.50E-17 5.99E-17 4.23E-17 3.68E-17
+4.5 1.14E-16 9.06E-17 6.49E-17 5.83E-17
+3.5 1.19E-16 9.49E-17 7.20E-17 6.53E-17
+2.5 1.28E-16 1.03E-16 7.57E-17 7.46E-17
+1.5 1.22E-16 9.39E-17 7.89E-17 7.12E-17
+0.5 1.09E-16 9.00E-17 7.30E-17 6.70E-17 i
-0.5 9.41E-16 8.29E-17 7.95E-17 7.24E-17
-1.5 9.10E-16 7.99E-17 7.48E-17 7.32E-17
-2.5 1.03E-16 8.35E-17 7.69t-17 7.10E-17 )
-3.5 1.05E-16 9.12E-17 7.06E-17 6.84E-17
-4.5 1.08E-16 8.44E-17 6.72E-17 6.05E-17
-5.5 7.22E-17 5.89E-17 4.43E-17 2.06E-17
-6.5 2.95E-17 2.35E-17 1.79E-17 1.86E-17 1
6-47 I
TABLE 6.3-4 Co-59 (n, y) REACTION RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS - CYCLE 17 IRRADIATION FEET REACTION RATE (rps/ nucleus) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.S 2.11E-14 2.19E-14 1.81E-14 1.55E-14
+5.5 3.93E-14 7.05E-14 5.06E-14 3.13E-14
+4.5 5.32E-14 9.92E-14 7.31E-14 4.20E-14
+3.5 6.16E-14 1.12E-13 8.56E-14 5.03E-14
+2.5 6.68E-14 1.18E-13 9.19E-14 5.58E-14
+1.5 6.63E-14 1.15E-13 9.45E-14 5.90E-14
+0.5 6.68E-14 1.07E-13 9.45E-14 5.90E-14
-0.5 6.63E-14 8.45E-14 7.62E-14 4.92E-14
-1.5 6.42E-14 7.93E-14 7.41E-14 4.83E-14
-2.5 6.47E-14 7.88E-14 7.15E-14 4.62E-14
-3.5 6.37E-14 7.72E-14 6.58E-14 4.31E-14
-4.5 5.32E-14 6.68E-14 5.32E-14 3.51E-14
-5.5 3.64E-14 4.92E-14 2.88E-14 1.13E-14
-6.5 3.39E-14 2.63E-14 2.03E-14 1.93E-14 c 6-48
-t
TABLE 6.3-5 DERIVED EXPOSURE RATES FROM THE CAPSULE BB DOSIMETRY EVALUATION 0.0 DEGREE AZIMUTH - CORE MIDPLANE TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1 0 MeV) 1.81E+09 1.52E+09 8%
((E > 0.1 MeV) 1.42E+10 1.32E+10 17%
$(E < 0.414 eV) 3.40E+09 1.54E+09 15%
dpa/sec 5.11E-12 4.62E-12 13% l l COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0.0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (rps/ nucleus) TRIAL ADJUSTED C/M C/M MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n,0f) Cd 9.04E-19 8.96E-19 9.00E-19 1.01 1.00 Ti-46 (n,p) Cd 1.34E-17 1.34E-17 1.32E-17 1.00 1.02 Fe- 54 (n, p) Cd 7.82E-17 8.71E-17 7.83E-17 0.90 1.00 Ni-58 (n,p) Cd 1.10E-16 1.25E-16 1.11E-16 0.88 0.99 U-238 (n, f) Cd 3.97E-16 5.17E-16 4.33E-16 0.77 0.92 Np-237(n,f) Cd 6.67E-15 7.03E-15 6.38E-15 0.95 1.05 l Co-59 (n, y) 9.31E-14 1.70E-13 9.34E-14 0.55 1.00 ) Co- 59 (n, y) Cd 5.52E-14 7.97E-14 5.54E-14 0.69 1.00 l U-235 (n, f) 7.83E-13 1.59E-12 7.93E-13 0.49 0.99 U-235 (n, f) Cd 2.63E-13 2.83E-13 2.56E-13 0.93 1.03 6-49
TABLE 6.3-6 , i DERIVED EXPOSURE RATES FROM THE CAPSULE DD DOSIMETRY EVALUATION I 15.0 DEGREE AZIMUTH - CORE MIDPLANE i l TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 1.55E+09 1.34E+09 8%
$(E > 0.1 MeV) 1.34E+10 1.18E+10 17%
3.09E+09
$(E < 0.414 eV) 2.13E+09 24%
dpa/sec 4.70E-12 4.15E-12 13% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 15.0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu-63(n,a) Cd 8.01E-19 7.47E-19 7.96E-19 1.07 1.01 Ti-4 6 (n, p) Cd 1.18E-17 1.11E-17 1.17E-17 1.06 1.01 Fe- 54 (n, p) Cd 6.82E-17 7.24E-17 6.89E-17 0.94 0.99 : Ni-58(n,p) Cd 9.61E-17 1.05E-16 9.71E-17 0.92 0.99 U-238(n,f) Cd 3.93E-16 4.37E-16 3.90E-16 0.90 1.01 Np-237(n,f) Cd 5.39E-15 6.32E-15 5.43E-15 0.85 0.99 Co- 59 (n,7) 1.17E-13 1.62E-13 1.20E-13 0.72 0.98 < Co- 59 (n, y) Cd 6.68E-14 8.14E-14 6.63E-14 0.82 1.01 U-235(n,f) 1.10E-12 1.46E-12 1.06E-12 0.75 1.04 U-235 (n, f) Cd 2.98E-13 2.86E-13 2.93E-13 1.04 1.02 l l l l 6-50 1
TABLE 6.3-7 ] DERIVED EXPOSURE RATES FROM THE CAPSULE EE DOSIMETRY EVALUATION 30.0 DEGREE AZIMUTH - CORE MIDPLANE TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
- $(E > l'0 MeV)
. 1.24E+08 1.10E+09 8%
$(E > 0.1 MeV) 1.08E+10 1.02E+10 17%
$(E < 0.414 eV) 2.63E+09 1.99E+09 24%
dpa/sec 3.80E-12 3.53E-12 13% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 30.0 DEGREE AZIMUTH - CORE MIDPLANE 4
- l. REACTION RATE (rps/ nucleus)
- TRIAL ADJUSTED M/C M/C 2
MEASURED CALC. CALC. TRIAL ADJUSTED
- Cu- 63 (n, at) Cd 6.79E-19 6.41E-19 6.78E-19 1.06 1.00 i Ti-46(n,p) Cd 1.02E-17 9.42E-18 1.00E-17 1.08 1.02 l Fe-54 (n, p) Cd 5.66E-17 5.99E-17 5.72E-17 0.95 0.99 Ni-58 (n, p) Cd 8.01E-17 8.59E-17 8.07E-17 0.93 0.99 U-238(n,f) Cd 3.08E-16 3.52E-16 3.17E-16 0.88 0.97 Np- 237 (n , f ) Cd 4.70E-15 5.07E-15 4.64E-15 0.93 1.01 Co-59 (n, y) 1.05E-13 1.38E-13 1.07E-13 0.76 0.98
- Co- 59 (n, y) Cd 5.76E-14 6.96E-14 5.73E-14 0.83 1.01 i U-235(n,f) 1.02E-12 1.24E-12 9.80E-13 0.82 1.01
, U-235(n,f) Cd 2.66E-13 2.44E-13 2.61E-13 1.09 1.04 6-51 r v.
I L I TABLE 6.3-8
- DERIVED EXPOSURE RATES FROM THE CAPSULE FF DOSIMETRY EVALUATION I 45.0 DEGREE AZIMUTH - CORE MIDPLANE ,
i ADJUSTED la ' TRIAL VALUE VALUE UNCERTAINTY 1
$(E > 1.0 MeV) .1.10E+09 9.66E+08 8%
9.03E+09 8.09E+09 17% ((E > O.1 MeV)
$(E < 0.414 eV) 2.58E+09 8.25E+08 27%
dpa/sec 3.23E-12 2.88E-12 13% i COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 45.0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, at) Cd 6.97E-19 6.28E-19 6.92E-19 1.11 1.01 Ti-46(n,p) Cd 1.00E-17 9.10E-18 9.87E-18 1.10 1.01 Fe-54(n,p) Cd 5.43E-17 5.60E-17 5.48E-17 0.97 0.99 Ni-58(n,p) Cd 7.65E-17 7.98E-17 7.68E-17 0.96 1.00 U-238(n,f) Cd 2.78E-16 3.18E-16 2.87E-16 0.87 0.97 Np-237(n,f) Cd 3.90E-15 4.36E-15 3.86E-15 0.89 1.01 Co-59(n,y) 6.50E-14 1.27E-13 6.48E-14 0.51 1.00 Co- 59 (n,7) Cd 4.40E-14 5.83E-14 4.43E-14 0.76 0.99 U-235(n,f) 4.52E-13 1.21E-12 4.70E-13 0.37 0.96 U-235(n,f)- Cd 1.97E-13 2.06E-13 1.90E-13 0.96 1.04 6-52
TABLE 6.3-9 DERIVED EXPOSURE RATES FROM THE CAPSULE AA DOSIMETRY EVALUATION 0.0 DEGREE AZIMUTH - CORE TOP TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY 4(E > 1.0 MeV) 8.26E+08 7.54E+08 10%
$ (E > 0.1 MeV) 6.49E+09 5.91E+09 22%
$(E < 0.414 eV) 1.55E+09 5.69E+08 26%
dpa/sec 2.33E-12 2.12E-12 17% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0.0 DEGREE AZIMUTH - CORE TOP REACTION RATE-(rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu-63(n,a) Cd 4.23E-19 4.09E-19 4.22E-19 1.03 1.00 Ti-46(n,p) Cd 6.68E-18 6.12E-18 6.51E-18 1.09 1.03 Fe-54(n,p) Cd 3.57E-17 3.98E-17 3.75E-17 0.90 0.95 Ni-58(n,p) Cd 5.67E-17 5.71E-17 5.58E-17 0.99 1.02 U-238(n,f) Cd 2.26E-16 2.36E-16 2.20E-16 0.96 1.03 Co-59 (n, y) 3.89E-14 7.74E-14 4.13E-14 0.50 0.94 Co-59(n,y) Cd 2.79E-14 3.64E-14 2.69E-14 0.77 1.04 U-235(n,f) 3.32E-13 7.27E-13 3.04E-13 0.46 1.09 l U-235(n,f) Cd 9.87E-14 1.29E-13 1.02E-13 0.77 0.97 l l 6-53 l
TABLE 6.3-10 DERIVED EXPOSURE RATES FROM THE CAPSULE CC DOSIMETRY EVALUATION ) 0.O DEGREE AZIMUTH - CORE BOTTOM j l TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 8.40E+08 6.66E+08 8%
((E > 0.1 MeV) 6.60E+09 4.89E+09 16%
$(E < 0.414 eV) 1.58E+09 7.10E+08 22%
dpa/sec 2.37E-12 1.79E-12 12% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0.0 DEGREE AZIMUTH - CORE BOTTOM REACTION RATE (rps/ nucleus) i TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, a) Cd 3.92E-19 4.16E-19 3.94E-19 0.94 1.00 Ti-46(n,p) Cd 6.19E-18 6.22E-18 6.09E-18 1.00 1.02 Fe-54(n,p) Cd 3.63E-17 4.04E-17 3.63E-17 0.90 1.00 J Ni-58(n,p) Cd 5.23E-17 5.81E-17 5.19E-17 0.90 1.01 U-238(n,f) Cd 1.85E-16 2.40E-16 1.97E-16 0.77 0.94 Np-237(n,f) Cd 2.51E-15 3.26E-15 2.49E-15 0.77 1.01 Co- 59 (n, y) 4.25E-14 7.87E-14 4.57E-14 0.54 0.93 Co- 59 (n, y) Cd 2.79E-14 3.70E-14 2.67E-14 0.75 1.05 U-235(n,f) 3.95E-13 7.39E-13 3.50E-13 0.54 1.13 U-235(n,f) Cd 7.12E-14 1.31E-13 7.73E-14 0.54 0.92 1 6-54 i
TABLE 6.3-11 FAST NEUTRON FLUX (E.> 1.0 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLE 17 IRRADIATION FEET NEUTRON FLUX (n/cm2 -sec) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG-
+6.5 4.21E+08 3.25E+08 2.40E+08 2.13E+08 +5.5 1.05E+09 8.23E+08 6.09E+08- 4.93E+08 +4.5 1.56E+09 1.12E+09 9.21E+08 7.71E+08 +3.5 1.74E+09 1.35E+09 1.08E+09 8.88E+08 +2.5 1.84E+09 1.50E+09 1.01E+09 9.87E+08 +1.5 1.75E+09 1.46E+09 1.03E+09 9.19E+08 +0.5 1.63E+09 1.37E+09 1.11E+09 8.97E+08 -0.5 1.41E+09 1.31E+09 1.09E+09 1.03E+09 -1.5 1.39E+09 1.18E+09 1.04E+09 9.58E+08 -2.5 1.37E+09 1.33E+09 1.06E+09 1.01E+09 -3.5 1.51E+09 1.31E+09 9.98E+08 9.27E+08 -4.5 1.47E+09 1.31E+09 9.46E+08 8.18E+08 -5.5 1.04E+09 8.69E+08 5.74E+08 5.68E+08 -6.5 4.31E+08 3.18E+08 2.31E+08 2.49E+08 i
I l 6-55
I l TABLE 6.3-12 FAST NEUTRON FLUX (E > 0.1 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLE 17 IRRADIATION FEET NEUTRON FLUX (n/cm2 -sec) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 3.66E+09 2.86E+09 2.23E+09 1.78E+09 +5.5 9.13E+09 7.25E+09 5.65E+09 4.13E+09 +4.5 1.36E+10 9.86E+09 8.54E+09 6.46E+09 +3.5 1.51E+10 1.19E+10 1.01E+10 7.44E+09 +2.5 1.59E+10 1.32E+10 9.41E+09 8.27E+09 +1.5 1.52E+10 1.28E+10 9.60E+09 7.70E+09 +0.5 1.42E+10 1.21E+10 1.02E+10 7.51E+09 -0.5 1.22E+10 1.15E+10 1.02E+10 8.67E+09 -1.5 1.20E+10 1.04E+10 9.69E+09 8.03E+09 -2.5 1.19E+10 1.17E+10 9.87E+09 8.43E+09 -3.5 1.31E+10 1.15E+10 9.25E+09 7.76E+09 -4.5 1.28E+10 1.15E+10 8.77E+09 6.85E+09 -5.5 9.04E+09 7.65E+09 5.32E+09 4.76E+09 -6.5 3.75E+09 2.80E+09 2.14E+09 2.09E+09 ;
I l 1 I 6-56 j I 1
TABLE 6.3-13 IRON DISPLACEMENT RATE AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLE 17 IRRADIATION i FEET DISPLACEMENT RATE (dpa/sec) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 1.28E-12 1.01E-12 7.70E-13 6.34E-13
+5.5 3.19E-12 2.55E-12 1.95E-12 1.47E-12
+4.5 4.74E-12 3.47E-12 2.96E-12 2.30E-12
+3.5- 5.30E 4.18E-12 3.48E-12 2.65E-12
, +2.5 5.58E-12 4.64E-12 3.26E-12 2.94E-12
+1.5 5.33E-12 4.51E-12 3.32E-12 2.74E-12
+0.5 4.96E-12 4.25E-12 3.55E-12 2.67E-12
-0.5 4.28E-12 4.05E-12 3.51E-12 3.09E-12
-1.5 4.22E-12 3.66E-12 3.35E-12 2.86E-12
-2.5 4.16E-12 4.12E-12 3.42E-12 3.00E-12
-3.5 4.59E-12 4.05E-12 3.20E-12 2.76E-12
-4.5 4.46E-12 4.05E-12 3.04E-12 2.44E-12
-5.5 3.16E-12 2.69E-12 1.84E-12 1.69E-12
-6.5 1.31E-12 9.85E-13 7.41E-13 7.43E-13
( e d 6-57 )
FIGURE 6.3-1 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 0.0 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 17 IRRADIATION 1.0E+10 f ~ sm 1.0E+09 m , [ \ ,
& t I I B Y \
w 4 L s E 8 y 1.0E+08 1.0E+07 ' ' ' '
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) 6-58
FIGURE 6.3-2 FAST NEUTRON FLUX (E > 1.O MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 15.0 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 17 IRRADIATION 1.0E+10 1
$ 1.0E+09
?
, \
m r
/ i, 1 E ; (
$ / \
~
l 1 5 II 8 u z51.0E+08 2 l i l a 1.0E +07
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) 6-59
FIGURE 6.3-3 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 30.0 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 17 IRRADIATION 1.0E +10
= "=- r"
[1.0E+09 ,
& / 1 5 r x 3 /
~
X J \ 2 ] 1 u. 8
=
z51.0E+08 1.0E+07
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) 6-60
FIGURE 6.3-4 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 45.0 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 17 IRRADIATION 1.0E+10 g 1.0E+09 -- - , N / \ E p x
$ ( \
~
/ \
.5 4 \
- u. L 8
3 z51.0E+08 1.0E+07
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) 6-61
6.4 - Cycles 18/20 Results 6.4.1 - Measured Reaction rates During the Cycles 18/20 irradiation, six multiple foil sensor sets and four stainless steel gradient chains were deployed in the reactor cavity as depicted in Figures 2.1-1 and 2.1-2. The capsule identifications associated with each of the multiple foil sensor sets were as follows: CAPSULE IDENTIFICATION AZIMUTH CORE CORE CORE (decrees) TOP MIDPLANE BOTTOM 0.O MM NN 00 15.0 PP 30.0 QQ 45.0 RR The contents of each of these irradiation capsules are specified in Appendix E to this report. The irradiation history of the Point Beach Unit 2 reactor during Cycles 18/20 is also listed in Appendix E. The irradiation ; history was obtained from NUREG-0020, " Licensed Operating ! Reactors Status Sut* mary Report" for the applicable operating period. Based on tnis reactor operating history, the individual sensor characteristics, and the measured specific activities I given in Appendix E, cycle average reaction rates referenced to a l core power level of 1518 MWt were computed for each multiple foil l sensor and gradient chain segment. l The computed reaction rates for the multiple foil sensor sets, including radiometric foils and solid state track recorders, irradiated during Cycles 18/20 are provided in Table 6.4-1. Corresponding reaction rate data from the four stainless steel gradient chains are recorded in Tables 6.4-2 through 6.4-4 for the Fe-54(n,p), Ni-58(n,p), and Co-59 (n,y) reactions, l respectively. l In regard to the data listed in Table 6.4-1, the Fe-54(n,p) reaction rates represent an average of the bare and cadmium 6-62 , I
1 1 I 1 covered measurements for each capsule. Likewise, the U-238 (n, f) reaction rates were obtained by averaging the results of the radiometric foil and solid state track recorder data. In addition, the fission rate measurements include corrections for U-235 impurities in the U-238 sensors as well as corrections for photo-fission reactions in both the U-238 and Np-237 sensors. 6.4.2 - Results of the Least Squares Adjustment Procedure l The results of the application of the least squares adjustment procedure to the six sets of multiple foil measurements obtained l from the Cycles 18/20 irradiation are provided in Tables 6.4-5 ) through 6.4-10. In these tables, the derived exposure i l experienced at each sensor set location along with data illustrating the fit of both the trial and adjusted spectra to the measurements are given. Also included in the tabulations are the la uncertainties associated with each of the derived exposure ! rates. j In regard to the comparisons listed in Tables 6.4-5 through 6.4-10, it should be noted that the columns labeled " trial calc" ; were obtained by normalizing the neutron spectral data from Table 4.1-1 to the absolute calculated neutron flux (E > 1.0 MeV) averaged over the Cycles 18/20 irradiation period as discussed in [Section 3.0. Thus, the comparisons illustrated in Tables 6.4-5 through 6.4-10 indicate the degree to which the calculated neutron energy spectra matched the measured data before and after adjustment. Absolute comparisons are discussed further in Saction 7.0 of this report. Complete traverses of fast neutron exposure rates in the reactor cavity were developed by combining the results of the least squares adjustment of the multiple foil data with the Fe-54(n,p) reaction rate measurements from the gradient chains. The gradient data were employed to establish relative axial distributions over the measurement range and these relative i distributions were then normalized to the FERRET results from the midplane sensor sets to produce axial distributions of exposure rates in terms of $(E > 1.0 MeV) , $(E > 0.1 MeV), and dpa/sec in the reactor cavity. 6-63
The resultant axial distributions of ((E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec from the gradient chain measurements are given in Tables 6.4-11, 6.4-12, and 6.4-13, respectively. The distributions of $(E > 1.0 MeV) are depicted graphically in Figures 6.4-1 through 6.4-4. In these graphical presentations, results for axial locations of -6.0, 0.0, and +6.0 feet relative to the core midplane represent the explicit results of the FERRET evaluations summarized in Tables 6.4-5 through 6.4-10, while results at the remaining axial locations depict the normalized data from the gradient chains. 6-64
TABLE 6.4
SUMMARY
OF REACTION RATES DERIVED FROM MULTIPLE FOIL SENSOR SETS CYCLES 18/20 IRRADIATION REACTION RATE fros/ nucleus) CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE NN PP OO RR MM OO Cu- 63 (n, a) Cd 8.97E-19 7.95E-19 7.09E-19 6.97E-19 3.41E-19 3.14E-19. Ti-46(n,p) Cd 1.37E-17 1.16E-17 1.04E-17 9.71E-18 5.56E-18 5.49E-18 Fe-54(n,p) Cd 7.56E-17 6.47E-17 5.57E-17 5.25E-17 2.85E-17 3.08E-17 Ni-58 (n,p) Cd 1.10E-16 9.34E-17 7.99E-17 7.57E-17 4.67E-17 4.53E-17 U-238(n,f) Cd 4.59E-16 4.16E-16 3.37E-16 3.15E-16 1.86E-16 1.80E-16 Np-237(n,f) Cd 5.76E-15 5.57E-15 4.73E-15 4.14E-15 2.80E-15 2.21E-15 Co- 59 (n, y) 8.74E-14 1.09E-13 9.75E-14 6.19E-14 3.47E-14 _3.66E-14 Co-59(n,7) Cd 5.34E-14 6.29E-14 5.43F-14 4.37E-14 2.50E-14 2.45E-14 U-235 (n, f) 8.74E-13 1.18E-12 1.10E-12 5.36E-13 2.70E-13 3.03E-13' U-235(n,f) Cd 2.19E-13 3.04E-13 2.47E-13 1.88E-13 6.83E-14 5.94E-14 Note: Cd indicates that the sensor was cadmium covered. 6-65 L
i l l l TABLE 6.4-2 i Fe- 54 (n, p) REACTION RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS - CYCLES 18/20 IRRADIATION i FEET REACTION RATE (rps/ nucleus) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 1.58E-17 1.19E-17 3.17E-17 9.24E-18
+5.5 4.04E-17 3.28E-17 2.40E-17 2.31E-17
+4.5 6.90E-17 5.95E-17 4.11E-17 3.75E-17
+3.5 7.81E-17 6.07E-17 4.95E-17 3.988-17
+2.5 8.61E-17 7.02E-17 4.91E-17 4.81E-17
+1.5 8.50E-17 6.41E-17 4.72E-17 4.66E-17
+0.5 7.51E-17 6.34E-17 5.10E-17 4.77E-17
-0.5 6.21E-17 5.53E-17 5.06E-17 4.85E-17
-1.5 5.91E-17 5.46E-17 5.33E-17 4.58E-17
-2.5 6.52E-17 6.30E-17 4.76E-17 4.51E-17
-3.5 7.16E-17 6.34E-17 5.10E-17 4.43E-17
-4.5 6.44E-17 4.96E-17 3.63E-17 3.60E-17
-5.5 4.04E-17 3.24E-17 2.54E-17 2.31E-17
-6.5 1.33E-17 1.18E-17 9.29E-18 8.83E-18 i
)
6-66 l l
I l TABLE 6.4-3 Ni-58(n,p) REACTION RATES-DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS - CYCLES 18/20 IRRADIATION I
-I.
FEET REACTION RATE (rps/ nucleus) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 2.25E-17 1.84E-17 4.17E-17 1.35E-17
+5.5 6.19E-17 4.77E 3.60E-17 3.13E-17
+4.5 9.87E-17 8.03E-17 5.95E-17 5.22E-17
+3.5 1.21E-16 9.50E-17 7.25E-17 6.16E-17
+2.5 1.28E-16 9.82E-17 7.25E-17 7.15E-17
+1.5 1 23E 9.56E-17 7.62E-17 6.22E-17
+0.5 1.13E-16 8.67E-17 7.75E-17 6.12E-17
-0.5 9.56E-17 8.86E-17 7.50E-17 7.03E-17
-1.5 8.41E-17 7.91E-17 7.50E-17 6.28E-17
-2.5 9.62E-17 8.35E-17 6.68E-17 7.03E-17
-3.5 1.07E-16 9.56E-17 6.06E-17 6.40E-17
-4.5 9.37E-17 7.65E-17 5.35E-17 5.43E-17
-5.5 6.07E-17 5.02E-17 3.33E-17 3.40E-17
-6.5 2.13E-17 2.05E-17 1.39E-17 1.42E-17 6-67
TABLE 6.4-4 Co- 59 (n, y) REACTION RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS - CYCLES 18/20 IRRADIATION FEET REACTION RATE (rps/ nucleus) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 1.87E-14 1.95E-14 5.13E-14 1.39E-14 ,
+5.5 4.25E-14 6.26E-14 4.56E-14 2.96E-14
+4.5 5.98E-14 9.34E-14 6.75E-14 3.99E-14 ;
+3.5 6.95E-14 1.09E-13 8.23E-14 4.80E-14
+2.5 7.76E-14 1.14E-13 9.04E-14 5.49E-14
+1.5 7.80E-14 1.11E-13 9.16E-14 5.77E-14
+0.5 8.00E-14 1.04E-13 9.12E-14 5.73E-14
-0.5 8.02E-14 9.89E-14 8.80E-14 5.75E-14
-1.5 7.74E-14 9.34E-14 8.66E-14 5.67E-14
-2.5 7.62E-14 9.16E-14 8.29E-14 5.45E-14
-3.5 7.27E-14 8.82E-14 7.42E-14 4.92E-14
-4.5 5.94E-14 7.28E-14 5.76E-14 3.99E-14
-5.5 3.83E-14 5.07E-14 3.08E-14 2.75E-14
-6.5 2.88E-14 2.66E-14 2.11E-14 2.05E-14 e
i 4 J 6-68
1 TABLE 6.4-5 DERIVED EXPOSURE RATES FROM THE CAPSULE NN DOSIMETRY EVALUATION 0.0 DEGREE AZIMUTH - CORE MIDPLANE TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 1.78E+09 1.51E+09 8%
4(E > 0.1 MeV) 1.40E+10 1.20E+10 16%
$(E < 0.414 eV) 3.34E+09 1.58E+09 24%
dpa/sec 5.02E-12 4.29E-12 13% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0.0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 6 3 (n, ot) Cd 8.97E-19 8.82E-19 8.92E-19 1.02 1.01 Ti-46(n,p) Cd 1.37E-17 1.32E-17 1.34E-17 1.04 1.02 Fe- 54 (n, p) Cd 7.56E-17 8.57E-17 7.76E-17 0.88 0.97 Ni-58(n,p) Cd 1.10E-16 1.23E-16 1.11E-17 0.89 0.99 - U-238(n,f) Cd 4.60E-16 5.09E-16 4.44E-16 0.90 1.04 Np-237(n,f) Cd 5.76E-15 6.92E-15 5.79E-15 0.83 1.00 Co- 59 (n, y) 8.74E-14 1.67E-13 9.20E-14 0.52 0.95 Co-59 (n, y) Cd 5.34E-14 7.85E-14 5.21E-14 0.68 1.03 U-235 (n, f) 8.74E-13 1.57E-12 7.96E-13 0.56 1.10 U-235 (n, f) Cd 2.19E-13 2.78E-13 2.23E-13 0.79 0.98 6-69
TABLE 6.4-6 DERIVED EXPOSURE RATES FROM THE CAPSULE PP DOSIMETRY EVALUATION 15.0 DEGREE AZIMUTH - CORE MIDPLANE TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 1.54E+09 1.36E+09 8%
$(E > 0.1 MeV) 1.33E+10 1.24E+10 17%
$(E < 0.414 eV) 3.07E+09 2.11E+09 24%
dpa/sec 4.67E-12 4.31E-12 13% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 15.0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, at) Cd 7.95E-19 7.42E-19 7.86E-19 1.07 1.01 Ti-46(n,p) Cd 1.16E-17 1.10E-17 1.14E-17 1.06 1.02 ; Fe-54(n,p) Cd 6.47E-17 7.20E-17 6.66E-17 0.90 0.97 l Ni- 5 8 (n, p) Cd 9.34E-17 1.04E-16 9.4?E-17 0.90 0.99 U-23 8 (n, f) Cd 4.16E-16 4.34E-16 3.93E-16 0.96 1.06 Np-237(n,f) Cd 5.57E-15 6.28E-15 5.62E-15 0.89 0.99 Co- 5 9 (n, y) 1.09E-13 1.61E-13 1.14E-13 0.68 0.96 l Co- 59 (n, y) Cd 6.29E-14 8.09E-14 6.17E-14 0.78 1.02 I U-235 (n, f) 1.18E-12 1.45E-12 1.07E-12 0.81 1.10 U-235 (n, f) Cd 3.04E-13 2.85E-13 3.02E-13 1.07 1.01 ) l I 1 6-70
TABLE 6.4-7 I DERIVED EXPOSURE RATES FROM THE CAPSULE QQ DOSIMETRY EVALUATION 30.0 DEGREE AZIMUTH - CORE MIDPLANE TRIAL ADJUSTED 1cr VALUE VALUE UNCERTAINTY 4(E > 1.0 MeV) 1.26E+09 1.13E+09 8%
$(E > 0.1 MeV) 1.10E+10 1.04E+10 - 17%
4(E < 0.414 eV) 2.68E+09 1.97E+09 23% dpa/sec 3.87E-12 3.62E-12 13% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 30.0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, cr) Cd 7.09E-19 6.53E-19 7.02E-19 1.09 1.01 Ti-46(n,p) Cd 1.04E-17 9.61E-18 1.02E-17 1.08 1.02 Fe-54(n,p) Cd 5.57E-17 6.10E-17 5.72E-17 0.91 0.97 Ni-58(n,p) Cd 7.99E-17 8.76E-17 8.09E-17 0.91 0.99 U-238(n,f) Cd 3.37E-16 3.59E-16 3.26E-16 0.94 1.03 Np-237(n,f) Cd 4.73E-15 5.16E-15 4.71E-15 0.92 1.00 i Co- 59 (n,y) 9.75E-14 1.41E-13 1.03E-13 0.69 0.95 i Co- 59 (n, y) Cd 5.43E-14 7.09E-14 5.31E-14 0.77 1.02 . U-235 (n, f) 1.10E-12 1.26E-12 9.76E-13 0.87 1.13 U-235 (n, f) Cd 2.47E-13 2.49E-13 2.48E-13 0.99 1.00 4 6-71
H TABLE 6.4-8 DERIVED EXPOSURE RATES FROM THE CAPSULE RR DOSIMETRY EVALUATION 45.0 DEGREE A?,IMUTH - CORE MIDPLANE TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY 4(E > 1.0 MeV) 1.13E+09 1.02E+09 8% 4(E > 0.1 MeV) 9.28E+09 8.74E+09 17% p(E < 0.414 eV) 2.65E+09 9.01E+08 27% dpa/sec 3.32E-12 3.09E-12 13% s 1 .) COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 45.0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C , MEASURED CALC. CALC. TRIAL ADJUSTED l Cu- 63 (n, a) Cd 6.97E-19 6.45E-19 6.86E-19 1.08 1.02 Ti-46(n,p) Cd 9.71E-18 9.35E-18 9.60E-18 1.04 1.01 l Fe-54(n,p) Cd 5.25E-17 5.76E-17 5.40E-17 0.91 0.97 Ni-58(n,p) Cd 7.57E-17 8.20E-17 7.64E-17 0.92 0.99 U-238(n,f) Cd 3.15E-16 3.27E-16 3.01E-16 0.96 1.05 Np-237(n,f) Cd 4.14E-15 4.48E-15 4.12E-15 0.92 1.01 Co-59 (n, y) 6.19E-14 1.30E-13 6.50E-14 0.48 0.95 Co- 59 (n, y) Cd 4.37E-14 5.99E-14 4.25E-14 0.73 1.03 U-235 (n, f) 5.36E-13 1.24E-12 5.02E-13 0.43 1.07 U-235 (n, f) Cd 1.88E-13 2.11E-13 1.89E-13 0.89 1.00 l l l 1 3 6-72 l l
TABLE 6.4-9 DERIVED EXPOSURE RATES FROM THE CAPSULE MM DOSIMETRY EVALUATION 0.0 DEGREE AZIMUTH - CORE TOP TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY
$(E > 1.0 MeV) 6.60E+08 6.50E+08 8%
$(E > 0.1 MeV) 5.18E+09 5.33E+09 16%
$(E < 0.414 eV) 1.24E+09 4.85E+08 25%
dpa/sec 1.86E-12 1.88E-12 13% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0.0 DEGREE AZIMUTH - CORE TOP REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C MEASURED -CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, cr) Cd 3.41E-19 3.26E-19 3.40E-19 1.05 1.00 l Ti-46(n,p) Cd 5.56E-18 4.88E-18 5.37E-18 1.14 1.04 Fe- 54 (n,p) Cd 2.85E-17 3.17E-17 3.03E-17 0.90 0.94 Ni-58(n,p) Cd 4.67E-17 4.56E-17 4.59E-17 1.02 1.02 l U-238 (n, f) Cd 1.86E-16 1.89E-16 1.84E-16 0.98 1.01 Np-237(n,f) Cd 2.80E-15 2.56E-15 2.68E-15 1.09 1.05 Co- 59 (n, y) 3.47E-14 6.18E-14 3.67E-14 0.56 0.95 Co-59 (n, y) Cd 2.50E-14 2.91E-14 2.40E-14 0.86 1.04 U-235(n,f) 2.70E-13 5.80E-13 2.50E-13 0.47 1.08 U-235(n,f) Cd 6.83E-14 1.03E-13 7.33E-14 0.66 0.93 6-73 I
- - . . ~ . . _. - ---
TABLE 6.4-10 DERIVED EXPOSURE RATES FROM THE CAPSULE 00 DOSIMETRY EVALUATION 4 0.0' DEGREE AZIMUTH - CORE BOTTOM 4 4 TRIAL ADJUSTED la VALUE VALUE UNCERTAINTY p(E > 1.0 MeV) 7.13E+08 6.07E+08 8% p(E > 0.1 MeV) 5.60E+09 4.32E+09 16%
- p(E < 0.414 eV) 1.34E+09 5.64E+08 23%
'dpa/sec 2.01E-12 1.59E-12 12%
f COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0.0 DEGREE AZIMUTH - CORE MIDPLANE l REACTION RATE (rps/ nucleus) TRIAL ADJUSTED M/C M/C ! MEASURED CALC. CALC. TRIAL ADJUSTED Cu- 63 (n, at) Cd 3.14E-19 3.53E-19 3.19E-19 0.89 0.98
- .Ti -46(n,p) Cd 5.49E-18 5.28E-18 5.32E-18 1.04 1.03 Fe-54(n,p) Cd 3.08E-17 3.43E-17 3.13E-17 0.90 0.98 i
- Ni-58(n,p) Cd 4.53E-17 4.93E-17 4.51E-17 0.92 1.00 ,
U-238(n,f) Cd 1.80E-16 2.04E-16 1.79E-16 0.88 1.01 ! Np-237(n,f) Cd 2.21E-15 2.77E-15 2.22E-15 0.80 1.00 Co-59(n,7) 3.66E-14 6.68E-14 3.87E-14 0.55 0.95 Co- 59 (n, y) Cd 2.45E-14 3.14E-14 2.36E-14 0.78 1.04 U-235 (n, f) 3.03E-13 6.27E-13 2.78E-13 0.48 1.09 U-235(n,f) Cd 5.94E-14 1.11E-13 6.44E-14 0.54 0.92 l l 4 1 6-74 l l
TABLE 6.4-11 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION
; OF AXIAL POSITION WITHIN THE REACTOR CAVITY ; CYCLES 18/20 IRRADIATION
. FEET NEUTRON FLUX (n/cm2 -sec) FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG l +6.5 3.48E+08 2.72E+08 1.96E+08
+5.5 8.89E+08 7.52E+08 5.33E+08 4.89E+08
+4.5 1.52E+09 1.36E+09 9.14E+08 7.95E+08
+3.5 1.72E+09 1.39E+09 1.10E+09 8.43E+08
+2.5 1.90E+09 1.61E+09 1.09E+09 1.02E+09
+1.5 1.87E+09 1.47E+09 1.05E+09 9.88E+08 l +0.5 1.65E+09 1.45E+09 1.13E+09 1.01E+09
-0.5 1.37E+09 1.27E+09 1;13E+09 1.03E+09
-1.5 1.30E+09 1.25E+09 1.19E+09 9.72E+08
-2.5 1.43E+09 1.44E+09 1.06E+09 9.56E+08 ,
-3.5 1.58E+09 1.45E+09 1.13E+09 9.40E+08
-4.5 1.42E+09 1.14E+09 8.08E+08 7.63E+08
] -5.5 8.89E+08 7.43E+08 5.64E+08 4.90E+08
-6.5 2.93E+08 2.70E+08 2.07E+08 1.87E+08 R
.I i 6-75 4-J
TABLE 6.4-12 FAST NEUTRON FLUX (E > 0.1 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLES 18/20 IRRADIATION FEET NEUTRON FLUX (n/cm2 -sec) - FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 2.77E+09 2.48E+09 1.68E+09
+5.5 7.07E+09 6.86E+09 4.90E+09 4.19E+09
+4.5 1.21E+10 '1.24E+10 8.41E+09 6.81E+09
+3.5 1.37E+10 1.27E+10 1.01E+10 7.23E+09
+2.5 1.51E+10 1.47E+10 1.01E+10 8.74E+09
+1.5 1.49E+10 1.34E+10 9.66E+09 8.46E+09
+0.5 1.31E+10 1.32E+10 1.04E+10 8.67E+09
-0.5 1.09E+10 1.16E+10 1.04E+10 8.81E+09
-1.5 1.03E+10 1.14E+10 1.09E+10 8.33E+09
-2.5 1.14E+10 1.32E+10 9.74E+09 8.19E+09
-3.5 1.25E+10 1.32E+10 1.04E+10 8.05E+09
-4.5 1.13E+10 1.04E+10 7.43E+09 6.54E+09
-5.5 7.07E+09 6.78E+09 5.19E+09 4.20E+09
-6.5 2.33E+09 2.46E+09 1.90E+09 1.60E+09
/ 6-76
Y l TABLE 6.4-13 IRON DISPLACEMENT RATE AS A FUNCTION
'OF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLES'18/20 IRRADIATION FEET DISPLACEMENT RATE (dpa/sec)
FROM MIDPLANE 0.0 DEG 15.0 DEG 30.0 DEG 45.0 DEG
+6.5 9.89E-13 8.62E-13 5.94E-13
+5.5 2.53E-12 2.38E-12 1.71E-12 1.48E-12
+4.5 4.31E-12 4.32E-12 2.93E-12 2.41E-12
+3.5 4.89E-12 4.41E-12 3.53E-12 2.55E-12
+2.5 5.39E-12 '5.10E-12 3.50E-12 3.09E-12
+1.5 5.31E-12 4.66E-12 3.36E-12 2.99E-12
+0.5 4.70E-12 4.60E-12 3.63E-12 3.07E-12 ]
-0.5 3.88E-12 4.02E-12 3.61E-12 3.11E-12
-1.5 3.69E-12 3.96E-12 3.80E-12 2.94E-12
-2.5 -4.08E-12 4.57E-12 3.39E-12 2.90E-12
-3.5 4.48E-12 4.60E-12 3.63E-12 2.85E-12
-4.5 4.03E-12 3.60E-12 2.59E-12 2.31E-12
-5.5 2.53E-12 2.36E-12 1.81E-12 1.48E-12
-6.5 8.32E-13 8.'56E-13 6.62E-13 5.67E-13 l
1 l J J J 4 6-77 l l 4-
-- - -c . - - - - - - - - - - - _ - - - _ - - _
FIGURE 6.4-1 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 0.0 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLES 18/20 IRRADIATION J 1.0E+10 l Ss r ~ 81.0E m
+09 .,
z m a o . g J i o \ h I
~ h l
5 E , 8 5 ze 1.0E+08 1 1.0E+07
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) 4 I
6-78 l
FIGURE 6.4-2 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 15.0 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLES 18/20 IRRADIATION 1.0E +10 n a M% E 1.0E+09 ,
- 9 .
i. m ' \ E f 1 o h A L T / \. 2
- u.
8 s y 1.0E+08 1.0E+07
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) 6-79
l FIGURE 6.4-3 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 30.0 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLES 18/20 IRRADIATION . l 1 l i 1.0E+10 A N
$w 1.0E+09 -
s ,/ \
$ a
~
h I
)
E i C 1 8 zE 1.0E+08 i 1.0E+07
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) 6-80
FIGURE 6.4-4 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 45.0 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLES 18/20 IRRADIATION 1.0E+10 1 81.0E+09 __ -
; m
? l
^
I \
/ \
w I
- r T,
4 x , 4 2
/ \ -
O h z51.0E+08 l 4
- 1.0E +0 7
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) 6-81
SECTION 7.0 COMPARISON OF CALCULATIONS WITH MEASUREMENTS As described in Section 3.3, the best estimate neutron exposure proj ections for the - Point Beach Unit 2 -pressure _ vessel' were based on a combination of plant specific neutron transport calculations and plant specific measurements. Direct comparisons of the transport calculations with the Point Beach Unit 2 measurement data base were used to quantify the biases that may exist due to the transport methodology, reactor modeling, and/or reactor operating characteristics over the respective irradiation periods. In this section, comparisons of the measurement results from surveillance capsule and reactor cavity dosimetry with corresponding analytical predictions at the measurement locations are presented. These comparisons are provided on two levels. In the first instance, predictions of fast neutron exposure rates in terms of $(E > 1.0 MeV), f(E > 0.1 MeV), and dpa/sec are compared with the results of the FERRET least squares adjustment procedure; while, in the second case, calculations of individual sensor reaction rates are compared directly with the measured data from the counting laboratories. It is shown that these two levels of' comparison yield consistent and similar results, indicating that the least squares adjustment methodology is producing accurate exposure results and that the measurement / calculation (M/C) comparisons yield an accurate plant specific bias factor that can be applied to neutron transport calculations performed for the Point Beach Unit 2 reactor to produce "best estimate" exposure projections for the pressure vessel wall. 7.1 Comparison of Least Squares Adjustment Results with Calculation In Table 7.1-1, comparisons of measured and calculated exposure rates for the four surveillance capsule dosimetry sets withdrawn to date as well as for the four cycles of reactor cavity midplane dosimetry sets irradiated during Cycles 15, 16, 17, and 18/20 are 7-1
given. In all cases, the calculated values were based on the fuel cycle specific exposure calculations averaged over the appropriate irradiation period. An examination of Table 7.1-1 indicates that, considering all of the available core midplane data, the measured exposure rates were less than calculated values by factors of 0.921, 0.975, and 0.959 for ((E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec, respectively. The standard deviations associated with each of the 20 sample data sets were 0.074 (8.0%), 0.161 (16.5%), and 0.120 (12.5%), respectively. 7.2 Comparisons of Measured and Calculated Sensor Reaction Rates In Table 7.2-1, measurement / calculation (M/C) ratios for each fast neutron sensor reaction rate from the surveillance capsule and reactor cavity irradiations are listed. This tabulation, provides a direct comparison, on an absolute basis, of calculation and measurement prior to the application of the least squares adjustment procedure as represented in the FERRET evaluations. An examination of Table 7.2-1 shows consistent behavior for all reactions and all measurement points. The overall average M/C ratio for the entire data set has an associated la standard deviation of 0.081 (8.3%). Furthermore, the average M/C bias of 0.974 observed in the reaction rate comparisons is in excellent agreement with the values of 0.921, 0.975, and 0.959 observed in the exposure rate comparisons shown in Table 7.1-1. l a I l l I 7-2
TABLE 7.1-1 COMPARISON OF MEASURED AND CALCULATED EXPOSURE RATES FROM SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY.. IRRADIATIONS
$(E > 1.0 MeV) [n/cm2 -sec]
CALCULATED MEASURED M/_C INTERNAL CAPSULES V (13 DEGREES) 1.35E+11 1.44E+11 1.067 T (23 DEGREES) 7.91E+10 8.21E+10 1.037 R (13 DEGREES) 1.34E+11 1.39E+11 1.031 S (33 DEGREES) 6.65E+10 6.81E+10 1.025 0 DEGREE CAYI E Cycle 15 2.22E+09 1.95E+09 0.877 Cycle 16 1.80E+09 1.53E+09 0.847 Cycle 17 1.81E+09 1.52E+09 0.841 Cycle 18/20 1.78E+09 1.51E+09 0.848 15 DEGREE CAVITY Cycle 15 1.89E+09 1.78E+09 0.942 Cycle 16 1.54E+09 1.35E+09 0.875 j Cycle 17 1.55E+09 1.34E+09 0.866 ; Cycle 18/20 1.54E+09 1.36E+09 0.885
.30 DEGREE CAVITY Cycle 15 1.42E+09 1.31E+09 0.924 Cycle 16 1.21E+09 1.06E+09 0.873 Cycle 17 1.24E+09 1.10E+09 0.886 Cycle 18/20 1.26E+09 1.13E+09 0.890 45 DEGREE CAVITY Cycle 15 1.17E+09 1.12E+09 0.959 Cycle 16 1.06E+09 1.03E+09 0.967 Cycle 17 1.10E+09 9.66E+08 0.878 Cycle'18/20 1.13E+09 1.02E+09 0.906 AVERAGE M/C BIAS FACTOR (K) 0.921 STANDARD DEVIATION (la) 0.074 7-3
TABLE 7.1-1 (continued) COMPARISON OF MEASURED AND CALCULATED EXPOSURE RATES FROM SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS
$(E > 0.1 MeV) [n/cm -sec)
CALCULATED MEASURED M/_.C l INTERNAL CAPSULES V (13 DEGREES) 5.12E+11 5.71E+11 1.114 ' T (23 DEGREES) 2.72E+11 2'.97E+11 1.091 R (13 DEGREES) 5 09E+11 5.61E+11 1.101 S (33 DEGREES) 2.35E+11 2.54E+11 1.084 0 DEGREE CAVITY > Cycle 15 1.74E+10 1.62E+10 0.928 Cycle 16 1.41E+10 1.31E+10 0.925 Cycle 17 1.42E+10 1.32E+10 0.928 l k Cycle 18/20 1.40E+10 1.20E+10 0.854 I f 15 DEGREE CAVITY l Cycle 15 1.63E+10 1.66E+10 1.020 l Cycle 16 1.33E+10 1.24E+10 0.934 i Cycle 17 1.34E+10. 1.18E+10 0.884 4 Cycle 18/20 1.33E+10 1.24E+10 0.934 i ! 30 DEGREE CAVITY Cycle 15 1.24E+10 1.21E+10 0.977 Cycle 16 1.05E+10 9.77E+09 0.928 1
- Cycle 17 1.08E+10 1.02E+10 0.941 Cycle 18/20 1.10E+10 1.04E+10 0.946 l
45 DEGREE CAVITY Cycle 15 9.60E+09 9.99E+09 1.040 Cycle 16 8.70E+09 9.06E+09 1.041 Cycle 17 9.03E+09 9.09E+09 0.896 8.74E+09 Cycle 18/20 9.28E+09 0.942 , AVERAGE M/C BIAS FACTOR (K) 0.975 , STANDARD DEVIATION (1ct) 0.161 i t 1 7-4 i
_ _ _ _ . . _ . . _ . . _ . _ _ . ~ _ . _. .. 4 TABLE 7.1-1 (continued) 4 i COMPARISON OF MEASURED AND CALCULATED EXPOSURE RATES FROM SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS Iron Displacements (dpa-sec] i CALCULATED MEASURED M/_C, INTERNAL CAPSULES V (13 DEGREES) 2.41E-10 2.60E-10 1.081 T (23 DEGREES) 1.35E-10 1.43E-10. 1.060 R (13 DEGREES) 2.39E-10 2.55E-10 1.065 S (33 DEGREES) 1.14E-10 1.20E-10 1.052 i 0 DEGREE CAVITY Cycle 15 6.26E-12 5.72E-12 0.914 ! Cycle 16 5.08E-12 4.59E-12 0.904 Cycle 17 5.11E-12 4.62E-12 0.906 Cycle 18/20 5.02E-12 4.29E-12 0.854 i 15 DEGREE CAVITY Cycle 15 5.73E-12 5.74E-12 1.002 Cycle 16 4.67E-12 4.29E-12 0.920 Cycle 17 4.70E-12 4.15E-12 0.883 Cycle 18/20 4.67E-12 4.31E-12 0.924 l 30 DEGREE CAVITY I Cycle 15 4.35E-12 4.20E-12 0.965 Cycle 16 3.71E-12 3.40E-12 0.918 i t Cycle 17 3.80E-12 3.53E-12 0.931 ) l l Cycle 18/20 3.87E-12 3.62E-12 0.935 45 DEGREE CAVITY
- Cycle 15 3.43E-12 3.49E-12 1.015 Cycle 16 3.11E-12 3.17E-12 1.018 Cycle 17 3.23E-12 2.88E-12 0.893 4 Cycle 18/20 3.32E-12 3.09E-12 0.931 )
l
, . AVERAGE M/C BIAS FACTOR (K) 0.959 STANDARD DEVIATION (la) 0,120 -
7-5 ]
TABLE 7.2-1 COMPARISON OF MEASURED AND CALCULATED NEUTRON SENSOR REACTION RATES FROM SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS Cu63 (n. or) Ti46(n.o) Fe54(n.0) NiS8(n.0) U238 (n. f) NO237 (n f) INTERNAL CAPSULES V (13 DEGREES) 0.910 0.923 1.108 1.125 T (23 DEGREES) 0.947 0.984 0.950 1.007 1.129 R (13 DEGREES) 0.987 0.937 0.957 1.044 1.117
-S (33 DEGREES) 1.070 0.883 1.113 1.089 0 DEGREE CAVITY Cycle 15 0.973 1.012 0.887 0.896 0.865 0.930 Cycle 16 0.964 0.977 0.863 0.863 0.823 0.923 Cycle 17 1.009 1.000 .0.898 0.880 0.768 0.949 Cycle 18/20 1.017 1.038 0.882 0.894 0.904 0.832 15 DEGREE CAVITY Cycle 15 1.081 1.118 0.965 0.953 0.942 1.016 Cycle 16 1.011 1.018 0.882 0.889 0.908 0.901 Cycle 17 1.072 1.063 0.942 0.915 0.899 0.853 Cycle 18/20 1.071 1.055 0.899 0.898 0.959 0.887 30 DEGREE CAVITY Cycle 15 1.015 1.056 0.961 0.917 0.926 0.966 Cycle 16 1.074 1.066 0.947 0.936 0.834 0.929 Cycle 17 1.059 1.083 0.945 0.932 0.875 0.927 Cycle 18/20 1.086 1.082 0.913 0.912 0.939 0.917 45 DEGREE CAVITY Cycle 15 1.069 1.116 0.926 0.949 0.976 1.052
~ Cycle 16 1.093 1.067 0.972 0.967 0.945 1.067 Cycle 17 1.110 1.099 0.970 0.959 '
O.874 0.893 Cycle 18/20 1.081 1.039 0.911 0.923 0.963 0.924 AVERAGE 1.034 1.055 0.925 0.921 0.933 0.972 ST. DEV. (la) 0.055 0.042 0.032 0.033 0.086 0.091 OVERALL AVERAGE M/C RATIO 0.974 STANDARD DEVIATION (la) 0.081 7-6
SECTION 8.0 BEST ESTIMATE NEUTRON EXPOSURE OF PRESSURE VESSEL MATERIALS In this section the measurement results provided in Sections 5.0 and 6.0 are combined with the results of the neutron transport calculations described in Section 4.0 to establish a mapping of the best estimate neutron exposure of the beltline region of the Point Beach Unit 2 reactor pressure vessel through the completion of Cycle 20. Based on the continued use of the Cycles 16-20 fuel loading patterns incorporating part length hafnium absorbers, projections of future vessel exposure to 32 and 48 effective full power years of operation are also provided. In addition to the spatial mapping over the beltline region, data pertinent to the maximum exposure experienced by the upper, intermediate, and lower shell forgings as well as the beltline circumferential welds are highlighted. 8.1 Exposure Distributions Within the Beltline Region As' described in Section 3.3 of this report, the best estimate j vessel exposure was determined from the following relationship: l I em-K4%, 2 where: e s.a u. - The best estimate fast neutron exposure at _ the location of interest. 3 K = The plant specific measurement / calculation l (M/C) bias factor derived from all available I surveillance capsule and reactor cavity dosimetry data. 4c u. - The absolute calculated fast neutron exposure at the location of interest. 8-1
From the data provided in Table 7.1-1, the plant specific bias factors (K) to be applied to the calculated exposure values given 1 in Section 4.2 were as follows: 4(E > 1.0 MeV) 0.921 A 0.074 (8.0%) 4(E > 0.1 MeV) 0.975 0.161 (16.5%) dpa 0.959 0.120 (12.5%) These bias factors were based on the results of the continuous monitoring program at Point Beach Unit 2 that has provide measured data from four internal surveillance capsules and sixteen reactor cavity sensor sets through the first 18.2 effective full power years of operation. The uncertainties listed with the individual bias factors are at the la level and are given on an absolute and percentage basis. Additional uncertainties associated with the evaluation of the l best estimate vessel exposure are discussed in Section 8.2. 8.1.1 Exposure Accrued During Cycles 1 through 20 To assess the incremental exposure resulting from irradiation during Cycles 1 through 20, the bias factors listed in Section 8.1 were applied directly to the calculated values from Section i 4.2 for the vessel clad / base metal interface to produce best estimate fluence levels characteristic of the midplane of the reactor core. The axial gradient chain measurements taken before and after implementation of part length hafnium were then employed to develop the complete axial traverse along the vessel l wall. The best estimate results applicable to the vessel inner surface are incorporated into Tables 8.1-1 through 8.1-12 to establish the exposure accrued by the reactor vessel through the end of Cycles 14, 15 and 20, respectively. Exposure distributions through the vessel wall, can be developed using these surface exposures and radial distribution functions , from Section 4.0. This exposure information, applicable through the end of Cycle 20, was derived from an extensive set of measurements and assures that embrittlement gradients can be established with a minimum uncertainty. Further, as the monitoring program continues and additional data become 8-2
1 I available, the overall plant specific data base for Point Beach i Unit 2 will expand resulting in reduced uncertainties and an improved accuracy in the assessment of vessel condition. 8.1.2 Projection of Future Vessel Exposure At the end of Cycle 20, the Point Beach Unit 2 reactor had accrued 18.2 effective full power years (EFPY) of operation. In order to establish a framework for the assessment of future vessel condition, exposure projections to 32 and 48 EFPY are also included in Tables 8.1-1 through 8.1-12 in addition to the plant specific exposure assessments through the end of Cycle 20. These extrapolations into the future were based on the assumption that the data averaged over the Cycles 16 through 20 irradiations were representative of all future fuel cycles. That_is, that future fuel designs would incorporate the low leakage fuel management concept employed during Cycles 16 through 20. Examination of these projected exposure levels establishes the long term effectiveness of the low leakage fuel management incorporated to date and can be used as a guide in assessing strategies for future vessel exposure management. The validity of these projections for future operation will be confirmed via the continued cavity monitoring program. l 8-3
TABLE 8.1-1 i
SUMMARY
OF BEST ESTIMATE FAST NEUTRON (E > 1.0 MeV)' EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF THE POINT BEACH UNIT 2 REACTOR PRESSURE VHF3EL l 0 DEGREE AZIMUTHAL ANGLE 4(E > 1.0 MeV) In/cmj2 Zfft) EOC 14 EOC 15 EOC 20 32 EFPY 48 EFPY j +6.5 3.15E+18 3.29E+18 4.02E+18 6.40E+18 9.15E+18
+5.5 7.67E+18 8.02E+18 9.84E+18 1.58E+19 2.26E+19
+4.5 1.12E+19 1.17E+19 1.46B+19 2.40E+19 3.49E+19 l +3.5 1.32E+19 1.38E+19 1.70E+19 2.75E+19 3.97E+19 j +2.5 1.47E+19 1.53E+19 1.88E+19 3.01E+19 4.32E+19
+1.5 1.45E+19 1.51E+19 1.85E+19 2.94E+19 4.20E+19
+0.5 1.41E+19 1.47E+19 1.77E+19 2.76E+19 3.90E+19 0.0 1.44E+19 1.50E+19 1.78E+19 2.68E+19 3.72E+19
-0.5 1.47E+19 1.53E+19 1.79E+19 2.60E+19 3.55E+19
-1.5 1.35E+19 1.41E+19 1.65E+19 2.45E+19 3.37E+19
-2.5 1.29E+19 1.34E+19 1.61E+19 2.46E+19 3.44E+19
-3.5 1.27E+19 1.32E+19 1.61E+19 2.52E+19 3.59E+19
-4.5 1.13E+19 1.18E+19 1.45E+19 2.33E+19 3.35E+19
-5.5 7.64E+18 7.99E+18 9.80E+18 1.57E+19 2.25E+19
-6.5 2.81E+18 2.94E+18 3.62E+18 5.85E+18 8.42E+18 Note: Axial location is provided relative to the axial midplane of the re. actor core 8-4
TABLE 8.1-2 -
SUMMARY
OF BEST ESTIMATE FAST NEUTRON (E > 1.0 MeV) EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF THE POINT BEACH UNIT 2 REACTOR PRESSURE VESSEL 15 DEGREE AZIMUTHAL ANGLE 4(E > 1.0 MeV) In/cmj2 Z(ft) EOC 14 EOC 15 EOC 20 32 EFPY 48 EFPY
+6.5 1.76E+18 1.85E+18 2.31E+18 3.79E+18 5.52E+18 +5.5 4.87E+18 5.11E+18 6.30E+18 1.01E+19 1.46E+19 +4.5 7.40E+18 7.78E+18 9.66E+18 1.58E+19 2.28E+19 +3.5 8.21E+18 8.63E+18 1.08E+19 1.77E+19 2.57E+19 +2.5 8.79E+18 9.24E+18 1.16E+19 1.91E+19 2.78E+19 +1.5 8.72E+18 9.16E+18 1.14E+19 1.85E+19 2.67E+19 +0.5 8.79E+18 9.24E+18 1.13E+19 1.80E+19 2.58E+19 0.0 9.09E+18 9.55E+18 1.15E+19 1.79E+19 2.54E+19 -0.5 9.38E+18 9.86E+18 1.17E+19 1.78E+19 2.49E+19 -1.5 9.08E+18 9.55E+18 1.14E+19 1.72E+19 2.40E+19 -2.5 9.01E+18 9.47E+18 1.14E+19 1.78E+19 2.52E+19- -3.5 8.43E+18 8.86E+18 1.09E+19 1.76E+19 2.54E+19 -4.5 7.69E+18 8.09E+18 9.97E+18 1.61E+19 2.32E+19 -5.5 4.94E+18 5.19E+18 6.41E+18 1.04E+19 1.50E+19 -6.5 1.89E+18 1.99E+18 2.46E+18 3.99E+18 5.76E+18 Note: Axial location is provided relative to the axial midplane of the reactor core 8-5
, - - ~ . -
.~ ._ . - . - - - , . _ . . .- - c TABLE 8.1-3 ,
SUB91ARY OF BEST ESTIMATE FAST NEUTRON (E > 1.0 MeV) EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF THE POINT DEACH UNIT 2~ REACTOR PRESSURE VESSEL 30 DEGREE AZIMUTHAL ANGLE 4(E > 1.0 MeV) In/cmj2 Z(ft) EOC 14 EOC 15 EOC 20 32 EFPY 48 EFPY
+6.5 1.50E+18 1.58E+18 1.95E+18 3.15E+18 4.55E+18
+5.5 3.60E+18 3.79E+18 4.68E+18 7.56E+18 1.09E+19
+4.5 5.56E+18 5.85E+18 7.23E+18 1.17E+19 1.69E+19
+3.5 6.29E+18 6.62E+18 8.17E+18 1.32E+19 1.91E+19
+2.5 6.43E+18 6.77E+18 8.36E+18 1.35E+19 1.95E+19
+1.5 6.25E+18 6.59E+18 8.13E+18 1.31E+19 1.90E+19
+0.5 6.57E+18 6.92E+18 8.55E+18 1.38E+19 1.99E+19 0.0 6.59E+18 6.94E+18 8.57E+18 1.39E+19 2.00E+19
-0.5 6.61E+18 6.97E+18 8.60E+18 1.39E+19 2.01E+19
-1.5 6.50E+18 6.85E+18 8.45E+18 1.37E+19 1.97E+19
-2.5 6.48E+18 6.83E+18 8.43E+18 1.36E+19 1.96E+19
-3.5 6.09E+18 6.42E+18 7.92E+18 1.28E+19 1.85E+19
-4.5 5.25E+18 5.53E+18 6.83E+18 1.10E+19 1.59E+19
-5.5 3.40E+18 3.58E+18 4.42E+18 7.15E+18 1.03E+19
-6.5 1.38B+18 1.45E+18 1.79E+18 2.90E+18 4.18E+18 Note: Axial location is provided relative to the axial midplane of the reactor core 8-6 4
. TABLE 8.1-4 _
SUlW4ARY OF BEST ESTIMATE FAST NEUTRON '(E'> 1.0 MeV) EXPOSURE PROJECTIONS _ FOR THE BELTLINE REGION OF THE POINT. BEACH UNIT 2 REACTOR PRESSURE VESSEL ~ 45 DEGREE AZIMUTHAL ANGLE 4(E >.1.0 MeV) In/cmj2 Z(ft) EOC 14 EOC 15 EOC 20 32 EFPY 48 EFPY
+6.5 1.24E+18 1.30E+18 1.62E+18 2.67E+18 3.88E+18 +5.5 2.94E+18 3.08E+18 3.85E+18 6.33E+18 9.21E+18 +4.5 4.66E+18 4.89E+18 6.10E+18 1.00E+19 1.46E+19 +3.5 5.10E+18 5.35E+18 6.67E+18 1.10E+19 1.60E+19 +2.5 5.79E+18 6.07E+18 7.58E+18 1.25E+19 1.81E+19 +1.5 5.56E+18 5.84E+18 7.29E+18 1.20E+19 1.74E+19 , +0.5 5.57E+18 5.84E+18 7.29E+18 1.20E+19 1.74E+19 0.0 5.69E+18 5.96E+18 7.45E+18 1.23E+19 1.78E+19 -0.5 5.81E+18 6.09E+18 7.61E+18 1.25E+19 1.82E+19 - -1.5 5.60E+18 5.87E+18 7.33E+18 1.21E+19 1.75E+19 ; -2.5 5.56E+18 5.83E+18 7.28E+18 1.20E+19 1.74E+19 i -3.5 5.33E+18 5.59E+18 6.98E+18 1.15E+19 1.67E+19 -4.5 4.78E+18 5.02E+18 6.26E+18 1.03E+19 1.50E+19 -5.5 3.14E+18 3.30E+18 4.12E+18 6.78E+18 9.86E+18 -6.5 1.28E+18 1.35E+18 1.68E+18 2.77E+18 4.02E+18 Note: Axial location is provided relative to the axial midplane of the reactor core 8-7
- i
_ . _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ . _ . _ . _ _ _ _ _ _ _ _ -._ _ ~,
TABLE 8.1-5 SUBG8.ARY OF BEST ESTIMATE FAST NEUTRON (E > 0.1'MeV) EXPOSURE PROJECTIONS 4 FOR THE BELTLINE REGION OF THE POINT BEACH UNIT 2 REACTOR PRESSURE VESSEL 0 DEGREE AZIMUTHAL ANGLE 2 4(E > 0.1 MeV) In/cm) Z(ft) EOC 14 EOC 15 MOC 20 32 EFPY 48 EFPY
+6.5 9.13E+18 9.54E+18 1.JiE+19 1.86E+19 2.66E+19
+5.5 2.23E+19 2.33E+19 2.85E+19 4.57E+19 6.56E+19
+4.5 3.26E+19 3.41E+19 4.24E+19 6.97E+19 1.01E+20
+3.5 3.82E+19 3.99E+19 4.93E+19 7.98E+19 1.15E+20
+2.5 4.26E+19 4.45E+19 5.46E+19 8.73E+19 1.25E+20
+1.5 4.20E+19 4.39E+19 5.36E+19 8.52E+19 1.22E+20
+0.5 4.08E+19 4.27E+19 5.14E+19 8.00E+19 1.13E+20 0.0 4.17E+19 4.36E+19 5.16E+19 7.78E+19 1.08E+20
-0.5 4.26E+19 4.45E+19 5.18E+19 7.55E+19 1.03E+20
-1.5 3.90E+19 4.08E+19 4.79E+19 7.10E+19 9.77E+19
-2.5 3.73E+19 3.90E+19 4.66E+19 7.12E+19 9.98E+19
-3.5 3.67E+19 3.84E+19 4.66E+19 7.32E+19 1.04E+20
-4.5 3.29E+19 3.44E+19 4.22E+19 6.76E+19 9.71E+19
-5.5 2.22E+19 2.32E+19 2.84E+19 4.55E+19 6.53E+19
-6.5 8.16E+18 8.53E+18 1.05E+19 1.70E+19 2.44E+19 Note: Axial location is provided relative to the axial midplane of the reactor core 8-8
. . . = . _~. - - - ~- . _ _ = .- . - . .- - , .
TABLE 8.1-6
SUMMARY
OF BEST ESTIMATE FAST NEUTRON (E > 0.1 MeV) EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF THE POINT BEACH UNIT 2 REACTOR PRESSURE VESSEL 15 DEGREE AZIMUTHAL ANGLE 2
@(E > 0.1 MeV) In/cm)
Z(ft) EOC 14 EOC 15 EOC 20 32 EFPY 48 EFPY
+6.5 5.42E+18 5.69E+18 7.11E+18 1.17E+19 1.70E+19
+5.5 1.50E+19 1.58E+19 1.94E+19 3.13E+19 4.50E+19
+4.5 2.28E+19 2.40E+19 2.97E+19 4.85E+19 7.03E+19
+3.5 2.53E+19 2.66E+19 3.31E+19 5.45E+19 7.92E+19
+2.5 2.71E+19 2.85E+19 3.56E+19 5.88E+19 8.57E+19
+1.5 2.69E+19 2.82E+19 3.50E+19 5.69E+19 8.22E+19
+0.5 2.71E+19 2.85E+19 3.49E+19 5.56E+19 7.96E+19 0.0 2.80E+19 2.94E+19 3.55E+19 5.53E+19 7.81E+19
-0.5 2.89E+19 3.04E+19 3.61E+19 5.49E+19 7.67E+19
-1.5 2.80E+19 2.94E+19 3.50E+19 5.31E+19 7.41E+19
-2.5 2.78E+19 2.92E+19 3.52E+19 5.49E+19 7.77E+19
-3.5 2.60E+19 2.73E+19 3.37E+19 5.44E+19 7.83E+19
-4.5 2.37E+19 2.49E+19 3.07E+19 4.96E+19 7.14E+19
-5.5 1.52E+19 1.60E+19 1.98E+19 3.20E+19 4.61E+19
-6.5 5.82E+18 6.12E+18 7.57B+18 1.23E+19 1.77E+19 5
Note: Axial location is provided relative to the axial midplane of the reactor core i 8-9 i i
I l l I TABLE 8.1-7
SUMMARY
OF BEST ESTIMATE FAST NEUTRON (E > 0.1 MeV) EXPOSURE PROJECTIONS l FOR THE BELTLINE REGION OF THE POINT BEACH UNIT 2 REAC'IOR PRESSURE VESSEL 30 DEGREE AZIMUTHAL ANGLE 4(E > 0.1 MeV) In/cmj2 Z(ft) EOC 14 EOC 15 EOC 20 32 EFPY 48 EFPY
+6.5 4.34E+18 4.578+18 5.64E+18 9.12E+18 1.31E+19
+5.5 1.04E+19 1.09E+19 1.35E+19 2.18E+19 3.15E+19
+4.5 1.61E+19 1.69E+19 2.09E+19 3.36E+19 4.87E+19
+3.5 1.82E+19 1.91E+19 2.36E+19 3.82E+19 5.51E+19
+2.5 1.86E+19 1.96E+19 2.42E+19 3.91E+19 5.63E+19
+1.5 1.81E+19 1.90E+19 2.35E+19 3.80E+19 5.48E+19
+0.5 1.90E+19 2.00E+19 2.47E+19 4.00E+19 5.76E+19 0.0 1.91E+19 2.01E+19 2.48E+19 4.01E+19 5.78E+19
-0.5 1.91E+19 2.01E+19 2.49E+19 4.02E+19 5.80E+19
-1.5 1.88E+19 1.98E+19 2.44E+19 3.95E+19 5.70E+19
-2.5 1.87E+19 1.97E+19 2.44E+19 3.94E+19 5.68E+19
-3.5 1.76E+19 1.85E+19 2.29E+19 3.70E+19 5.34E+19
-4.5 1.52B+19 1.60E+19 1.97E+19 3.19E+19 4.60E+19
-5.5 9.82E+18 1.03E+19 1.28E+19 2.07E+19 2.98E+19
-6.5 3.99E+18 4.20E+18 5.18E+18 8.38E+18 1.21E+19 Note: Axial location is provided relative to the axial midplane of the reactor core 8-10
TABLE 8.1-8 SUM 4ARY OF BEST ESTIMATE FAST NEUTRON (E > 0.1 MeV) EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF THE POINT BEACH UNIT 2 REACTOR PRESSURE VESSEL 45 DEGREE AZIMUTHAL ANGLE 4(E > 0.1 MeV) In/cIn 2j Zfft) EOC 14 EOC 15 EOC 20 32 EFPY 48 EFPY
+6.5 3.44E+18 3.61E+18 4.51E+18 7.43E+18 1.08E+19
+5.5 8.18E+18 8.58E+18 1.07E+19 1.76E+19 2.57E+19
+4.5 1.30E+19 1.36E+19 1.70E+19 2.80E+19 4.07E+19
+3.5 1.42E+19 1.49E+19 1.86E+19 3.06E+19 4.45E+19
+2.5 1.61E+19 1.69E+19 2.11E+19 3.47E+19 5.05E+19
+1.5 1.55E+19 1.63E+19 2.03E+19 3.34E+19 4.86E+19
+0.5 1.55E+19 1.63E+19 2.03E+19 3.34E+19 4.86E+19 0.0 1.58E+19 1.66E+19 2.07E+19 3.41E+19 4.96E+19
-0.5 1.62E+19 1.70E+19 2.12E+19 3.49E+19 5.07E+19
-1.5 1.56E+19 1.63E+19 2.04E+19 3.36E+19 4.88E+19
-2.5 1.55E+19 1.62E+19 2.03E+19 3.34E+19 4.85E+19
-3.5 1.48E+19 1.56E+19 1.94E+19 3.20E+19 4.65E+19
-4.5 1.33E+19 1.40E+19 1.74E+19 2.87E+19 4.17E+19
-5.5. 8.75E+18 9.18E+18 1.15E+19 1.89E+19 2.74E+19
-6.5 3.57E+18 3.75E+18 4.68E+18 7.70E+18 1.12E+19 Note: Axial location is provided relative to the axial midplane of the reactor core 8-11 D
r e TABLE 8.1-9 i
SUMMARY
OF BEST ESTIMATE IRON ATOM DISPLACEMENT [dpa] EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF THE POINT BEACH UNIT 2 REACTOR PRESSURE VESSEL 0 DEGREE AZIMUTHAL ANGLE DISPLACEMENTS Idoal Z(ft) EOC 14 EOC 15 EOC 20 32 EFPY 48 EFPY
+6.5 5.34E-03 5.58E-03 6.83E-03 1.09E-02' 1.55E-02
+5.5 1.30E-02 1.36E-02 1.67E-02 2.68E-02 3.84E-02
+4.5 1.91E-02 1.99E-02 2.48E-02 4.08E-02 5.92E-02
+3.5 2.23E-02 2.33E-02 2.88E-02 4.67E-02 6.74E-02
+2.5 2.49E-02 2.60E-02 3.19E-02 5.11E-02 7.33E-02
+1.5 2.46E-02 2.57E-02 3.14E-02 4.99E-02 7.13E-02
+0.5 2.39E-02 2.50E-02 3.01E-02 4.68E-02 6.61E-02 0.0 2.44E-02 2.55E-02 3.02E-02 4.55E-02 6.32E-02
-0.5 2.49E-02 2.60E-02 3.03E-02 4.42E-02 6.03E-02
-1.5 2.288-02 2.39E-02 2.80E-02 4.15E-02 5.72E-02
-2.5 2.18E-02 2.28E-02 2.72E-02 4.17E-02 5.84E-02
-3.5 2.15E-02 2.24E-02 2.72E-02 4.29E-02 6.09E-02
-4.5 1.92E-02 2.01E-02 2.47E-02 3.96E-02 5.68E-02
-5.5 1.30E-02 1.36E-02 1.66E-02 .2.66E-02 3.82E-02
-6.5 4.78E-03 4.99E-03 6.15E-03 9.92E-03 1.43E-02 Note: Axial location is provided. relative to the axial midplane of the reactor core 8-12
TABLE 8.1-10 SUle4ARY OF BEST ESTIMATE IRON ATOM DISPLACEMENT [dpa] EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF THE POINT BEACH UNIT 2 REACTOR PRESSURE VESSEL-15 DEGREE AZIMUTHAL ANGLE DISPLACEMENTS Idoal Zift) EOC 14 EOC 15 EOC 20 32 EFPY 48 EFPY
+6.5 3.08E-03 3.23E-03 4.03E-03 6.64E-03 9.65E-03 +5.5 8.51E-03 8.95E-03 1.10E-02 1.77E-02 2.55E-02 +4.5 1.29E-02 1.36E-02 1.69E-02 2.76E-02 3.99E-02 +3.5 1.44E-02 1.51E-02 1.88E-02 3.09E-02 4.50E-02 +2.5 1.54E-02 1.62E-02 2.02E-02 3.34E-02 4.87E-02 +1.5 1.53E-02 1.60E-02 1.99E-02 3.23E-02 4.67E-02 +0.5 1.54E-02 1.62E-02 1.98E-02 3.16E-02 4.52E-02 0.0 1.59E-02 1.67E-02 2.02E-02 3.14E-02 4.44E-02 -0.5 1.64E-02 1.72E-02 2.05E-02 3.12E-02 4.35E-02 -1.5 1.59E-02 1.57E-02 1.99E-02 3.01E-02 4.21E-02 -2.5 1.58E-02 1.66E-02 2.00E-02 3.12E-02 4.41E-02 , -3.5 1.47E-02 1.55E-02 1.91E-02 3.09E-02 4.45E-02 -4.5 1.35E-02 1.41E-02 1.74E-02 2.82E-02 4.05E-02 -5.5 8.64E-03 9.08E-03 1.12E-02 1.82E-02 2.62E-02 t -6.5 3.31E-03 3.48E-03 4.30E-03 6.98E-03 1.01E-02 ;
Note: Axial location is provided relative to the axial midplane of the reactor core ! r 8-13 '
TABLE 8.1-11 SUbMARY OF BEST ESTIMATE IRON ATOM DISPLACEMENT [dpa] EXPOSTJRE PROJECTIONS FOR THE BELTLINE REGION OF TriE ' POINT BEACH UNIT 2 REACTOR PRESSURE VESSEL
.30 DEGREE AZIMUTHAL ANGLE DISPLACEMENTS Idoal Zift) EOC 14 EOC 15 EOC 20 32 EFPY 48 EFPY
+6.5 2.56E-03 2.70E-03 3.33E 5.39B-03 7.77E-03
+5.5 6.14E-03 6.47E-03 7.98E-03 1.29E-02 1.86E-02
+4.5 9.49E-03 1.00E-02 1.23E-02 2.00E-02 .2.88E +3.5 1.07E-02 1.13E-02 1.40E-02 2.26E-02 3.25E-02
+2.5 1.10E-02 1.16E-02 1.43E-02 2.31E-02 3.33E-02
+1.5 1.07E-02 1.12E-02 1.39E-02 2.25E-02 3.24E-02
+0.5 1.12E-02 1.18E-02 1.46E-02 2.36E-02 3.40E-02 0.0 1.13E-02 1.19E-02 1.46E-02 2.37E-02 3.41E-02
-0.5 1.13E-02 1.19E-02 1.47E-02 E-02 3.42E-02
-1.5 1.11E-02 1.17E-02 1.44E-02 '2 3.37E-02
-2.5 1.11E-02 1.17E-02 1.44E-02 2. 2 3.36E-02
-3.5 1.04E-02 1.10E-02 1.35E-02 2 .1. 2 3.15E-02
-4.5 8.96E-03 9.44E-03 1.17E-02 1.88. 1 .2.72E-02
-5.5 5.80E-03 6.11E-03 7.55E-03 1.22r 1.76E-02
-6.5 2.36E-03 2.48E-03 3.06E-03 4.?"~-03 7.14E-03 i
Note: Axial location is provided relative to-the axial midplane of the reactor core 8-14
TABLE 8.1-12 SUBEARY OF BEST ESTIMATE IRON ATOM DISPLACEMENT [dpa] EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF THE POINT BEACH UNIT 2 REACTOR PRESSURE VESSEL' 45 DEGREE AZIMUTHAL ANGLE DISPLACEMENTS Idoal Z(ft) EOC 14 EOC 15 EOC 20 32 EFPY 48 EFPY
+6.5 2.09E-03 2.19E-03 2.73E-03 4.50E-03 6.54E-03
+5.5 4.96E-03 5.20E-03 6.49E-03 1.07E-02 1.55E-02
+4.5 7.86E-03 8.25E-03 1.03E-02 1.60E-02 2.46E-02
+3.5 8.60E-03 9.02E-03 1.13E-02 1.85E-02 2.69E-02
+2.5 9.76E-03 1.02E-02 1.28E-02 2.10E-02 3.06E-02
+1.5 9.39E-03 9.85E-03 1.23E-02 2.02E-02 2.94E-02
+0.5 9.39E-03 9.85E-03 1.23E-02 2.02E-02 2.94E-02 0.0 9.59E-03 1.01E-02 1.26E-02 2.07E-02 3.01E-02
-0.5 9.80E-03 1.03E-02 1.28E-02 2.11E-02 3.07E-02
-1.5 9.44E-03 9.90E-03 1.24E-02 2.03E-02 2.96E-02
-2.5 9.38E-03 9.84E-03 1.23E-02 2.02E-02 2.94E-02
-3.5 8.99E-03 9.43E-03 1.18E-02 1.94E-02 2.82E-02
-4.5 8.06E-03 8.46E-03 1.06E-02 1.74E-02 2.53E-02
-5.5 5.30E-03 5.56E-03 6.94E-03 1.14E-02 1.66E-02
-6.5 2.27E-03 2.27E-03 2.83E-03 4.6-6E-03 6.78E-03 Note: Axial location is provided relative to the axial midplane of the reactor core 8-15
- _ - _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ = _ - _ _ _ _ _ _ _ - - _ - _ - _ - - -
8.2 Exposure of Specific Beltline Materials As shown in Figure 2.1-2, the beltline region of the Point Beach Unit 2 reactor pressure vessel is comprised of an intermediate shell forging (123V500), a lower shell forging (122W195), a circumferential weld (SA-1484) joining these two ring forgings, and a circumferential weld (Heat No. 21935) joining the intermediate shell forging to the nozzle shell course. The lower circumferential weld is centered 15.06 inches below the axial- i midplane of the active core; while the intermediate shell forging extends upward to an elevation 8.44 inches above the active fuel-and the lower shell forging extends downward to an elevation 39.87 inches below the bottom of the active fuel. The maximum neutron exposure experienced by each of these beltline materials can be extracted from the data provided in Tables 8.1-1 through 8.1-12. 8.2-1 Circumferential Weld (SA-1484) The current (End of Cycle 20) and projected maximum exposures of the beltline circumferential weld are listed in Table 8.2-1 and illustrated graphically in Figures 8.2-1 through 8.2-3. In this , table and the accompanying figures, the weld exposure is I expressed in terms of $(E > 1.0 MeV), $(E > 0.1 MeV), and dpa. 1 In developing the exposure profiles for the circumferential weld, it is noted that, although the flux reduction afforded by the Cycle 16-20 fuel loading patterns with part length hafnium absorbers has lessened the exposure rates within the 0-15 degree azimuthal sector, the maximum exposure point on the weld remains 1 at the o degree azimuth throughout the service life of the unit. However, the magnitude of the projected exposures are j significantly lower than would be the case had the flux ' reduction measures not been implemented. ; 8.2-2 Intermediate Shell Forging (123V500) l The current and projected maximum exposures of the intermediate shell forging are given in Table 8.2-2. Again, all three 8-16
l exposure parameters are provided. In the case of the intermediate forging, it can be noted from Table 8.1-1, that, due to the introduction of the part length absorbers and the corresponding reduction in exposure rates in the vicinity of the circumferential weld, the axia) location of the maximum exposure at the 0 and 15 degree azimuthal angles shifts from an elevation near core midplane to an elevation approximately 2.5 ft. above core midplane as the lifetime of the unit increases. Corresponding variations at the 30 and 45 degree azimuths are less evident. Since the maximum exposure pc;..t for the intermediate shell forging is variable due to the flux reduction measures, these values are not illustrated graphically, but are presented only in tabular form. 8.2-3 Lower Shell Forging (122W195) The current and and projected exposures for the lower shell forging are listed in Table 8.2-3. As in the case of the intermediate forging, all three exposure parameters are tabulated. In the case of the lower shell forging, the part length absorbers cause the maximum exposure location at 0 and 15 degrees to shift from the top of the forging to a position 3.5 feet below the active core midplane. However, the absorbers have a negligible impact at the 30 and 45 degree azimuths, resulting in the maximum exposure location remaining at the top of the forging adjacent to the circumferential weld. Again, due to this shift in the maximum exposure elevation, the data applicable to the lower shell forging are not illustrated graphically, but, rather, are presented only in tabular form. 8.2-4 Upper to Intermediate Shell Weld (Heat No. 21935) and Upper Shell Forging The current and and projected exposures for the upper to intermediate shell circumferential weld and the adjacent upper shell forging are listed in Table 8.2-4. These materials located above the reactor core are seen to be unaffected by the introduction of the part length hafnium absorbers. 8-17
1 TABLE 8.2-1 MAXIMUM FAST NEUTRON EXPOSURE OF THE POINT BEACH UNIT 2 BELTLINE CIRCUMFERENTIAL WELD (SA-1484 ) 2 4(E > 1.0 MeV) In/cm) AZIMUTHAL
~
ANGLE flTiC2 BFPY3 IE$E5EMr 48.O EFPY O' 65INEi19El 02f49E+19] 3.42E+19 15* 1.19E+19 1.74E+19 2.42E+19 30' 8.49E+18 1.38E+19 1.98E+19 45' 7.40E+18 1.22E+19 1.77E+19 4(E > 0.1 MeV) In/cmj 2 AZIMUTHAL ANGLE 18.2 EFPY 32.0 EFPY 48.0 EFPY 0" 4.89E+19 7.21E+19 9.90E+19 15* 3.53E+19 5.36E+19 7.48E+19 30' 2.45E+19 3.97E+19 5.73E+19 45* 2.06E+19 3.39E+19 4.93E+19 IRON DISPLACEMENTS Idoal AZIMUTHAL ANGLE 18.2 EFPY 32.0 EFPY 48.0 EFPY , O' 2.86E-02 4.22E-02 5.80E-02 15* 2.01E-02 3.04E-02 4.25E-02 I 30* 1.45E-02 2.34E-02 3.38E-02 45* 1.25E-02 2.05E-02 2.99E-02 : 1 1 4 i 8-18
TABLE 8.2-2 MAXIMUM FAST NEUTRON EXPOSURE OF THE POINT BEACH UNIT 2 INTERMEDIATE SHELL FORGING (123V500) 2 4(E > 1.0 MeV) In/cm) AZIMUTHAL ANGLE .18.2 EFPY 32.0 EFPY 48.0 EFPY O' 1.88E+19 3.01E+19 4.32E+19 15' 1.16E+19 1.91E+19 2.78E+19 30 8.60E+18 1.39E+19 2.01E+19 45' 7.61E+18 1.25E+19 1.82E+19 2 4(E > 0.1 MeV) in/cm) AZIMUTHAL ANGLE 18.2 EFPX 32.0 EPPY 48.0 EFPY 0 5.46E+19 8.73E+19 1.25E+20 15* 3.61E+19 5.88E+19 8.57E+19 l 30' 2.49E+19 4.02E+19 5.80E+19 l 45' 2.12E+19 3.49E+19 5.07E+19 , i l IRON DISPLACEMENTS Ideal AZIMUTHAL ANGLE 18.2 EFPY 32.0 EFPY 48.0 EFPY l O' 3.19E-02 5.11E-02 7.33E-02 15* 2.05E-02 3.34E-02 4.87E-02 30* 1.47E-02 2.37E-02 3.42E-02 45' 1.28E-02 2.11E-02 3.07E-02 I l l
)
8-19
.. _- . . . - . . . . - . . . . __ _ _ = _ .
l TABLE 8.2-3 MAXIMUM PAST NEUTRON EXPOSURE OF THE POINT BEACH UNIT 2 LOWER SHELL FORGING (122W195) 4 3 4(E > 1.0 MeV) In/cmj 2 AZIMUTHAL ANGLE 18.2 EFPY 32.0 EEEX 48.0 EFPY O' 1.69E+19 2.52E+19 3.59E+19 1 15' 1.19E+19 1.78E+19 2.54E+19 30' 8.49E+18 1.38E+19 1.98E+19 45' 7.40E+18 1.22E+19 1.77E+19 4 2 4(E > 0.1 MeV) In/cm) AZIMUTHAL ANGLE 18.2 EPPY 32.0 EFPY 48.0 EFPY O' 4.89E+19 7.32E+19 1.04E+20 15* 3.53E+19 5.49E+19 7.83E+19 30' 2.45E+19 3.97E+19 5.73E+19 45' 2.06E+19 3.39E+19 4.93E+19 IRON DISPLACEMENTS Idpal AZIMUTHAL ANGLE E dX 32.0 EFPY 48.0 EFPY l O' 2. B ui- 02 4.29E-02 6.09E-02
; 15* 2.01E-02 3.12E-02 4.45E-02 30' 1.45E-02 2.34E-02 3.38E-02 45' 1.25E-02 2.05E-02 2.99E-02 l
l l 8-20 l l
. . - .- - _. _ - . . ~ . _ - .
TABLE 8.2-4 MAXIMUM FAST NEUTRON EXPOSURE OF THE POINT BEACH UNIT 2 UPPER / INTERMEDIATE SHELL CIRCUMFERENTIAL WELD (Heat NO. 21935) AND THE UPPER SHELL FORGING 2 4(E > 1.0 MeV) In/cm) AZIMUTHAL ANGLE 18.2 EFPY 32.O EFPY 48.O EFPY 0* 3.45E+18 5.48E+18 7.84E+18 15' 2.00E+18 3.29E+18 4.78E+18 30* 1.69E+18 2.74E+18 3.95E+18 45* 1.39E+18 2.29E+18 3.33E+18 2 4(E > 0.1 MeV) In/cm) AZIMUTHAL ANGLE 18.2 EFPY 32.0 EFPY 48.0 EFPY O' 1.00E+19 1.59E+19 2.27E+19 15* 6.16E+18 1.01E+19 1.47E+19 30' 4.90E+18 7.92E+18 1.14E+19 45' 3.87E+18 6.37E+18 9.27E+18 1 IRON DISPLACEMENTS Idoal l AZIMUTHAL ' ANGLE 18.2 EFPY 32.0 EFPY 48.0 EFPY 0* 5.85E-03 9.31E-03 1.33E-02 15' 3.50E-03 5.75E-03 8.36E-03 30* 2.89E-03 4.68E-03 6.75E-03 l 45" 2.34E-03 3.86E-03 5.61E-03 8-21
d FIGURE 8.2-1 , FAST NEUTRON FLUENCE (E > 1.0 MeV) AS A FUNCTION OF AZIMUTHAL ANGLE AT THE INNER RADIUS OF THE BELTLINE CIRCUMFERENTIAL WELD l 1 4 1.0E+20 5
%[%w ' "
S % % . $1.0E+19 __ 2 g :: 8 i 5 z 1 1.0E+18 i 0 5 10 15 20 25 30 35 40 45 ! Azimuthal Angle (Degrees) i
+ 18.2 EFPY
- 32.0 EFPY + 48.0 EFPY !
i
)
8-22 1
FIGURE 8.2-2 FAST NEUTRON FLUENCE (E > 0.1 MeV) AS A FUNCTION OF AZIMUTHAL ANGLE AT THE INNER RADIUS OF THE BELTLINE CIRCUMFERENTIAL WELD l i 1 1.0E+21 R 5 1 8 l 1.0E+20:.- _ l E - C .. e ;% - m .. g '= . z ~ - 1.0 E+19
- O 5 10 15 20 25 30 35 40 45 Azimuthal Angle (Degrees)
+ 18.2 EFPY + 32.0 EFPY
- 48.0 EFPY i 1
8-23
FIGURE 8.2-3 IRON ATOM DISPLACEMENTS (dpa] AS A FUNCTION OF AZIMUTHAL ANGLE AT THE INNER RADIUS OF THE BELTLINE CIRCUMFERENTIAL WELD 1.0 E+00 a E 5m E 8 { 1.0E-01
.e Q am-h 1% - w% __--
D6 f _8
<p: ::---
~
---n
%i 1.0E-02 0 5 10 15 20 25 30 35 40 45 Azimuthal Angle (Degrees)
+ 18.2 EFPY + 32.0 EFPY + 48.0 EFPY 8-24
8.3 Uncertainties in Exposure Projections The overall uncertainty in the best estimate exposure projections within the pressure vessel wall stem primarily from two sources;
- 1) the uncertainty in the bias factor (K) derived from the plant specific measurement data base; and
- 2) the analytical uncertainty associated with relating the results at the measurement locations to the desired results within the pressure vessel wall.
Uncertainty in the bias factor derives directly from the individual uncertainties in the measurement process, in the least squares adjustment procedure, and in the location of the surveillance capsule and cavity dosimetry sensor sets. The h analytical uncertainty in the relationship between the exposure of the pressure vessel and the exposure at the measurement l locations are based on the vessel thickness tolerance relative to l the cavity data and on downcomer water density variations and vessel inner radius tolerance relative to the surveillance capsule data. 1 The la uncertainties associated with the bias factors applicable , to 4(E > 1.0 MeV), 4 (E > 0.1 MeV) , and dpa are given in Section l 8.1 of this report. The additional information pertinent to the l required analytical uncertainty for vessel locations has been obtained from benchmarking studies using the Westinghouse neutron transport methodology and from several comparisons of power reactor internal surveillance capsule dosimetry and reactor cavity dosimetry for which the irradiation history of all sensors was the same. Based on these benchmarking evaluations the additional uncertainty associated with the tolerances in dosimetry positioning, vessel thickness, vessel inner radius and downcomer temperature was estimated to be approximately 6% for all exposure parameters. These uncertainty components were then combined as follows: 8-25 i
i la UNCERTAINTY . 4(E > 1.0 MeV) 4(E > 0.1 MeV) dpa
- Bias Factor 8.0% 16.5% 12.5%
l Analytical 6.0% 6.0% 6.0% Combined 10.0% 17.6% 13.9% Thus, the total uncertainty associated with the neutron exposure projections at the pressure vessel clad / base metal interface for Point Beach Unit 2 was estimated to be: 4 la Uncertainty 4(E > 1.0 MeV) 10% 4(E > 0.1 MeV) 18% dpa 14% These uncertainty values are well within the 20% la uncertainty in vessel fluence projections required by the PTS rule. .I e l 8-26 4
1 1 SECTION 9.0 l REFERENCES
- 1. Anderson, S. L. and Fero, A. H., " Reactor Cavity Neutron
. Measurement Program for Wisconsin" Electric Power Company Point BeachrNuclear Plant-Unit 1 and Unit 2," WCAP-11944, Rev 1, Junet1989.
- 2. Yanichko, S. E., et. al., " Wisconsin Michigan Power Co. and the Wisconsin Electric Power Company Point Beach Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-7712, June 1971.
- 3. Anderson, S. L., et. al., " Reactor Cavity Neutron Measurement Program for Wisconsin Electric Power Company Point. Beach Unit 2 -
Evaluations Through Cycle 15", WCAP-12795, November 1990. . 4. Anderson, S. L., et. al., " Reactor Cavity Neutron
-Measurement Program for Wisconsin Electric Power Company Point Beach Unit 2", WCAP-12795, Rev. 1, October 1991.
i 5. Anderson, S. L., et. al., " Reactor Cavity Neutron Measurement Program for Wisconsin Electric Power Company Point Beach Unit 2", WCAP-12795, Rev. 2, September 1992. I- 6. RSIC Computer Code Collection CCC-543, " TORT-DORT Two- and Three-Dimensional Discrete Ordinates Transport, Version , 2.8.14", January 1994. l l
- 7. RSIC Data Library Collection DLC-175, " BUGLE-93, Production
- and Testing of the VITAMIN-B6 Fine Group and the BUGLE-93 j Broad Group Neutron / Photon Cross-Section Libraries Derived
- from ENDF/B-VI Nuclear Data", April 1994.
1 I
- 8. Maerker, R. E., et. al., " Accounting for Changing Source
, Distributions in Light Water Reactor Surveillance Dosimetry , Analysis", Nuclear Science and Engineering, Volume 94, pp 291-308, 1986. ~ 9-1 . Y 4
, - . , - y - - - -,v.- w m,.--,v_._L
- 9. ASTM Designation E706-87, " Standard Master Matrix for l Light-Water Reactor Pressure Vessel Surveillance Standards,"
in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1993
- 10. ASTM Designation E853-87, " Standard Practice for Analysis and Interpretation of Light -Water Reactor Surveillance Results," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1993.
- 11. ASTM Designation E261-90, " Standard Method for Determining Neutron flux, Fluence, and Spectra by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa. 1993.
- 12. ASTM Designation E262-86 (Reapproved 1991), " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa. 1993.
- 13. ASTM Designation E263-88, " Standard Method for Determining Fast Neutron Flux Density by Radioactivation of Iron," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1993.
- 14. ASTM Designation E264-92, " Standard Method for Determining Fast Neutron Flux Density by Radioactivation of Nickel," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1993.
- 15. ASTM Designation E481-86 (Reapproved 1991), " Standard Method for Measuring Neutron Flux Density by Radioactivation of Cobalt and Silver," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1993.
- 16. ASTM Designation E523-92, " Standard Method for Determining Fast Neutron Flux Density by Radioactivation of Copper," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1993.
9-2
- 17. ASTM Designation E704-90, " Standard Method for Measuring Reaction Rates.by Radioactivation of Uranium-238," in ASTM ,
Standards, Section'12, American Society'for Testing and-Materials, Philadelphia, Pa., 1993.
- 18. ASTM Designation E705-90, " Standard Method for Determining Fast Neutron Flux Density by Radioactivation of Neptunium-237," in' ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,
1993.
- 19. ASTM Designation E1005-84 (Reapproved 1991), " Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance," in ASTM Standards, Section 12, American Society.for Testing and Materials, Philadelphia, Pa., 1993.
- 20. Schmittroth, E. A., " FERRET Data Analysis Code",
HEDL-TME-79-40, Hanford Engineering Development Laboratory, Richland, Washington, September ~1979.
- 21. McElroy, W. N., et. al., "A Computer-Automated Iterative Method of Neutron Flux Spectra Determined by Foil Activation," AFWL-TR-67-41, Volumes I-IV, Air Force Weapons J Laboratory, Kirkland AFB, NM, July 1967.
- 22. RSIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section Compendium", July 1994.
-23. Maerker, R. E. as reported by Stallman, F. W., " Workshop on Adjustment Codes and Uncertainties - Proc. of the 4th j ASTM / EURATOM Symposium on Reactor Dosimetry," NUREG/CP-0029, l NRC, Washington, D.C., July 1982.
- 24. Scherpereel, L. R., " Core Physics Characteristics of the i Point Beach Nuclear Plant - Unit 1, Cycle 1," WCAP-7430, December 1969. [ proprietary)
- 25. Hawrylak, J. P., " Revised Cycle 2 Nuclear Design
. Characteristics for Point Beach Unit 2," WCAP-8418, Rev 1,_ November,1974. [ proprietary) 9-3
4
- 26. Hawrylak, J. P., et. al., "The Nuclear Design - Core i Management of the Point Beach Unit 2 Nuclear Reactor - Cycle !
3, " WCAP-8759, March 1976. [ proprietary]
- 27. Hawrylak, J. P., et. al., "The Nuclear Design - Core Management of the Point Beach Unit'2 Nuclear Reactor - Cycle 4," WCAP-8934, February 1977. [ proprietary]
- 28. Hawrylak, J. P., et. al., "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 5," WCAP-9275, February 1978. [ proprietary]
- 29. Pilzer, E. H., et. al., "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 6," WCAP-9493, April 1979. [ proprietary]
- 30. Scherder, W. J., et. al., "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 7," WCAP-9667, February 1980. [ proprietary]
- 31. Smith R. T., et. al., "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 8,"
WCAP-9846, March 1981. [ proprietary]
- 32. Smith R. T., et. al., "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 9,"
WCAP-10048, March 1982. [ proprietary] l I
- 33. Smith R. T., et. al., "The Nuclear Design - Core Management )
of the Point Beach Unit 2 Nuclear Reactor - Cycle 10," , WCAP-10278, March 1983. [ proprietary).
- 34. Smith R. T., "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 11, " WCAP-10583, August 1984. [ proprietary).
- 35. Smith R. T., "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 12 Rev 1,"
WCAP-10897, November 1985. [ proprietary] . 1 9-4 I
- 36. -Smith-R. T., "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor . Cycle 13 " WCAP-11288, November 1986.- [ proprietary) .
I
- 37. Smith R. T., "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Rea'ctor - Cycle 14 " WCAP-11571, September 1987. [ proprietary].
- 38. Smith R. T., "The Nuclear Design - Core Management of the-Point Beach Unit 2 Nuclear Reactor - Cycle 15 , -" WCAP-11903, September 1988. [ proprietary).
- 39. Smith R' . T., "The Nuclear Design - Core Management of.the Point Beach Unit 2 Nuclear Reactor - Cycle 16," WCAP-12362, !
September 1989. [ proprietary) .
'40. -Smith R. T., "The Nuclear Design - Core' Management of the .
Point Beach Unit 2 Nuclear Reactor - Cycle 17," WCAP-12683, September 1990. [ proprietary) .
- 41. Hoerner J. A., et.al, "The Nuclear Design and Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 18," WCAP-13063, September 1991. [ proprietary) .
, 42. Hoerner J. A., et.al, "The Nuclear Design and Core l Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 19," WCAP-13467, September 1992. [ proprietary). l
- 43. Beebe B. R., et.al, "The Nuclear Design and Core Management i of the Point Beach Unit 2 Nuclear Reactor - Cycle 20,"
f WCAP-13843, September 1993. [ proprietary). I 44. Davidson, J. A., et. al., " Analysis of Capsule T from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 2 Reactor Vessel Radiation Surveillance Program," [ WCAP-9331, August 1978.
- - 45. Yanichko, S. E., et. al., " Analysis of Capsule R from the Wisconsin Electric Power Company Point Beach Nuclear Plant
~
. Unit No. 2 Reactor Vessel Radiation Surveillance Program,"
WCAP-9635, December 1979. p 9-5 4 i
, , , , - .-w-, , ,w.m - - . . , . _. ., ,, . . . - . . , - - . . ,
- 46. Lowe, A. L., et. al., " Analysis of Capsule S, Wisconsin Electric Power Company, Point Beach Unit No. 2, Reactor Vessel Material Surveillance Program", BAW-2140, August 1991.
- 47. Perrin, J. S.,.et. al., " Point Beach Nuclear Plant Unit No.
2' Pressure Vessel Surveillance Program: Evaluation of
-Capsule V," Battelle Memorial Institute Report, June 1975.
9-6
APPENDIX A SPECIFIC ACTIVITIES AND IRRADIATION HISTORY OF SENSORS FROM SURVEILLANCE CAPSULES V, T, R AND S In this appendix, the irradiation history as extracted from NUREG-0020 and the measured specific activities of radiometric sensors irradiated in surveillance Capsules V, T, R and S are provided. i The irradiation history of capsules withdrawn to date was as follows: f CYCLE NO. STARTUS. SHUTDOWN COMMENT I 1 05/30/72 10/17/74 CAPSULE V WITHDRAWN 2 12/20/74 02/26/76
- 3 03/26/76 03/04/77 CAPSULE T WITHDRAWN
) 4 04/19/77 03/22/78 4 5 04/17/78 03/23/79 CAPSULE R WITHORAWN 6 04/19/79 04/11/80 ! 7 05/13/80 04/17/81 !. 8 05/21/81 04/16/82 4 9 05/26/82 03/25/83 10 07/05/83 09/28/84 11 11/19/84 10/05/85 i 12 11/24/85 09/27/86 1 13 12/01/86 10/03/87 1 14 11/17/87 10/08/88 i 15 11/21/88 09/23/89 i 4- 16 11/24/89 10/06/90 CAPSULE S WITHORAWN , J REF. CORE POWER = 1518 MWt a The monthly thermal generation applicable to the Point Beach Unit 2 reactor is l provided on pages A-2 and A-3. Pages A-4 through A-7 contain the measured , specific activities ofsensors removed from Capsules T, R, and S. l A-1
i l 4 MONTHLY THERMAL GENERATION DURING THE FIRST SIXTEEN FUEL CYCLES OF THE POINT 8EACH UNIT 2 REACTOR j THERMAL THERMAL THERMAL THERMAL j GENERATION GENERATION GENERATION GENERATION
} M (W-hr) M (W-hr) M (W-hr) M (W-hr)
J 5/72-3/74 13508112 4/76 972825 5/78 1099610 6/80 1065238 ) 4/74 1076568 5/76 956959 6/78 1044078 7/80 1114363
; 5/74 1111056 6/76 1007721 7/78' 1040240 8/80 1124200
} 6/74 885000 7/76 1026604 8/78 1054425 9/80 1047595 l 7/74 954648 8/76 1022317 9/78 1059006 10/80 1112514
! 8/74 1111608 9/76 1005046 10/78 1104830 11/80 989299
) 9/74 1054224 10/76 1116240 11/78 1079867 12/80 1114432
! 10/74 557784 11/76 1064445 12/78 1074069 1/81 1115599
- 11/74 0 12/76 1102576 1/79 1116477 2/81 1008189 l 12/74 302016 1/77 1102848 2/79 1010047 3/81 1112552 I 1/75 1113456 2/77 1001354 3/79 746785 4/81 559549 i 2/75 881295 3/77 100292 4/79 585747 5/81 186873 l 3/75 1081779 4/77 214373 5/79 1071794 6/81 1047500
, 4/75 916898 5/77 1108075 6/79 1021650 7/81 1112509 5/75 880266 6/77 1066583 7/79 408680 8/81 1092410 l i 6/75 914234 7/77 1072410 8/79 1114720 9/81 988920 7/75 1063799 8/77 973371 9/79 1032935 10/81 1088553 8/75 748416 9/77 1039145 10/79 1117434 11/81 1080909 l
- 9/75 997380 10/77 1109781 11/79 1045777 12/81 1073535 j 10/75 989176 11/77 1062668 12/79 1095273 1/82 1106250 l 11/75 974925 12/77 1106636 1/80 1111502 2/82 1005969 12/75- 1114475 1/78 1066993 2/80 974189 3/82 1091347 l 1/76 1070693 2/78 1000903 3/80 613022 4/82 528202 2/76 929464 3/78 723452 4/80 328351 5/82 124874 3/76 115451 4/78 352782 5/80 539505 6/82 1074941 L
p A-2
1 1 MONTHLY THERMAL GENERATION DURING THE FIRST SIXTEEN FUEL CYCLES OF THE POINT BEACH UNIT 2 REACTOR THERMAL THERMAL THERMAL THERMAL GENERATION GENERATION GENERATION GENERATION M (MW-hr) M (MW-hr) M (W-hr) M (MW-hr) 7/82 1117700 8/84 11269;93 9/86 945202 10/88 229686 8/82 1052625 9/84 921284 10/86 0 11/88 229879 9/82 864026 10/84 0 11/86 10240 12/88 1113148
, 10/82 1094393 11/84 252959 12/86 1088521 1/89 1127540 11/82 1089345 12/84 1062541 1/87 1037881 2/89 1018557 12/82 1109078 1/85 1103110 2/87 1017595 3/89 1030496 1/83 1121499 2/85 1015875 3/87 1066066 4/89 1008555 2/83 1016200 3/85 1121631 4/87 1091705 5/89 1117762 3/83 869556 4/85 1089086 5/87 1031289 6/89 1089331 4/83 0 5/85 1125089 6/87 1090934 7/89 1126185 5/83 0 6/85 1081247 7/87 1120329 8/89 1078103 6/83 0 7/85 1104022 8/87 1033949 9/89 776161 883944 1125504 0 l 7/83 8/85 9/87 1092960 10/89 )
- 8/83 1075582 9/85 1090774 10/87 70587 11/89 141690 l 9/83 1087367 10/85 141568 11/87 382736 12/89 1123322 10/83 1115645 11/85 95661 12/87 1067563 1/90 1127322
- 11/83 1088861 12/85 1033331 1/88 1123573 2/90 1018430
! 12/83 1120529 1/86 1096374 2/88 1055151 3/90 1118045 1/84 1109752 2/86 1001623 3/86 1127373 4/90 1090485 2/84 1052908 3/86 1120825 4/88 949345 5/90 1127775
- 3/84 1125590 4/86 1037164 5/88 1127476 6/90 1080806 4/84 1068842 5/86 1110585 6/88 1091026 7/90 1128256 5/84 934923 6/86 1024233 7/88 1126029 8/90 1129766
, 6/84 1083029 7/86 1069310 8/88 1129392 9/90 1120491 7/84 1119374 8/86 1083054 9/88 1072861 10/90 162384 I 4 1 4 A-3
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.A_ 'ff .
{ 1 -- ; -- A ! l
I Since surveillance Capsules T, R, and S were irradiated for multiple fuel cycles, the flux adjustment factors, C, j defined in Section 3.0 were employed in the reaction rate calculations for the-individual sensor sets. The quantity C; is defined as the calculated ratio of ((E > 1.0 MeV) during the irradiation period j to the time weighted average $(E > 1.0 MeV) over the entire irradiation period. .The values of Cj used in the evaluation of the Point Beach Unit 2 surveillance capsules were as follows: FLUX ADJUSTMENT FACTOR C; CAPSULE V CAPSULE T CAPSULE R CAPSULE S CYCLE 1 1.000 0.973 1.008 1.080 CYCLE 2 1.029 1.022 1.179 CYCLE 3 1.010 1.008 1.168 CYCLE 4 0.955 1.116 CYCLE 5 1.008 1.119 CYCLE 6 1.155 CYCLE 7 0.971 CYCLE 8 0.926 CYCLE 9 0.938 CYCLE 10 0.961 CYCLE 11 0.949 CYCLE 12 0.905 CYCLE 13 0.859 CYCLE 14 0.934 CYCLE 15 0.856 CYCLE 16 0.753 9 2 1 A-8
APPENDIX B MEASURED SPECIFIC ACTIVITY AND IRRADIATION HISTORY OF REACTOR CAVITY SENSOR SETS - CYCLE 15 In this appendix, the irradiation history as extracted from NUREG-0020 and the measured specific activities of radiometric i sensors irradiated in the reactor cavity during Cycle 15 are i provided. The irradiation history of Cycle 15 was as follows: Cycle No. STARTUP SHUTDOWN COMME'NT 15 11/21/88 11/24/89 Ref. Core Power = 1518 MWt THERMAL GENERATION MONTH (MW-Hr) 11/88 229879 12/88 1113148 1/89 1127540 2/89 1018557 3/89 1030496 4/89 1008555 5/89 1117762 6/89 1089331 7/89 1126185 8/89 1078103 9/89 776161 TOTAL 10715717 l The irradiation capsule loading diagram and the measured specific activities of the radiometric monitors from the Cycle 15 irradiation are provided on pages B-2 through B-9. For the multiple foil sensor sets, the individual foil ID can be correlated with the capsule loading diagram provided in Section 6.1-1 in order to determine the location of the foil within the reactor cavity during irradiation. ; B-1
)
CollTENTS OF $ULTIPLE FOIL SDISOR SETS i 1 CAPSULE 10 BARE OR RADIOMETRIC MONITOR ID and CADMIUM SSTR j P05fif0N DifELDED 7,3 gj, gg Ij, g f,g y,g}g g ggggig G-1 8 CG - - - - IN - - PS-48 G-2 Cd 0G AG AG G H OH G - -- G-3 Cd - - - - - - - 2 PB-6C 1 H-1 8 CH - - - - BL - - P8-188 H-2 Cd OH AN AH H L OL M - -
. H-3 Cd - - - - - - - 3 N-16C l I-1 8 CI - - - - IK - - M-78 I-2 Cd 0! A! A! ! K OK ! - --
I-3 Cd - - - -- - - 4 PS-7C J-1 8 CJ - - - - SI - - PB-178 j J-2 Cd DJ AJ M J ! O! J - - 1 J-3 Cd - - - -- - - 5 PS-17C l 1 K-1 8 CK -- -- BJ - - PS-188 K-2 Cd (K AK AK K J DJ K - -. K-3 Cd - - --- - - 6 PB-lac l L-1 8 CL - --- IN - - PS-198 L-2 Cd OL AL AL L M OH L - - L-3 Cd - - - - - - - 7 PB-19C l XX-1 8 CT - - -- IA - - PS-18 l XX-2 Cd DT AT AT T A DA T - - XX-3 Cd - - - - - - - .1 PB-1C i B-2 I
1
'lestir.g.tcuse Elec ric Ccrpcratien Advar.ced Er.ergy Systerra - Analytical Laboratcry F.EICPT Walt: Fill Site Request # 13910 TO: A.H. Fero (W)NMD, Energy Center (East 4-17)
Received: 1/19/90 Reported: 2/20/90 l (RESULTS OF ANALYSIS] Point Beach Unit 2 Cycle 15 Reactor Cavity Dosimetry Lab Dosimeter ($ 2/1/90) Foil ID Staplet Material Nuclide dps/ag foil 2 sigma AT 90-233 Ni Co-58 10.5 +/- 0.1 1 AG 90-243 Ni Co-58 94.9 +/- 1.1 ! AE 90-253 Ni co-58 245.4 +/- 3.1 l AI 90-263 Ni Co-58 93.0 +/- 1.4 I AJ 90-273 Ni co-58 215.3 +/- 2.1 1 AK 90-283 Ni Co-58 160.3 +/- 1.4 l AL 90-293 Ni Co-58 143.0 +/- 1.7 ! AT 90-234 Cu co-60 0.0233 +/- 0.0005 AG 90-244 Cu . Co-60 0.242 +/- 0.007 AE 90-254 Cu co-60 0.670 +/- 0.011 AI 90-264 Cu Co-60 0.240 +/- 0.007 AJ 90-274 Cu Co-60 0.618 +/- 0.011 AK 90-284 Cu Co-60 0.467 +/- 0.010 AL 90-294 Cu Co-60 0.448 +/- 0.007 T 90-235 Ti Sc-46 0.190 +/- 0.011 G 90-245 Ti sc-46 2.00 +/- 0.05 5 90-255 Ti Sc-46 5.05 +/- 0.15 I 90-265 Ti Sc-46 1.98 +/- 0.07 J 90-275 Ti Sc-46 4.64 +/- 0.12 K 90-285 Ti Sc-46 3.49 +/- 0.11 L 90-295 Ti sc-46 3.30 +/- 0.08
)
Remarks: AL File: 13910
References:
Lab.Booke41p74. IJ435p213-216. Proce& ares: A-524. Analyst: WIF, NM, CAB. Approved: - - #-20-9o l B-3 i 1
. _ .. ~- - . . - - .. . - - . . . . _ _ _ _ _ _ _ _ _ _ _ _
Westinghcuse Electric Cer;cratien Advanced Energy Systees - Analytical Laboratory FIPCP- Waltz Mill Site Requesti 1391C TO: A.H. Fero (W)pdD, Energy Center (East 4-17) Received: 1/19/90 Reported: 2/20/90 (RESULTS OF ANALYSIS) Point Beach Unit 2 Cycle 15 Reactor Cavity Dosimetry Lab Dosimeter ($ 2/1/90) Foil ID Sample 4 Material Nuclide dps/mg foil 2 signa BA 90-237 A1Co Co-60 65.1 +/- 1.1 DA 90-238 A1Co Co-60 47.0 +/- 0.9 BM 90-247 A1Co Co-60 178.8 +/- 4.6 DM 90-248 A1Co Co-60 131.5 +/- 1.6 BL 90-257 AlCo co-60 525.6 +/- 8.3 DL 90-258 A1Co Co-60 310.6 +/- 6.1 BK 90-267 A1Co Co-60 209.8 +/- 3.6 DK 90-268 A1Co Co-60 136.3 +/- 3.3 BI 90-277 A1Co Co-60 657.6 +/- 8.9 DI. 90-278 A1Co Co-60 375.8 +/- 7.0 BJ 90-287 A1Co Co-60 537.8 +/- 8.2 DJ 90-288 AlCo Co-60 305.9 +/- 6.1 BH 90-297 AlCo co-60 339.9 +/- 6.6 DH 90-298 A1Co Co-60 222.8 +/- 5.3 CT 90-231 Fe Mn-54 0.757 +/- 0.016 DT 90-232 Fe Mn-54 0.912 +/- 0.016 CG 90-241 Fe Mn-54 6.72 +/- 0.12 DG 90-242 Fe Mn-54 8.33 +/- 0.09 CB 90-251 Fe Mn-54 20.81 +/- 0.37 DE 90-252 Fe Mn-54 21.31 +/- 0.29 CI 90-261 Fe Mn-54 8.39 +/- 0.19 DI 90-262 Fe Mn-54 7.73 +/- 0.14 CJ 90-271 Fe Mn-54 19.07 +/- 0.27 ELT 90-272 Fe Mn-54 18.72 +/- 0.24 CK 90-281 Fe Mn-54 15.15 +/- 0.13 DK 90-282 Fe Mn-54 13.98 +/- 0.17 CL 90-291 Fe Mn-54 12.41 +/- 0.22 DL 90-292 Fe Mn-54 12.10 +/- 0.24 Bauarks: AL File: 13910
References:
Lab.Bocke41p74. LB435p213-216. Procedures: A-524. Analyst: WIF, NRM, CAB. Approved: g g,3,g B-4
I i < 1 Wstir.g.'.ctse Electric Cct;cratien Adiar.ced Energy Systeers - Analytical laboratory PI?CIC Waltz Mill Site Pequesti 1391C
- To: A.H. Fero (W)MD, Energy Center (East 4-17)
Received: 1/19/90 Reported: 2/20/90 j [RESULTS OF ANALYSIS] Point Beach Unit 2 Cycle 15 Reactor Cavity Dosimetry i Lab Dosimeter (f 2/1/90) Foil ID Samplet Material Nuclide dps/ag foil 2 sigma T 90-239 U-238 Ir-95 1.40 +/- 0.01 G 90-249 U-238 Ir-95 7.74 +/- 0.07 8 90-259 U-238 Ir-95 19,59 +/- 0.13 I 90-269 U-238 3r-95 7.90 +/- 0.07 , J 90-279 U-238 Ir-95 17.76 +/- 0.13 i 90-289 U-238 K L 90-299 Ir-95 13.32 +/- 0.11 U-238 3r-95 11.69 +/- .0.08 4 T 90-239 U-238 Ru-103 0.544 +/- 0.005
- G 90-249 U-238 Ru-103 3.780 +/- 0.040 r
H 90-259 U-238 Ru-103 9.308 +/- 0.075 , I - 90-269 U-238 Ru-103 3.801 +/- 0.038 , J 90-279 U-238 Ru-103 8.259 +/- 0.074 i 90-289 U-238 K Ru-103 6.082 +/- 0.063 l L 90-299 U-238 Ru-103 5.451 +/- 0.048 T 90-239 U-238 Cs-137 0.114 +/- 0.003 l j G 90-249 U-238 Cs-137 90-259 U-238 0.680 +/- 0.019 l i l H Cs-137 1.843 +/- 0.036 I 90-269 U-238 Cs-137 0.677 +/- 0.021 ! 90-279 U-238 J Cs-137 1.683 +/- 0.036 i K 90-289 U-238 Cs-137 1.241 +/- 0.030 L 90-299 U-238 Cs-137 1.102 +/- 0.021 i 1 90-240 Np Cs-137 2.33 +/- 0.32 2 90-250 Np Cs-137 10.59- +/- 0.53 ! 3 90-250 Np Cs-137 24.07 +/- 0.64 4 90-270 Np Cs-137 9.69- +/- 0.58 5 90-280 Np Cs-137 23.84 +/- 0.78 6 90-290 Np Cs-137 16.37 +/- 0.59 7 90-300 Np Cs-137 14.66 +/- 0.55 i l nearks: a J AL File: 13910 i
References:
Lab. Books 41p74.18035p213-216. Procedures: A-524. Analyst: WIT, Nim, CAB. Approved: g*
- A ~20* E0 l
B-5
i I WestircMure E*.t: ric C r;craien Advanced Erargy Syste:.s - Analytical Lateratcry F.EKIC Waltz . Mill Site Fecuest( 13910 TO: A.B. Fero (W)lpdD, Diergy Center (East 4-17) Received: 1/19/90 Reported: 2/20/90 [REELTS T AlfLYSIS] Point Beach Unit 2 Cycle 15 Beactor Cavity Dosimetry Bead Chain Tag ID: S-2, O degree. Feet (< @ s/ag of chain 0 2/1/90 >] fra Lab sti-54 Co-58 Co-60 Mi@ lane Sampleo dps/ag 2 signa dpshg 2 sigma dps/mg 2 signa j
+6.5 90-301 3.11+/- 0.11 5.20 +/- 0.12 42.4 +/- 0.3 I
+5.5 90-302 7.58 +/- 0.40 12.89+/- 0.53 95.6 +/- 1.1 l
+4.5 90-303 11.07 +/- 0.53 18.96+/- 0.63 122.7 +/- 1.1 1
+3.5 90-304 12.96 +/- 0.85 21.20+/- 0.84 151.0 +/- 1.7
+2.5 90-305 14.50+/- 0.63 23.24+/- 0.72 168.4 +/- 1.4
+1.5
- 96-306 14.26 +/- 0.82 22.83+/- 0.97 173.3 +/- 1.8
+0.5 90-307 13.88+/- 0.69 21.38+/- 0.65 177.5 +/- 1.4
-0.5 90-308 14.53 +/- 0.86 22.45 +/- 0.97 200.7 +/- 2.0
-1.5 90-309 13.29 +/- 0.60 20.72 +/- 0.65 187.0 +/- 1.4
-2.5 90-310 12.70 +/- 0.63 21.29+/- 0.64 178.1 +/- 1.4
-3.5 90-311 12.50+/- 0.60 20.34+/- 0.64 163.6 +/- 1.3
-4.5 90-312 11.24 +/- 0.52 18.61+/- 0.57 131.6 +/- 1.2
-5.5 90-313 7.55+/- 0.42 12.54+/- 0.44 87.4 +/- 1. 0 l
-6.5 90-314 2.78+/- 0.22 4.90+/- 0.24 67.4 +/- 0.8 t
IME: For the gasse counts, 6 beads were cut fece the chain at the desipated points of " Feet fra M141ane" (3 beads on endt side of the point). l 6 beads are about 1.4 inches long and weigh about 1.2 gras.
)
Remarks: I i AL File: 13910 l
References:
Imb.Bookf41p74. Ia435p21>216. MystI* CDs. W: N */ 88 ****'" # ~ 3# ~i# l 1 B-6
> estire.cuse Electric Ccrpcratien Advar.ced Er.ergy Syster.s - Analytical Laboratory F.!KIC Waltz Fill Site Requesti 13910 l
TO: A.H. Fero (W)NMD, Energy Center (East 4-17) Received: 1/19/90 Reported: 2/20/90 l 1 [REMETS OF ANALYSIS) Point Beach Unit 2 Cycle 15 Reactor Cavity Dosimetry Bead Chain Tag ID: 5-2, 15 degree. Feet [< dps/ag of chain # 2/1/90 >} frcui Iab h 54 Co-58 Co-60 Midplane Sample 4 dps/ag 2 sigma W 2 sips dps/mg 2 signa
+6.5 90-315 2.40+/- 0.20 4.11 +/- 0.23 43.7 +/- 0.5
+5.5 9&316 6.64 +/- 0.51 10.19 +/- 0.53 141.4 +/- 1.3
+4.5 9&317 10.05 +/- 0.62 16.52 +/- 0.63 203.6 +/- 1.6
+3.5 90-318 11.15 +/- 0.90 17.72 +/- 0.94 236.1 +/- 2.3
+2.5 90-319 12.09 +/- 1.03 20.12 +/- 1.08 253.7 +/- 2.3
+1.5 90-320 11.87+/- 0.92 19.71+/- 0.99 253.9 +/- 2.3 I
+0.5 90-321 12.04 +/- 0.97 19.37 +/- 1.06 247.4 +/- 2.3
-0.5 90-322 12.84 +/- 1.03 20.33 +/- 1.08 253.9 +/- 2.3
-1.5 90-323 12.41+/- 0.97 19.61 +/- 1.06 241.9 +/- 2.3
-2.5 90-324 12.26 +/- 0.92 18.61 +/- 1.05 227.6 +/- 2.2
-3.5 90-325 11.54 +/- 0.85 18.61 +/- 0.93 212.3 +/- 2.1
-4.5 90-326 10.52 +/- 0.87 17.06 +/- 0.94 176.8 +/- 1.9
, -5.5 90-327 6.74 +/- 0.40 11.32 +/- 0.37 125.9 +/- 0.8 i -6.5 90-328 2.58 +/- 0.21 4.50 +/- 0.22 68.2 +/- 0.5 4 i I e IDIT: For the gasma counts, 6 beads were cut frcut the chain at the designated points of " Feet frcui Midplane" (3 beads en each side of the point). 6 beads are about 1.4 inches long ard weigh about 1.2 grams. i 1 I j i . Remarks: 1 AL File: 13910 neferences: Imb.Bockt41p74. Ia435p213-216. ; Procedures: A-524. I
- Analyst
- WIF,1891, QB. Jgproved: .
4- M -fd s i B-7 i
Westirgrase Electric Cor;craticn Advanced Er.ergy Systems - Analytical Laboratcry REXPT Kalt2 Mill Site Recuesti 12910
'10: A.H. Fero Of)lpdD, Energy Center (East 4-17)
Received: 1/19/90 Reported: 2/20/90 [REELTS OF AB8LYSISj Point Beach Unit 2 Cycle 15 Retctor Cavity namhtry Bead Chain Tag ID: S-2, 30 degree. Feet [< % s/ag of chain 0 2/1/90 ->J from Imb Iti-54 Co-58 CHO Midplane susplet dps/ag 2 signa 4 sang 2 sigma 4 sang 2 signa
+6.5 90-329 2.03 +/- 0.11 3.36+/- 0.11 38.5 0.3
+5.5 90-330 4.95 +/- 0.33 8.58 +/- 0.37 108.2 0.8
+4.5 90-331 7.60+/- 0.38 12.21 +/- 0.51 154.5 0.9
+3.5 90-332 8.37 +/- 0.67 13.71 +/- 0.54 183.4 1.2
+2.5 90-333 8.93 +/- 0.46 14.84 +/- 0.54 195.0 1.1
+1.5 90-334 8.51 +/- 0.78 14.97+/- 0.98 204.0 2.0
+0.5 90-335 8.60 +/- 0.84 14.18 +/- 0.93 210.6 2.1
-0.5 90-336 9.30 +/- 0.65 14.69 +/- 0.70 206.6 1.6
-1.5 90-337 8.88 +/- 0.68 14.16 +/- 0.70 201.7 1.5
-2.5 90-338 9.42 +/- 0.66 14.13+/- 0.61 192.8 1.5
-3.5 90-339 8.06 +/- 0.54 13.38 +/- 0.65 175.1 1.4
-4.5 90-340 7.05 +/- 0.35 11.97 +/- 0.43 138.9 0.9
-5.5 90-341 4.45 +/- 0.27 7.94 +/- 0.32 74.5 0.7
-6.5 90-342 1.94 +/- 0.20 3.31 +/- 0.26 51.9 0.6 IGE: For the gassa counts, 6 beads are cut fremt the chain at the designated points of
- Feet from Midplane" (3 beads on each side of the point).
6 beads are about 1.4 inches long and wigh about 1.2 grams. i 1 mamarks: i f AL File: 13910
References:
IAb.Bocki41p74. Ia035p213-216. Procedures: A-524. // Analyst: WIF, W M, CAB. W: dd*/ W h A-8#-fC t B-8
l I , Westirghcuse Cectric Cor;cratien Advanced Er.ergy Systerrs - Analytical Laboratory REPOPC Faltz Eill Site Pequesti 13910 i 4
- TO: A.H. Fero Of)WdD, Energy Center (East 4-17)
Hoceived: 1/19/90 Reported: 2/20/90 [REEELTS CF AELYSIS] Point Beach Unit 2 Cycle 15 Reactor Cavity Dosimetry Beed Chain Tag ID: 8-2, 45 degree. Feet [< 4 s/ag of chain f 2/1/90 >] from Imb Mn-54 Co-58 Co40 M141ane Sample # @s/sq 2 sigma dyeWg 2 sipa dpating 2 sigma
+6.5 90-343 1.79 +/- 0.14 3.09 +/- 0.19 32.4 +/- 0.3
+5.5 90-344 4.14 +/- 0.24 7.15 +/- 0.29 66.4 +/- 0.6
+4.5 90-345 6.64 +/- 0.43 10.80 +/- 0.49 92.1 +/- 1.0
+3.5 90-346 7.39 +/- 0.32 12.00+/- 0.47 105.9 +/- 0.8
+2.5 90-347 8.16 +/- 0.38 12.88 +/- 0.43 119.2 +/- 0.9
+1.5 90-348 7.79 +/- 0.28 12.58 +/- 0.35 125.1 +/- 0.8
+0.5 90-349 7.83+/- 0.35 12.51 +/- 0.37 124.9 +/- 0.8
-0.5 90-350 8.05 +/- 0.39 13.16 +/- 0.45 127.1 +/- 0.9
-1.5 90-351 7.89+/- 0.40 12.74 +/- 0.45 123.5 +/- 0.9
-2.5 90-352 7.76 +/- 0.47 12.49 +/- 0.55 119.3 +/- 1.2
-3.5 90-353 7.40 +/- 0.50 11.63 +/- 0.56 109.2 +/- 1.1
-4.5 90-354 7.09 +/- 0.44 10.93 +/- 0.49 90.4 +/- 1.0
-5.5 90-355 4.44 +/- 0.22 7.55+/- 0.25 64.3 +/- 0.5 ,
-4.5 90-356 1.83+/- 0.17 3.02 +/- 0.17 48.9 +/- 0.4 j IEE: For the gamma counts, 6 beads were cut from the chain at the designated points of " Feet from Midplane" (3 beads on each side of the point).
6 beads are about 1.4 inches long and weigh about 1.2 gems. l l l l l l Asmarks: AL File: 13910
References:
Imb.Booke41p74. Ia#35p213-216. proced= = : A-524. Analyst: wir, wet, CnB. Approved: 6d.Afgr,4/% a,tho B-9
l > 1 APPENDIX C l MEASURED SPECIFIC ACTIVITY AND IRRADIATION HISTORY OF REACTOR CAVITY SENSOR SETS - CYCLE 16 In this appendix, the irradiation history as extracted from-NUREG-0020 and the measured specific activities of radiometric
- sensors irradiated in the reactor cavity during Cycle 16 are provided.
The irradiation history of Cycle 16 was as follows: i Cycle No. STARTUP SHUTDOWN COMMENT 3 16 11/24/89 10/06/90 Hf Absorbers Inserted i Ref. Core Power = 1518 MWt THERMAL GENERATION MONTH (MW-Hr) 11/89 141690 12/89 1123322 1/90 1127322 2/90 1018430 3/90 1118045 4/90 1090485 , 5/90 1127775 6/90 1080806
- 7/90 1128256 8/90 1129766 9/90 1120491 10/90 162384
- TOTAL 11368772 The irradiation capsule loading diagram and the measured specific activities of the radiometric monitors from the Cycle 16 irradiation are provided on pages C-2 through C-8. For the multiple foil sensor sets, the individual foil ID can be correlated with the capsule loading diagram provided in Section j 6.2-1 in order to determine the location of the foil within the reactor cavity during irradiation.
C-1 5 1
CONTENTS OF MULTIPLE FOIL SENSOR SETS CYCLE 16 IRRADIATION i CAPSULE ID BARE OR RADIOMETRIC MONITOR ID j and CADMIUM SSTR POSITION SHIELDED fa Hi EM 11 Bh G2 U-238 PACKAGE M-1 B CM -- -- -- -- BG -- PB-4B M-2 Cd DM AM AM M G DG M -- M-3 Cd -- -- -- -- -- -- -- PB-4C N-1 B CN -- -- -- -- BF -- PB-12B N-2 Cd DN AN AN N F DF N -- N-3 Cd -- -- -- -- -- -- -- PB-12C 0-1 B CO -- -- -- -- BE -- PB-5B 0-2 Cd DO A0 A0 0 E DE O -- 0-3 Cd -- -- -- -- -- -- -- PB-5C P-1 B CP -- -- -- -- BD -- PB-138 P-2 Cd DP AP AP P D DD P -- P-3 Cd -- -- -- -- -- -- -- PB-13C Q-1 B CS -- -- -- -- BC -- PB-14B Q-2 Cd DS AS AS S C DC S -- Q-3 Cd -- -- -- -- -- -- -- PB-14C R-1 B CR -- -- -- -- BB -- PB-15B R-2 Cd DT AT AT T A DA T -- R-3 Cd -- -- -- -- -- -- -- PB-15C C-2
Westinghouse Advanood Energy Systems l RERRr Analytical Laboratory - Maltz Mill Site l Esquesti 14245 Originator: S.L. Anderson (W)MID, Energy Center (4-36) Received: 1/14/91 Reported: 3/26/91 [RESULTS OF ANALYSIS] Point Beach Reactor Cavity Dosimetry Lab Dosimeter (6 12/12/90) Foil ID Samplee Material Nuclide dps/ag
- 2 sigma CM 91-254 Fe Mn-54 9.09E+00 +/- 1.3E-01 DM 91-255 Fe Mn-54 9.83E+00 +/- 1.4E-01 CN 91-263 Fe Mn-54 2.04E+01 +/- 2.0E-01 DN 91-264 Fe Mn-54 1.97E+01 +/- 1.9E-01 CO 91-272 Fe Mn-54 9.93E+00 +/- 1.4E-01 DO 91-273 Fe Mn-54 8.85E+00 +/- 1.3E-01 CP 91-281 Fe Mn-54 1.75E+01 +/- 2.0E-01 DP 91-282 Fe Mn-54 1.66E+01 +/- 1.7E-01 CS 91-290 Fe Mn-54 1.47E+01 +/- 1.7E-01 DS 91-291 Fe Mn-54 1.50E+01 +/- 1.7E-01 CR 91-299 Fe Mn-54 1.42E+01 +/- 1.7E-01 DR 91-300 Fe Mn-54 1.40E+01 +/- 1.6E-01 AM 91-257 Cu Co-60 2.79E-01 +/- 4.7E-03 AN 91-266 Cu Co-60 5.84E-01 +/- 8.4E '3 AO 91-275 Cu Co-60 2.51E-01 +/- 5.6E 3 AP 91-284 Cu Co-60 5.10E-01 +/- 6. 4E-' d 3 AS 91-293 Cu Co-60 4.56E-01 +- 7.5E-03 AR 91-302 Cu Co-60 4.49E-01 +- 7.3E-03 BG 91-260 AICo Co-60 1.85E+02 +/- 2.9E+00 DG 91-261 AlCo Co-60 1.32E+02 +/- 2.4E+00 BF 91-269 AlCo Co-60 4.39E+02 +/- 4.6E+00 DF 91-270 AICo Co-60 2.64E+02 +/- 3.5E+00 BE 91-278 AlCo Co-60 2.05E+02 +/- 3.1E+00 DE 91-279 AlCo Co-60 1.33E+02 +/- 2.5E+00 BD 91-287 Alco Co-60 5.63E+02 +/- 5.2E+00 DD 91-288 A1Co Co-60 3.37E+02 +/- 3.3E+00 BC 91-296 AlCo Co-60 5.03E+02 +/- 4.2E+00 DC 91-297 A1Co Co-60 2.73E+02 +/- 3.0E+00 BB 91-305 AICo Co-60 3.13E+02 +/- 2.6E+00 DB 91-306 A1Co Co-60 2.08E+02 +/- 2.0E+00 Remarks:
- Results are in units of dps/(ag of Dosimeter Material).
AL File: 14245
References:
Lab.Bookt 46 pages 103-104. f Procedures: A-524. Analyse nr, me Appraedi ,
, g/ '
C-3
- REW SID WestirrJhru== Advanced Energy Systems l REFGtr Analytical Laboratory - Maltz Mill Site l Rogueste 14245 Originator: S.L. Anderson (W)N!D, Energy Center (4-36)
Received: 1/14/91-Reported: 3/27/91 [RESULTS OF ANALYSIS] Point Beach Reactor Cavity Dosimetry Lab Dosimeter (6 12/12/90) Foil ID Sample # Material Nuclide dps/ag
- 2 sigma M 91-262 0-238 Cs-137 7.63E-01 +/- 2.7?-02 N 91-271 U-238 Cs-137 1.61E+00 +/- 8.7.1-02 0 91-280 0-238 Cs-137 7.30E-01 +/- 3. 9.'-0 2 P 91-289 U-238 Cs-137 1.47E+00 +/- 6.2E M S 91-298 U-238 Cs-137 1.12E+00 +/- 2.2E-02 R 91-307 U-238 Cs-137 1.03E+00 +/- 2.0E-02 M 91-262 U-238 Ru-103 1.26E+01 +/- 2.3E-01 N 91-271 0-238 Ru-103 2.28E+01 +/- 5.2E-01 0 91-280 U-238 Ru-103 1.11E+01 +/- 3.2E-01 P 91-289 U-238 Ru-103 1.92E+01 +/- 4.2E-01 S 91-298 U-238 Ru-106 1.58E+01 +/- 2.0E-01 R 91-307 U-238 Ru-103 1.48E+01 +/- 1.9E-01 M 91-262 U-238 2:-95 1.63E+01 +/- 2.lE-01 N 91-271 U-238 Ir-95 3.10E+01 +/- 4.1E-01 0 91-280 U-238 Er-95 1.52E+01 +/- 2.3E-01 P 91-289 U-238 Ir-95 2.72E+01 +/- 3.8E-01 S 91-298 U-238 Ir-95 2.24E+01 +/- 2.0E-01 R 91-307 U-238 Ir-95 2.07E+01 +/- 1.9E-01 M 91-258 Ti Sc-46 3.50E+00 +/- 8.lE-02 N 91-267 Ti Sc-46 6.94E+00 +/- 9.6E-02 0 91-276 Ti Sc-46 3.16E+00 +/- 6.3E-02 P 91-285 Ti .Sc-46 5.99E+00 +/- 8.8E-02 S 91-294 Ti Sc-46 5.26E+00 +/- 8.3E-02
- R 91-303 Ti Sc-46 5.02E+00 +/- 7.9E-02 AM 91-256 Ni Co-58 1.89E+02 +/- 3.7E+00 I AN 91-265 Ni Co-58 3.67E+02 +/- 5.1E+00
, A0 91-274 Ni Co-58 1.74E+02 +/- 3.5E+00 AP 91-283 Ni Co-58 3.17E+02 +/- 4.8E+00 { As 91-292 Ni Co-58 2.69E+02 +/- 4.4E+00 t AR 91-301 Ni Co-58 2.55E+02 +/- 4.lE+00 l
- Remarks:
- Results are in units of dps/(ag af Dosimeter Material).
- Sample 491-280,Zr-95 data merected.
AL File: 14245
References:
Lab.Booke 49 pages 32-37. . ! Procedures: A-524. 4 . Analystt WIF, 5 Approved: _
*h- --
1 I' 4 C-4
~ . _ ..__ . . . _ . - . . . __ _ _ . _ . . ~ . _ _ .._ _ _ _ _ -
l l i Westinghouse advanced Energy systems , PJIlPCRP ' Analytical Laboratory - Waltz Mill Site Requestl 14220 ! Originator: S.L. Anderson (W)NID, Energy Center (4-36) 4 Received: 1/14/91 Reported: 3/27/91 [RESLTS T ANALYSIS] 1 Point Beach Reactor Cavity Dosimetry i Bead Chain Tag ID: 0 deg. dps/mg of chain 0 12/12/90 Feet [< fran Lab Mn-54 Co-58 Co-60 Midplane Samplef dps/mg 2 signa dps/mg 2 sigma dpa/mg 2 sigre
. +7.5 91-65-A 1.15E+00 +/- 1.5E-01 3.26E+00 +/- 2.7E-01 3.01E+01+/-2.4E-0:
+6.5 91-65-9 3.65E+00 +/- 1.9E 9.89E+00 +/- 4.2E-01 4.31E+01 +/- 2.8E-0:
+5.5 91-65-C 8.87E+00 +/- 3.lE-01 2.41E+01 +/- 7.6E-01 9.86E+01 +/- 4.3E-0:
+4.5 91-65-D 1.40E+01 +/- 5.6E- 01 3.75E+01 +/- 1.2E+00 1.32EM2 +/- 7.0E-0:
+3.5 91-65-E 1.56E+01 +/- 8.0E-01 4.00E+01 +/- 1.8E+00 1.53E+02 +/- 1.1E+0!
+2.5 91-65-F 1.66E+01+/-7.2E-Cl 4.19E+01 +/- 1.8E+00 1.66E+02 +/- 9.9E-0; 4 +1.5 91-65-G 1.58E+01 +/- 7.9E-01 3.89E+01 +/- 1.6E+00 1.61E+02 +/- 9.2E-0;
- +0.5 91-65-H 1.42E+01 +/- 7.2E-01 3.54E+01 +/- 1.7E+00 1.63E+02 +/- 9.8E-01 i -0.5 91-45-I 1.14E+01 +/- 6.5E-01 2.95E+01 +/- 1.5E+00 1.60E+02 +/- 9.7E-0;
-1.5 91-65-J 1.12E+01 +/- 6.6E-01 2.84E+01 +/- 1.4E+00 1.54E+02 +/- 9.6E-0;
-2.5 91-65-K 1.26E+01 +/- 7.9E-01 3.12E+01 +/- 1.6E+00 1.51E+02+/-8.5E-0;
-3.5 91-65-L 1.31E+01+/-6.9E-01 3.57E+01 +/- 1.5E+00 1.55E+02 +/- 9.6E-0;
-4.5 91-65-M 1.30E+01 +/- 7.2E-01 3.33E+01 +/- 1.5E+00 1.27E+02 +/- 7.8E-0;
-5.5 91-65-N 8.83E+00 +/- 5.5E-01 2.37E+01 +/- 1.2E+00 9.07E+01 +/- 7.3E-0:
i -6.5 91-65-0 3.36E+00 +/- 5.0E-01 9.73E+00 +/- 9.9E-01 6.68E+01 +/- 5.7E-0 1 I i
, Remarks:
- Results are in units of dps/(ag of Dosimeter Material).
4 AL File: 14220
References:
Lab.Bookt 49 pages 32-37. Procedures: A-524. Analyst: WT, 'IK Approved:_ e i C-5
l i WestL W = Advanced Energy systems I i REPGE Analytical Laboratory - Waltz Mill Site l Requesti 14220
. i
- Originator
- S.L. Anderson (W)NID, Energy Center (4-36)
Received: 1/14/91 Reported: 3/27/91 [REE0MS T AHRLYSIS] i Point Beach Reactor Cavity Dosimetry t. ! Bead Chain Tag ID: 15 deg. Feet [< @ sang of chain 0 12/12/90 Mn-54 Co-58 Co-60 frein Lab 2 sigtw Midplane Samples dps/mg 2 sigma dps/mg 2 sigma dps/mg
- +7.5 91-66-A 8.48E-01+/-1.7E-01 2.47E+00 +/- 3.954 1 2.71E+01+/-2.9E-0 4 +6.5 91-66-B 2.68E+00 +/- 3.5E-01 7.15E+00 +/- 7.2E-01 4.32E+01+/-4.1E-0
+5.5 91-66-C 6.66E+00 +/- 5.5E-01 1.88E+01 +/- 1.3E+00 1.42E+02 +/- 9.lE-0.
2.03E+02 +/- 1.1E+0t
+4.5 91-66-D 1.07E+01 +/- 7.3E41 2.93E+01 +/- 1.7E+00
+3.5 91-66-E 1.28E+01 +/- 1.4E+00 3.10E+01 +/- 2.6E+00 2.28E+02 +/- 1.7E+0(
i +2.5 91-66-F 1.29E+01 +/- 9.8E-01 3.23E+01 +/- 2.3E+00 2.40E+02 +/- 1.5E+0C
+1.5 91-66-G 1.22E+01 +/- 1.2E+00 3.03E+01 +/- 2.6E+00 2.29E+02+/-1.6E+0C
+0.5 91-66-e 1.llE+01 +/- 9.9E-01 2.86E+01 +/- 1.8E+00 2.17E+02 +/- 1.5E+0C
-0.5 91-66-I 9.90E+00 +/- 1.3E+00 2.49E+01+/-2.4E+00 2.00E+02 +/- 1.5E+0C
- -1.5 91-66-J 9.97E+00+/-6.5E-01 2.58E+01 +/- 1.6E+00 1.91E+02 +/- 1.lE+0C
- -2.5 91-66-K 9.75E+00+/-9.6E-01 2.60E+01 +/- 2.2E+00 1.86E+02 +/- 1.2E+0C 4 -3.5 91-66-L 1.16E+01+/-7.3E-01 2.86E+01 +/- 1.6E+00 1.90E+02 +/- 1.lE+0C j -4.5 91-46-M 1.llE+01 +/- 6.8E-01 2.89E+01 +/- 1.6E+00 1.66E+02 +/- 1.0E+0C
-5.5 ' 91-66-N 6.95E+00 +/- 5.5E-01 1.84E+01 +/- 1.3E+00 1.22E+02 +/- 8.5E-0:
-4.5 91-66-0 2.97E+00+/-4.2E-01 7.37E+00 +/- 8.lE-01 6.52E+01 +/- 5.lE-0:
{ i l I t h 4 I Remarks:
- Results are in units of dps/(mg of Dosimeter Material).
AL File.: 14220
References:
Lab. Books 49 pages 32-37. , Procedures: A-524. Approved g -
. Analyst: WIF, '!K --
1 l C-6 l l
)
l Westinghouse Advanced EnerqY Systems l REKRP l Analytical Laboratory - Waltz Mill Site l Requesti 14220 Originator S.L. Anderson (W)NID, Energy Center (4-36) Received: 1/14/91
, Reported: 3/27/91
[REEELTS OF ANhLYSIS] Point Beach Reactor Cavity namimetry Bead Chain Tag ID: 30 deg. Feet [< dygdag of chain 0 12/12/90 > from Lab Mn-54 Co-58 Co-60 Midplane Sanpled dps/mg 2 sigma aps/mg 2 sigma dpg4ag 2 sigma
+7.5 91-67-A 7.58E-01 +/- 1.1E-01 2.13B+00 +/- 3.1E-01 2.29E+01+/-2.1E-01
+6.5 91-67-B 2.25E+00+/-2.2E-01 5.96E+00 4/- 5.3E-01 3.74E+01 +/- 2.9E-01
+5.5 91-67-< 5.66E+00 +/- 5.3E-01 1.47E+01 +/- 1.0EMO 1.06E+02+/-7.9E-01
+4.5 91-67-0 8.21E+00 +/- 6.1E-01 2.08E+01 +/- 1.4EMO 1.50E+02 +/- 9.4E-01
+3.5 91-47-E 8.59E+00+/-9.0E-01 2.26E+01 +/- 2.0E+00 1.72E+02 +/- 1.1E+00
+2.5 91-67-F 9.49E+00+/-6.7E-01 2.49EMI +/- 1.5EMO 1.86E+02 +/- 1.1E+00
+1.5 91-67-G 9.11E+00+/-9.0E-01 2.39E+01+/-2.0E+00 1.84E+02 +/- 1.2E+00
+0.5 91-47-H 9.54E+00 +/- 6.4E-01 2.40E+01 +/- 1.5E+00 1.88E+02 +/- 1.1E+00
-0.5 91-67-I 9.199+00 +/- 6.7E-01 2.41E+01 +/- 1.5E+00 1.83E+02 +/- 1.0E+00
-1.5 91-67-J 8.91E+00 +/- 9.2E-01 2.27E+01+/-2.1E+00 1.73E+02 +/- 1.1E+00
-2.5 91-67-K 9.13E+00+/-6.9E-01 2.29E+01 +/- 1.6E+00 1.72E+02 +/- 1.0E+00
-3.5 91-67-L 8.27E+00 +/- 6.3E-01 2.19E+01 +/- 1.4E+00 1.58E+02 +/- 9.8E-01
-4.5 91-67-M 7.70E+00 +/- 4.0E-01 2.04E+01+/-9.4E-01 1.24E+02 +/- 5.2E-01
-5.5 '91-67-N 5.10E+00+/-2.3E-01 1.32E+01+/-5.2E-01 7.01E+01 +/- 3.6E-01
-6.5 91-67-0 2.13E+00 +/- 2.1E-01 5.42E+00 +/- 4.2E-01 4.97E+01 +/- 3.0E-01 1
i
- Renarks
- Results are in units of dps/(ag of Dosimeter Material).
i 4 AL File: 14220 References Lab.Booke 49 pages 32-37. , Procedures: A-524. . Analyst: W17, Mt Approved: d l C-7
l l l Nestinghouse Advanood Energy Systems REEORE l Analytical Laboratory - Waltz Mill Site Requesti 14220 Originator: S.L. Anderson (W)NID, Energy Center (4-36) Received: 1/14/91 Reported: V27/91
)
[REELTS & AINJSIS] Point Beach Reactor Cavity Dosimetry Bead Chain Tag ID: 45 deg. Feet [< dpvagofchain 6 12/12/90 >; from Lab un-54 Co-58 Co-60 Midplane Samplef dpe/ag 2 signa dps/ag 2 sigma dpg4sg 2 signa
+7.5 91-68-A 6.41E-01 +/- 9.2E-02 1.78E+00 +/- 1.9E-01 2.05E+01 +/- 1.3E41 M.5 91-68-B 2.04E+00 +/- 1.7E-01 5.23E+00 +/- 3.7E-01 3.20E+01 +/- 2.4E-01 +5.5 91-68-C 4.86E+00 +/- 3.4E-01 1.24E+01 +/- 7.1E-01 6.48E+01+/-3.9E-01 +4.5 91-68-D 7.55E+00 +/- 5.4E-01 1.95E+01 +/- 1.3E+00 8.97E+01 +/- 7.3E-01 +3.5 91-68-E 7.94E+00 +/- 5.9E-01 2.09E+01 +/- 1.3E+00 1.0$E+02 +/- 7.95-01 +2.5 91-68-F 8.91E+00 +/- 4.3E-01 2.24E+01 +/- 9.1E-01 1.13E+02+/-5.15-01 +1.5 91-68-G 8.84E+00 +/- 5.9E-01 2.15E+01 +/- 1.2E+00 1.20E+02 +/- 8.58-01 +0.5 91-68-H 8.78E+00 +/- 6.1E41 2.15E+01 +/- 1.3E+00 1.195+02 +/- 8.4E-01 -0.5 91-68-I 8.65E+^0 +/- 6.7E-01 2.27E+01 +/- 1.5E+00 1.15E+02 +/- 7.4E-01 -1.5 91-68-J 8.71E+00+/-5.8E-01 2.13B+01 +/- 1.1E+00 1.16E+02 +/- 8.3E-01 -2.5 91-68-K 8.34E+00 +/- 6.3E41 2.19E+01 +/- 1.3E+00 1.07E+02 +/- 7.2E-01 -3.5 91-68-L 8.15E+00 +/- 3.5E-01 2.085+01+/-7.2E-01 1.01E+02+/-4.4E-01 -4.5 91-68-M 7.68E+00 +/- 5.0E-01 2.03E+01 +/- 1.4E+00 8.34E+01 +/- 6.3E- 01 -5.5 .91-68-N 5.12E+00 +/- 1.4E-01 1.35E+01 +/- 3.4E-01 6.22E+01 +/- 2.0E-01 -4.5 91-68-0 2.03E+00 +/- 2.0E-01 5.78E+00 +/- 4.5E-01 4.63E+01 +/- 2.7E-01 i
t i Renarks:
- Results are in units of dpe/(ag of Dosimeter Material).
1 AL File: 14220
References:
Lab.Booke 49 pages 32-37. , Proceduress A-524. Analyst: WIT, TK Approved: gh -- - J C-8 s 0
APPENDIX D MEASURED SPECIFIC ACTIVITY AND IRRADIATION HISTORY OF REACTOR CAVITY SENSOR SETS - CYCLE 17 In this appendix, the irradiation history as extracted from. NUREG-0020 and che measured specific cctivities of radiometric sensors irradiated in the reactor cavity during Cycle 17 are provided. ,
. The irradiation history of Cycle 17 was as follows:
Cycle No. STARTUP SHUTDOWN COMMENT 17 11/17/90 09/27/91 Hf Absorbers Inserted Ref. Core Power' = 1518 MWt THERMAL GENERATION MQEM (MW-Hr) , 11/89 405311 l 12/89 1120800 1/90 1123755 2/90 1013700 3/90 1110255 4/90 1087195 5/90 1124595 l 6/90 1080040 7/90 1100758 8/90 1087688 9/90 941710
, TOTAL 11368772 The irradiation capsule loading diagram and the measured specific activities of the radiometric monitors from the Cycle 17 irradiation are provided on pages D-2 through D-9. For the
, multiple foil sensor sets, the individual foil ID can be correlated with the capsule loading diagram provided in Section 6.3-1 in order to determine the location of the foil within the reactor cavity during irradiation. 1 D-1 1
'Jestinghouse Advanced Energy Systems REPORT Analytical Laboratory - Waltz Mill Site Requestf 14477 Originator: S. Anderson (W)NTD, Energy Center Received: 10/18/91 Reported: 12/12/91 POINT SEACH UNIT 2 CYCLE 17 REACTOR CAVITY 0051 METRY Lab Dosimeter (910/24/91)
Foil 10 Sample # Material Nuclide dps/ag
- 2 sigma BG 91-1822 Fe Mn 54 9.82E+00 +/- 1.1E-01 AG 91-1823 Fe Mn 54 1.08E+01+/- 1.1E-01 8H 91-1831 Fe Mn 54 2.30E+01 +/. 1.6E-01 AH 91-1832 Fe Mn 54 2.22E+01 +/. 1.6E-01 AI 91 1840 Fe Mn 54 1.11E+01 +/- 1.2E-01 81 91 1841 Fe Mn 54 9.91E+00 +/- 1.1E 01 BJ 91-1849 Fe Mn 54 1.98E+01 +/- 1.7E-01 AJ 91-1850 Fe Mn 54 1.96E+01 +/- 1.5E-01 BK 91-1858 Fe Mn-54 1.64E+01 +/- 1.5E 01 AK 91-1859 Fe Mn 54 1.63E+01 +/- 1.4E-01 BL 91 1867 Fe Mn 54 1.58E+01 +/- 1.4E-01 AL 91 1868 Fe Mn 54 1.56E+01+/- 1.3E-01 G 91 1824 Ni Co 58 2.83E+02 +/- 1.6E+00 H 91 1833 Ni Co-58 5.51E+02 +/- 2.2E+00 I 91-1842 Ni Co-58 2.61E+02 +/- 1.5E+00 J 91-1851 Ni Co 58 4.80E+02+/- 2.0E+00 K 91-1860 Ni Co 58 4.00E+02 +/- 1.5E+00 L 91 1869 Ni Co 54 3.82E+02+/. 1.8E+00 BG 91 1828 A1Co Co-60 1.95E+02+/- 1.9E+00 AG 91-1829 AICo Co-60 1.40E+02 +/- 1.6E+00 J
BH 91-1837 A1Co Co-60 4.67E+02 +/- 3.0E+00 , AH 91 1838 A1Co Co-60 2.77E+02 +/- 2.3E+00 l A1 91 1846 AICo Co-60 2.13E+02 +/- 2.0E+00 t 81 91-1847 A1Co Co-60 1.40E+02 +/- 1.6E+00 BJ 91-1855 AlCo Co 60 5.89E+02 +/- 3.3E+00 AJ 91 1856 A1Co Co 60 3.35E+02+/- 2.6E+00 8K 91 1864 A1Co Co 60 , 5.25E+02 +/- 3.2E+00
- AK 91 1865 A1Co Co 60 2.89E+02 +/- 2.3E+00
- BL 91 1873 AICo Co 60 3.26E+02 +/. 2.5E+00 AL 91-1874 A1Co Co-60 2.21E+02+/- 2.1E+00 l
- i i Remarks:
- Results are in units of dps/(ag of Dostmeter Material).
i i AL File: 14477 l
References:
Lab Bookf46 pages 246 247 $ Procedures: A 524. Analyst: WTF. TRK, MRK Approved: [ F. i' i E D-2 ; k 'l
1 Westinghouse Advanced Energy Systems l REPORT Analytical Laboratory - WaFtz Mill Site Request # 14477 ' Originator: S. Anderson (W)NTD, Energy Center Received: 10/18/91 Reported: 12/12/91 POINT 8EACH UNIT 2 CYCLE 17 REACTOR CAVITY 005! METRY j i Lab Dostmeter (8 10/24/91) Foil ID dps/ag
- 2 signa Sample # Material Nuclide G 91 1830 U 238 Cs 137 7.59E-01 +/. 2.21E 02 H 91 1839 U 238 Cs 137 1.49E+00 +/. 3.09E 02 I 91 1848 U-238 Cs 137 6.87E.01 +/. 2.34E 02 1 J 91 1857 U 238 Cs 137 1.37E+00 +/.2.77E.02 1 4
K 91 1866 U 238 Cs 137 1.15E+00 +/ 2.52E 02 L 91 1875 U 238 Cs-137 1.01E+00 +/ 2.62E 02 4 G 91 1830 U 238 Ru 103 2.61E+01 +/. 1.09E 01 i H 91 1839 U 238 Ru 103 4.67E+01 +/- 1.52E-01 i I 91 1848 U-238 Ru 103 2.32E+01 +/. 1.18E-01 J 91 1857 U 238 Ru 103 4.22E+01 +/. 1.43E 01
- K 91 1866 U 238 Ru 103 3.52E+01 +/. 1.32E 01 L 91 1875 U 238 Ru 103 3.19E+01 +/. 1.53E-01 l
l G 91-1830 U 238 Zr 95 2.55E+01 +/. 1.25E 01 - H 91 1839 U 238 Zr.95 4.70E+01 +/. 1.75E 01 l I 91 1848 U 238 Zr 95 2.31E+01 +/- 1.35E 01 J 91 1857 U 238 Zr.95 4.38E+01 +/.1.63E 01 ) K 91 1866 U 238 Zr 95 3.69E+01 +/.1.50E-01 i . L 91-1875 U 238 Zr 95 3.23E+01 +/.1.65E01 l , G 91 1826 Tl Sc 46 4.91E+00 +/. 4.79E 02 ! H 91 1835 Tl Sc 46 9.84E+00 +/. 6.94E 02 I 91 1844 Tl Sc 46 4.55E+00 +/. 4.62E 02 J 91 1853 Tl Sc 46 8.68E+00 +/. 6.50E 02 i K 91 1862 Tl Sc 46 7.50E+00 +/.6.02E02 i L 91 1871 Ti Sc 46 7.36E+00 +/. 5.95E 02 Remarks:
- Results are in units of dps/(ag of Dostmeter Material).
4 AL File: 14477
References:
Lab Bookf46 pages 246 247 Procedures: A 524. *
/ 7
. Analyst: WTF, TRK, MRK Approved:
- 1. I D-3
Westinghouse Advanced Energy Systems REPORT Analytical Laboratory - Waltz Mill Site Requesti 14477 Originator: S. Anderson (W)NTD, Energy Center Received: 10/18/91 Reported: 12/12/91 [RESULTSOFANALYSIS) POINT 8EACH UNIT 2 CYCLE 17 REACTOR CAVITY 00$1 METRY Lab Dosimeter (8 10/24/91) Foil ID Samplef Material Nuclide dps/ag
- 2 sigma G 91 1825 Cu Co 60 2.87E.01 t/.3.79E.03 H 91-1834 Cu Co 60 6.14E.01 +/ 5.53E-03 I 91 1843 Cu Co.60 2.66E-01 +/. 3.63E 03 J 91 1852 Cu Co 60 5.44E-01 +/ 5.19E.03 K 91 1861 Cu Co 60 4.61E 01 +/.4.85E.03 L 91 1870 Cu Co 60 4.73E.01 +/. 4.83E.03 I
1 i l l i 1 I Remarks:
- Results are in units of dps/(ag of Dostmeter Material).
AL File: 14477
References:
Lab Sookf46 pages 246 247 "'h l Procedures: A 524. / l Analyst: WTF, TRK, MRK Approved: fl - i b i [)-4 l l i (
.__ _- . . . ~ ._. -. -_ - _ - - . . .- - - _ - .
l i 1 Westinghouse Advanced Energy Systems REPORT Analytical Laboratory . Waltz Mill Site Requesti'14477 Originator: S. Anderson (W) NATO, Energy Center Radiation Engineering & Analysis Received: 10/18/91 Westinghouse Electric Corporation Reported: 2/18/92 POINT BEACH UNIT 2 CYCLE 17 REACTOR CAVITY DOSIMETRY 8ead Chain Tag ID: 0 deg. Feet [< ...----.--..-- dps/mg of chain 9 10/24/91 .... .
-- .>}
from Lab - . . . . - Nn 5 4 . . . - . . . .....-- Co 58 . . -. ------ Co 60 --
. Midplane Sample # dps/mg 2 sigma dps/ag 2 sigma dps/mg 2 sigma
+7.5 91 1818A 1.25E+00 +/- 1.2E.01 4.79E+00 +/- 2.9E 01 2.80E+01 +/. 1.6E-01
+6.5 91 18188 4.13E+00 +/. 2.4E 01 1.55E+01 +/ 5.9E 01 4.0$E+01 +/. 2.7E 01
+5.5 91 1818C 1.03E+01 +/- 3.5E 01 3.71E+01 +/. 8.8E 01 7.52E+01 +/. 3.8E 01
+4.5 91-18180 1.53E+01 +/. 9.9E.01 5.65E+01 +/. 2.2E+00 1.02E+02 +/ 1.0E+00
+3.5 91-1818E 1.71E+01 +/- 1.1E+00 5.90E+01 +/- 2.5E+00 1.18t+02 +/ 1.lE+00
+2.5 91 1818F 1.80E+01 +/. 1.lE+00 6.33E+01 +/ 2.6E+00 1.28E+02 +/.1.1E+00
+1.5 91 1818G 1.72E+01 +/. 1.lE+00 6.01E+01 +/- 2.5E+00 1.27E+02 +/- 1.1E+00
+0.5 91 1818H 1.60E+01 +/. 1.0E+00 5.39E+01 +/. 2.5E+00 1.28t+02 +/. 1.lE+00 0.5 91 18181 1.38E+01 +/ 1.!E+00 4.65E+01 +/- 2.5E+00 1.27E+02 +/ 1.lE+00 1.5 91-1818J 1.36E+01 +/- 1.1E+00 4.50E+01 +/. 2.6E+00 1.23E+02 +/- 1.lE+00 2.5 91 1818K 1.34E+01 +/. 9.4E.01 5.07E+01 +/. 2.4E+00 1.24E+02 +/. 1.lE+00
-3.5 91 1818L 1.44E+01 +/. 9.8E.01 5.20E+01 +/. 2.3E+00 1.22E+02 +/. 1.lE+00 4.5 91 1818M 1.44E+01 +/ 8.9E.01 5.34E+01 +/. 2.4E+00 1.02E+02 +/. 1.0E+00 5.5 91 1818N 1.02E+01 +/- 4.5E.01 3.57E+01 +/ 1.0E+00 6.97E+01 +/- 4.7E.01 6.5 91 18180 4.23E+00 +/ 4.lE.01 1.46E+01 +/. 8.4E.01 6.49E+01 +/- 4.5E 01 :
i Remarlis:
- Results are in units of dps/(eg of Dosimeter Material).
AL File: 14477
References:
Lab Bookf46 pages 246 247 . Procedures: A.524. Analyst: WTF, TRK, MRK Approved: !k; [' g.g *r . C I D.5 4 _. _ . . _ ~ _ . . _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ __ __ ___
, Westinghouse Advanced Energy Systems REPORT - Analytical Laboratory - Waltz Mill Site Requestf 14477 Originator: S. Anderson (W)NATD, Energy Center Radiation Engineering & Analysis Received: 10/18/91 Westinghouse Electric Corporation Reported: 2/18/92 POINT BEACH UNIT 2 CYCLE 17 REACTOR CAVITY 00SIMETRY Bead Chain Tag 10: 15 deg.
Feet [< - -- -----..-- dps/ag of chain 9 10/24/91 ----------->} from Lab ----. - Mn.54 -- ---. ....... Co-58 -...... ....... Co 60 .- . - Midplane Samplef dps/ag 2 sigma dps/ag 2 signa dps/mg 2 sigma
+7.5 91 1819A 1.11E+00 +/- 1.1E-01 4.14E+00 +/- 2.6E.01 2.56E+01 +/. 1.5E 01
+6.5 91-18198 3.10E+00 +/- 2.3E 01 1.15E+01 +/ 5.5E-01 ,4.19E+01 +/- 2.8E-01
+5.5 91-1819C 7.86E+00 +/- 7.1E 01 2.M E+01 +/- 1.8E+00 1.35E+02 +/- 9.0E 01
+4.5 91-18190 1.07E+01 +/. 7.8E 01 4.48E+01 +/- 2.2E+00 1.90E+02 +/- 1.1E+00
+3.5 91 1819E 1.29E+01 +/- 7.0E-01 4.69E+01 +/- 1.9E+00 2.15E+02 +/- 9.3E 01
+2.5 91 1819F 1.43E+01 +/- 1.3E+00 5.09E+01 +/- 3.0E+00 2.26E+02 +/. 1.5E+00
+1.5 91-1819G 1.39E+01 +/- 9.3E-01 4.64E+01 +/. 2.2E+00 2.20E+02 +/- 1.2E+00
+0.5 91 1819H 1.31E+01 +/- 8.9E 01 4.45E+01 +/ 2.2E+00 2.05E+02 +/- 1.1E+00 0.5 91-18191 1.25E+01 +/ 9.0E-01 4.10E+01 +/. 1.9E+00 1.62E+02 +/- 9.9E.01 1.5 91-1819J 1.13E+01 +/- 7.4E 01 3.95E+01 +/- 2.0E+00 1.52E+02 +/- 9.5E 01
-2.5 91 1819K 1.27E+01 +/. 7.7E-01 4.13E+01 +/. 1.8E+00 1.51E+02 +/. 9.6E 01 3.5 91 1819L 1.25E+01 +/- 8.4E 01 4.51E+01 +/ 2.0E+00 1.48E+02 +/ 9.5E 01 4.5 91-1819M 1.25E+01 +/- 7.6E-01 4.17E+01 +/- 1.9E+00 1.28E+02 +/- 8.8E 01 5.5 91 1819N 8.30E+00 +/- 6.7E 01 2.91E+01 +/. 1.5E+00 9.42E+01 +/ 7.5E 01 6.5 91-18190 3.04E+00 +/- 2.4E 01 1.16E+01 +/- 6.2E-01 5.03E+01 +/- 3.1E 01 i
Remarks:
- Results are in units of dps/(eg of Dosimeter Material).
AL Flie: 14477 -
References:
Lab Bookf46 pages 246 247 / Procedures: A-524. f/.* gl /f./g , Analyst: WTF, TRK, MRK Approved: f
/ .
1 D-6 , 1 l l l
Westinghouse Advanced Enerny Systems REPORT Analytical Laboratory - Waftz Mill Site Requesti 14477 Originator: S. Anderson (W)NATD, Energy Center Radiation Engineering & Analysis Received: 10/18/91 Westinghouse Electric Corporation Reported: 2/18/92 (RESULTSOFANALYS!$] P0 INT BEACH UNIT 2 CYCLE 17 REACTOR CAVITY DOSIMETRY J10 8ead Chain Tag 10: 40 deg. Feet [<...........---- dps/ag of chain 9 10/24/91 ------------>] from Lab - --.- Mn 54 ------ ------- Co-58 ---.... ------- Co 60 - ----- Midplane Samplef dps/ag 2 sigma dps/mg 2 signa dps/mg 2 sigma
+7.5 91-1820A f.05E 01 +/- 1.1E 01 3.07E+00 +/- 2.7E-01 2.10E+01 +/ 1.4E-01
+6.5 91-1820B 2.39E+00 +/. 1.5E 01 9.26E+00 +/- 4.0E 01 3.47E+01 +/ 1.8E 01
+5.5 91-1820C 6.06E+00 +/. 5.7E 01 2.09E+01 +/- 1.4E+00
+4.5 91-18200 9.70E+01 +/ 7.7E-01 9.17E+00 +/- 7.7E-01 3.21E+01 t/- 1.8E+00 1.40E+02 +/- 9.3E 01
+3.5 91-1820E 1.08E+01 +/. 8.0E 01 3.56E+01 +/- 1.8E+00 1.64E+02 +/- 9.lE-01
+2.5 91-1820F 1.01E+0! +/- 8.3E-01 3.74E+01 +/- 2.1E+00 1.76E+02 +/- 1.0E+00
+1.5 91-1820G 1.03E+01 +/ 7.9E-01 3.90E+01 +/- 2.3E+00 1.81E+02 +/ 1.0E+00
+0.5 .91-1820H 1.10E+01 +/. 8.3E 01 3.61E+01 +/- 2.1E+00 1.81E+02 +/- 1.0E+00 0.5 91-18201 1.09E+01 +/- 8.7E-01 3.93E+01 t/- 2.1E+00 1.46E+02 +/ 9.4E-01
-1.5 91-1820J 1.04E+01 t/ 7.2E 01 3.70E+01 +/- 1.8E+00 1.42E+02 +/- 9.2E-01
-2.5 91 1820K 1.06E+01 +/ 8.2E 01 3.80E+01 +/- 2.0E+00 1.37E+02 +/- 9.1E-01
-3.5 91-1820L 9.93E+00 +/- 7.1E 01 3.49E+01 t/- 1.9E+00 1.26E+02 +/- 8.8E-01
-4.5 91-1820M 9.42E+00 +/. 6.7E-01 3.32E+01 +/- 1.7E+00 1.02E+02 +/ 7.8E 01 5.5 91-1820N 5.71E+00 +/- 3.2E-01 2.19E+01 +/- 7.0E 01 5.51E+01 +/- 3.2E 01
-6.5 91-18200 2.30E+00 +/ 2.1E-01 8.87E+00 +/- 5.1E 01 - 3.89E+01 +/ 2.7E 01 3
5 l 1 ) i + Remarks:
- Results are in units of dps/(ag of Dosimeter Material).
AL File: 14477 -
References:
Lab Bookf46 pages 246 247 "5 Procedures: A 524. Analyst: WTF TRK, MRK Approved: (s' . g.,y fy 4 D-7
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originator: S. Anderson (W)NATD, Energy Center Radiation Engineering & Analysis Received: 10/18/91 < Westinghouse Electric Corporation Reported: 2/18/92 (RESULTSOFANALYSIS] POINT BEACH UNIT 2 CYCLE 17 REACTOR CAVITY 00$1 METRY 8ead chain Tag 10: 45 deg.
- Feet (<........-...... dps/ag of chain 9 10/24/91 .....--.-...->]
, . from Lab ....... Mn.54 .... .. ....... Co.58 ....... ....... Co 60 ....... Midplane Samplef dps/ag 2 sigma dps/ag 2 signa dps/ag 2 signa
+7.5 91 1821A 7.41E-01 +/. 9.0E-02 2.43E+00 +/- 1.9E.01 2.01E+01 +/- 1.2E 01
+6.5 91 18218 2.22E+00 +/. 1.2E-01 8.10E+00 +/- 2.8E-01 2.97E+01 +/ 1.5E-01 i +5.5 91 1821C 5.15E+00 +/. 2.9E 01 1.82E+01 +/- 7.0E.01 6.00E+01 +/. 3.4E-01
+4.5 91 18210 8.05E+00 +/. 3.5E 01 2.88E+01 +/. 8.3E.01 8.04E+01 +/- 3.9E-01
+3.5 91 1821E 9.27E+00 +/. 5.8E 01 3.23E+01 +/. 1.7E+00 9.63E+01 +/- 7.6E 01
+2.5 91 1821F 1.03E+01 +/. 7.2E 01 3.69E+01 +/. 1.9E+00 1.07E+02 +/. 8.0E 01
, +1.5 91 1821G 9.59E+00 +/. 6.5E.01 3.52E+01 +/. 1.8E+00 1.13E+02 +/- 8.3E 01
+0.5 91 1821H 9.36E+00 +/. 6.4E 01 3.31E+01 +/. 1.7E+00 1.13E+02 +/. 8.2E 01 0.5 91 18211 1.08E+01 +/. 6.5E-01 3.58E+01 +/.1.6E+00 9.42E+01 +/. 7.5E 01 1.5 91-1821J 1.00E+01 +/. 6.0E 01 3.62E+01 +/. 1.7E+00 9.26E+01 +/ 7.5E 01 2.5 91-1821K 1.05E+01 +/. 7.5E.01 3.51E+01 +/. 1.5E+00 8.86E+01 +/. 7.3E-01 3.5 91 1821L 9.67E+00 +/. 6.2E 01 3.38E+01 +/. 1.6E+00 8.25E+01 +/- 7.0E-01 4.5 91 1821M 8.53E+00 +/. 3.5E-01 2.99E+01 +/. 8.5E-01 6.73E+01 +/. 3.6E.01 5.5 91-1821N 5.93E+00 +/. 2.7E-01 1.02E+01 +/- 1.2E+00 2.17E+01 +/- 6.9E.01 6.5 91 18210 2.60E+00 +/. 1.5E 01 9.18E+00 +/- 3.8E.01 3.69E+01 +/. 1.9E-01 l
l Remarks:
- Results are in units of dps/(ag of Dosimeter Material).
AL File: 14477 l
References:
Lab Sookf46 pages 246 247 -
/
Procedures: A 524. i
$, g q//gf <
Analyst: WTF, TRK, MRK Approved: l 1 D-8 l l 1
APPENDIX E MEASURED SPECIFIC ACTIVITY AND IRRADIATION HISTORY OF REACTOR CAVITY SENSOR SETS - CYCLES 18 THROUGH 20 In this appendix, the irradiation history as extracted from NUREG-0020 and the measured specific activities of radiometric sensors irradiated in the reactor cavity during Cycles 18 through 20 are provided. The irradiation history of Cycles 18 through 20 was as follows: Cycle No. STARTUP SHUTDOWN COMMENT 18 11/14/91 09/26/92 Hf Absorbers Inserted 19 11/18/92 09/25/93 Hf Absorbers Inserted 20 10/31/93 09/24/94 Hf Absorbers Inserted Ref. Core Power = 1518 MWt THERMAL THERMAL THERMAL GENERATION GENERATION GENERATION MONTH (MW-Hr) MONTH (MW-Hr) MONTH (MW-Hr) 11/91 475898 11/92 393930 10/93 13664 12/91 1064971 12/92 1100982 11/93 1044717 1/92 1127958 1/93 1116955 12/93 1112456 2/92 1056750 2/93 1003337 1/94 1113719 3/92 1050400 3/93 1065780 2/94 992516 1 ^ 4/92 1084547 4/93 1073515 3/94 1115386 l 5/92 1126747 5/93 1121723 4/94 1082174 6/92 1126747 6/93 1071687 5/94 1114573 7/92 1124583 7/93 1116051 6/94 1092180 8/92 1100487 8/93 1118139 7/94 1126859 9/92 927929 9/93 866899 8/94 1129707 4 9/94 821279 TOTAL 11267017 TOTAL 11048998 TOTAL 11759230 The irradiation capsule loading diagram and the measured specific i activities of the radiometric monitors from the Cycles 18 through 1 20 irradiation are provided on pages E-2 through E-9. For the l multiple foil sensor sets, the individual foil ID can be correlated with the capsule loading diagram provided in Section 6.4-1 in order to determine the location of the foil within the reactor cavity during irradiation. t E-1 { 4
. . - . ~ , -, -
I
- CONTENTS OF MULTIPLE FOIL SENSOR SETS l CYCLES 18/20 IRRADIA*110N a
i : 1 Capsula I.D. Bare or - and Cadmium Radiometric Monitor I.D. SSTR Position Shielded fg gi G Ii HD, Q M-g}g Packnoe . M-1 8 AA - - - - 8A - PS-398 M-2 Cd AK AA AA AK AA AA AA - l m-3 Cd - - - - - - - PB-39C i 4 W-1 8 AB - - - - 88 - PB-41B NN-2 Cd AL AB AB AL A8 A8 AB - NN-3 Cd - - - - - - - PB-41C 1 00-1 B AC - - - - 8C - PB-408 AC - 00-2 Cd AN AC AC AN AC AC . 00-3 Cd - - - - - - - PB-40C ! PP-1 B AD - - - - BD - PB-428 t AN AD AD -
- PP-2 Cd AN AD AD AD PP-3 Cd - - - - - - -
PB-42C QQ-1 B AE - - - - BE - PB-438 QQ-2 Cd A0 AE AE AD AE AE AE - l QQ-3 Cd - - - - - - - PB-43C I RR-1 8 AF - - - - BF - PB-448 i RR-2 Cd AP AF AF AP AF AF T - RR-3 Cd - - - - - - - PB-44C e E-2
i-e l West 4a N e Electric Corporatico REIORT Chemistry Operations Technology & Analysis Requests 15576 Waltz Mill Site ? Originator: A.H. Fero (W)lrTD, Energy Center (4-36) 4 ' Received: 12/12/94
- Reported: 3/8/95 (RESULTS OF ANALYSIS] !
- Point Beach Reactor Cavity Dosimetry '
j Lab Dosimeter (G 2/20/95) Foil ID Sample # Material Nuclide -dye /as
- 2 sissa AA 95-016 Fe b-54 1.00E+01 +/- 1.2F-01 AK 95-017 Fe Mn-54 1.12E+01 +/- 1.2E-01 AB 95-025 Fe th-54 2.85E+01 +/- 2.6E-01 i AL 95-026 Fe Mn-54 2.78E+01 +/- 2.4E-01 AC 96-034 Fe Mn 1.21E+01 +/- 1.7E-01 1 95-035 AM Fe Mn-54 1.08E+01 +/- 1.5E-01 AD 95-043 Fe Mn-54 2.38E+01 +/- 2.5E-01 AN 95-044 Fe Mn-54 2.43E+01 +/- 2.6E-01
, AE 95-052 Fe Mn-54 -2.08E+01 +/- 2.4E-01 A0 95-053 Fe Mn-54 2.07E+01 +/- 2.2E-01 AF- 95-061 Fe Mn-54 1.95E+01 +/- 2.2E-01 AP 95-062 Fe Mn-54 1.98E+01 +/- 2.12-01 AA 95-019 Cu Co-60 5.97E-01 +/- 1.8E-02 AB 95-028 Cu Co-60 1.57E+00 +/- 2.9E-02 _l AC 95-037 Cu Co-60 5.49E-01 +/- 1.7E-02 i AD 95-046 Cu Co-60 1.39E+00 +/- 2.8E-02 95-055 ( AE Cu Co-60 1.24E+00 +/- 2.7E-02 i AF 95-064 Cu Co-60 1.22E+00 +/- 2.7E-02 l BA 95-022 A1Co Co-60 4.49E+02 +/- 5.4E+40 , AA 95-023 AICo Co-60 3.23E+02 +/- 4.6E+00 4 BB 95-031 A1Co Co-60 1.13E+03 +/- 1.4E+01 l AB 95-032 AlCo Co-60 6.90E+02 +/- 1.1E+01 BC 95-040 A1Co Co-60 4.73E+02 +/- 5.6E+00 AC 95-041 A1Co Co-60 3.17E+02 +/- 4.5E+00 BD 95-049 AlCo Co-60 1.41E+03 +/- 1.6E+01 I f 95-050 AD AlCo Co-60 8.13E+02 +/- 1.2E+01 i BE 95-058 A1Co Co-60 1.26E+03 +/- 1.5E+01 - AE 95-059 A1Co Co-60 7.01E+02 +/- 1.1E+01 BF 95-067 A1Co Co-60 8. 01E+42 +/- 1.2E+01 AF 95-068 AlCo Co-60 5.65E+02 +/- 9.6E+00 Remarks:
- Results are in units of dPs/(ag of Dosimeter Material).
,. / /
AL File: 15576 t Procedures: A-524. Analyst: WTF, TRK Approved: / I. fig / ( /v34
/
E-3
West 4= M m Electric Corporation REPORT Chemistry Operations Technologr & Analysis Requests 15576 Waltz Mill Site Originator: A.H. Fero (W)NTD. Energy Center (4-36) Received: 12/12/94 Reported: 3/8/95 [RESULTS OF ANALYSIS] Point Beach Reactor Cavity Dosimetry Lab Dosimeter (G 2/20/95) Foil ID Sample # Material Nuclide dps/ag
- 2 signa AK 95-020 Ti So-46 1.60E+00 +/- 4.98-02 AL 95-029 Ti Sc-46 3.94E+00 +/- 7.9E-02 -
AM 95-038 Ti Sc-46 1.58E+40 +/- 4.9E-02 A4 95-047 Ti Sc-46 3.34E+00 +/- 7.38-02 A0 95-056 Ti Sc-46 3.01E+00 +/- 6.9E-02 AP 95-065 Ti Sc-46 2.86E+00 +/- 6.7E-02 AA 95-018 Ni C0-58 7.41E+01 +/- 1.0E+00 AB 95-027 Ni CO-58 1.74E+02 +/- 1.6E+00 AC 95-036 Ni CO-58 7.18E+01 +/- 1.0E+00 AD 95-045 Ni CO-58 1.48E+02 +/- 1.4H+00 AE 95-054 Ni CO-58 1.28t+02 +/- 1.4E+00 AF 95-063 Ni CO-58 1.23E+02 +/- 1.3E+00 Remarks:
- Results are in units of dps/(ma of Dosimeter Material).
AL File: 15576 '\ // Procedures: A-524. Analyst: WTF. TRK
,)> / / . /v 4 7/'5p(---
Approved: I
/
E-4
I I i ! Westinghouse Electric Corporation j REPORT Chemistry Operations Technology & Analysis Requests 15576 Waltz Mill Site 1 l Orjginator: A.H. Fero (W)NTD. Energy Center (4-36) l Received: 12/12/94 Reported: 3/8/95 [RESULTS OF ANALYSIS) Point Beach Reactor Cavity Dosimetry ; Lab Dosiseter (9 2/20/95) Foil ID Samples Material Nuclide dpe/mg
- 2 sissa AA 95-024 U-238 Cs-137 1.79E+00 +/- 8.5E-02 i AB 95-033 U-238 Co-137 4.53E+00 +/- 1.1E-01 l AC 95-042 U-238 Co-137 1.73E+00 +/- 6.7E-02 AD 95-051 U-238 Ca-137 4.13E+00 +/- 9.2E-02 AE 95-060 U-238 Ca-137 3.37E+00 +/- 1.0E-01 T 95-069 U-238 Ca '37 3.14E+00 +/- 9.7E-02 AA 95-024 U-238 Ru-103 2.51E+00 +/- 1.1E-01 AB 95-033 U-238 Ru-103 5.63E+00 +/- 1.3E-01 AC 95-042 U-238 Ru-103 2.44E+00 +/- 9.2E-02
'AD 95-051 U-238 Ru-103 4.95E+00 +/- 1.4E-01 AE 95-060 U-238 Ru-103 4.14E+00 +/- 1.4E-01 T 95-069 U-238 Ru-103 3.92E+00 +/- 1.2E-01 AA 95-024 U-238 Zr-95 5.78E+00 +/- 1.7E-01 AB 95-033 U-238 Zr-95 1.35E+01 +/- 2.3E-01 AC 95-042 U-238 Zr-95 5.66E+00 +/- 1.6E-01 AD 95-051 U-238 Zr-95 1.22E+01 +/- 2.3E-01 AE 95-060 U-238 Zr-95 9.89E+00 +/- 2.3E-01 4 T 95-069 U-238 Zr-95 9.58E+00 +/- 2.2E-01 d
Remarks:
- Results are in units of dpe/(mg of Dosimeter Material).
1 , AL File: 15576 Procedures: A-524. - 4 4t4 / Analyst: RTF, TRK Approved:
- E-5
l l West 4a s - -- Electric Corporation 4 REPORT Chemistry Operations Technology & Analysis Requests 15576 Waltz Mill Site Originator: A.H. Fero (W)NTD, Energy Center (4-36) Received: 12/12/94 Reported: 3/25/95 l , [RESULTS OF ANALYSIS) Point Beach Reactor Cavity Dosimetry Bead Chain Tag ID: O deg. Feet (< dpe/ag of chain 0 2/20/95 ->] from Iab Mn-54 Co-58 Co-60 Midplane Sample # dpe/as 2 signa dpe/ag 2 signa dpe/ag 2 signa
+7.5 95-012-A 1.08E+00 +/- 1.6E-01 1.05E+00 +/- 1.7E-01 6.32E+01 +/- 4.7E-01 l +0.5 95-012-B 4.15E+00 +/- 2.9E-01 3.53E+00 +/- 3.3E-01 9.23E+01 +/- 7.1E-01
, +5.5 95-012-C 1.06E+01 +/- 6.5E-01 9.72E+00 +/- 7.9E-01 2.10E+02 +/- 1.5E+00
+4.5 95-012-D 1.81E+01 +/- 8.6E-01 1.55E+01 +/- 9.3I-01 2.95E+02 +/- 2.0E+00
+3.5 95-012-E 2.05E W 1 +/- 1.1E+00 1.90E+01 +/- 1.2E+00 3.43E+02 +/- 2.2E+00
+2.5 95-012-F 2.26E+01 +/- 1.1E+00 2.01E+01 +/- 1.2E+00 3.83E+02 +/- 2.3E+00
+1.5 95-012-G 2.23E+01 +/- 1.5E+00 1.93EM1 +/- 1.5E+00 3.85E+02 +/- 3.2I+00
+0.5 95-012-H 1.97E+01 +/- 1.1E+00 1.77E+01 +/- 1.2E+00 3.95E+02 +/- 2.7E+00
-0.5 95-012-I 1.63E+01 +/- 1.0E+00 1.50E+01 +/- 1.2E+00 3.96E+02 +/- 2.6E+00
-1.5 95-012-J 1.55E+01 +/- 1.0E 4 0 1.32E+01 +/- 1.1E+00 3.82E+02 +/- 2.5E+00
-2.5 95-012-K 1.71E+01 +/- 1.1E+00 1.51E+01 +/- 1.2E+00 3.76E+02 +/- 2.5E+00
-3.5 95-012-L 1.88E+01 +/- 1.1E+00 1.68E+01 +/- 1.1E+00 3.59E+02 +/- 2.2E+00
-4.5 95-012-M 1.69E+01 +/- 8.0E-01 1.47E+01 +/- 9.7E-01 2.93E+02 +/- 2.0E+00
-5.5 95-012-N 1.06E+01 +/- 4.5E-01 9.52E+00 +/- 4.5E-01 1.89E+02 +/- 1.0E+00
-6.5 95-012-0 3.49E+00 +/- 3.4E-01 3.35E+00 +/- 3.5E-01 1.42E+02 +/- 8.8E-01 l
l 1 s i Remarks:
- Results are in units of dps/(mg of Dosimeter Material).
t AL File: 15576 ' Procedures: A-524 Analyst: WTF. TEK Approved:
/
/
E-6
5 Westinghouse Electric Corporation REPORT Chemistry Operations Technology & Analysis Requests 15576 Waltz Mill Site l l Originator: A.H. Fero (W)lfrD, Energy Center (4-36) l Received: 12/12/94 l Reported: 3/25/95 l (RESULTS OF ANALYSIS] Point Beach Reactor Cavity Dosimetry Bead Chain Tag ID: 15 deg. Feet (< dpe/ag of chain 9 2/20/95 >} from Lal Hn-54 Co-58 Co-60 Midplane Sample # dpe/ag 2 signa dpe/mg 2 signa dps/mg 2 sigma
+7.5 95-013-A 1.11E+00+/-1.5E-05 9.14E-01 +/- 1.7E-01 5.84E+01 +/- 4.1E-01
+6.5 95-013-B 3.11E+00 +/- 4.0E-01 2.88E+00 +/- 4.3E-01 9.63E+01 +/- 9.3E-01 <
+5.5 95-013-C 8.60E+00 +/- 8.3E-01 7.48E+03 +/- 9.0E-01 3.09E+02 +/- 2.1E+00 I
+4.5 95-013-D 1.56E+01 +/- 1.6E+00 1.26E+01 +/- 1.5E+00 4.61E+02 +/- 3.6E+00
+3.5 95-013-E 1.59E+01 +/- 1.2E+00 1.49E+01 +/- 1.4E+00 5.39E+02 +/- 3.0E+00
+2.5 95-013-F 1.84E+01 +/- 1.4E+00 1.54E+01 +/- 1.4E+00 5.64E+02 +/- 3.1E+00
+1.5 95-013-G 1.68E+01 +/- 1.7E+00 1.50E+01 +/- 1.8E+00 5.49E+02 +/- 3.9E+00
+0.5 95-013-H 1.66E+01 +/- 1.5E+00 1.36E+01 +/- 1.9E+00 5.11E+02 +/- 3.8E+00
. -0.5 95-013-I 1.45E+01 +/- 1.5E+00 1.39E+01 +/- 2.2E+00 4.88E+02 +/- 3.7E+00
-1.5 95-013-J 1.43E+01 +/- 1.6E+00 1.24E+01 +/- 1.8E+00 4.61E+02 +/- 3.6E+00
-2.5 95-013-K 1.65E+01 +/- 1.7E+00 1.31E+01 +/- 1.8E+00 4.52E+02 +/- 3.5E+00
-3.5 95-013-L 1.66E+01 +/- 1.7E+00 1.50E+01 +/- 1.7E+00 4.35E+02 +/- 3.5E+00
-4.5 95-013-M 1.30E+01 +/- 9.7E-01 1.20E+01 +/- 1.2E+00 3.59E+02 +/- 2.3E+00
-5.5 95-013-N 8.50E+00 +/- 5.0E-01 7.88E+00 +/- 5.8E-01 2.50E+02 +/- 1.2E+00 !
-6.5 95-013-0 3.09E+00 +/- 3.3E-01 3.21E+00 +/- 4.3E-01 1.31E+02 +/- 8.5E-01 l
l l l ) - l Remarks:
- Resulta are in units of dps/(ms of Dosimeter Material).
l I
. 1 AL File: 15576 / J Procedures: A-524 f A4t, & y Analyst: WTF, TRK Approved:
/
E-7 l i
Westinghouse Electric Corporation REPORT Chemistry Operations Technology & Analysis Requesta 15576 Waltz Mill Site Originator: A.H. Fero (W)NTD, Energy Center (4-36) Received: 12/12/94 Reported: 3/25/95 Y (RESULTS OF ANALYSIS] Point Beach Reactor Cavity Dosimetry Bead Chain Tag ID: 30 deg. Feet (< dpe/ag of chain e 2/20/95 >] from Lab - - - Mn-54 Co-58 Co-60 Midplane Sample # dps/ag 2 sissa dpe/ag 2 sissa dps/ag 2 signa
+7.5 95-014-A 2.90E+00 +/- 4.0E-01 2.50E+00 +/- 5.1E-01 1.53E+02 +/- 1.2E+00
+6.5 95-014-B 8.32E+00 +/- 6.8E-01 6.62E+00 +/- 6.7E-01 2.53E+02 +/- 1.5E+00
+5.5 95-014-C 6.29E+00 +/- 5.5E-01 5.72E+00 +/- 6.0E-01 2.25E+02 +/- 1.4E+00
+4.5 95-014-D 1.08E+01 +/- 8.7E-01 9.44E+00 +/- 1.2E+00 3.33E+02 +/- 2.2E+00
+3.5 95-014-E 1.30E+01 +/- 1.4E+00 1.15E+01 +/- 1.7E+00 4.06E+02 +/- 3.3E+00
+2.5 95-014-F 1.29E+01 +/- 1.4E+00 1.15E+01 +/- 1.4E+00 4.46E+02 +/- 3.5E+00
+1.5 95-014-G 1.24E+01 +/- 1.3E+00 1.21E+01 +/- 1.7E+00 4.52E+02 +/- 3.5E+00
+0.5 95-014-H 1.34E+01 +/- 1.4E+00 1.23E+01 +/- 1.6E+00 4.50E+02 +/- 3.5E+00
-0.5 95-014-I 1.33E+01 +/- 1.3E+00 1.19E+01 +/- 1.5E+00 4.34E+02 +/- 3.5E+00
-1.5 95-014-J 1.40E+01 +/- 1.4E+00 1.19E+01 +/- 1.6E+00 4.27E+02 +/- 3.4E+00
-2.5 95-014-K 1.25E+01 +/- 1.2E+00 1.06E+01 +/- 1.6E+00 4.09E+02 +/- 2.7E+00
-3.5 95-014-L 1.34E+01 +/- 1.3E+00 9.62E+00 +/- 1.3E+00 3.66E+02 +/- 2.3E+00
-4.5 95-014-M 9.54E+00 +/- 1.1E+00 8.49E+00 +/- 1.2E+00 2.84E+02 +/- 2.0E+00
, -5.5 95-014-N 6.66E+00 +/- 6.3E-01 5.28E+00 +/- 6.7E+00 1.52E+02 +/- 1.2E+00 t
-6.5 95-014-0 2.44E+00 +/- 4.4E-01 2.21E+00 +/- 5.6E-01 1.04E+02 +/- 9.9E-01 Remarks:
- Results are in units of dps/(as of Dosimeter Material).
AL File: 15576 Procedures: A-524. Analyst: WTF. TRK Approved: h' 44"
/
E-8
I West 4 W == Electric Corporation REPORT Chemistry Operations Technologr & Analysis Request # 15578 Waltz Mill Site Originator: A.H. Fero (H)trfD, Energy Center (4-36) Received: 12/12/94 Reported: 3/25/95 [RESULTS OF ANALYSIS) Point Beach Reactor Cavity Dosimetry Bead Chain Tag ID: 45 deg.
. Feet [< dps/ag of chain 0 2/20/95 >l from Lab Mn-54 Co-58 Co-60 Midplane Samples dps/ag 2 signa dpe/ag 2 sissa dpe/ag 2 sissa
+7.5 95-015-A 8.68E-01 +/- 1.3E-01 8.88H-01 +/- 1.8E-01 4.59E+01 +/- 3.6E-01
+6.5 95-015-B 2.44E+00 +/- 1.8E-01 2.17E+00 +/- 2.21-01 6.87E+01 +/- 4.5E-01
+5.5 95-015-C 6.09E+00 +/- 5.1E-01 5.03E+00 +/- 5.8E-01 1.46E+02 +/- 1.2E+00
+4.5 95-015-D 9.90E+00 +/- 5.8E-01 8.40E+00 +/- 7.3E-01 1.97E+02 +/- 1.4E+00 1
+3.5 95-015-E 1.05E+01 +/- 6.6E-01 9.91E+00 +/- 7.6E-01 2.37E+02 +/- 1.5E+00
+2.5 95-015-F 1.27E+01 +/- 8.2E-01 1.15E+01 +/- 1.7E+00 2.71E+02 +/- 1.9E+00
+1.5 95-015-G 1.23E+01 +/- 8.8E-01 1.00E+01 +/- 1.1E+00 2.85E402 +/- 2.0E+00
+0.5 95-015-H 1.26E+01 +/- 9.3E-01 9.85E+00 +/- 1.0E+00 2.83E+02 +/- 2.0E+00
-0.5 95-015-I 1.28E+01 +/- 8.8E-01 1.13E+01 +/- 1.1E+00 2.84E+02 +/- 2.0E+00
-1.5 95-015-J 1.21E+01 +/- 8.3E-01 1.01E+01 +/- 8.9E-01 2.80E+02 +/- 2.0E+00
-2.5 95-015-K 1.19E+01 +/- 9.2E-01 1.13E+01 +/- 1.3E+00 2.69E+02 +/- 2.0E+00
-3.5 95-015-L 1.17E+01 +/- 7.1E-01 1.03E+01 +/- 9.0E-01 2.43E+02 +/- 1.5E+00
-4.5 95-015-M 9.50E+00 +/- 6.4E-01 8.74E+00 +/- 6.9E-01 1.97E+02 +/- 1.4E+00
-5.5 95-015-N 6.10E+00 +/- 4.9E-01 5.471+00 +/- 5.7E-01 1.36E+02 +/- 1.1E+00 ;
-6.5 95-015-0 2.33E+00 +/- 3.0E-01 2.29E+00 +/- 3.4E-01 1.01E+02 +/- 7.5E-01 j l
Remarks:
- Resulb are in units of dps/(as of Dosimeter Material).
Proce e 524 t Analyst: WTF. TRK Approved: / r i l E-9}}