ML20077K074
| ML20077K074 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/30/1992 |
| From: | Shaun Anderson WISCONSIN ELECTRIC POWER CO. |
| To: | |
| Shared Package | |
| ML20077K067 | List: |
| References | |
| WCAP-12795, WCAP-12795-R02, WCAP-12795-R2, NUDOCS 9501100152 | |
| Download: ML20077K074 (231) | |
Text
{{#Wiki_filter:r s WCAP-12795, Rev. 2 i i s WESTINGHOUSE CLASS 3 REACTOR CAVITY NEUTRON MEASUREMENT PROGRAM FOR WISCONSIN ELECTRIC POWER COMPANY POINT BEACH UNIT 2 Stanwood L. Anderson Arnold H. Fero September 1992 Work performed under Shop Order No. WLKP-450
-s APPROVED: [ [""-
F. L. Lau, Manager , Radiation and Systems Analysis ; t Prepared by Westinghouse for the Wisconsin Electric Power Company i Purchase Order No. C-46250-C l 1
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WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Business Unit 1 P.O. Box 355 l Pittsburgh, Pennsylvania 15230 , l l l l I
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- 1992 WESTINGHOUSE ELECTRIC CORPORATION 9501100152 950105 PDR ADOCK 05000301
._ P ,_ . _ _ . PDR_ l
L ( EXECUTIVE
SUMMARY
At the conclusion of Fuel Cycle 14, a reactor cavity measurement program was instituted at Point Beach Unit 2 to provide continuous monitoring of ! the beltline region of the reactor pressure vessel and reactor vessel support structure. The use of the cavity measurement program coupled with available surveillance capsule measurements provides a plant specific data
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l base that enables the evaluation of the vessel exposure and the . uncertainty associated with that exposure over the service life of the unit. , During the first cycle of irradiation (Cycle 15), the reactor was operating with a conventional low leakage fuel management strategy. At l 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, and 17 cavity measurements demonstrated
. the following incremental flux reduction for the circumferential weld. ,
MEASURED CAVITY FLUX (E > 1.0 MeV) l [n/cm2-sec] i CYCLE 15 CYCLE 16 16/15 CYCLE 17 17/15 0 Degrees 1.85E+09 1.28E+09 0.692 1.37E+09 0.741 15 Degrees 1.67E+09 1.22E+09 0.731 1.15E+09 0.689 30 Degrees 1.21E+09 9.32E+08 0.770 1.05E+09 0.868 45 Degrees 1.08E+09 9.52E+08 0.881 9.84E+08 0.911
- 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 and 17) 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: ; t l
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NEUTRON FLUENCE (E > 1.0 MeV) -t [n/cm2) { r v, 32 EFPY 48 EFPY Beltline Circumferential Weld 2.52E+19 3.39E+19 ! Intermediate Shell Forging 2.88E+19 4.12E+19 ,
, Lower Shell Forging 2.62E+19 3.64E+19 *l Upper / Int. Shell Cire. Weld 3.70E+18 5.00E+18 Lower Shell/ Head Cire. Weld < l.00E+17 < l.00E+17 As further data are accumulated from subsequent 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.
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3' TABLE OF CONTENTS '" i LIST OF FIGURES iii LIST OF TABLES v 1.0 OVERVIEW 0F THE PROGRAM 1-1 !
2.0 DESCRIPTION
OF THE MEASUREMENT PROGRAM 2-l' 2.1 Description of Reactor Cavity Dosimetry 2-1 2.2 Description of Surveillance Capsule Dosimetry 2-8 3.0 NEUTRON TRANSPORT AND 00SIMETRY EVALUATION METHODOLOGY 3-1 3.1 Neutron Transport Analysis Methods 3-1 3.2 Neutron Dosimetry Evaluation Methodology 3-8 4.0 RESULTS OF NEUTRON TRANSPORT CALCULATIONS 4-1 4.1 Reference Forward Calculation 4-1 4.2 Fuel Cycle Specific Adjoint Calculations 4-15 5.0 EVALUATIONS OF SURVEILLANCE CAPSULE DOSIMETRY 5-1 I 5.1 Measured Reaction rates 5-1 5.2 Results of the Least Squares Adjustment Procedure 5-2 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 Cycle 17 Results 6-42 7.0 COMPARIS0N OF CALCULATIONS WITH MEASUREMENTT 7-1 7.1 Comparison of Least Squares Adjustment ks.ults 7-2 with Calculation i
w . a ..v. : : , .a ;u. 7.2 Comparisons of Measured and Calculated Sensor .7-3 , Reaction Rates 8.0 BEST ESTIMATE NEUTRON EXPOSURE OF PRESSURE. VESSEL MATERIALS 8-1 8.1 Exposure Distributions Within the Beltline Region 8-1 f
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8.2 Exposure of Specific Beltline Materials 8-17 f' 8.3 Uncertainties in Exposure Projections 8-27
9.0 REFERENCES
9-1 APPENDIX A MEASURED SPECIFIC ACTIVITY AND IRRADIATION HISTORY A-1 0F SURVEILLANCE CAPSULE SENSOR SETS APPENDIX B MEASURED SPECIFIC ACTIVITY AND IRRADIATION HISTORY B-1 ! 0F REACTOR CAVITY SENSOR SETS 9
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LIST OF FIGURES
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l Fiaure Title Paae i 2.1-1 Azimuthal Location of Sensor Strings 2-5 2.1-2 Axial Location of Multiple Foil Sensor Sets 2-6 2.1-3 Irradiaticn Capsule for Cavity Sensor Sets 2-7 2.2-1 Neutron Sensor Locations Within Internal 2-9 ; Surveillance Capsules
, 3.1-1 Reactor Geometry Showing a 45 Sector 3-6 .
3.1-2 Internal Surveillance Capsule Geometry 3-7 [ 6.1-1 Fast Neutron Flux (i > s.0 MeV) as a Function 6-18 of Axial Position Along the O Degree Traverse j
- Cycle 15 Irradiation
{ 6.1-2 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-19 { of Axial Position Along the 15 Degree Traverse !
- Cycle 15 Irradiation
. 6.1-3 Fast Neutron Flux.(E > 1.0 MeV) as a Function 6-20 of Axial Position Along the 30 Degree Traverse I
- Cycle 15 Irradiation f
6.1-4 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-21 l of Axial Position Along the 45 Degree Traverse
- Cycle 15 Irradiation j 6.2-1 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-38 of Axial Position Along the O Degree Traverse
- Cycle 16 Irradiation 6.2-2 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-39 !
of Axial Position Along the 15 Degree Traverse
- Cycle 16 Irradiation 6.2-3 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-40 .
of Axial Position Along the 30 Degree Traverse
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- Cycle 16 Irradiation 6.2-4 Fast Neutron Flux (E > 1.0 MeV) as a Function 6 -
of Axial Position Along the 45 Degree Traverse j
- Cycle 16 Irradiation 1
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LIST;0FiFIGURES Fiaure Title Paae 6.3-1 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-58 4 of Axial Position Along the 0 Degree Traverse
- Cycle 17 Irradiation 6.3-2 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-59 i of Axial Position Along the 15 Degree Traverse
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- Cycle 17 Irradiation 6.3-3 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-60 of Axial Position Along the 30 Degree Traverse
- Cycle 17 Irradiation 6.3-4 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-61 '
of Axial Position Along the 45 Degree Traverse
- Cycle 17 Irradiation 8.2-1 Fast Neutron Fluence (E > 1.0 MeV) as a Function 8-24 of Azimuthal Angle at the Inner Radius of the , i Beltline Circumferential Weld 8.2-2 Fast Neutron Fluence (E > 0.1 MeV) as a Function 8-25 -
t of Azimuthal Angle at the Inner Radius of the 8eltline Circumferential Weld 8.2-3 Iron Atom Displacements [dpa) as a Function 8-26 of Azimuthal Angle at the Inner Radius of the Beltline Circumferential Weld i l .. t 1
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LIST OF TABLES Table Title Paae 4.1-1 Calculated Reference Neutron Energy Spectra at 4-5 Can ty Sensor Set Locations 4.1-2 Calculated Neutron Sensor Reaction Rates and 4-6 Exposure Rates at the Cavity Sensor Set Locations 4.1-3 Calculated Reference Neutron Energy spectra at 4-7
". Surveillance Capsule Locations l t
4.1-4 Calculated Neutron Sensor Reaction Rates and 4-8 Exposure Rates at the Center of the Surveillance ' Capsules ; 4.1-5 Radial Gradient Corrections for Sensors Contained 4-9 in Point Beach Unit 2 Internal Surveillance Capsules 4.1-6 Azimuthal Variation of Fast Neutron Flux 4-10 (E > 1.0 MeV) at the Pressure Vessel Inner Radius
. 4.1-7 Summary of Exposure Rates at the Pressure Vessel 4-11 Clad / Base Metal Interface 4.1-8 Relative Radial Distribution of Neutron Flux 4-12 (E > 1.0 MeV) Within the Pressure Vessel Wall 4.1-9 Relative Radial Distribution of Neutron Flux 4-13 '
(E > 0.1 MeV) Within the Pressure Vessel Wall 4.1-10 Relative Radial Distribution of Iron Displacement 4-14 Rate (dpa) Within the Pressure Vessel Wall 4.2-1 Calculated Fast Neutron Flux (E > 1.0 MeV) at the 4-17 Center of Reactor Vessel Surveillance Capsules
. 4.2-2 Calculated Fast Neutron Fluence (E > 1.0 MeV) at 4-18 .
the Center of Reactor Vessel Surveillance Capsules 4.2-3 Calet ated Fast Neutron Flux (E > 1.0 MeV) at the 4-19 Pressure Vessel Clad / Base Metal Interface 4.2-4 Calculated Fast Neutron Fluence (E > 1.0 MeV) at 4-20 the Pressure Vessel Clad / Base Metal Interface 4.2-5 Calculated Fast Neutron Flux (E > 1.0 MeV) at the 4-21 Cavity Sensor Set Locations v
LIST OF TABLES Table Title Paae 4.2-6 Calculated Fast Neutron Fluence (E > 1.0 MeV) at 4-22 the Cavity Sensor Set Locations 4.2-7 Calculated Fast Neutron Flux (E > 0.1 MeV) at the 4-23
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Center of Reactor Vessel Surveillance Capsules 4.2-8 Calculated Fast Neutron Fluence (E > 0.1 MeV) at 4-24 the Center of Reactor Vessel Surveillance Capsules .' 4.2-9 Calculated Fast Neutron Flux (E > 0.1 MeV) at the 4-25 : Pressure Vessel Clad / Base Metal Interface 4.2-10 Calculated Fast Neutron Fluence (E > 0.1 MeV) at 4-26 the Pressure Vessel Clad / Base Metal Interface ; 4.2-11 Calculated Fast Neutron Flux (E > 0.1 MeV) at the 4-27 Cavity Sensor Set Locations 4.2-12 Calculated Fast Neutron Fluence (E > 0.1 MeV) at 4-28 the Cavity Sensor Set Locations ,! 4.2-13 Calculated Iron Atom Displacement Rate at the 4-29 Center of Reactor Vessel Surveillance Capsules ! 4.2-14 Calculated Iron Atom Displacements at the Center 4-30 , of Reactor Vessel Surveillance Capsules 4.2-15 Calculated Iron Atom Displacement Rate at the 4-31 Pressure Vessel Clad / Base Metal Interface 4.2-16 Calculated Iron Atom Displacements at the Pressure 4-32 Vessel Clad / Base Metal Interface 4.2-17 Calculated Iron Atom Displacement Rate at the 4-33 Cavity Sensor Set Locations - 4.2-18 Calculated Iron Atom Displacements at the Cavity 4-34 Sensor Set Locations - 5,1-1 Summary of Reaction Rates Derived from Multiple 5-4
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Foil Sensor Sets Withdrawn from Internal Surveillance Capsules vi i I
LIST OF TABLES Table Title Pace 5.2-1 Derived Exposure Rates from Surveillance Capsule V 5-8 Withdrawn at the End of Fuel Cycle 1 5.2-2 Derived Exposure Rates from Surveillance Capsule T 5-9 Withdrawn at the End of Fuel Cycle 3 l 5.2-3 Derived Exposure Rates from Surveillance Capsule R 5-10 !
. Withdrawn at the End of Fuel Cycle 5 5.2-4 Derived Exposure Rates from Surveillance Capsule S i-11 Withdrawn at the End of Fuel Cycle 16 l 6.1-1 Summary of Reaction Rates derived from Multiple 6-4 Foll Sensor Sets Irradiated During Cycle 15 6.1-2 Fe-54 (n,p) Reaction Rates Derived from the 6-5 Stainless Steel Gradient Chains Irradiated During l Cycle 15 6.1-3 Ni-58 (n p) Reaction Rates Derived from the 6-6 Stainless Steel Gradient Chains Irradiated During Cycle 15 6.1-4 Co-59 (n,1) Reaction Rates Derived from the 6-7
> Stainless Steel Grac'ient Chains Irradiated During l Cycle 15 l 6.1-5 Derived Exposure Rates from the Capsule H 6-8 Dosimetry Evaluation 0 Degree Azimuth
- Core Midplane 6.1-6 Derived Exposure Rates from the Capsule J 6-9 Dosimetry Evaluation 15 Degree Azimuth
- Core Midplane 6.1-7 Derived Exposure Rates from the Capsule K 6-10 Dosimetry Evaluation 30 Degree Azimuth
- Core Midplane 6.1-8 Derived Exposure Rates from the Capsule L 6-11
( Dosimetry Evaluation 45 Degree Azimuth
- Core Midplane vii
r LIST OF TABLES '! L Table Title Paae 6.1-9 Derived Exposure Rates from the Capsule G 6-12 Dosimetry Evaluation 0 Degree Azimuth !
- Core Top 6.1-10 Derived Exposure Rates from the Capsule I 6-13 ,
Dosimetry Evaluation 0 Degree Azimuth
- Core Bottom .'
6.1-11 Derived Exposure Rates from the Capsule XX 6-14 [ Dosimetry Evaluation 0 Degree Azimuth .
- Vessel Support Elevation 6.1-12 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-15 of Axial Position Within the Reactor Cavity ,
- Cycle 15 Irradiation 6.1-13 Fast Neutron Flux (E > 0.1 MeV) as a Function 6-16 !
of Axial Position Within the Reactor Cavity
- Cycle 15 Irradiation i 6.1-14 Iron Atom 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 Multiple 6-18 Foil Sensor Sets Irradiated During Cycle 16 6.2-2 Fe-54 (n,p) Reaction Rates Derived from the 6-19 Stainless Steel Gradient Chains Irradiated During ;
Cycle 16 ; 6.2-3 Ni-58 (n,p) Reaction Rates Derived from the 6-20 - Stainless Steel Gradient Chains Irradiated During Cycle 16 - 6.2-4 C0-59 (n,7) Reaction Rates Derived from the 6-21 ,
- Stainless Steel Gradient Chains Irradiated During
- j Cycle 16 v
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t LIST OF TABLES Table Title Paae 6.2-5 Derived Exposure Rates from the Capsule N 6-2'9 Dosimetry Evaluation 0 Degree Azimuth
- Core Midplane 6.2-6 Derived Exposure Rates from the Capsule P 6-30 Dosimetry Evaluation 15 Degree Azimuth r
". - Core Midplane 6.2-7 Derived Exposure Rates from the Capsule Q 6-31
- Dosimetry Evaluation 30 Degree Azimuth ;
- Core Midplane ,
6.2-8 Derived Exposure Rates from the Capsule R 6-32 Dosimetry Evaluation 45 Degree Azimuth
- Core Midplane 6.2-9 Derived Exposure Rates from the Capsule M 6-33
, Dosimetry Evaluation 0 Degree Azimuth
- Core Top 6.2-10 Derived Exposure Rates from the Capsule 0 6-34 Dosimetry Eyaluation 0 Degree Azimuth
- Core Bottomi 6.2-11 Fast Neutron Flux (E > 1.0 MeV) as a function 6-35 of Axial Position Within the Reactor Cavity
- Cycle 16 Irradiation 6.2-12 Fast Neutron Flux (E > 0.1 MeV) as a Function 6-36 of Axial Position Within the Reactor Cavity
- - Cycle 16 Irradiation 6.2-13 Iron Atom 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 Multiple 6-45 Foil Sensor Sets Irradiated During Cycle 17 6.3-2 Fe-54 (n.p) Reaction Rates Derived from the 6-46 j Stainless Steel Gradient Chains Irradiated During {
Cycle 17 ix l
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LIST OF TABLES ! t I Table Title Paae 6.3-3 Ni-58 (n.p) Reaction Rates Derived from the 6-47 l Stainless Steel Gradient Chains Irradiated During > Cycle 17 l 6.3-4 Co-59 (n,1) Reaction Rates Derived from the 6-48 ; Stainless Steel Gradient Chains Irradiated During : Cycle 17 . 6.3-5 Derived Exposure Rates from the Capsule BB 6-49 Dosimetry Evaluation 0 Degree Azimuth - ,
- Core Midplane -
6.3-6 Derived Exposure Rates from the Capsule DD 6-50 Dosimetry Evaluation 15 Degree Azimuth
- Core Midplane ;
6.3-7 Derived Exposure Rates from the Capsule EE 6-51 l Dosimetry Evaluation 30 Degree Azimuth .
- Core Midplane j 6.3-8 Derived Exposure Rates from the Capsule FF 6-52 ;
Dosimetry Evaluation 45 Degree Azimuth
- Core Midplane 6.3-9 Derived Exposure Rates from the Capsule AA 6-53 Dosimetry Evaluation 0 Degree Azimuth l
- Core Top 6.3-10 Derived Exposure Rates from the Capsule CC 6-54 ;
Dosimetry Evaluation 0 Degree Azimuth !
- Core Bottom -
6.3-11 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-55 of Axial Position Within the Reactor Cavity -
- Cycle 17 Irradiation ;
6.3-12 Fast Neutron Flux (E > 0.1 MeV) as a Function 6-56 l of Axial Position Within the Reactor Cavity
- Cycle 17 Irradiation -
LIST OF TABLES f
- Table Title Paae ,
6.3-13 Iron Atom Displacement Rate as a Function 6-57 ! of Axial Position Within the Reactor Cavity. l
- Cycle 17 Irradiation 7.1-1 Comparison of Measured and Calculated Exposure 7-4
- Rates from Surveillance Capsule and Cavity l
, Dosimetry Irradiations :
7.2-2 Comparison of Measured and Calculated Neutron 7-7
. Sensor Reaction Rates From Surveillance Capsule ,
and Cavity Dosimetry Irradiations l 8.1-1 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) 8-5 i Exposure Projections for the Beltline Region of the Point Beach Unit 2 Reactor Pressure Vessel 0 Degree Azimuthal Angle
, 8.1-2 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) 8-6 Exposure Projections for the Beltline Region of the Point Beach Unit 2 Reactor Pressure Vessel l
- 15 Degree Azimuthal Angle ,
8.1-3 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) 8-7 Exposure Projections for the Beltline Region of ! the Point Beach Unit 2 Reactor Pressure Vessel i
- 30 Degree Azimuthal Angle ;
8.1-4 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) 8-8 Exposure Projections for the Beltline Region of the Point Beach Unit 2 Reactor Pressure Vessel
- 45 Degree Azimuthal Angle 8.1-5 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) 8-9 Exposure Projections for the Beltline Region of the Point Beach Unit 2 Reactor Pressure Vessel 0 Degree Azimuthal Angle xi
.. . ~_ . -- h LIST OF TABLES l Table Title Paae
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8.1-6 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) 8-10 Exposure Projections for the Beltline Region of ; the Point Beach Unit 2 Reactor Pressure Vessel f
- 15 Degree Azimuthal Angle f 8.1-7 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) 8-11 I Exposure Projections for the Beltline Region of .' !
the Point Beach Unit 2 Reactor Pressure Vessel ;
- 30 Degree Azimuthal Angle .
8.1-8 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) 8-12 [' Exposure Projections for the Beltline Region of the Point Beach Unit 2 Reactor Pressure Vessel
- 45 Degree Azimuthal Angle :
8.1-9 Summary of Best Estimate Iron Atom Displacement 8-13 ; Exposure Projections for the Beltline Region of .I the Point Beach Unit 2 Reactor Pressure Vessel 0 Degree Azimuthal Angle
-l u e Proj t fr h B i R of the Point Beach Unit 2 Reactor Pressure Vessel
- 15 Degree Azimuthal Angle !
8.1-11 Summary of Best Estimate Iron Atom displacement 8-15 Exposure Projections for the Beltline Region of the Point Beach Unit 2 Reactor Pressure Vessel
- 30 Degree Azimuthal Angle - r 8.1-12 Summary of Best Estimate Iron Atom Displacement 8-16 Exposure Projections for the Beltline Region of -
the Point Beach Unit 2 Reactor Pressure Vessel
- 45 Degree Azimuthal Angle 8.2-1 Maximum Fast Neutron Exposure of Point Beach 8-20 Unit 2 Beltline Circumferential Weld (SA-1484) -
8.2-2 Maximum Fast Neutron Exposure of Point Beach 8-21
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Unit 2 Intermediate Shell Forging (123V500) xii j l
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[:- LIST OF TABLES l l l Table Title Paae - 8.2-3 Maximum Fast Neutron _ Exposure of Point Beach 8-22 Unit 2 Lower Shell Forging (122W195) 8.2-4 Maximum Fast Neutron Exposure of Point Beach 8-23 Unit 2 Upper / Intermediate Shell Circumferential j Weld
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SECTION 1.0 OVERVIEW 0F THE PROGRAM The Reactor Cavity Neutron Measurement Program [1] initiated at Point Beach Unit 2 at the start of Fuel Cycle 15 was designed to provide a
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mechanism for the long term continuous monitoring of the neutron exposure of those portions of the reactor vessel and vessel support structure which i
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. may experience radiation induced increases in reference nil ductility ;
transition temperature (RTNDT) over the nuclear power plant lifetime. ;
. When used in conjunction with dosimetry from previously withdrawn internal j surveillance capsules and with the results of neutron transport calculations, the reactor cavity neutron dosimetry provides the means for determination of the neutron exposure of the pressure vessel and the projection of 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 with respect to 1 - 10CFR50 Appendix G pressure / temperature limit curves for normal heatup and cooldown of the reactor coolant system.
2 - Emergency Response Guidline (ERG) pressure / temperature limit curves. 3 - Pressurized Thermal Shock (PTS) RTPTS screening critevia. In addition, an accurate measure of the neutron exposure of the reactor
. vessel and support structure can provide a sound basis for requalification l should operation of the plant beyond the current design and/or licensed !
lifetime prove to be desirable. l - Within the nuclear industry it has been common practice to base estimates l of the fast neutron exposure of pressure vessels either directly on the l 1-1
l l 1 l results of neutron transport calculations or on the' analytical results normalized to measurements obtained from internal surveillance capsules. l However, there are potential drawbacks associated with both of these approaches to exposure assessment. *' In performing neutron transport calculations for pressurized water reactors, several design and operational variables have an impact on the ;
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magnitude of the analytical prediction of exposure rates within the pressure vessel wall as well as on the uncertainties associated with that [ prediction. Of particular note in this regard are cycle to cycle [ l variations in core power distributions (particularly with implementation i of low leakage loading patterns), variations of water temperature in the - i i downcomer regions of the reactor internals, and deviations in as-built j versus design dimensions for the reactor internals and pressure vessel. l The manner in which these important variables are treated in the analysis ! may lead to an increased uncertainty in the exposure evaluations for the ! pressure vessel; and, these increased uncertainties may well result in the ( use of overly conservative estimates of vessel embrittlement in the .! assessment of pressure temperature limitations as well as of the expected l service life of the component. .- The reactor vessel materials surveillance program [2] consisting of ! several surveillan:e capsules attached to the thermal shield in the ! downcomer region near the pressure vessel wall has been in service since ! the initial startup of the reactor. The neutron dosimetry contained in [ these capsules provides measurement capability to determine the fast l neutron exposure of the materials test specimens also located within the capsules, but at the same time produces measured data only at a single - location within the reactor geometry. Therefore, the surveillance capsule . dosimetry, by itself, cannot provide information regarding the azimuthal, - radial, and axial gradients of neutron exposure within the pressure ! vessel. Furthermore, data from internal surveillance capsules are, by
- design, obtained at rather infrequent intervals; and surveillance !
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measurement locations may not be in proximity to critical areas on the j pressure vessel. These limitations place a heavy reliance on analytical ; l 1-2 i
results to project exposure levels to the vessel wall as well as to l provide predictions of vessel exposure for time periods beyond the last l scheduled capsule withdrawal. l With the addition of supplementary passive neutron sensors in the reactor j cavity annulus between the reactor vessel wall and the biological shield, I the deficiencies in both surveillance dosimetry and analytical prediction can be mitigated and the uncertainties associated with exposure estimates l for the pressure vessel can be minimized. With state of the art neutron
". sensors deployed to establish the absolute magnitude of the azimuthal and axial exposure rate distributions in the reactor cavity, the burden placed
. on the neutron transport calculation is reduced to the determination of relative neutron energy spectra for sensor set interpretation and relative spatial distributions for extrapolation of the measurement results to positions at the inner radius and through the thickness of the pressure vessel wall. Studies have shown that the operational and design variables cited above that have a strong impact on the calculated magnitude of
. exposure rates have only a minor effect on both the interpretation of cavity dosimetry and on the extrapolation of measurement results to key
. vessel locations. It is possible, therefore, to employ cavity measurements and a set of reference neutron transport calculations to produce vessel exposure projections with a reduced uncertainty over that inherent in an approach based on analysis alone. Furthermore, since the cavity neutron measurements are not directly tied to the materials surveillance program, measurement intervals can be chosen to easily provide integral vessel exposure over plant lifetime.
The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could letad to an improvement in the 1-3
E 1 uncertainties associated with damage trend curves as well as to a more , accurate evaluation of damage gradients through the pressure vessel wall. l Because of this potential shift away from threshold fluence toward an
*l energy dependent damage function for data correlation, ASTM Standard' i Practice E853, " Analysis and Interpretation of Light Water Reactor *!
Surveillance Results", recommends reporting displacements per iron atom ; (dpa) along with fluence (E > 1.0 MeV) to provide a data base for future [ reference. The energy dependent dpa function to be used for this , evaluation is specified in ASTM Standard Practice E693, " Characterizing - Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom". The application of the dpa parameter to the assessment of embrittlement - gradients has already been promulgated in Revision 2 to Regulatory Guide 1.99, " Radiation Damage to Reactor Vessel Materialt". I With the aforementioned views in mind, the Reactor Cavity Measurement -[ Program was established to meet the following objectives-1 - Determine azimuthal and axial gradients of fast neutron exposure over the beltline region of the reactor pressure - vessel. I 2 - Provide measurement capability sufficient to allow the ! determination of pressure vessel exposure in terms of both fluence (E > 1.0 MeV) and iron displacements per atom. f 3 - Establish a methodology for the projection of exposure ! gradients through the thickness of the pressure vessel wall. - 4 - Provide a long term continuous monitoring capability for the - ; beltline region of the pressure vessel. This report provides the results of neutron dosimetry evaluations e performed subsequent to the completion of Fuel Cycle 17. Fast neutron - exposure in terms of fast neutron fluence (E > 1.0 MeV) and dpa is
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l 1-4 , l
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t established for all measurement locations in the reactor cavity. The analytical formalism describing the relationship among the measurement ; points and locations within the pressure vessel wall is described and used !
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to project the Cycle 17 exposure of the vessel itself. , Results of exposure evaluations from surveillance capsule dosimetry i withdrawn at the end of Fuel Cycles 1, 3, 5, and 16 as well as cavity dosimetry results from Cycles 15 and 16 are incorporated to provide the integrated exposure of the pressure vessel from plant startup through the ;
. end of Cycle 17. 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 provided based on the measured flux reduction achieved by the insertion of part length hafnium absorbers in selected fuel assemblies at .
. the onset of Cycle 16. Current evaluations and future projections are '
provided for the beltline circumferential weld as well as for both the intermediate and lower shell forgings that comprise the highly irradiated portions of the reactor vessel.
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e a 1-5
SECTION
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. The intent here was to determine changes in spectra caused by 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 multiple foil packages were positioned opposite the top and bottom of the active core at the azimuthal angle corresponding to 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 i 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 2-1 l
l i l L provide fast neutron exposure assessments for materials comprising the 1 beltline region of the reactor pressure vessel. During the Cycle 15 irradiation an additional multiple foil sensor set was included in the vicinity of the reactor 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. 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 crient 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 ir. 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 relativo to the core centerline. 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, " l 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 l extended 7 feet relative to the midplane of the active core. - 2-2
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 of the support bars were positioned { 26.625 inches above the top of the active fuel. The sensor sets and l i gradient chains were retained and supported at the bottom by chain clamps l' atti.ched to stainless steel eye nuts with stainless steel threaded chain connectors. The eye nuts were, in turn, attached 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 i
'. 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 posi+ ion 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. l
- i i
2.1.3 - Description of Gradient Chains i Along with the multiple foil sensor sets placed at discrete locations 2-3
- ~ ~ . - _ - _ - . __ _ _ _ _ _ _ . _
i within the reactor cavity, gradient chains were employed to obtain, axial i 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 chains consisted of Type 304 stainless steel bead chain of 0.188 inch
- diameter. When coupled with a chemical analysis, the stainless steel yielded activation results for the Fe-54 (n,p), Ni-58 (n.p), and .
Co-59 (n,1) 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. l
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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 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 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 included in Capsules V, T, R, and S are provided in the appendices to this report.
e 6 2-8 i
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FIGURE 2.2-1 NEUTRON SENSOR LOCATIONS WITHIN INTERNAL SURVEILLANCE CAPSULES l I l 4
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2-9
SECTION 3.0 NEUTRON TRANSPORT AND DOSIMETRY EVALUATION METHODOLOGIES
~
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 l . transport techniques. A single forward calculation provided the relative energy distribution of neutrons and gamma rays 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 l power distributions; thus, providing a direct comparison with all i dosimetry results obtained over the operating history of the reactor. In combination, the absolute cycle specific data from the adjoint l evaluations together with relative neutron energy spectra distributions I from the forward calculation provided the'means to: 1 - Evaluate neutron dosimetry from reactor cavity and I surveillance capsule locations. 2 - Enable a direct comparison of analytical prediction with measurement. 3 - Extrapolate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel. 4 - Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves.
~
f 3-1
1 i l 1 3.1.1 - Reference Forward Calculation l l A plan view of the reactor geometry at the core midplane elevation is shown in Figure 3.1-1. 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. 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 and gamma ray energy spectra at dosimetry locations within the capsules, a meaningful ccmparison 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 sur';eillea:a capsules, the small aluminum capsules used in the reactor cavity measurersent program were designed to minimize perturbations in the neutron flux and, thus, to provide free field data at the 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 and 3-2 was carried out in R,0 geometry using the DOT ' two-dimensional discrete ordinates code [3] and the SAILOR cross-section library [4). The SAILOR library is a 67 group coupled neutron-gamma ray
- ENDFB-IV based data set produced specifically for light water reactor applications. 17 these analyses, anisotropic scatter'm was treated with -
a P3 expansion cf the cross-sections and the angular discretization was 3-2
modeled with an S8 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 !
- i
. 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 .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 result in exposure rates well below these conservative predictions. This difference between the conservative forward calculation and the fuel cycle specific best estimate computations is illustrated by a cnmparison of the analytical results given in Section ! 4.0 of this report. 3.1.2 - Cycle Specific Adjoint Calculations All adjoint analyses were also carried out using an S8 order of angular quadrature and the P3 cross-section approximation from the SAILOR library. - Adjoint source locations were chosen at each of the azimuthal locatio n J containing cavity dosimetry (0,15, 30, and 45 degrees) as well as at. the l corresponding azimuths on the pressure vessel inner radius. In addition, adjoint calculations were carried out for sources positioned at the 3-3
geometric center of surveillance capsub s located at 13, 23, and 33 degrees relative to the core cardinal axes. Again, these calculations were run in R,0 geometry to provide neutron source distribution importance functions i for the exposure parameter of interest; in this case, p (E > 1.0 MeV). - The importance functions generated from these eleven 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 i evaluations for all subsequent' fuel cycles. - Having the importance functions and appropriate core source distributions, , the response of interest can be calculated as: i ((R,0)-[R[0[E I(R,0,E) S(R,0,E) R dR de dE . where: p(R,0) - Flux (E > 1.0 MeV) at radius R and azimuthal angle 0. I(R,0,E) - Adjoint importance function at radius R, azimuthal r angle 0, and neutron source energy E. i S(R,0,E) - Neutron source strength at core location R,0 and energy E. It is important to note that the cycle specific neutron source distributions, - S(R,0,E), utilized with the adjoint importance functions, I(R,0,E), permitted 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 -l
[ spectrum introduced by the build-in of plutonium isotopes as the burnup of l 3-4
individual fuel assemblies increased. i
. Although the adjoint importance functions used in these analyses were based on a response function defined by the threshold neutron flux (E > 1.0 MeV), prior calculations have shown that, while the ;
implementation of low leakage loading patterns significantly impact the l magnitude and the spatial distribution of the neutron field, changes in
~
the relative neutron energy spectrum are of second order. Thus, for a given location the exposure parameter ratios such as
'. [dpa/sec]/[p (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 on a cycle specific basis by using the appropriate
[dpa/sec]/[p (E > 1.0 MeV)] and [p (E > 0.1 MeV)]/[4 (E > 1.0 MeV)] ratios from the reference forward analysis in conjunction with the cycle specific p (E > 1.0 MeV) solutions from the individual adjoint ' evaluations. . In particular, after defining the following exposure rate ratios, R - IdDa/seCl i [p (E > 1.0 MeV)] R " I6 (E > 0.1 MeVil 2 [p (E > 1.0 MeV)] the corresponding fuel cycle specific exposure rates at the adjoint source locations were computed from the following relations: l dpa/sec = [p (E > 1.0 MeV)] adjoint R 3 i p (E > 0.1 MeV) = [p (E > 1.0 MeV)] adjoint R 2 6 3-5
n 4 4- " 9 & 9 o FIGURE 3.1-1 I REACTOR GEOMETRY SHOWING A 45' R,0 SECTOR l ,
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INTERNAL SVRVEILLANCE CAPSVLE GE0 METRY I (13, 23, e 33) i l pi tif tstilitit,r ~ 100*30 '
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M 3.2 - Neutron Dosimetry Evaluation Methodology The use of p;ssive neutron sensors such as those included in the internal surveillance capsule and reactor cavity dosimetry sets does not yield a direct measure of the energy dependent neutron 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 - following variables are of interest: 1 - The measured specific activity of each sensor 2 - The physical characteristics of each sensor 3 - The operating history of the reactor 4 - The energy response of each sensor . 5 - The neutron energy spectrum at the sensor location In this section the procedures used by Westinghouse to determine sensor specific activities, to develop reaction rates for 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. For the most part these procedures apply to all of the evaluations provided in this report. However, in some cases, the specific activities - pertaining to individual internal serveillance capsules were determined from prior analysis by a radiochemical laboratory other than Westinghouse. In those cases, 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. i
.l 3-8
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 tne Westinghouse Scier.ce 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 [5 through 15]. 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
- j. 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 of U-238 and Np-237 fission monitors from internal surveillance capsules, by direct counting preceded by dissolution and chemical separation of cesium l from the sensor. For the stainless steel gradient chains used in the cavity irradiations, individual sensors were obtained by cutting the chains 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.
4 The irradiation history of the reactor over its operating lifetime was i obtained from NUREG-0020, " Licensed Operating Reactors Status Summary
*- Report". In particular, operating data were extracted from that report on a monthly bases from reactor startup to the end of the current evaluation l
period. For the sensor sets utilized in surveillance capsule and reactor l 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 3-9
the reactions of interest in the exposure evaluations. 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: R = A NO FY Z[Pj/ Pref] Cj [1-e-Atjj[,-Atd ]
)
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 P ref (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). . Pref = maximum or reference core power level of the reactor (MW). Cj - calculated ratio of p (E > 1.0 MeV) during irradiation - period j to the time weighted average p (E > 1.0 MeV) over the entire irradiation period. A = decay constant of the product isotope (sec-1), tj - length of irradiation period j (sec). td - decay time following irradiation period j (sec). and the summation is carried out over the total number of monthly intervals comprising the total irradiation period. - In the above equation, the ratio P /jPref accounts for month by month - variation of power level within a given fuel cycle. The ratio C3s i 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 -
3-10
1 l
\
fuel cycle. For a single cycle irradiation Cj = 1.0. However, for multiple cycle irradiations, particularly those employing low leakage fuel i management the additional Cj correction must be utilized. ! J 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 [(Pj/ Pref] tj ' j 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. The SSTR fissionable deposits were designed for reuse in the long term [ monitoring program. Therefore, following processing each sensor was carefully examined to assure that the deposits were neither damaged nor I contaminated during irradiation, handling, and post-irradiation processing. i ss h f s on r ck nf ned an re rre po g o e active portion of the ftissionable deposit and that the edges of the active area were sharply defined with a sufficient drop-off in track density to
~
3-11 i
indicate acceptable signal to background ratios for the measurements. Each mica SSTR 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 isotopes over the course of the irradiation. Likewise, corrections were made to both U-238 and Np-237 sensors to account for gamma ray induced fission reactions occuring 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 performing the dosimetry evaluations for the internal surveillance capsules, the sensor reaction rates measured at the 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 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 value at the geometric center of the surveillance ' capsule. In the case of the reactor cavity sensors, all of the monitors were located at the same radial location. Thus, gradient corrections were - not required in the evaluation of these dosimetry sets. 3-12 l
. . . . - _- _ . - - . _ = . -
I F 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 [16]. The FERRET approach used the measured reaction rate data and the calculated neutron energy spectrum at the sensor set locations' as input and proceeded to adjust a priori (calculated) group fluxes to produce a best fit (in a least squares sense) to the reaction rate data. The ,
- exposure parameters along with associated uncertainties were then obtained l 1 from the adjusted spectra.
s In the FERRET evaluations, a log-normal least-squares algorithm weights l both the a priori values and the measured data in accordance with the : assigned uncertainties and correlations. In general, the measured values ! f are linearly related to the flux p by some response matrix A: I (s,a) (s) (a) f i - E Ajg p g {
, 9 i where i indexes the measured values belonging to a single data set s, g f designates the energy group and a delineates spectra that may be simultaneously adjusted. For example, l i
Rj - E aj g p g ; 9 ' I relates a set of measured reaction rates Rj to a single spectrum l pgby the multigroup cross section ajg. (In this case, FERRET i also adjusts the cross-sections.) The log-normal approach automatically l accounts for the physical constraint of positive fluxes, even with l s. 3-13 1
, - _ - - _______A
I 1 large assigned uncertainties. In the FERRET analysis of the dosimetry data, the continuous quantities (i.e., fluxes and cross-sections) were approximated in 53 groups. The - calculated fluxes from the reference forward calculation were expanded into the FERRET group structure using the SAND-II code [17]. This ~ procedure was carried out by first expanding the a priori spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not coincide. The 620-point spectrum was then easily collapsed to the group scheme used in .' FERRET. The cross-sections were also collapsed into the 53 energy-group structure using SAND II with calculated spectra (as expanded to 620 groups) as weighting functions. The cross sections were taken from the ENDF/8-V dosimetry file. Uncertainty estimates and 53 x 53 covariance matrices were constructed for each cross section. Correlations between cross sections were neglected due to data and code limitations, but this . omission does not significantly impact the results of the adjustment. For each set of data or a priori values, the inverse of the corresponding relative covariance matrix M is used as a statistical weight. In some cases, as for the cross sections, a multigroup covariance matrix is used. More often, a simple parameterized form is employed: 2 M gg. = Rn+Rg RP g gg i where R n specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the corresponding set of values. The fractional uncertainties Rgspecify additional random uncertainties for
- group g that are correlated with a correlation matrix:
O 3-14 1
i P gg. - (1-0) 6 99 + 0 el-fo-o')2 27 2 The first term specifies purely random uncertainties while the second term ' describes short-range correlations over a range 7 (0 specifies the strength of the latter term). I For the a priori calculated fluxes, a short-range correlation of 7=6 i groups was used. This choice implies that neighboring groups are strongly correlated when 0 is close to 1. Strong long-range correlations (or j anticorrelations) were justified based on information presented by R.E. Maerker[18]. Maerker's results are closely duplicated when y = 6. I For the integral reaction rate covariances, simple normalization and j random uncertainties were combined as deduced from experimental j uncertainties. ! In performing the least squares adjustment with the FERRET code, the input j spectra from the reference forward transport calculation were normalized , to the measured Fe-54 (n p) Mn-54' reaction rates to remove any constant calculation to measurement bias and, thus, to permit the adjustment to take place on a relative basis. The specific normalization factors for i individual evaluations depended on the location of the sensor set as well ! as on the neutron flux level at that location. l The specific assignment of uncertainties in the measured reaction rates and the input (a priori) spectra used in the FERRET evaluations was as l follows- i REACTION RATE UNCERTAINTY 5% l 1 FLUX NORMALIZATION UNCERTAINTY 30% l l 3-15
l l l FLUX GROUP UNCERTAINTIES li (E > 0.0055 MeV) 30% (0.68 ev < E < 0.0055 MeV) 58% (E < 0.68 ev) 104%
~
SHORT RANGE CORRELATION (E > 0.0055 MeV) 0.9 (0.68 ev < E < 0.0055 MeV) 0.5 (E < 0.68 ev) 0.5 FLUX GROUP CORRELATION RANGE (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 se, . ' 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.
.O
- I 3-16
l 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 that the results of the reference forward transport j l 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. l l
- 4.1.1 - Cavity Sensor Set Locations Data from the reference forward calculation pertinent to cavity 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 sensor locations at 0,15, 30, and 45 degrees relative to the core cardinal axes are listed. These data represent the a priori 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 l distributions establish a baseline against which adjusted spectra may be 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
.. In Table 4.1-2, the calculated neutron sensor reaction rates associated with the spectra from Table 4-1 are provided along with the reference ! exposure rates in terms of p (E > 1.0 MeV), p (E < 0.1 MeV) and , l dpa/sec. Also listed are the associated exposure rate ratios calculated
~
for each of the cavity sensor set locations. ! i 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 p (E > 0.1 MeV) I l and dpa/sec for each irradiation period. In addition, the ratios of I U238(1,f)/U238(n,f) and Np237(7,f)/Np237(n,f) were used to make - photo-fission corrections to measured reaction rates in the U238 and Np237 l fission monitors prior to use in the FERRET adjustment procedure. . 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 l rates in terms of ( (E > 1.0 MeV), p (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 l fashion as was the case for the cavity sensor sets. . i 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. ; i 4-2 i f
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. These factors were used in a multiplicative fashion to relate measured reaction rates to the corresponding value at the geometric center of the capsules. 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 4.1-10. In Table 4.1-6, the calculated azimuthal distribution of fast neutron flux (E > 1.0 MeV) is listed for the center of the vessel cladding, at the pressure vessel clad / base metal interface, and at the center of the first mesh interval in the base metal. The interface information (base metal
. inner radius) was obtained by averaging the two sets of data obtained directly from the reference forward calculation. In this detailed tabulation, calculated flux levels are given for each of the 51 azimuthal mesh intervals included in the analytical model model.
In Table 4.1-7, the calculated azimuthal distribution of exposure rates in terms of p (E > 1.0 MeV), p (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 4.1-7 are the exposure rate ratios [p (E > 0.1 MeV)]/[p (E > 1.0 MeV)] and [dpa/sec]/[p (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 p (E > 1.0 MeV), p (E > 0.1 Mev), and dpa/sec is given in Tables 4.1-8, 4.1-9, and 4.1-10, 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 4-3
rc i L - . distributions within the vessel wall were obtained by normalizing the : calculated or projected exposure at the vessel inner radius to the ; gradient data given.in Tables 4.1-8 through'4.1-10. ! i f 1 i e i
}
I i t i t 5 i
.. j t
. I i
f I t 4-4 !
TABLE 4.1-1 CALCULATED REFERENCE NEUTRON ENERGY SPECTRA AT CAVITY SENSOR SET LOCATIONS , o 1518 MWt; Fa - 1.2
~
LOWER AZIMUTHAL ANGLE LOWER AZIMUTHAL ANGLE ENERGY ENERGY (Mev) 0 DEG 15 DEG 30 DEG 45 DEG (Mev) 0 DEG_ 15 DEG 30 DEG 45 DEG 1.42E+1 9.07E+5 7.07E+5 5.54E+5 5.30E+5 2.97E-1 3.68E+9 3.13E+9 2.20E+9 1.72E+9
. 1.22E+1 3.25E+6 2.53E+6 2.00E+6 1.89E+6 1.83E-1 3.87E+9 3.47E+9 2.48E+9 1.86E+9 1.00E+1 9.58E+6 7.25E+6 5.61E+6 5.24E+6 1.11E-1 4.67E49 4.14E+9 2.97E+9 2.27E+9
. 8.61E+0 1.59E+7 1.19E+7 9.14E+6 8.45E+6 6.74E-2 3.87E+9 3.45E+9 2.49E+9 1.92E+9 7.41E+0 2.33E+7 1.73E+7 1.31E+7 1.20E+7 4.09E-2 2.20E+9 2.02E+9 1.47E+9 1.11E+9 6.07E+0 4.56E+7 3.37E+7 2.53E+7 2.28E+7 3.18E-2 6.53E+8 6.11E+8 4.47E+8 3.34E+8 4.97E+0 5.56E+7 4.11E+7 3.05E+7 2.69E+7 2.61E-2 3.79E+8 3.58E+8 2.64E+8 1.97E+8 3.68E+0 1.04E+8 7.70E+7 5.58E+7 4.83E+7 2.42E-2 1.39E+9 1.18E+9 8.48E+8 6.84E+8 3.01E+0 9.63E+7 7.14E+7 5.10E+7 4.36E+7 2.19E-2 9.09E+8 8.06E+8 5.76E+8 4.45E+8
. 2.73E+0 8.52E+7 6.31E+7 4.23E+7 3.79E+7 1.50E-2 1.78E+9 1.66E+9 1.19E+9 8.77E+8 2.47E+0 1.05E+8 7.89E+7 5.56E+7 4.67E+7 7.10E-3 1.67E+9 1.60E+9 1.17E+9 8.67E+8 2.37E40 5.44E+7 4.12E+7 2.89E+7 2.42E+7 3.36E-3 2.33E+9 2.16E+9 1.59E+9 1.22E+9 2.35E+0 1.87E+7 1.38E+7 9.56E+6 8.07E+6 1.59E-3 1.98E+9 1.84E+9 1.36E+9 1.05E+9 2.23E+0 9.34E+7 6.97E+7 4.85E+7 4.09E+7 4.54E-4 2.70E+9 2.51E+9 1.87E+9 1.45E+9 1.92E+0 2.28E+8 1.71E+8 1.20E+8 1.01E+8 2.14E-4 1.47E+9 1.37E+9 1.03E+9 7.96E+8 1.65E+0 3.56E+8 2.69E+8 1.87E+8 1.56E+8 1.01E-4 1.50E+9 1.39E+9 1.04E+9 8.18E+8 1.35E+0 5.64E+8 4.37E+8 3.04E+8 2.49E+8 3.73E-5 1.96E+9 1.80E+9 1.35E+9 1.07E+9 1.00E+0 1.34E+9 1.06E+9 7.34E+8 5.95E+8 1.07E-5 2.29E+9 2.10E+9 1.58E+9 1.26E+9 8.21E-1 1.25E+9 1.00E+9 6.95E+8 5.57E+8 5.04E-6 1.27E+9 1.16E+9 8.71E+8 6.95E+8 7.43E-1 5.98E+8 5.06E+8 3.55E+8 2.72E+8 1.86E-6 1.46E+9 1.34E+9 1.01E+9 8.05E+8 6.08E-1 2.96E+9 2.41E+9 1.67E+9 1.34E+9 8.76E-7 9.28E+8 8.52E+8 6.42E+8 5.15E+8 4.98E-1 2.25E+9 1.92E+9 1.35E+9 1.04E+9 4.14E-7 7.66E+8 7.04E+8 5.30E+8 4.28E+8 3.69E-1 2.96E+9 2.76E+9 1.77E+9 1.38E+9 1.00E-7 1.35E+9 1.21E+9 9.04E+8 7.49E+8 0.00 3.68E+9 2.90E+9 2.14E+9 2.04E+9 NOTE: The upper energy of group 1 is 17.33 Mev. j
/ 4-5 l i
<j ~ '
TABLE.4.1-2 CALCULATED NEUTRON SENSOR REACTION RATES AND EXPOSURE RATES ) AT THE CAVITY SENSOR SET LOCATIONS *I 1518 MWt; F a
= 1.20
)
AZIMUTHAL ANGLE i O DEGREES 15 DEGREES 30 DEGREES 45 DEGREES j Reaction Rate (ros/ nucleus) Cu63(n,a) 1.84E-18 1.39E-18 1.05E-18 9.72E-19 - Ti46(n.p) 2.79E-17 2.10E-17 1.56E-17 1.41E-17 ; Fe54(n,p) 1.63E-16 1.22E-16 8.88E-17 7.81E-17 l NiS8(n,p) 2.28E-16 1.72E-16 1.24E-16 1.08E-16 i U238(n,f) (Cd) 9.55E-16 7.31E-16 5.12E-16 4.32E-16 j l Np237(n,f) (Cd) 1.41E-14 1.15E-14 8.01E-15 6.42E-15 i CoS9(n,7) 2.60E-13 2.30E-13 1.68E-13 1.44E-13 -l CoS9(n,1) (Cd) 1.46E-13 .l.36E-13 1.01E-13 7.97E-14 ! U238(1,f) 3.88E-17 3.13E-17 2.32E-17 2.12E-17 -! Np237(1,f) 1.10E-16 8.84E-17 6.54E-17 5.98E-17 ! Neutron Flux (n/cmEl ( (E > 1.0 MeV) 3.20E+09 2.53E+09 1.73E+09 1.43E+09 p (E > 0.1 MeV) 2.54E+10 2.17E+10 1.52E+10 1.19E+10 I doa/sec j Displacement rate 9.71E-12 8.14E-12 5.70E-12 4.52E-12 - l Exposure Rate Ratios ** l p(E > 0.1)/p(E > 1.0) 7.94 8.58 8.79 8.32 l [dpa/sec]/d(E > 1.0) 3.03E-21 3.22E-21 3.29E-21 3.16E-21 U238(y,f)/U238(n,f) 0.0406 0.0428 0.6453 0.0491 -' Np237(7,f)/Np237(n,f) 0.00780 0.00769 0.00816 0.00931 4-6 - w - - . . - - -
TABLE 4.1-3 CALCULATED REFERENCE NEUTRON ENERGY SPECTRA AT SURVEILLANCE CAPSULE LOCATIONS 1518 MWt; Fa - 1.2 i LOWER AZIMUTHAL ANGLE LOWER AZIMUTHAL ANGLE l ENERGY ENERGY (Mev) 13 DEG 23 DEG 33 DEG (Mev) 13 DEG 23 DEG 33 DEG l 1.42E+1 2.31E+07 1.89E+07 1.66E+07 2.97E-1 4.86E+10 2.53E+10 2.34E+10 1.22E+1 8.46E+07 6.90E+07 6.05E+07 1.83E-1 6.29E+10 3.28E+10 3.04E+10 1.00E+1 3.02E+08 2.38E+08 2.08E+08 1.llE-1 6.llE+10-3.15E+10 2.92E+10 8.61E+0 5.52E+08 4.34E+08 3.79E+08 6.74E-2 4.91E+10 2.52E+10 2.34E+10 7.41E+0 9.48E+08 7.19E+08 6.30E+08 4.09E-2 3.71E+10 1.90E+10 1.77E+10 6.07E+0 2.14E+09 1.60E+09 1.40E+09 3.18E-2 1.28E+10 6.56E+09 6.09E+09 4.97E+0 3.03E+09 2.19E+09 1.93E+09 2.61E-2 7.27E+09 3.74E+09 3.47E+09 ; 3.68E+0 7.83E+09 4.30E+09 3.84E+09 2.42E-2 1.38E+10 7.01E+09 6.52E+09 3.01E+0 5.90E+09 3.76E+09 3.40E409 2.19E-2 8.52E+09 4.31E+09 4.01E+09 ; 2.73E+0 4.83E+09 3.05E+09 2.76E+09 1.50E-2 2.17E+10 1.llE+10 1.03E+10 l 2.47E+0 5.75E+09 3.57E+09 3.26E+09 7.10E-3 3.39E+10 1.74E+10 1.61E+10 ! 2.37E+0 2.86E+09 1.78E+09 1.62E+09 3.36E-3 4.51E+10 2.31E+10 2.14E+10 ; 2.35E+0 8.75E+08 5.44E+08 4.95E+08 1.59E-3 4.00E+10 2.04E+10 1.89E+10 2.23E+0 4.31E+09 2.66E+09 2.43E+09 4.54E-4 6.16E+10 3.12E+10 2.90E+10 i 1.92E+0 1.12E+10 6.84E+09 6.25E+09 2.14E-4 3.55E+10 1.80E+10 1.67E+10 : 1.65E40 1.45E<10 8.86E+09 8.13E+09 1.01E-4 3.75E+10 1.90E+10 1.76E+10 1.35E+0 2.25E+10 1.31E+10 1.21E+10 3.73E-5 5.02E+10 2.53E+10 2.35E+10 1.00E+0 4.64E+10 2.62E+10 2.42E+10 1.07E-5 6.10E+10 3.07E+10 2.86E+10 ! 8.21E-1 3.49E+10 1.92E+10 1.78E+10 5.04E-6 3.52E+10 1.76E+10 1.64E+10 7.43E-1 1.69E+10 9.27E+09 8.60E+09 1.86E-6 4.31E+10 2.16E+10 2.01E+10 6.08E-1 5.52E+10 2.94E+10 2.72E+10 8.76E-7 2.90E+10 1.46E+10 1.36E+10 ! 4.98E-1 4.10E+10 2.30E+10 2.13E+10 4.14E-7 2.52E+10 1.27E+10 1.18E+10 l 3.69E-1 5.17E+10 2.72E+10 2.52E+10 1.00E-7 5.48E+10 2.79E+10 2.57E+10 0.00 1.25E+11 6.52E+10 5.91E+10 ! NOTE: The upper energy of group 1 is 17.33 Mev. 4-7
F' TABLE 4.1-4 l CALCULATED NEUTRON SENSOR REACTION RATES-AND EXPOSURE RATES I AT THE CENTER OF THE SURVEILLANCE CAPSULES- *i AZIMUTHAL ANGLE - 13 DEGREES 23 DEGREES 33 DEGREES
~
Reaction Rate (ros/ nucleus) - Cu63(n,a) 6.82E-17 5.25E-17 4.60E-17 Fe54(n,p) 8.40E-15 5.80E-15 5.17E-15 - NiS8(n.p) 1.16E-14 7.82E-15 7.00E-15 U238(n,f) (Cd) 4.42E-14 2.78E-14 2 '.'i 14 Np237(n,f) (Cd) 4.04E-13 2.31E-13 2.iEL-13 CoS9(n,1) 8.81E-12 4.53E-12 4.16E-12 CoS9(n,1) (Cd) 3.76E-12 1.90E-12 1.77E-12 U238(1,f) 2.23E-15 1.29E-15 1.19E-15 . i1p237(1,f) 6.23E-15 3.60E-15 3.32E-15 Neutron Flux (n/cmIl ( (E > 1.0 MeV) 1.33E+11 7.99E+10 7.31E+10 ( (E > 0.1 MeV) 5.08E+11 2.78E+11 2.56E+11 doa/sec Displacement rate 2.46E-10 1.41E-10 1.29E-10 Exposure Rate Ratios ** p(E > 0.1)/p(E > 1.0) 3.82 3.48 3.50 (dpa/sec]/p(E > 1.0) 1.85E-21 1.76E-21 1.76E-21 U238(1, f)/U238(n, f) 0.0505 0.0464 0.0472 -
?;p237(1,f)/Np237(n,f) 0.0154 0.0156 0.0157 4-8
TABLE 4.1-5 ; i RADIAL GRADIENT CORRECTIONS FOR SENSORS CONTAINED IN ! POINT BEACH UNIT 2 INTERNAL SURVEILLANCE CAPSULES RADIAL LOCATION (cm) j 157.59 158.35- 158.59 ; 13 DEGREE CAPSULE
'. Cu63(n,a) 0.866 1.000 1.040 l Fe54(n,p) 0.856 1.000 1.045 [
. NiS8(n,p) 0.857 1.000 1.046 ;
U238(n,f) (Cd) 0.856 1.000 1.049 Np237(n,f) (Cd) 0.862 1.000 1.050 CoS9(n,1) 0.950 1.000 0.977 ; CoS9(n,1) (Cd) 0.860 1.000 1.047 i i . 23 DEGREE CAPSULE l Cu63(n,a) 0.865 1.000 1.040 l . Fe54(n,p) 0.356 1.000 1.044 ! NiS8(n,p) 0.856 1.000 1.045 ; U238(n,f) (Cd) 0.858 1.000 1.048 f Np237(n,f) (Cd) 0.866 1.000 1.050 , CoS9(n,1) 0.963 -' 1.000 0.972 CoS9(n,1) (Cd) 0.873 1.000 1.042
- 33 DEGREE CAPSULE
- Cu63(n,a) 0.867 1.000 1.040 Fe54(n,p) 0.856 1.000 1.045 NiS8(n,p) 0.856 1.000 1.046 U238(n,f) (Cd) 0.856 1.000 1.048 Np237(n,f) (Cd) 0.863 1.000 1.050 CoS9(n,y) 0.956 1.000 0.975 CoS9(n,1) (Cd) 0.868 1.000 1.045 4-9
TABLE 4.1-6 AZIMUTHAL VARIATION OF FAST NEUTRON FLUX (E > 1.0 MeV) AT THE PRESSURE VESSEL INNER RADIUS - THETA RADIUS (cm) THETA RADIUS (cm) (Dea) 167.84 168.04 168.27 (Deo) 167.84 168.04 168.27 0.25 4.47E+10 4.41E+10 4.35E+10 23.00 1.97E+10 1.95E+10 1.93E+10 O.75 4.46E+10 4.40E+10 4.34E+10 23.43 1.97E+10 1.94E+10 1.92E+10 2.00 4.42E+10 4.37E+10 4.31E+10 23.72 1.98E+10 1.95E+10 1.91E+10 4.00 4.32E+10 4.27E+10 4.22E+10 23.90 1.97E+10 1.94E+10 1.91E+10 6.00 4.18E+10 4.13E+10 4.07E+10 24.11 1.96E+10 1.93E+10 1.90E+10 . 8.00 3.96E+10 3.91E+10 3.86E+10 24.61 1.94E+10 1.92E+09 1.90E+10 10.00 3.66E+10 3.62E+10 3.57E+10 26.00 1.93E+10 1.91E+09 1.89E+10 11.39 3.41E+10 3.37E+10 3.32E+10 28.00 1.91E+10 1.89E+09 1.86E+10 11.84 3.32E+10 3.27E+10 3.22E+10 30.00 1.86E+10 1.84E+09 1.82E+10 12.06 3.29E+10 3.24E+10 3.19E+10 31.39 1.81E+10 1.79E+10 1.76E+10 12.28 3.24E+10 3.19E+10 3.14E+10 31.89 1.79E+10 1.77E+10 1.74E+10 . 12.57 3.16E+10 3.12E+10 3.08E+10 32.10 1.79E+10 1.76E+10 1.73E+10 13.00 3.03E+10 3.00E+10 2.96E+10 32.28 1.79E+10 1.76E+10 1.73E+10 . 13.44 2.94E+10 2.91E+10 2.87E+10 32.57 1.77E+10 1.75E+10 1.72E+10 13.72 2.90E+10 2.85E+10 2.80E+10 33.00 1.74E+10 1.72E+10 1.70E+10 13.90 2.86E+10 2.82E+10 2.77E+10 33.44 1.73E+10 1.71E+10 1.69E+10 14.12 2.81E+10 2.77E+10 2.73E+10 33.72 1.74E+10 1.71E+10 1.68Et10 14.61 2.71E+10 2.68E+10 2.65E+10 33.90 1.73E+10 1.70E+10 1.67E+10 16.00 2.50E+10 2.47E+10 2.44E+10 34.12 1.72E+10 1.70E+10 1.67E+10 18.00 2.29E+10 2.26E+10 2.23E+10 34.61 1.71E+10 1.69E+10 1.67E+10 20.00 2.12E+10 2.10E+10 2.07E+10 36.00 1.69E+10 1.67E+10 1.65E+10 - 21.39 2.06E+10 2.04E+10 2.01E+10 38.00 1.66E+10 1.64E+10 1.62E+10 21.88 2.04E+10 2.01E+10 1.98E+10 40.00 1.63E+10 1.61E+10 1.59E+10 - 22.10 2.03E+10 2.00E+10 1.97E+10 42.00 1.60E+10 1.58E+10 1.56E+10 22.28 2.03E+10 2.00E+10 1.96E+10 44.00 1.58E+10 1.56E+10 1.54E+10 - 22.56 2.00E+10 1.98E+10 1.95E+10 l . Note: The vessel clad / base metal interface is located at a radius of 168.04 cm. ' 1 4-10
l 1 TABLE 4.1-7
SUMMARY
OF EXPOSURE RATES AT THE PRESSURE VESSEL . CLAD / BASE METAL INTERFACE !
'~
THETA Flux (n/cm2-sec) (E > 0.1) doa/sec l (Dea) (E > 1.0) (E > 0.1) doa/sec (E > 1.0) (E > 1.0) i 0.25 4.41E+10 1.20E+11 7.26E-Il 2.72 1.65E-21 4.00 4.27E+10- 1.16E+11 7.04E-Il 2.72 1.65E-21
. 10.00 3.62E+10 1.01E+11 6.04E-11 2.79 1.67E-21 i 14.61 2.68E+10. 7.83E+10 4.57E-Il 2.92 1.71E-21 20.00 2.10E+10 5.84E+10 3.53E-11 2.78 1.68E-21 24.61 1.92E+10 5.32E+10 3.22E-11 2.77 1.68E-21 30.00 1.84E+10 4.96E+10 3.05E-11 2.70 1.66E-21 !
34.61 1.69E+10 4.66E+10 2.82E-11 2.76 1.67E-21 I 40.00 1.61E+10 4.19E+10 2.64E-11 2.60 1.64E-21 44.00 1.56E+10 4.06E+10 2.56E-11 2.60 1.64E-21 l l t P e , I 4-11 i
.i
TABLE 4.1-8 1 RELATIVE RADIAL DISTRIBUTION OF NEUTRON FLUX (E > 1.0 MeV) WITHIN THE PRESSURE VESSEL WALL RADIUS AZIMUTHAL ANGLE
~l (cm) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES i I
168.04 1.000 1.000 1.000 1.000 168.27 0.986 0.988 0.987 0.988 ! 168.88 0.972 0.976 0.975 0.975 - l 169.75 0.922 0.929 0.927 0.928 : 170.89 0.841 0.850 0.G47 0.850 - 172.17 0.734 0.744 0.740 0.745 173.49 0.621 0.631 0.627 0.633 174.90 0.519 0.530 0.525 0.531 : 176.30 0.426 0.437 0.431 0.438 l i$:l! l:U9 ll l'26 l .! 180.42 0.235 0.247 0.242 0.246 : 181.51 0.187 0.197 0.192 0.196 - 182.60 0.156 0.167 0.163 0.165 183.90 0.129 0.141 0.137 0.139 l 184.55 0.101 0.113 0.111 0.112 Note: Base Metal Inner Radius = 168.04 cm. 1/4 T Location - 172.17 cm. 1/2 T Location = 176.30 cm. - 3/4 T Location - 180.42 cm. Base Metal Outer Radius = 184.55 cm. t
.r
)
4-12
--- q I
I i TABLE 4.1-9 RELATIVE RADIAL' DISTRIBUTION OF-NEUTRON FLUX (E > 0.1 MeV) ' WITHIN THE PRESSURE VESSEL WALL RADIUS AZIMUTHAL ANGLE -i (cm) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES l 168.04 1.000 1.000 1.000 1.000 l 168.27 1.005 1.006 1.007 1.007 I
. 168.88 1.009 1.013 1.014 1.014 !
169.75 1.006 1.012 1.014 1.015 i 170.89 0.981 0.992 0.996 0.998 j 172.17 0.934 0.949 0.955 0.959 173.49 0.872 0.891 0.898 0.903 ; 174.90 0.803 0.827 0.834 0.840 176.30 0.729 0.757 0.764 0.769 177.50 0.657 0.687 0.695 0.700
, 178.91 0.595 0.629 0.638 0.641 180.42 0.524 0.562 0.572 0.572 '
. 181.51 0.452 0.492 0.503 0.500 ,
182.60 0.400 0.442 0.454 0.450 183.90 0.347 0.393 0.407 0.400 184.55 0.282 0.331 0.350 0.343
)
l Note: Base Metal Inner Radius = 168.04 cm. 1/4 T Location - 172.17 cm. i I 1/2 T Location = 176.30 cm. 3/4 T Location = 180.42 cm. Base Metal Outer Radius = 184.55 cm. ! O e 4-13 I
i TABLE 4.1-10 ! i RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa) WITHIN THE PRESSURE VESSEL WALL RADIUS AZIMUTHAL ANGLE (cm) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES 168.04 1.000 1.000 1.000 1.000 + 168.27 0.988 0.990 0.990 0.990 168.88 0.977 0.981 0.980 0.980 - 169.75 0.937 0.945 0.943 0.943 170.89 0.874 0.886 0.882 0.883 - 172.17 0.790 0.806 0.801 0.303 ; 173.49 0.701 0.720 0.713 0.715 174.90 0.617 0.639 0.630 0.632 , 176.30 0.536 0.561 0.552 0.553 177.50 0.465 0.491 0.482 0.482 : 178.91 0.409 0.437 0.429 0.428 - 180.42 0.350 0.380 0.372 0.370 181.51 0.294 0.324 0.318 0.314 j 182.60 0.256 0.287 0.282 0.277 183.90 0.219 0.251 0.249 0.243 184.55 0.177 0.210 0.211 0.205 Note: Base Metal Inner Radius - 168.04 cm. [ 1/4 T Location - 172.17 cm. j 1/2 T Location = 176.30 cm.
- 3/4 T Location - 180.42 cm.
Base Metal Outer Radius - 184.55 cm.
- t I
i 4-14 I
'4.2 - Fuel Cycle Specific Adjoint Calculations Results of the fuel cycle specific adjoint transport calculations for the first a
17 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, and 17 cavity <icsimetry irradiations as well as for the four sets of surveillance capsule dosimetry withdrawn to date. These results also provide the fuel cycle specific relationship among the surveillance capsule and reactor cavity measurement locations and key positions at the inner radius of the pressure vessel wall. l The core power distributions used in the cycle specific fast neutron exposure calculations for Fuel Cycles 1 through 17 were taken from the fuel cycle design reports applicable to Point Beach Unit 2 [19 through 35). The data extracted from the fuel cycle design reports represented cycle averaged relative fuel assembly powers and burnups as well as cycle averaged relative axial distributions. Therefore, the results of the adjoint evaluation 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. 4
]
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 17 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 4.2-3 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 I o 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 I i* I l 4-15 _ _ _ _ _ _ _ _ - _ --- - - - - - - - - - - - ^ - - - - - - - - - - - - - - - - - - ^ - - - - - - - ' ' - - - - ' - - -
l
.! l displacements (Tables 4.2.13 through 4.2.18).
O O I e Q L e i e GO 1 O 4 4-16
t TABLE 4.2-1 l
.e CALCULATED FAST NEUTRON FLUX (E >.1.0 MeV) AT THE -
CENTER OF REACTOR. VESSEL SURVEILLANCE CAPSULES , d-(E > 1.0 MeV) In/cm2-sec1 " i CYCLE No 13 DEGREES 23 DEGREES '33 DEGREES I I 1- 1.10E+11 6.45E+10 5.86E+10
. 2 1.12E+11 6.83E+10 6.41E+10 l 3.. 1.llE+11 6.70E+10 6.34E+10 i
. . 4 1.05E+11 6.46E+10 6.07E+10 i 5 1.10E+11 6.77E+10 6.09E+10 6 8.41E+10 6.36E+10 6.24E+10 :
7 8.50E+10 5.58E+10 5.27E+10 8 8.59E+10 5.48E+10 5.04E+10 l 9 8.93E+10 5.58E+10 5.10E+10 i
.. 10 8.05E+10 5.47E+10 5.22E+10 l 11 7.62E+10 5.64E+10 5.15E+10 i
. 12 -8.00E+10 5.61E+10 4.92E+10 j 13 7.39E+10 5.29E+10 4.65E+10 14 7.60E+10 5.51E+10 5.07E+10 I 15 7.34E+10 5.32E+10 4.64E+10 l 16 5.91E+10 4.31E+10 4.05E+10 I I
17 5.89E+10 4.38E+10 4.17E+10 i 9 y e.
.. . j k
4-17 - i i
. c. ..
.I j
TABLE 4.2-2 i CALCULATED FAST NEUTRON FLUENCE- (E > 1.0 MeV)-AT.THE. ! i
~ CENTER OF REACTOR VESSEL SURVEILLANCE' CAPSULES-
~'
CYCLE 4 (E > 1.0 MeV) In/cm21 LENGTH CYCLE No (EFPS) 13 DEGREES '23 DEGREES 33 DEGREES l 1 4.81E+07 5.30E+18. 3.10E+18 2.82E+18
~
2 3.32E+07 9.01E+18 5.37E+18 4.95E+18 - 3 2.75E+07 1.21E+19 7.21E+18 6.69E+18 4 2.74E+07 1.49E+19 8.98E+18 8.35E+18 - 5 2.79E+07 1.80E+19 1.09E+19 1.00E+19 l 6 2.73E+07 2.03E+19 1.26E+19 1.17E+19 7 2.82E+07 2.27E+19 .1.42E+19 1.32E+19 l 8 2.70E+07 2.50E+19 1.57E+19 1.46E+19 ! 9 2.50E+07 2.72E+19 1.71E+19 1.59E+19 ! 10 3.77E+07 3.03E+19 1.9]E+19 1.78E+19 . 11 2.68E+07 3.23E+19 2.06E+19- 1.92E+19 , 12 2.52E+07 3.43E+19 2.20E+19 2.05E+19 --1 13 2.55E+07 3.62E+19 2.34E+19 2.16E+19 l 14 2.72E+07 3.83E+19 2.49E+19 2.30E+19 ! 15 2.54E+07 4.01E+19- 2.'62E+19 2.42E+19 ! 16 2.70E+07 4.18E+19 2.74E+19 2.53E+19 l 17 2.66E+07 4.33E+19 2.86E+19 2.64E+19 1 i r 4-18 ! r
TABLE 4.2-3 CALCULATED FAST NEUTRON FLUX (E > 1.0 MeV) AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE 6 (E > 1.0 MeV) In/cm2-sec1 i CYCLE No 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES 1 3.69E+10 2.16E+10 1.48E+10 1.29E+10
. 2 3.76E+10 2.21E+10 1.60E+10 1.41E+10 3 3.74E+10 2.18E+10 1.58E+10 1.41E+10
. 4 3.51E+10 2.07E+10 1.51E+10 1.33E+10 5 3.58E+10 2.17E+10 1.54E+10 1.26E+10 6 2.66E+10 1.74E+10 1.55E+10 1.31E+10 7 2.92E+10 1.72E+10 1.33E+10 1.26E+10 8 3.01E+10 1.74E+10 1.28E+10 1.20E+10 9 3.10E+10 1.80E+10 1.29E+10 1.22E+10
. 10 2.76E+10 1.65E+10 1.31E+10 1.22E+10 11 2.30E+10 1.60E+10 1.32E+10 1.13E+10 12 2.50E+10 1.66E+10 1.27E+10 1.06E+10 13 2.31E+10 1.54E+10 1.21E+10 1.02E+10 14 2.33E+10 1.58E+10 1.30E+10 1.17E+10 15 2.26E+10 1.54E+10 1.21E+10 9.94E+09 16 1.84E+10 1.24E+10 1.03E+10 9.36E+09 17 1.84E+10 1.24E+10 1.05E+10 9.84E+09 6 4 4-19
=
TABLE 4.2-4 CALCULATED FAST NEUTRON FLUENCE (E > 1.0 MeV) AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE
~
CYCLE 4 (E > 1.0 MeV) In/cm21 LENGTH
- CYCLE No (EFPS) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES 1 4.81E+07 1.78E+18 1.04E+18 7.09E+17 6.19E+17
~
2 3.32E+07 3.02E+18 1.77E+18 1.24E+18 1.09E+18 - 3 2.75E+07 4.05E+18 2.37E+18 1.68E+18 1.48E+18 4 2.74E+07 5.02E+18 2.94E+18 2.09E+18 1.84E+13 - 5 2.79E+07 6.01E+18 3.54E+18 2.52E+18 2.19E+18 6 2.73E+07 6.74E+18 4.02E+18 2.95E+18 2.55E+18 7 2.82E+07 7.56E+18 4.51E+18 3.32E+18 2.91E+18 8 2.70E+07 8.38E+18 4.97E+18 3.66E+18 3.23E+18 9 2.50E+07 9.15E+18 5.43E+18 3.99E+18 3.53E+18 10 3.77E+07 1.02E+19 6.05E+18 4.48E+18 3.99E+18 . 11 2.68E+07 1.08E+19 6.47E+18 4.83E+18 4.30E+18 12 2.52E+07 1.14E+19 6.89E+18 5.15E+18 4.56E+18 - 13 2.55E+07 1.20E+19 7.29E+18 5.46E+18 4.82E+18 14 2.72E+07 1.27E+19 7.72E+18 5.82E+18 5.14E+18 15 2.54E+07 1.32E+19 8.11E+18 6.12E+18 5.40E+18 16 2.70E+07 1.37E+19 8.44E+18 6.40E+18 5.65E+18 17 2.66E+07 1.42E+19 8.77E+18 6.68E+18 5.91E+18 e . e 4-20 l
TABLE 4.2-5 CALCULATED FAST NEUTRON FLUX (E > 1.0 MeV) AT THE CAVITY SENSOR SET LOCATIONS d (E > 1.0 MeV) In/cm2-seci l CYCLE No O DEGREES 15 DEGREES 30 DEGREES 45 DEGREES 1 2.68E+09 2.04E+09 1.40E+09 1.17E+09
, 2 2.73E+09 2.10E+09 1.50E+09 1.27E+09' 3 2.71E+09 2.07E+09 1.48E+09 1.27E+09 4 2.55E+09 1.97E+09 1.41E+09 1.20E+09 5 2.62E+09 2.04E+09 1.43E+09 1.17E+09 6 1.99E+09 1.66E+09 1.39E+09 1.18E+09 7 2.11E+09 1.66E+09 1.24E+09 1.10E+09 8 2.17E+09 1.66E+09 1.21E+09 1.05E+09 9 2.24E+09 1.72E+09 1.23E+09 1.07E+09
- l. 10 2.00E+09 1.58E+09 1.22E+09 1.07E+09 11 1.75E+09 1.49E+09 1.20E+09 1.02E+09
- 12 1.88E+09 1.55E+09 1.18E+09 9.71E+08 13 1.73E+09 1.43E+09 1.llE+09 9.28E+08 14 1 76E+09 1.48E+09 1.19E+09 1.04E+09 15 1.71E+09 1.43E+09 1.llE+09 9.15E+08 16 1.39E+09 1.17E+09 9.44E+08 8.28E+08 17 1.40E+09 1.17E+09 9.68E+08 8.60E+08 l
e 4-21
p., .. . .. .
=-
TABLE 4.2-6 CALCULATED FAST NEUTRON FLUENCE (E > 1.0 MeV) AT THE CAVITY SENSOR SET LOCATIONS
~
CYCLE & (E > 1.0 MeV) In/cm21 LENGTH CYCLE No (EFPS) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES 1 4.81E+07 1.29E+17 9.82E+16 6.75E+16 5.61E+16 3.32E+07 2.19E+17 1.68E+17 1.17E+17 9.83E+16 .' 2 3 2.75E+07 2.94E+17 2.25E+17 1.58E+17 1.33E+17 4 2.74E+07 3.64E+17 2.79E+17 1.97E+17 1.66E+17 - 5 2.79E+07 4.37E+17 3.36E+17 2.37E+17 1.99E+17 6 2.73E+07 4.91E+17 3.81E+17 2.75E+17 2.31E+17 7 2.82E+07 5.51E+17 4.28E+17 3.10E+17 2.62E+17 8 2.70E+07 6.09E+17 4.73E+17 3.42E+17 2.90E+17 9 2.50E+07 6.65E+17 5.16E+17 3.73E+17 3.17E+17 10 3.77E+07 7.41E+17 5.75E+17 4.19E+17 3.57E+17 9 11 2.68E+07 7.88E+17 6.15E+17 4.51E+17 3.85E+17 12 2.52E+07 8.35E+17 6.54E+17 4.81E+17 4.09E+17 . 13 2.55E+07 8.79E+17 6.90E+17 5.09E+17 4.33E+17 14 2.72E+07 9.27E+17 7.31E+17 5.42E+17 4.61E+17 15 2.54E+07 9.71E+17 7.67E+17 5.70E+17 4.84E+17 16 2.70E+07 1.01E+18 7.99E+17 5.95E+17 5.07E+17 17 2.66E+07 1.05E+18 8.30E+17 6.21E+17 5.30E+17 9 4 4-22 i
TABLE 4.2-7 CALCULATED FAST NEUTRON FLUX (E > 0.1 MeV) AT THE l CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES I 6 (E > 0.1 MeV) In/cm2-seci CYCLE No 13 DEGREES 23 DEGREES 33 DEGREES l 1 4.21E+11 2.25E+11 2.05E+11 '[
. 2 4.26E+11 2.38E+11 2.24E+11 3 4.23E+11 2.33E+11 2.22E+11 l
. 4 4.00E+11 2.25E+11 2.12E+1.
5 4.18E+11 2.35E+11 2.12E+11 ! 6 3.21E+11 2.21E+11 2.18E+11 i 7 3.25E+11 1.94E+11 1.84E+11 8 3.28E+11 1.91E+11 1.76E+11 l 9 3.41E+11 1.94E+11 1.79E+11
. 10 3.08E+11 1.90E+11 1.83E+11 l 11 2.91E+11 1.96E+11 1.80E+11 l 12 3.06E+11 1.95E+11 1.72E+11 ;
13 2.82E+11 1.84E+11 1.63E+11 i 14 2.90E+11 1.92E+11 1.78E+11 l 15 2.80E+11 1.85E+11 1.63E+11 ; 16 2.26E+11 1.50E+11 1.42E+11 17 2.25E+11 1.52E+11 1.46E+11 I l
. 1 4-23
_au .- a a u a-- -. .- w - A ..a.a __u3 -,..a,.n ..o.u_,. s. I t TABLE 4.2-8 f CALCULATED FAST NEUTRON FLUENCE (E > 0.1 MeV) AT THE CENTER OF REACTOR VESSEL SURVEPMNCE CAPSULES CYCLE + (E > 0.1 MeV) In/cm21 - LENGTH f CYCLE No (EFPS.)_ 1LDfGREES 23 DEGREES 33 DEGREES ; I 4.81E+07 2.03E+19 1.08E+19 9.87E+18 ;
~
2 3.32E+07 3.44E+19 1.87E+19 1.73E+19 - f 3 2.75E+07 4.60E+19 2.51E+19 2.34E+19 4 2.74E+07 5.70E+19 3.13E+19 2.92E+19 - l 5 2.79E+07 6.87E+19 3.78E+19 3.52E+19 i 6 2.73E+07 7.74E+19 4.39E+19 4.11E+19 f 7 2.82E+07 8.66E+19 4.94E+19 4.63E+19 I 8 2.70E407 9.55E+19 5.45E+19 5.llE+19 ! 9 2.50E+07 1.04E+20 5.94E+19 5.56E+19 10 3.77E+07 1.16E+20 6.65E+19 6.24E+19 . L 11 2.68E+07 1.23E+20 7.18E+19 6.73E+19 I i 12 2.52E+07 1.31E+20 7.67E+19 7.16E+19 -- i 13 2.55E+07 1.38E+20 8.14E+19 7.58E+19 [ 14 2.72E+07 1.46E+20 8.66E+19 8.06E+19 ! 15 2.54E+07 1.53E+20 9.13E+19 8.47E+19 f 16 2.70E+07 1.60E+20 9.54Et19 -8.86E+19 ; 17 2.66E+07 1.66E+20 9.94E+19 9.24E+19 i-i
. i i
i 4-24 i
?
[ t TABLE 4.2-9 ; CALCULATED FAST NEUTRON FLUX (E > 0.1 MeV) AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE 6 (E > 0.1 MeV) In/cm2-sec1 l CYCLE No O DEGREES 15 DEGREES 30 DEGREES 45 DEGREES 1 1.00E+11 6.31E+10 3.98E+10 3.34E+10
. 2 1.02E+11 6.45E+10 4.33E+10 3.68E+10 l 3 1.02E+11 6.38E+10 4.26E+10 3.67E+10 4 9.56E+10 6.04E+10 4.08E+10 3.45E+10
~
5 9.74E+10 6.33E+10 4.17E+10 3.27E+10 6 7.24E+10 5.08E+10 4.20E+10 3.42E+10 , 7 7.93E+10 5.03E+10 3.58E+10 3.28E+10 8 8.20E+10 5.07E+10 3.45E+10 3.12E+10 - 9 8.44E+10 5.27E+10 3.47E+10 3.17E+10 , 10 7.49E+10 4.81E+10 3.54E+10 3.17E+10 11 6.25E+10 4.66E+10 3.56E+10 2.95E+10 12 6.81E+10 4.84E+10 3.44E+10 2.75E+10 13 6.27E+10 4.51E+10 3.28E+10 2.65E+10 14 6.33E+10 4.63E+10 3.50E+10 3.05E+10 l 15 6.14E+10 4.49E+10 3.28E+10 2.59E+10 i 16 5.00E+10 3.60E+10- 2.77E+10 2.43E+10 17 5.02E+10 3.61E+10 2.84E+10 2.56E+10 l t G 9 4-25
i TABLE 4.2-10
- CALCULATED FAST NEUTRON FLUENCE (E > 0.1 MeV) AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE I
CYCLE + (E > 0.1 MeV) In/cm21 l LENGTH CYCLE No (EFPS) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES 1 4.81E+07 4.83E+18 3.03E+18 1.92E+18 1.61E+18 , 2 3.32E+07 8.22E+18 5.18E+18 3.35E+18 2.83E+18 .' ! 3 2.75E+07 1.10E+19 6.93E+18 4.52E+18 3.84E+18 j 4 2.74E+07 1.36E+19 8.58E+18 5.64E+18 4.78E+18 . 5 2.79E+07 1.64E+19 1.03E+19 6.81E+18 5.70E+18 ; 6 2.73E+07 1.83E+19 1.17E+19 7.95E+18 6.63E+18 l 7 2.82E+07 2.06E+19 1.32E+19 8.96E+18 7.55E+18 ! 8 2.70E+07 2.28E+19 1.45E+19 9.89E+18 8.40E+18 j 9 2.50E+07 2.49E+19 1.58E+19 1.08E+19 9.19E+18 . 10 3.77E+07 2.77E+19 1.77E+19 1.21E+19 1.04E+19 . < 11 2.68E+07 2.94E+19 1.89E+19 1.30E+19 1.12E+19 12 2.52E+07 3.11E+19 2.01E+19 1.39E+19 1.19E+19 - l 13 2.55E+07 3.27E+19 2.13E+19 1.48E+19 1.25E+19 14 2.72E+07 3.44E+19 2.25E+19 1.57E+19 1.34E+19 ; 15 2.54E+07 3.60E+19 2.37E+19 1.65E+19 1.40E+19 , 16 2.70E+07 3.73E+19 2.46E+19 1.73E+19 1.47E+19 ; 17 2.66E+07 3.87E+19 2.55E+19 1.80E+19 1.54E+19 4-26 i
- _ ~ _ - . . - . -
TABLE ~4.2-11 CALCULATED FAST NEUTRON FLUX (E > 0.1 MeV) AT-THE CAVITY SENSOR SET LOCATIONS 6 (E > 0.1 MeV) In/cm2-sec1 ! CYCLE No 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES ; 1 2.13E+10 1.75E+10 1.23E+10 9.71E+09
. 2 2.17E+10 1.80E+10 1.32E+10 1.06E+10 3 2.15E+10 1.78E+10 1.30E+10 1.05E+10 4 2.02E+10 1.69E+10 1.24E+10 9.96E+09 '
5 2.08E+10 1.75E+10 1.26E+10 9.75E+09 6 1.58E+10 1.43E+10 1.22E+10 9.84E+09 7 1.68E+10 1.42E+10 1.09E+10 9.12E+09 ! 8 1.72E+10 1.43E+10 1.06E+10 8.76E+09 9 1.78E+10 1.47E+10 1.08E+10 8.89E+09 . 10 1.59E+10 1.35E+10 1.07E+10 8.90E+09 11 1.39E+10 1.27E+10 1.05E+10 8.52E+09 12 1.49E+10 1.33E+10 1.03E+10 8.08E+09 < 13 1.37E+10 1.23E+10 9.79E+09 7.72E+09 14 1.40E+10 1.27E+10 1.04E+10 8.63E+09 15 1.36E+10 1.23E+10 9.73E+09 7.61E+09 L 16 1.10E+10 1.00E+10 8.30E+09 6.89E+09 17 1.llE+10 1.01E+10 8.51E+09 7.16E+09 i 4-27 ,
i TABLE 4.2-12 CALCULATED FAST NEUTRON FLUENCE (E > 0.1 MeV) AT THE CAVITY SENSOR SET LOCATIONS CYCLE + (E > 0.1 MeV) In /ctn21 j LENGTH
~
j CYCLE No (EFPS) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES i 1 4.81E+07 1.02E+18 8.42E+17 5.94E+17 4.67E+17
, {
2 3.32E+07 1.74E+18 1.44E+18 1.03E+18 8.18E+17 . 3 2.75E+07 2.33E+18 1.93E+18 1.39E+18 1.llE+18 4 2.74E+07 2.89E+18 2.39E+18 1.73E+18 1.38E+18 - 5 2.79E+07 3.47E+18 2.88E+18 2.08E+18 1.65E+18 l 6 2.73E+07 3.90E+18 3.27E+18 2.42E+18 1.92E+18 i 7 2.82E+07 4.37E+18 3.67E+18 2.72E+18 2.18E+18 P 2.70E+07 4.84E+18 4.06E+18 3.01E+18 2.41E+18 9 2.50E+07 5.28E+18 4.42E+18 3.28E+18 2.64E+18 10 3.77E+07 5.88E+18 4.93E+18 3.68E+18 2.97E+18 . i 11 2.68E+07 6.25E+18 5.28E+18 3.97E+18 3.20E+18 ! 12 2.52E+07 6.63E+18 5.61E+18 4.23E+18 3.40E+18 - 13 2.55E+07 6.98E+18 5.92E+18 4.48E+18 3.60E+18 14 2.72E+07 7.36E+18 6.27E+18 4.76E+18 3.84E+18 15 2.54E+07 7.71E+18 6.58E+18 5.01E+18 4.03E+18 16 2.70E+07 8.00E+18 6.85E+18 5.23E+18 4.22E+18 4 17 2.66E+07 8.30E+18 7.12E+18 5.46E+18 4.41E+18 l .. l m I e 4-28 l _ _ _ . - - _ _ _ _ _ _ _ _ _ _ _ .
TABLE 4.2-13 ! CALCULATED IRON ATOM DISPLACEMENT RATE AT THE .
- . CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES l
~ Disolacement Rate Idoa/sec1 CYCLE No I 13 DEGREES 23 DEGREES 33 DEGREES 1 2.04E-10 1.14E-10 1.03E-10 .
. 2 2.06E-10 1.20E-10 1.13E-10 [
3 .2.05E-10 1.18E-10 1.12E-10 f 4 1.94E-10 1.14E-10 1.07E-10 l 5 2.03E-10 1.19E-10 1.07E-10 6 1.56E-10 1.12E-10 1.10E-10 7 1.57E 9.82E-11 9.28E-Il 8 1.59E-10 9.64E-Il 8.87E-11 . 9 1.65E-10 9.82E-Il 8.98E-11 ! . 10 1.49E-10 9.62E-Il 9.19E-11 ! 11 1.41E-10 9.93E-Il 9.06E-11 l 12 1.48E-10 9.87E-11 8.66E-11 ! 13 1.37E-10 9.32E-Il 8.19E-11 14 1.41E-10 9.69E-11 8.93E-Il f 15 1.36E-10 9.37E-11 8.17E-Il I 16 1.09E-10 7.59E-Il 7.14E-Il ! 17 1.09E-10 7.70E-11 7.33E-Il i I I f 4-29 l i
TABLE 4.2-14 CALCULATED 1RON ATOM DISPLACEMENTS AT THE : CENTER OF REACTOR VESSEL SURVEILLANCE CAPSULES CYCLE Disolacements Idoal LENGTH
~
CYCLE No (EFPS) 13 DEGREES 23 DEGREES 33 DEGREES 1 4.81E+07 9.81E-03 5.46E-03 4.96E-03
~
2 3.32E+07 1.67E-02 9.45E-03 8.71E-03 3 2.75E+07 2.23E-02 1.27E-02 1.18E-02 4 2.74E+07 2.76E-02 1.58E-02 1.47E 5 2.79E+07 3.33E-02 1.91E-02 1.77E-02 6 2.73E+07 3.75E-02 2.22E-02 2.07E-02 7 2.82E+07 4.19E-02 2.50E-02 2.33E-02 8 2.70E+07 4.62E-02 2.76E-02 2.57E-02 9 2.50E+07 5.04E-02 3.00E-02 2.79E-02 10 3.77E+07 5.60E-02 3.36E-02 3.14E-02 . 11 2.68E+07 5.98E-02 3.63E-02 3.38E-02 12 2.52E+07 6.35E-02 3.88E-02 3.60E-02 -; 13 2.55E+07 6.70E-02 4.12E-02 3.81E-02 14 2.72E407 7.08E-02 4.38E-02 4.05E-02 15 2.54E+07 7.42E-02 4.62E-02 4.26E-02 16 2.70E+07 7.73E-02 4.82E-02 4.45E-02 17 2.66E+07 8.02E-02 5.03E-02 4.65E-02 O O 4-30 i
TABLE 4.2-15 CALCULATED 1RON ATOM DISPLACEMENT RATE AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE ~ Displacement Rate Idoa/seci CYCLE No 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES 1 6.09E-11 3.67E-Il 2.45E-Il 2.llE-Il
. 2 6.20E-ll 3.76E-Il 2.66E-11 2.32E-11 3 6.17E-11 3.71E-Il 2.62E-Il 2.31E-11 4 5.80E-11 3.51E-Il 2.51E-Il 2.18E-Il 5 5.91E-Il 3.69E-11 2.56E-11 2.07E-Il 6 4.39E-11 2.96E-Il 2.58E-11 2.15E-11 7 4.81E-Il 2.93E-11 2.20E-Il 2.07E-Il 8 4.97E-Il 2.95E-Il 2.12E-11 1.97E-Il 9 5.12[-11 3.07E-Il 2.14E-11 2.00E-Il
. 10 4.55E-11 2.80E-11 2.18E-Il 2.00E-Il 11 3.79E-Il 2.71E-11 2.19E-Il 1.86E-Il 12 4.13E-11 2.82E-Il 2.llE-11 1.73E-Il 13 3.80E-Il 2.63E-11 2.01E-11 1.67E-Il 14 3.84E-Il 2.69E-11 2.15E-Il 1.93E-11 15 3.72E-Il 2.61E-11 2.02E-Il 1.63E-Il 16 3.04E-11 2.12E-Il 1.75E-Il 1.61E-11 17 3.04E-11 2.12E-11 1.75E-Il 1.61E-Il O m 4-31
TABLE 4.2-16 CALCULATED IRON ATOM DISPLACEMENTS AT THE PRESSURE VESSEL CLAD / BASE METAL INTERFACE CYCLE Disolacements Idoal LENGTH CYCLE No (EFPS) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES 1 4.81E+07 2.93E-03 1.77E-03 1.18E-03 1.01E-03
~
2 3.32E+07 4.99E-03 3.01E-03 2.06E-03 1.78E-03 - 3 2.75E+07 6.69E-03 4.03E-03 2.78E-03 2.42E-03 4 2.74E+07 8.28E-03 5.00E-03 3.47E-03 3.02E-03 - 5 2.79E+07 9.92E-03 6.03E-03 4.18E-03 3.59E-03 6 2.73E+07 1.11E-02 6.83E-03 4.89E-03 4.18E-03 7 2.82E+07 1.25E-02 7.66E-03 5.51E-03 4.76E-03 8 2.70E+07 1.38E-02 8.46E-03 6.08E-03 5.30E-03 9 2.50E+07 1.51E-02 9.22E-03 6.62E-03 5.80E-03 10 3.77E+07 1.68E-02 1.03E-02 7.44E-03 6.55E-03 . 11 2.68E+07 1.78E-02 1.10E-02 8.02E-03 7.05E-03 12 2.52E+07 1.89E-02 1.17E-02 8.56E-03 7.49E-03 - 13 2.55E+07 1.98E-02 1.24E-02 9.07E-03 7.91E-03 14 2.72E+07 2.09E-02 1.31E-02 9.66E-03 8.44E-03 15 2.54E+07 2.18E-02 1.38E-02 1.02E-02 8.85E-03 16 2.70E+07 2.27E-02 1.44E-02 1.06E-02 9.26E-03 17 2.66E+07 2.35E-02 1.50E-02 1.llE-02 9.69E-03 l 4-32 l
TABLE 4.2-17 CALCULATED IRON ATOM DISPLACEMENT RATE AT THE CAVITY SENSOR SET LOCATIONS ,
. 1 Displacement rate Idoa/seci ,
CYCLE No 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES ! 1 8.12E-12 6.57E-12 4.62E-12 3.69E-12 -
'. 2 8.27E-12 6.76E-12 4.94E-12 4.02E-12 !
3 8.21E-12 6.67E-12 '4.86E 'E 4.00E-12
. 4 7.71E 6.34E-12 4.65E-12 3.78E-12 5 7.93E-12 6.56E-12 4.72E-12 3.70E-12 6 6.04E-12 5.36E-12 4.58E-12 3.74E-12 .i 7 6.39E-12 5.34E-12 4.07E-12 3.46E-12 i 8 6.57E-12 5.35E-12 3.98E-12 3.33E-12 ;
9 6.80E-12 5.53E-12 4.05E-12 3.38E-12 !
.. 10 6.07E-12 5.09E-12 4.02E-12 3.38E-12 11 5.30E-12 4.78E-12 3.94E-12 3.24E-12 ! . 12 5.69E-12 4.99E-12~ 3.87E-12 3.07E-12 l 13 5.25E-12 4.62E-12 3.67E-12 2.93E-12 14 5.34E-12 4.78E-12 3.90E-12 3.28E-12 -!
15 5.18E-12 4.60E-12 3.64E-12 2.89E-12 16 4.22E-12 3.75E-12 -3.10E-12 2.62E-12 1 17 4.23E-12 3.77E-12 3.18E-12 2.72E-12 ! f l i i I i
~
i 4-33 ! w - . . . . , , . ,
TABLE 4.2-18 e CALCULATED IRON ATOM DISPLACEMENTS AT THE CAVITY SENSOR SET LOCATIONS CYCLE __pisolacements Idoal LENGTH
~
CYCLE No (EFPS) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES 1 4.81E+07 3.90E-04 3.16E-04 2.22E-04 1.77E-04
~
2 .3.32E+07 6.65E-04 5.41E-04 3.86E-04 3.11E-04 - 3 2.75E+07 8.91E-04 7.24E-04 5.20E-04 4.21E-04 4 2.74E+07 1.10E-03 8.98E-04 6.48E-04 5.24E-04 - 5 2.79E+07 1.32E-03 1.08E-03 7.79E-04 6.28E-04 6 2.73E+07 1.49E-03 1.23E-03 9.04E-04 7.30E-04 7 2.82E+07 1.67E-03 1.38E-03 1.02E-03 8.27E-04 8 2.70E+07 1.85E-03 1.52E-03 1.13E-03 9.17E-04 9 2.50E+07 2.02E-03 1.66E-03 1.23E-03 1.00E-03 10 3.77E+07 2.24E-03 1.85E-03 1.38E-03 1.13E-03 . , 11 2.68E+07 2.39E-03 1.98E-03 1.48E-03 1.22E-03 12 2.52E+07 2.53E-03 2.11E-03 1.58E-03 1.29E-03 - ! 13 2.55E+07 2.66E-03 2.22E-03 1.6BE-03 1.37E-03 ! 14 2.72E+07 2.81E-03 2.35E-03 1.78E-03 1.46E-03 ! 15 2.54E+07 2.94E-03 2.47E-03 1.87E-03 1.53E-03 16 2.70E+07 3.05E-03 2.57E-03 1.96E-03 1.60E-03 17 2.66E+07 3.17E-03 2.67E-03 2.04E-03 1.67E-03 O e 4-34
l SECTION 5.0 . EVALUATIONS OF SURVEILLANCE CAPSULE 00SIMETRY : l ;
~ !
In this section, the results of the evaluations of the four neutron sensor l 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 WITHDRA'.*AL ' IRRADIATION CAPSULE ID LOCATION TIME TIME (EFPS) - V 13 DEGREES END OF CYCLE 1 4.81E+07 T 23 DEGREES END OF CYCLE 3 1.09E+08 R 13 DEGREES END OF C': *.E 5 1.64E+08 S 33 DEGREES END OF Licti 16 4.66E+08
. I i
5.1 - Measured Reaction Rates ! With the exception of Capsule V, radiometric counting of each of these [ data sets was accomplished by Westinghouse [35, through 37] using t? ' procedures discussed in Section 3.0 of this report. 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 Battelle Memorial Institute [38]. However, in this case, the i
- measured specific activities were not published.
The irradiation history of the Point Beach Unit 2 reactor during the first i 16 fuel cycles is also listed in Appendix A. The irradiation history was l obtained from NUREG-0020, " Licensed Operating Reactors Status Summary l Report" for the applicable operating periods. In addition to the reactor power history, for the multiple cycle irradiations (Capsules T, R, and S), l the flux level adjustment factors for each cycle are also tabulated in 5-1
- .- . . . = - _- - -- .-
l t i Appendix A. These adjustment factors were determined from the fuel cycle specific adjoint calculations described in Section 4.0 of this report. j Based on the irradiation history and associated flux level adjustment
- l factors, the individual sensor characteristics, and the measured specific activities, reaction rates averaged over the appropriate irradiation ;
periods and referenced to a core power level of 1518 MWt were computed for- I the sensor sets removed from Capsules T, R, and S. In the case of Capsule V, reaction rates were developed directly from the derived neutron flux # and spectrum averaged reaction cross-sections reported in Reference 38. [ 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. - l 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, for 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 include adjustments for 7,f reactions occuring over the . f respective irradiation periods. 5.2 - Results of the Least Squares Adjustment Procedure The results of the application of the least squares adjustment procedure l to the four sets of surveillance capsule dosimetry are provided in Tables : 5.2-1 through 5.2-4. In these tables, the derived exposure experienced by each capsule along with data illustrating the fit of both the a priori 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 Tables 5.2-1 through 5.2-4, it ' should be noted that the columns labeled "a priori calc" were obtained by normalizing the neutron spectral data from Table 4.1-3 to the measured - 9 5-2
1 i Fe-54 (n.p) reaction rates'from each sensor set as discussed in Section ; 3.0. Thus, the comparisons illustrated in Tables 5.2-1 through 5.2-4 ! indicate only the degree to which the relative neutron energy spectra { matched the measured sensor data before and after adjustment. These data l are not meant to provide an absolute comparison of calculation and :
- i measurement. Absolute comparisons are discussed in Section 7.0 of'this !
report. 3 b 9
~
l i a i l l l 1
*6 1
i 4 I I l 5-3 i
t TA8LE 5.1-1
SUMMARY
OF REACTION RATES DERIVED FROM MULTIPLE F0Il set!SOR SETS i WITHDRAWN FROM INTERNAL SURVEILLANCE CAPSULES
~
REACTION RATE (ros/ nucleus) CAPSULE CAPSULE CAPSULE CAPSULE , REACTION V T R S ,
~
*Cu63(na)Co60 6.28E-17 4.86E-17 6.78E-17 4.29E-17 -
*Fe54(np)Mn54 7.74E-15 5.53E-15 7.79E-15
*NiS8(np)CoS8 7.29E-15 1.11E-14 7.14E-15 .
U238(nf)Csl37 5.04E-14 2.71E-14 4.64E-14 2.52E-14 Np237(nf)Cs137 3.93E-13 2.30E-13' 4.12E-13 1.88E-13
*CoS9(ny)Co60 7.63E-12 5.08E-12 9.39E-12 3.78E-12 CoS9(ny)Co60 3.21E-12 2.02E-12 3.93E-12 1.68E-12 -
* - Bare foil, all others were cadmium covered
.l I
I t 1 9 5-4
TABLE 5.2-1 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE V DOSIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 1 ~ A PRIORI ADJUSTED PARAMETER VALUE VALUE UNCERTAINTY ( (E > 1.0 Mev) 1.24E+11 1.48E+11 6% ( (E > 0.1 Mev) 4.76E+11 5.36E+11 13% ,
. p (E < 0.414 ev) 1.65E+11 1.82E+11 19% ;
p (Total) 1.23E+12 1.28E+12- 13% dpa/sec 2.24E-10 2.53E-10 9% COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE V . REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 6.28E-17 8.67E-17 6.41E-17 1.38 1.02 Fe-54 (n.p) 7.74E-15 8.14E-15 7.88E-15 1.05 1.02 U-238 (n,f) (Cd) 5.04E-14 4.08E-14 4.78E-14 0.81 0.95 Np-237 (n,f) (Cd) 3.93E-13 3.65E-13 4.01E-13 0.93 1.02 Co-59 (n.7) 7.63E-12 7.09E-12 7.62E-12 0.93 1.00 Co-59 (n,7) (Cd) 3.21E-12 3.15E-12 3.21E-12 0.98 1.00 M e 5-8 i n,- -
TABLE 5.2-2 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE T DOSIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 3
~
A PRIORI ADJUSTED PARAMETER VALUE VALUE UNCERTAINTY
~
p (E > 1.0 Mev) 7.43E+10 8.23E+10 6% p (E > 0.1 Mev) 2.62E+11 2.84E+11 13%
~
( (E < 0.414 ev) 8.57E+10 1.23E+11 18% - u (Total) 6.48E+11 7.26E+11 13% dpa/sec 1.29E-10 1.40E-10 8% - COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE T REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 4.86E-17 6.64E-17 5.04E-17 1.37 1.04 Fe-54 (n,p) 5.50E-15 5.31E-15 5.34E-15 0.97 0.97 Ni-58 (n p) 7.29E-15 7.08E-15 7.21E-15 0.97 0.99 U-238 (n,f) (Cd) 2.71E-14 2.53E-14 2.73E-14 0.93 1.01 Np-237 (n,f) (Cd) 2.30E-13 2.10E-13 2.31E-13 0.91 1.00 Co-59 (n,1) 5.08E-12 3.66E-12 5.07E-12 0.72 1.00 - Co-59 (n,1) (Cd) 2.02E-12 1.60E-12 2.02E-12 0.79 1.00 O e 5-6
I TABLE 5.2-3 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE R DOSIMETRY WITHDRAWN AT THI FND OF FUEL CYCLE 5 ! t' A PRIORI ADJUSTED PARAMETER VALUE VALUE UNCERTAINTY 4 (E > 1.0 Mev) 1.28E+11 1.42E+11 6% ; p (E > 0.1 Mev) 4.90E+11 5.40E+11 13% ! 4 (E < 0.414 ev) 1.69E+11 2.22E+11 18% ( (Total) 1.26E+12 1.40E+12 13% dpa/sec 2.31E-10 2.51E-10 9% i 4 COMPARIS0N OF MEASURED AND CALCULATED SENSOR RFACTION RATES SURVEILLANCE CAPSULE R f REACTION RATE (ros/ nucleus) C/M ; i A PRIORI ADJUSTED j REACTJQR_ MEASURED CALC. CALC. A PRIORI ADJUSTED i Cu-63 (n,a) 6.78E-17 8.92E-17 6.90E-17 1.32 1.02 ! Fe-54 (n.p) 7.79E-15 8.37E-15 7.89E-15 1.07 1.01 ' Ni-58 (n,p) 1.11E-14 1.13E-14 1.10E-14 1.01 0.99 l U-238 (n,f) (Cd) 4.64E-14 4.19E-14 4.52E-14 0.90 0.98 Np-237 (n,f) (Cd) 4.06E-13 3.76E-13 4.10E-13 0.93 1.01 l
- Co-59 (n,1) 9.39E-12 7.30E-12 9.37E-12 0.78 1.00 Co-59 (n,y) (Cd) 3.93E-12 3.24E-12 3.92E-12 0.82 1.00 i O
e e 6 l 5-7
)
i j
TABLE 5.2-4 DERIVED EXP03URE RATES FROM SURVEILLANCE CAPSULE S DOSIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 16
~
A PRIORI ADJUSTED PARAMETER VALUE VALUE UNCERTAINTY d (E > 1.0 Mev) 7.48E+10 7.44E+10 6% ( (E > 0.1 Mev) 2.66E+11 2.41E+11 13%
.4 (E < 0.414 ev) 8.58E+10 8.64E+10 19% .'
( (Total) 6.58E+11 6.01E+11 13% dpa/sec 1.30E-10 1.22E-10 8% - COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE S REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 4.29E-17 6.39E-17 4.45E-17 1.49 1.04 Ni-58 (n,p) 7.14E-14 6.96E-14 6.86E-14 0.98 0.96 U-238 (n,f) (Cd) 2.52E-14 2.52E-14 2.50E-14 1.00 0.99 Np-237 (n,f) (Cd) 1.88E-13 2.12E-13 1.92E-13 1.13 1.02 Co-59 (n,1) 3.78E-12 3.69E-12 3.78E-12 0.98 1.00 Co-59 (n,1) (Cd) 1.68E-12 1.63E-12 1.68E-12 0.97 1.00 - e ! 5-8
SECTION 6.0 i [ EVALUATIONS OF REACTOR CAVITY D0SIMETRY L 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 at the beginning of Fuel Cycle 15; and, to date, has included measurement
. enluations at the conclusion of Cycles 15,16, and 17. The evaluation of each set of data was accomplished using a consistent approach based on the
. methodology discussed in Section 3.0, resulting in an accurate data base defining the exposure of the reactor vessel wall.
i 6.1 - Cycle 15 Results t
, 6.1.1 - Measured Reaction rates i
During the Cycle 15 irradiation, seven multiple foil sensor set; and four stainless steel gradient chains were deployed in the reactor 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 [1]: CAPSULE IDENTIFICATION AZIMUTH VESSEL CORE CORE CORE (dearees) SUPPORT TOP MIDPLANE BOTTOM
. 0 XX G H I '
15 J
.. 30 K 45 L l
The contents of each of these irradiation capsules is specified in Reference 1 and, for completeness, is also included in Appendix B to this report. , 6-1 I
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 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 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,1) 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) and Np-237 (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 and the effects of 7,f reactions in the U-238 sensors as well for the effects of 7,f reactions in the Np-237 monitors. L 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 measurements 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 a priori 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. 6-2
i I In regard to the comparisons listed in Tables 6.1-5 through 6.1-11, it should be noted that the columns labeled "a priori cale" were obtained by normalizing the neutron spectral data from Table 4.1-1 to the measured ; Fe-54 (n,p) reaction rates from each sensor set as discussed in Section 3.0. Thus, the comparisons illustrated in Tables 6.1-5 through 6,1-11 indicate only the degree to which the relative neutron energy spectra matched the measured data before and after adjustment. These data tre not maant to provide an absolute comparison of calculation and measurement. Absolute comparisons are discussed 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 t the multiple foil data with the Fe-54 (n,p) and Ni-58 (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), p (E > 0.1 MeV), and dpa/sec in the reactor cavity. The resultant axial distributions of p (E > 1.0 MeV), p (E > 0.1 MeV), and dpa/sec 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 Figures 6.1-1 through 6.1-4. In these graphical presentations, the solid symbols represent the explicit results of the FERRET evaluations, while the open symbols depict the normalized data from the gradient chains. . l l 1
I TABLE 6.1-1. !
SUMMARY
OF REACTION RATES DERIVED FROM MULTIPLE FOIL SENSOR SETS IRRADIATED DURING CYCLE 15
~
REACTION RATE (ros/ nucleus) , I
~
CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE ; REACTION. H J K L G I XX l Cu63(n,a) 1.07E-18 9.86E-19 7.45E-19 7.15E-19 3.86E-19 3.83E-19 3.72E-20 ." Ti46(n,p) 1.62E-17 1.49E-17 1.12E-17 1.06E-17 6.44E-18 6.37E-18 6.11E-19 Fe54(n p) 9.46E-17 8.49E-17 6.56E-17 5.51E-17 3.38E-17 3.62E-17 3.75E-18 . NiS8(n,p) 1.38E-16 1.21E-16 9.00E-17 8.04E-17 5.34E-17 5.23E-17 5.91E-18 U?38(n,f) 5.21E-16 4.54E-16 3.35E-16 3.11E-16 2.08E-16 1.87E-16 2.38E-17 ; Np237(n,f) 7.60E-15 7.55E-15 5.31E-15 4.63E-15 3.21E-15 3.06E-15 7.20E-16
*CoS9(n,1) 1.13E-13 1.42E-13 1.16E-13 7.34E-14 3.86E-14 4.53E-14 1.41E-14 CoS9(n,1). 6.71E-14 8.12E-14 6.60E-14 4.81E-14 2.85E-14 2.94E-14 1.01E-14 !
*U235(nf) 9.61E-13 1.18E-12 1.19E-12 5.45E-13 3.09E-13 3.52E-13 1.16E-13 ..
U235(nf) 3.12E-13 4.05E-13 3.12E-13 2.05E-13 8.25E-14 7.52E-14 5.16E-14 .
* - Bare foil, all others were cadmium covered I
i l . 6-4 l ____.__---?
1 TA8LE 6.1-2 Fe-54 (n.p) REACTION RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS IRRADIATED DURING CYCLE 15 l l FEET REACTION RATE (ros/nocleus) ' FROM I MIDPLANE O DEG 15 DEG 30 DEG 45 DEG
'. +6.5 1.98E-17 1.53E-17 1.30E-17 1.14E-17
+5.5 4.83E-17 4.24E-17 3.16E-17 2.64E-17
. +4.5 7.08E-17 6.44E-17 4.85E-17 4.24E-17
+3.5 8.29E-17 7.14E-17 5.34E-17 4.71E-17 ;
+2.5 9.25E-17 7.72E-17 5.70E-17 5.20E-17 j
+1.5 9.12E 7.59E-17 5.43E-17 4.97E-17
+0.5 8.87E-17 7.65E-17 5.49E-17 4.99E-17
-0.5 9.25E-17 8.16E-17 5.93E-17 5.13E-17
, -1.5 8.48E-17 7.91E-17 5.66E-17 5.03E-17 !
-2.5 8.10E-17 7.84E-17 6.01E-17 4.95E-17
. -3.5 7.97E-17 7.33E-17 5.14E-17 4.72E-17 *
-4.5 7.14E-17 6.70E-17 4.50E-17 4.52E-17 i
-5.5 4.82E-17 4.30E-17 2.84E-17 2.83E-17 !
-6.5 1.77E-17 1.65E-17 1.24E-17 1.17E-17 !
i t i
+
1 6-5 l l I l
- v i TABLE 6.1-3 !
- e Ni-58 (n,p) REACTI0h RATES DERIVED FROM THE STAINLESS STEEL l GRADIENT CHAINS IRRADIATED DURING CYCLE 15 i
FEET REACTION RATE (ros/ nucleus) !
. i FROM t MIDPLANE' 0 DEG 15 DEG 30 DEG 45 DEG I
+6.5 2.95E-17 2.34E-17 1.91E-17 1.76E-17 - I
+5.5 7.33E-17 5.80E-17 4.87E-17 4.06E-17
+4.5 1.08E-16 9.37E-17 6.93E-17 6.14E-17 -
+3.5 1.20E-16 1.01E-16 7.78E-17 6.82E-17 :
+2.5 1.32E-16' 1.14E-16 8.41E-17 7.33E-17
+1.5 1.30E-16 1.12E-16 8.52E-17 7.16E-17
+0.5 1.22E-16 1.10E-16 8.07E-17 7.10E-17 ,
-0.5 1.28E-16 1.15E-16 8.35E-17 7.50E-17
-1.5 1.18E-16 1.11E-16 8.07E-17 7.22E-17 .
-2.5 1.21E-16 1.06E-16 8.01E-17 7.10E-17
-3.5 1.15E-16 1.06E-16 7.61E-17 6.59E-17 -l
-4.5 1.06E-16 9.72E-17 6.82E-17 6.19E-17 i
-5.5 7.10E-17 6.42E-17 4.51E-17 4.29E-17 l
-0.5 2.78E-17 2.56E-17 1.88E-17 1.72E-17 I
[ 1 l
.I I
i 6-6 ! l
i. I TABLE 6.1-4
- Co-59 (n,1) REACTION RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS IRRADIATED DURING CYCLE 15 l FEET REACTION RATE fros/ nucleus) i FROM MIDPLANE O DEG 15 DEG 30 DEG 45 DEG l
. +6.5 2.40E-14 2.47E-14 2.18E-14 1.83E-14
+5.5 5.40E-14 7.97E-14 6.11E-14 3.75E-14 t
. +4.5 6.95E-14 1.15E-13 8.76E-14 5.21E-14
+3.5 8.54E-14 1.33E-13 1.03E-13 5.99E-14
+2.5 9.50E-14 1.44E-13 1.10E-13 6.73E-14 !
+1.5 9.78E-14 1.44E-13 1.15E-13 7.07E-14 :
+0.5 1.01E-13 1.40E-13 1.19E-13 7.07E-14 ;
-0.5 1.14E-13 1.44E-13 1.17E-13 7.18E-14 l
, -1.5 1.06E-13 1.37E-13 1.14E-13 7.01E-14 i
-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
-5.5 4.94E-14 7.12E-14 4.21E-14 3.63E-14
-6.5 3.81E-14 3.86E-14 2.93E-14 2.76E-14 !
'O 4
6-7
~
i r TABLE 6.1-5 DERIVED EXPOSURE RATES FROM THE CAPSULE H DOSIMETRY EVALVATION' 0 DEGREE AZIMUTH - CORE MIDPLANE A PRIORI ADJUSTED PARAMETER VALUE VALUE UNCERTAINTY ; p (E > 1.0 Mev) 1.89E+09 1.87E+09 G d (E > 0.1 Mev) 1.52E+10 1.38E+10 14a
~
p (E < 0.414 ev) 2.89E+09 1.83E+09 24% - ( (Total) 4.13E+10 3.77E+10 14% dpa/sec 6.35E-12 5.87E-12 12% - COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES 0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION , MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 1.07E-18 1.41E-18 1.12E-18 1.32 1.04 Ti-46 (n p) 1.62E-17 1.70E-17 1.60E-17 1.05 0.99 Fe 54 (n,p) 9.46E-17 9.35E-17 9.28E-17 0.99 0.98 Ni-58 (n.p) 1.38E-16 1.28E-16 1.33E-16 0.93 0.97 U-238 (n,f) (Cd) 5.21E-16 5.47E-16 5.37E-16 1.05 1.03 I Np-237 (n,f) (Cd) 7.60E-15 8.llE-15 7.62E-15 1.07 1.00 - Co-59 (n,1) 1.13E-13 1.54E-13 1.14E-13 1.37 1.01 Co-59 (n,1) (Cd) 6.71E-14 7.73E-14 6.72E-14 1.15 1.00 - f U-235 (n,f) 9.61E-13 1.40E-12 9.61E 1.46 1.00 l U-235 (n,f) (Cd) 3.12E-13 2.96E-13 3.09E-13 0.95 0.99
- e 6-8
TABLE 6.1-6 DERIVED EXPOSURE RATES FROM THE CAPSULE J D0SIMETRY EVALUATION 15 DEGREE AZIMUTH - CORE MIDPLANE A PRIORI ADJUSTED PARAMETER VALUE VALUE UNCERTA!NTY p (E > 1.0 Mev) 1.76E+09 1.69E+09 6% p (E > 0.1 Mev) 1.58E+10 1.45E+10 15%
. p (E < 0.414 ev) 2.86E+09 2.29E+09 24%
( (Total) 3.85E+10 3.82E+10 12%
. dpa/sec 5.60E-12 5.21E-12 11%
COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES 15 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 9.86E-19 1.28E-18 1.03E-18 1.30 1.05 Ti-46 (n.p) 1.49E-17 1.53E-17 1.46E-17 1.03 0.98 Fe-54 (n p) 8.49E-17 8.41E-17 8.26E-17 0.99 0.97 Ni-58 (n,p) 1.21E-16 1.16E-16 1.17E-16 0.96 0.97 U-238 (n,f) (Cd) 4.54E-16 5.01E-16 4.73E-16 1.10 1.04
. Np-237 (n,f) (Cd) 7.55E-15 7.97E-15 7.54E-15 1.06 1.00 Co-59 (n,7) 1.42E-13 1.61E-13 1.40E-13 1.14 0.98
- Co-59 (n,7) (Cd) 8.12E-14 8.63E-14 8.26E-14 1.06 1.02 U-235 (n,f) 1.18E-12 1.38E-12 1.19E-12 1.17 1.01 U-235 (n,f) (Cd) 4.05E-13 3.16E-13 3.98E-13 0.78 0.98 e
6-9
TABLE 6.1-7 DERIVED EXPOSURE RATES FROM THE CAPSULE K 00SIMETRY EVALUATION 30 DEGREE AZIMUTH - CORE MIDPLANE 'i A PRIORI ADJUSTED PARAMETER VALUE VALUE UNCERTAINTY
~
p (E > 1.0 Mev) 1.30E+09 1.23E+09 6% ; p (E > 0.1 Mev) 1.16E+10 1.02E+10 15%
~
p (E < 0.414 ev) 2.22E+09 2.24E+09 22% - p (Total) 2.92E+10 2.87E+10 12% dpa/sec 4.llE-12 3.69E-12 11% - COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES 30 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 7.45E-19 1.03E-18 7.83E-19 1.38 1.05 Ti-46 (n,p) 1.12E-17 1.21E-17 1.11E-17 1.08 0.99 Fe-54 (n,p) 6.56E-17 6.45E-17 6.28E-17 0.98 0.96 Ni-58 (n,p) 9.00E-17 8.83E-17 8.77E-17 0.98 0.97 U-238 (n,f) (Cd) 3.35E-16 3.72E-16 3.49E-16 1.11 1.04 Np-237 (n,f) (Cd) 5.31E-15 5.85E-15 5.32E-15 1.10 1.00 - Co-59 (n,1) 1.16E-13 1.26E-13 1.23E-13 1.09 1.06 Co-59 (n,y) (Cd) 6.60E-14 6.80E-14 6.44E-14 1.03 0.98 - U-235 (n,f) 1.19E-12 1.08E-12 1.15E-12 0.90 0.97 U-235 (n,f) (Cd) 3.12E-13 2.49E-13 3.11E-13 0.80 1.00 6-10
TABLE 6.1-8 DERIVED EXPOSURE RATES FROM THE CAPSULE L 00SIMETRY EVALUATION 45 DEGREE AZIMUTH - CORE MIDPLANE A PRIORI ADJU$ FED PARAMETER VALUE VALUE UNCERTAINTY p (E > 1.0 Mev) 1.03E+09 1.10E+09 6% p (E > 0.1 Mev) 8.66E+09 8.66E+09 14%
'. p (E < 0.414 ev) 1.95E+09 9.89E+08 25%
( (Total) 2.20E+10 2.12E+10 12%
. dpa/sec 3.12E-12 3.15E-12 11%
COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES 45 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CfLC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 7.15E-19 9.03E-19 7.44E-19 1.26 1.04 Ti-46 (n,p) 1.06E-17 1.05E-17 1.04E-17 0.99 0.98 Fe-54 (n,p) 5.51E-17 5.41E-17 5.50E-17 0.98 1.00 Ni-58 (n,p) 8.04E-17 7.37E-17 7.83E-17 0.92 0.97 U-238 (n,f) (Cd) 3.'1E-16 3.01E-16 3.16E-16
. 0.97 1.02 Np-237 (n,f) (Cd) 4.63E-15 4.48E-15 4.64E-15 0.97 1.00 Co-59 (n,7) 7.34E-14 1.03E-13 7.34E-14 1.41 1.00
. Co-59 (n,1) (Cd) 4.81E-14 5.13E-14 4.82E-14 1.07 1.00 U-235 (n,f) 5.45E-13 9.35E-13 5.47E-13 1.72 1.00 U-235 (n,f) (Cd) 2.05E-13 1.88E-13 2.03E-13 0.92 0.99 6-11
TABLE 6.1-9 DERIVED EXPOSURE RATES FROM THE CAPSULE G DOSIMETRY EVALUATION 0 DEGREE AZIMUTH - CORE TOP
~
A PRIORI ADJUSTED PARAMETER VALUE VALUE UNCERTAINTY p (E > 1.0 Mev) 6.75E+08 7.56E+08 6% p (E > 0.1 Mev) 5.43E+09 5.60E+09 14% ( (E < 0.414 ev) 1.03E+09 5.63E+08 22% - p (Total) 1.47E+10 1.33E+10 14% dpa/sec 2.27E-12 2.33E-12 12% - COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES 0 DEGREE AZIMUTH - CORE TOP REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 3.86E-19 5.04E-19 4.05E-19 1.31 1.05 Ti-46 (n,p) 6.44E-18 6.07E-18 6.21E-18 0.94 0.96 Fe-54 (n,p) 3.38E-17 3.34E-17 3.44E-17 0.99 1.02 Ni-58 (n p) 5.34E-17 4.58E-17 5.12E-17 0.86 0.96 U-238 (n,f) (Cd) 2.08E-16 1.95E-16 2.12E-16 0.94 1.02 Np-237 (n,f) (Cd) 3.21E-15 2.89E-15 3.19E-15 0.90 0.99 . Co-59 (n,1) 3.86E-14 5.51E-14 4.22E-14 1.43 1.09 Co-59 (n,1) (Cd) 2.85E-14 2.76E-14 2.69E-14 0.97 0.94 - U-235 (n,f) 3.09E-13 5.00E-13 2.98E-13 1.62 0.96 U-235 (n,f) (Cd) 8.25E-14 1.06E-13 8.41E-14 1.28 1.02 - 9 6-12 j
TABLE 6.1-10 1 DERIVED EXPOSURE RATES FROM THE CAPSULE I DOSIMETRY EVALUATION 0 DEGREE AZIMUTH - CORE BOTTOM A PRIORI ADJUSTED PARAMETER VALUE VALUE UNCERTAINTY ; p (E > 1.0 Mev) 7.24E+08 .7.04E+08 6% i ( (E > 0.1 Mev) 5.83E+09 5.16E+09 14%
- p (E < 0.414 ev) 1.11E+09 6.86E+08 20%
( (Total) 1.58E+10 1.26E+10 14% ' dpa/sec 2.43E-12 2.17E-12 12% : i COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES ; O DEGREE AZIMUTH - CORE BOTTOM [ REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n a) 3.83E-19 5.41E-19 4.05E-19 1.41 1.06 Ti-46 (n,p) 6.37E-18 6.50E-18 6.20E-18 1.02 0.97 Fe-54 (n,p) 3.52E-17 3.58E-17 3.51E-17 0.99 0.97 Ni-58 (n p) 5.23E-17 4.91E-17 5.04E-17 0.94 0.96 U-238 (n,f) (Cd) 1.87E-16 2.09E-16 1.98E-16 1.12 1.06 ,
- Np-237 (n,f) (Cd) 3.06E-15 3.10E-15 3.02E-15 1.01 0.99 Co-59 (n,7) 4.53E-14 5.91E-14 4.74E-14 1.30 1.05 Co-59 (n,7) (Cd) 2.94E-14 2.96E-14 2.84E-14 1.01 0.97 :
U-235 (n,f) 3.52E-13 5.36E-13 3.45E-13 1.52 0.98 U-235 (n,f) (Cd) 7.52E-14 1.13E-13 7.67E-14 1.51 1.02 i 9 6-13 i
- - -- ..n--
j TABLE 6.1-11 l DERIVED EXPOSURE RATES FROM THE CAPSULE XX DOSIMETRY EVALUATION - 0 DEGREE AZIMUTH - VESSEL SUPPORT ELEVATION -! t A PRIORI ADJUSTED ~l PARAMETER VALUE VALUE UNCERTAINTY-f ( (E > 1.0 Mev) 7.50E+07 1.08E+08 7% p (E > 0.1 Mev) 6.04E+08 1.26E+09 15% ! ( (E < 0.414 ev) 1.15E+08 1.73E+08 24% .' ! ( (Total) 1.64E+09 3.97E+09 12% , dpa/sec 2.52E-13 4.85E-13
~
14% - j COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES 0 DEGREE AZIMUTH - VESSEL SUPPORT ELEVATION , REACTION RATE (ros/ nucleus) C/M
~
A PRIORI ADJUSTED ! REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n a) 3.72E-20 5.60E-20 3.95E-20 1.51 1.06 Ti-46 (n.p) 6.11E-19 6.74E-19 5.98E-19 1.10 0.98 Fe-54 (n,p) 3.75E-18 3.71E-18 3.70E-18 0.99 0.99 Ni-58 (n p) 5.91E-18 5.09E-18 5.64E-18 0.86 0.95 ! U-238 (n,f) (Cd) 2.38E-17 2.17E-17 2.57E-17 0.91 1.08 Np-237 (n,f) (Cd) 7.20E-16 3.22E-16 6.92E-16 0.45 0.96 - f Co-59 (n,7) 1.41E-14 6.12E-15 1.46E-14 0.43 1.03 i Co-59 (n,7) (Cd) 1.01E-14 3.07E-15 9.92E-15 0.30 0.98 - U-235 (n,f) 1.16E-13 5.56E-14 1.14E-13 0.48 0.99 , U-235 (n,f) (Cd) 5.16E-14 1.17E-14 5.07E-14 0.23 0.98 6-14 i s l s P
TABLE 6.1-12 FAST NEUTRON FLUX (E >'1.C 1eV) AS A FUNCTION i 0F AXIAL POSITION WITHIN '~ REACTOR CAVITY CYCLE 15 IRRAE 10N AZIMUTHAL A =LE HEIGHT
~(ft) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES
. +7.8 1.08E+08
+6.5 4.18E+08 3.78E+08 2.75E+08 2.46E+08
. +6.0 7.56E+08
+5.5 1.03E+09 9.28E+08 6.76E+08 6.04E+08
+4.5 1.56E+09 1.41E+09 1.02E+09 9.15E+08
+3.5 1.73E+09 1.57E+09 1.14E+09 1.02E+09
+2.5 1.90E+09 1.71E+09 1.25E+09 1.12E+09-
+1.5 1.86E+09 1.68E+09 1.22E+09 1.09E+09
. +0.5 1.82E+09 1.65E+09 1.20E+09 1.07E+09 0.0 1.87E+09 1.69E+09 1.23E+09 1.10E+09
. -0.5 1.92E+09 1.73E+09 1.26E409 1.13E+09
-1.5 1.83E+09 1.65E+09 1.20E+09 1.07E+09 'l
-2.5 1.82E+09 1.64E+09 1.19E+09 1.07E+09
-3.5 1.71E+09 1.55E+09 1.13E+09 1.01E+09
-4.5 1.57E+09 1.42E+09 1.03E+09 9.22E+08
-5.5 1.03E+09 9.30E+08 6.77E+08 6.05E+08
-6.0 7.04E+08
-6.5 4.13E+08 3.73E+08 2.72E+08 2.43E+08
=.
O p 9 6-15 a
. = - . -- .. .
r-TABLE 6.1-13 FAST NEUTRON FLUX (E > 0.1 MeV) AS A FUNCTION .
~
OF AXIAL POSITION WITHIN THE REACTOR CAVITY ! CYCLE 15 IRRADIATION i AZIMUTHAL ANGLE ., [ HEIGHT , (ft) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES !
~
+7.8 1.26E+09 -
[
+6.5 3.09E+09 3.24E+09 2.28E+09 1.94E+09 i
+6.0 5.60E+09 -
+5.5 7.58E+09 7.96E+09 5.60E+09 4.76E+09
+4.5 1.15E+10 1.21E+10 8.49E+09 7.20E+09
+3.5 1.28E+10 1.34E+10 9.45E+09 8.03E+09
+2.5 1.40E+10 1.47E+10 1.03E+10 8.78E+09
+1.5 1.37E+10 1.44E+10 1.01E+10 8.60E+09
+0.5 1.34E+10 1.41E+10 9.94E+09 8.44E+09 .t 0.0 1.38E+10 1.45E+10 1.02E+10 8.66E+09 l
-0.5 1.41E+10 1.49E+10 1.05E+10 8.87E+09 ;
-1.5 1.35E+10 1.42E+10 9.97E+09 8.46E+09
-2.5 1.34E+10 1.41E+10 9.90E+09 8.41E+09 !
-3.5 1.26E+10 1.33E+10 9.35E+09 7.94E+09 [
-4.5 1.16E+10 1.22E+10 8.55E+09 7.26E+09 )
-5.5 7.59E+09 7.98E+09 5.61E+09 4.76E+09 l
-6.0 5.16E+09
-6.5 3.05E+09 3.20E+09 2.25E+09 1.91E+09 v
6-16 [
l
-TABLE 6.1-14 l IRON ATOM DISPLACEMENT RATE (dpa/sec) AS A FUNCTION ;
0F AXIAL POSITION WITHIN THE REACTOR CAVITY i CYCLE 15 IRRADIATION ! AZIMUTHAL ANGLE i HEIGHT ! (ft) 0 DEGREES J5 DEGREES 30 DEGREES 45 DEGREES
. +7.8 4.85E-13
+6.5 '
1.31E-12 1.16E-12 8.25E-13 7.04E-13
. +6.0 2.33E-12 !
+5.5 3.22E-12 2.86E-12 2.03E-12 1.73E-12 !
+4.5 4.88E-12 4.33E-12 3.07E-12 2.62E-12 !
+3.5 5.44E-12 4.83E-12 3.42E-12 2. 92.E-12 !
+2.5 5.95E-12 5.28E-12 3.74E-12 3.19E-12 !
+1.5 5.83E-12 5.17E-12 3.66E 3.13E-l'2 -
'- f
+0.5 5.72E-12 5.08E-12 3.60E-12 3.07E-12 !
i 0.0 5.87E-12 5.21E-12 3.69E-12 3.15E-12 !
. -0.5 6 01E-12 5.34E-12 3.78E-12 3.23E-12 i
-1.5 5.74E-12 5.09E-12 3.61E-12 3.08E-12 !
-2.5 5.70E-12 5.06E-12 3.58E-12 3.06E-12 l
-3.5 5.38E-12 4.77E-12 3.38E-12 2.89E-12 !
-4,5 4.92E-12 4.37E-12 3.09E-12 2.64E-12 l
-5.5 3.23E-12 2.87E-12 2.03E-12 1.73E-12 l
-6.0 2.17E-12
-6.5 1.30E-12 1.15E-12 8.15E-13 6.96E-12
{ i
.. l i
i e 6-17
FIGURE 6.1-1 : FAST NEUTRON Fl.UX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION
~
ALONG THE O DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 15 IRRADIATION
~
Neutron Flux (n/cm2-sec) f a Ce u C O 1.000E + 09 1 1.000E + 08 ..
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) -
O ~! CYCLE 15 - 0 DEG 6-18 1
FIGURE 6.1-2 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 15 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 15 IRRADIATION Neutron Flux (n/cm2-sec) 1.000E + 10 _ 3 a C;u C ' 1.000E + 09 _
.. 1.000E + 08
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft)
O CYCLE 15 - 15 DEG 6-19 j
i FIGURE 6.1-3 , FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION i
~
ALONG THE 30 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 15 IRRADIATION . 1.000E + 10
" " * ~** '
I M
-- C C n:C C n N
1.000E + 09 _ i 1.000E + 08 e ,
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) -
O CYCLE 15 - 30 DEG 6-20 l
FIGURE 6.1-4 FAST NEUTRON FLUX (E > 1.0 MeV) AS A' FUNCTION OF AXIAL POSITION ALONG THE 45 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 15 IRRADIATION Neutron Flux (n/cm2-sec) 1.000E + 10 _ 1.000E + 09 C C C UC 5
,, 1.000E + 08
-8 -6 -4 -2 0 2 4 6 8
. Distance From Core Midplane (ft)
O CYCLE 15 - 45 DEG 6-21
6.2 - Cycle 16 Results 6.2.1 - Measured Reaction rates During the Cycle 16 irradiation, six multiph foil sensor sets and four
~
stainless steel gradient chains were deployeo in the reactor cavity as depicted in Figures 2.1-1 and 2.1-2. The capsule identifications associated with each
, l of the multiple foil sensor sets were as follows (1):
CAPSULE IDENTIFICATION - AZIMUTH CORE CORE CORE (dearees) TOP MIDPLANE BOTTOM - 0 M N 0 15 P 30 Q 45 R The contents of each of these irradiation capsules is specified in Reference 1 - and, for completeness, is also included in Appendix B to this report. The irradiation history of the Point Beach Unit 2 reactor during Cycle 16 is i also listed in Appendix B. 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 aci.ivities given l 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 ~~ 4' radiometric foils and solid state track recorders, irradiated during Cycle 16 are provided in Table 6.2-1. Corresponding reaction rate data from the the ~ four stainless steel gradient gradient chains are recorded in Tables 6.2-2 through 6.2-4 for the Fe-54 (n.p), Ni-58 (n,p), and Co-59 (n,y) reactions, ' respectively. 6-22
-i 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 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 and-the effects of 7,f reactions in the U-238 sensors as well for the effects of 7,f reactions in the Np-237 monitors. .
". 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 a priori 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. t . In regard to the comparisons listed in Tables 6.2-5 through 6.2-10, it should j be noted that the columns labeled "a priori calc" were obtained by normalizing i the neutron spectral data from Table 4.1-1 to the measured Fe-54 (n,p) reaction rate,s from each sensor set as discussed in Section 3.0. Thus, the comparisons : iilustrated in Tables 6.2-5 through 6.2-10 indicate only the degree to which !
'he relative neutron energy spectra matched the measured data before and after adjustment. These data are not meant to provide an absolute comparison of calculation and measurement. Absolute comparisons are discussed in Section 7.0 , . of this report. ;
E
.. 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) and Ni-58 (n,p) reaction rate measurements from the gradient chains. The gradient data were employed to est%1ish relative axial distributions over the measurement range and these relmtive distributions were then normalized to the FERRET results from the 6-23
I i
?
midplane' sensor sets to produce axial distributions of exposure rates in terms f of ( (E > 1.0 MeV), p (E > 0.1 MeV), and dpa/sec in the reactor cavity. [ t The resultant axial distributions of ( (E > 1.0 MeV), d (E > 0.1 MeV), - and dpa/sec are given in Tables 6.2-11, 6.2-12, and 6.2-13, respectively. The
' l distributions' of ( (E > 1.0 MeV) are depicted graphically in Figures 6.2-1 :
through 6.2-4. In these graphical presentations, the solid symbols represent the explicit results of the FERRET evaluations, while the open symbols depict , the normalized data from the gradient chains.
, l r
S t 7 F I I l r I I I [ e 6-24 { t t t
TABLE 6.2-1
SUMMARY
OF REACTION PATES DERIVED FROM MULTIPLE F0IL SENSOR SETS IRRADIATED DURING CYCLE 16 l REACTION RATE f ros/ nucleus) CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE REACTION N P 0 R M 0
". Cu63(n,a) 8.59E-19 7.51E-19 6.7'E-19 6.61E-19 4.11E-19 3.69E-19 ,
Ti46(n p) 1.27E-17 1.10E-17 9.64E-18 9.20E-18 6.41E-18 5.79E-18 )
. Fe54(n.p) 7.44E-17 6.33E-17 5.51E-17 5.23E-17 3.51E-17 3.48E-17 NiS8(n p) 1.07E-16 9.24E-17 7.84E-17 7.43E-17 5.51E-17 5.07E-17 l U238(n,f) 4.01E-16 3.60E-16 2.63E-16 2.66E-16 2.15E-16 1.84E-16 )
Np237(n,f) 6.01E-15 5.28E-15 4.27E-15 4.13E-15 2.95E-15 2.92E-15
*CoS9(n,1) 8.74E-14 1.12E-13 1.00E-13 6.23E-14 3.68E-14 4.08E-14 CoS9(n,1) 5.26E-14 6.71E-14 5.44E-14 4.14E-14 2.63E-14 2.65E-14 l
. *U235(n,f) 8.28E-13 1.01E-12 9.96E-13 4.40E-13 2.61E-13 3.66E-13 , U235(n,f) 2.66E-13 3.08E-13 2.32E-13 1.95E-13 7.73E-14 7.28E-14
* - Bare foil, all others were cadmium covered i
i Y ~ 6-25
l
)
TABLE 6.2-2 Fe-54 (n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS IRRADIATED DURING CYCLE 16 FEET REACTION RATE fros/ nucleus) FROM MIDPLANE O DEG 15 DEG 30 DEG 45 DEG
~
+6.5 1.92E-17 1.41E-17 1.19E-17 1.07E-17 -
+5.5 4.67E-17 3.51E-17 2.98E-17 2.56E-17
+4.5 7.37E-17 5.63E-17 4.32E-17 3.98E-17 -
+3.5 8.21E-17 6.74E-17 4.52E-17 4.18E-17
+2.5 8.74E-17 6.79E-17 5.00E-17 4.69E-17
+1.5 8.32E-17. 6.42E-17 4.80E-17 4.65E-17 40.5 7.48E-17 5.84E-17 5.02E-17 4.62E-17
-0.5 6.00E-17 5.21E-17 4.84E-17 4.55E-17
-1.5 5.90E-17 5.25E-17 4.69E-17 4.59E-17 -
-2.5 6.63E-17 5.13E-17 4.81E-17 4.39E-17
-3.5 6.90E-17 6.11E-17 4.35E-17 4.29E-17 -
-4.5 6.85E-17 5.84E-17 4.05E-17 4.04E-17
-5.5 4.65E-17 3.66E-17 2.69E-17 2.70E-17
-6.5 1.77E-17 1.56E-17 1.12E-17 1.07E-17 l .
6-26 1
. . - . . . . - _ . .= -. -. . .
I i TABLE 6.2-3 Ni-58 (n,p) REACTION RATES DERIVED.FROM THE STAINLESS STEEL : GRADIENT CHAINS' IRRADIATED DURING CYCLE 16 FEET REACTION RATE (ros/ nucleus) FRM MIDPLANE O DEG 15 DEG _30 DEG 45 DEG~
'. : +6.5 2.91E-17 2.11E-17 1.75E-17 1.54E-17 [
+5.5 7.09E-17 5.53E-17 4.33E-17 3.65E-17 :i
^
+ +4.5 1.10E-16 8.62E-17 6.12E-17 5.74E-17
+3.5 'l.18E-16 9.13E-17 6.65E-17' 6.15E-17 f
+2.5 1.23E-16 9.51E 7.33E-17 6.59E-17 ,
+1.5 1.15E-16 8.92E-17 7.04E-17 6.33E-17 [
+0.5 1.04E-16 8.42E-17 7.06E-17 6.33E-17. l
-0.5 8.68L 17 7.33E-17 7.09E-17 6.68E-17 .;
. -1.5 8.36E-17 7.59E-17 6.68E-17 6.27E-17 ;
-2.5 9.18E-17 7.65E-17 6.74E-17 6.45E-17 !
l
-3.5 1.05E-16 8.42E-17 6.45E-17 6.12E-17 !
-4.5 9.80E-)? 8.51E-17 6.01E-17 5.98E-17 f
-5.5 6.98E-17 5.42E-17 3.89E-17 3.97E-17 i
-6.5 2.86E-17 2.17E-17 1.60E-17 1.70E-17 ;
i i i t 1 i 6-27 '
TABLE 6.2-4 , Co-59 (n,7) REACTION RATES DERIVED FROM THE STAINLESS STEEL
~
GRADIENT CHAINS IRRADIATED DURING CYCLE 16
~
I FEET REACTION RATE (ros/ nucleus) l FROM l MIDPLANE O DEG 15 DEG 30 DEG 45 DEG
~
+6.5 2.25E-14 2.25E-14 1.95E-14 1.67E-14 - I
+5.5 5.14E-14 7.41E-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 9.60E-14 6.26E-14
+0.5 8.50E-14 1.13E-13 9.81E-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 7.88E-14 9.70E-14 8.97E-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 I
-5.5 4.73E-14 6.36E-14 3.66E-14 3.24E-14 l
-6.5 3.48E-14 3.40E-14 2.59E-14 2.42E-14 t
I h D
=b 1
6-28 i
TA8LE 6.2-5 i DERIVED EXPOSURE RATES FROM THE CAPSU' E N DOSIMETRY EVALUATION 0 DEGREE AZIMUTH - CORE MIDPLANE .
~~
A PRIORI ADJUSTED. PARAMETER VALUE VALUE UNCERTAINTY l p (E > 1.0 Mev) 1.49E+09 1.45E+09 6% f ( (E > 0.1 Mev) 1.20E+10 1.10E+10 14%
. p (E < 0.414 ev) 2.27E+09 1.52E+09 23% i
( (Total) 3.25E+10 3.07E+10 14%
. dpa/sec 4.99E-12 4.66E-12 12%
i l , COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0 DEGREE AZIMUTH - CORE MIDPLANE : i i REACTION RATE (ros/ nucleus) C/M i A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 8.59E-19 1.11E-18 8.94E-19 1.29 1.04 Tf-46 (n,p) 1.27E-17 1.34E-17 1.26E-17 1.05 0.99 Fe-54 (n.p) 7.44E-17 7.35E-17 7.26E-17 0.99 0.98 Ni-58 (n p) 1.07E-16 1.01E-16 1.04E-16 0.94 0.97 U-238 (n,f) (Cd) 4.01E-16 4.30E-16 4.15E-16 1.07 1.03 i
. Np-237 (n,f) (Cd) 6.01E-15 6.38E-15 6.02E-15 1.06 1.00 00-59 (n,1) 8.74E-14 i.21E-13 9.05E-14 1.39 1.04 '
. Co-59 (n,y) (Cd) 5.26E-14 6.08E-14 5.18E-14 1.16 0.99 !
U-235 (n,f) 8.28E-1., 1.10E-12 8.13E-13 1.33 0.98 ! U-235 (n,f) (Cd) 2.66E-13 2.33E-13 2.64E-13 0.88 0.99 i
\
i I i 6-29 !
i TABLE 6.2-6 DERIVED EXPOSURE RATES FROM THE CAPSULE P DOSIMETRY EVALUATION 15 DEGREE AZIMUTH - CORE MIDPLANE A PRIORI ADJUSTED PARAMETER VALUE VALUE UNCERTAINTY
~
p (E > 1.0 Mev) 1.31E+09 1.28E+09 6% p (E > 0.1 Mev) 1.18E+10 1.03E+10 15% p (E < 0.414 ev) 2.13E+09 1.91E+09 23% - ( (Total) 2.86E410 2.81E+10 12% dpa/sec 4.17E-12 3.76E-12 11% - COMPARIS0N Of MEASURED AND CALCULATED SENSOR REACTION RATES 15 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 7.51E-19 9.54E-19 7.80E-19 1.27 1.04 Ti-46 (n,p) 1.10E-17 1.14E-17 1.09E-17 1.04 0.99 Fe-54 (n p) 6.33E-17 6.27E-17 6.24E-17 0.99 0.99 Ni-58 (n,p) P.24E-17 8.62E-17 8.95E-17 0.93 0.97 U-238 (n,f) (Cd) 3.60E-16 3.73E-16 3.67E-16 1.04 1.02 Np-237 (n,f) (Cd) 5.28E-15 5.93E-15 5.33E-15 1.12 1.01 - C0-59 (n,7) 1.12E-13 1.20E-13 1.15E-13 1.07 1.03 Co-59 (n,1) (Cd) 6.71E-14 6.43E-14 6.63E-14 0.96 0.99 U-235 (n,f) 1.01E-12 1.03E-12 9.96E-13 1.02 0.99 U-235 (n,f) (Cd) 3.08E-13 2.36E-13 3.06E-13 0.77 0.99 e 6-30
TABLE 6.2-7 DERIVED EXPOSURE RATES FROM THE CAPSULE Q 00SIMETRY EVALUATION 30 DEGREE AZIMUTH - CORE MIDPLANE A PRIORI ADJUSTED PARAMETER VALUE VALUE UNCERTAINTY p (E > 1.0 Mev) 1.09E+09 9.73E+08 6% ( (E > 0.1 Mev) 9.75E+09 8.16E+09 15% 4 (E < 0.414 ev) 1.87E+09 1.95E+09 21% ( (Total) 2.45E+10 2.29E+10 12%
. dpa/sec 3.45E-12 2.95E-12 11%
COMPARIS0N OF MEASURED AND CALCULATE 3 SENSOR REACTION RATES 30 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n.o) 6.71E-19 8.63E-19 7.02E-19 1.29 1.05 Ti-46 (n,p) 9.64E-18 1.02E-17 9.57E-18 1.05 0.99 Fe-54 (n,p) 5.51E-17 5.41E-17 5.29E-17 0.98 0.96 Ni-58 (n,p) 7.84E-17 7.41E-17 7.52E-17 0.95 0.96 U-238 (n,f) (Cd) 2.63E-16 3.12E-16 2.79E-16 1.19 1.06
. Np-237 (n,f) (Cd) 4.27E-15 4.91E-15 4.26E-15 1.15 1.00 Co-59 (n,y) 1.00E-13 1.06E-li 1.05E-13 1.06 1.05 *. Co-59 (n,y) (Cd) 5.44E-14 5.70E-14 5.33E-14 1.05 0.98 U-235 (n,f) 9.96E-13 9.03E-13 9.70E-13 0.91 0.97 U-235 (n,f) (Cd) 2.32E-13 2.09E-13 2.32E-13 0.90 1.00 O
6 31
~ i TABLE 6.2-8 ! DERIVED EXPOSURE RATES FROM THE CAPSULE R DOSIMETRY EVALUATION f 45 DEGREE AZIMUTH - CORE MIDPLANE ;
-j A PRIORI ADJUSTED 'I PARAMETER VALUE VALUE UNCERTAINTY l
~
p (E > 1.0 Mev) 9.76E+08 9.63E+08 6% p (E > 0.1 Mev) 8.22E+09 7.65E+09 15% ( (E < 0.414 ev) 1.85E+09 7.53E+08 26%
- 1
( (Total). 2.09E+10 1.91E+10 12% dpa/sec ?. 96E-12 2.78E-12 11% COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 45 DEGREE AZIMUTH - CORE MIDPLANE I REACTION RATE (ros/ nucleus) C/M t A PRIORI ADJUSTED REACTION MEASURED
)
CALC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 6.61E-19 8.57E-19 6.87E-19 1.30 1.04 Ti-46 (n.p) f 9.20E-18 9.92E-18 9.15E-18 1.08 0.99 ! Fe-54 (n,p) 5.23E-17 5.14E-17 5.08E-17 0.98 0.97 Ni-58 (n,p) 7.43E-17 7.00E-17 7.18E-17 0.94 0.97 U-238 (n,f) (Cd) 2.66E-16 2.86E-16 2.77E-16 1.07 1.04 i Np-237 (n,f) (Cd) 4.13E-15 4.25E-15 4.12E-15 1.03 1.00
- i Co-59 (n,1) 6.23E-14 9.81E-14 6.11E-14 1.57 0.98 Co-59 (n,1) (Cd) 4.14E-14 4.87E-14 4.22E-14 1.18 1.02 -**
l U-235 (n,f) 4.4CE-13 8.87E-13 4.47E-13 2.02 1.02 ) U-235 (n,f) (Cd) 1.95E-13 1.78E-13 1.92E-13 0.91 0,.98 5 6-32 ! 3
l TABLE 6.2-9 ! l DERIVED EXPOSURE RATES FROM THE CAPSULE M D0SIMETRY EVALUATION
; E O DEGREE AZIMUTH - CORE TOP :
A PRIORI ADJUSTED [ i PARAMETER VALUE VALUE UNCERTAINTY p (E > 1.0 Mev) 7.01E+08 7.53E+08 6% + p (E > 0.1.Mev) 5.64E+09 5.19E+09 14%
". p (E < 0.414 ev) 1.07E+09 '4.86E+08 23%
( (Total) 1.53E+10 1.23E+10 14% ,
. dpa/sec 2.35E-12 2.19E-12 12%
t i COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES 0 DEGREE AZIMUTH - CORE TOP
. t REACTION RATE (ros/ nucleus) C/M._
A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI AQJyllED [ Cu-63 (n,a) 4.llE-19 5.23E-19 4.27E-19 1.21 1.04 ! Ti-46 (n p) 6.41E-18 6.30E-18 6.27E-18 0.98 0.98 i Fe-54 (n p) 3.51E-17 3.47E-17 3.56E-17 0.99 1.01 Ni-58 (n,p) 5.51E-17 4.75E-17 5.28E-17 0.86 0.96 i U-238 (n,f) (Cd) 2.15E-16 2.03E-16 2.17E-16 0.94 1.01 ;
. Np-237 (n,f) (Cd) 2.95E-15 3.01E-15 2.96E-15 1.02 1.00 Co-59 (n,7) 3.68E-14 5.72E-14 3.84E-14 1.55 1.04
. Co-59 (n,1) (Cd) 2.63E-14 2.87E-14 2.54E-14 1.09 0.97 U-235 (n,f) 2.61E-13 5.19E-13 2.57E-13 1.99 0.98 U-235 (n,f) (Cd) 7.73E-14 1.10E-13 7.84E-14 1.42 1.01 !
l
. s s
f 6-33
)
--_----a
TA8LE 6.2-10 i DERIVED EXPOSURE RATES FROM THE CAPSULE O DOSIMETRY EVALUATION
~
0 DEGREE AZIMUTH - CORE BOTTOM A PRIORI ADJUSTED PARAMETER VALUE VALUE UNCERTAINTY
~
p (E > 1.0 Mev) 6.95E+08 6.88E+08 6% p (E > 0.1 Mev) 5.59E+09 4.98E+09 14% 4 (E < 0.414 ev) 1.06E+09 6.93E+08 19% - p (Total) 1.52E+10 1.22E+10 13% dpa/sec 2.33E-12 2.09E-12 12% - COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0 DEGREE AZIMUTH - CORE BOTTOM REACTION RATE (ros/ nucleus) C/M__ A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu 63 (n,o) 3.69E-19 5.19E-19 3.87E-19 1.41 1.05 Ti-46 (n.p) 5.79E-18 6.24E-18 5.71E-18 1.08 0.99 - Fe-54 (n,p) 3.48E-17 3.43E-17 3.37E-17 0.99 0.97 Ni-58 (n,p) 5.07E-17 4.71E-17 4.87E-17 0.93 0.96 U-238 (n,f) (Cd) 1.84 E- 16 2.01E-16 1.93E-16 1.09 1.05 Np-237 (n,f) (Cd) 2.92E-15 2.98E-15 2.89E-15 1.02 0.99 - Co-59 (n,1) 4.08E-14 5.67E-14 4.45E-14 1.39 1.09 Co-59 (n,1) (Cd) 2.65E-14 2.84E-14 2.51E-14 1.07 0.95
~
U-235 (n,f) 3.66E-13 5.15E-13 3.51E-13 1.41 0.96 U-235 (n,f) (Cd) 7.28E-14 1.09E-13 7.44E-14 1.49 1.02 9 6-34
i TABLE 6.2-11 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION , OF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLE 16 IRRADIATION AZIMUTHAL ANGLE HEIGHT , (ft) 0 DEGREES 15 DEGREES 10 DEGREES 45 DEGREES
". +6.5 4.28E+08 3.34E+08 2.38E+08 2.26E+08
+6.0 7.53E+08
. +5.5 1.04E+09 8.56E+08 5.91E+08 5.39E+08
+4.5 1.63E+09 1.35E+09 8.47E+08 8.46E+08
+3.5 1.78E+09 1.52E+09 9.03E+08 8.94E+08
+2.5 1.87E+09 1.56E+09 9.97E+08 9.80E+08
+1.5 1.77E+09 1.47E+09 9.58E+08 9.56E+08
+0.5 1.59E+09 1.36E+09 9.81E+08 9.53E+08 ;
. 0.0 1.45E409 1.28E+09 9.73E408 9.63E+08 ,
-0.5 1.31E+09 1.20E+09 9.64E+08 9.72E+08
. -1.5 1.27E+09 1.22E+09 9.21E+08 9.46E+08
-2.5 1.41E+09 1.22E409 9.38E+08 9.38E+08
-3.5 1.54E+09 1.39E+09 8.72E+08 9.03E+08 i
-4.5 1.48E+09 1.37E+09 8.12E+08 8.66E+08
-5.5 1.03E+09 8.64E+08 5.33E+08 5.77E+08-
-6.0 6.88E+08
-6.5 4.08E+08 3.57E+08 2.21E+08 2.38E+08 l h
6-35
I TABLE 6.2-12 l i FAST NEUTRON FLUX (E > 0.1 MeV) AS A FUNCTION . OF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLE 16 IRRADIATION AZIMUTHAL ANGLE HEIGHT (ft) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES
~ +6.5 3.25E+09 2.69E+09 2.00E+09 1.80E+09 .- +6.0 5.19E+09 +5.5 7.90E+09 6.89E+09 4.96E+09 4.28E+09 -
i
+4.5 1.24E+10 1.09E+10 7.10E+09 6.72E+09 + +3.5 1.35E+10 1.22E+10 7.57E+09 7.10E+09 +2.5 1.42E+10 1.25E+10 8.36E+09 7.79E+09 +1.5 1.34E+10 1.18E+10 8.03E+09- 7.60E+09 3 +0.5 1.21E+10 1.09E+10 8.22E+09 7.57E+09 +
0.0 1.10E+10 1.03E+10 8.16E+09 7.65E+09 -
-0.5 9.90E+09 9.64E+09 8.09E+09 7.72E+09 -1.5 9.63E+09 9.86E+09 7.73E+09 7.51E+09 - -2.5 1.07E+10 9.78E+09 7.86E+09 7.45E+09 ; -3.5 1.17E+10 1.12E+10 7.32E+09 7.17E+09 -4.5 1.12E+10 1.10E+10 6.81E+09 6.88E+09 -5.5 7.82E+09 6.95E+09 4.47E+09 4.58E+09 -6.0 4.98E+09 -6.5 3.09E+09 2.88E+09 1.85E+09 1.89E+09 '
i e 6-36
TABLE 6.2-13 IRON ATOM DISPLACEMENT RATE (dpa/sec) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLE 16 IRRADIATION AZIMUTHAL ANGLE HEIGHT i (ft) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES
. +6.5 1.38E-12 9.82E-13 7.22E-13 6.53E-13
+6.0 2.19E-12
. +5.5 3.35E-12 2.51E-12 1.79E-12 1.56E-12
+4.5 5.23E-12 3.97E-12 2.57E-12 2.44E-12
+3.5 5.72E-12 4.47E-12 2.74E-12 2.58E-12 :
+2.5 6.02E-12 4.58E-12 3.02E-12 2.83E-12
+1.5 5.69E-12 4.31E-12 2.90E-12 2.76E-12
+0.5 5.13E-12 4.00E-12 2.97E-12 2.75E-12
. 0.0 4.66E-12 3.76E-12 2.95E-12 2.78E-12 .
-0.5 4.19E-12 3.52E-12 2.92E-12 2.80E-12
-1.5 4.08E-12 3.60E-12 2.79E-12 2.73E-12 ;
-2.5 4.54E-12 3.57E-12 2.84E-12 2.71E-12 >
-3.5 4.95E-12 4.09E-12 2.65E-12 2.61E-12
-4.5 4.76E-12 4.02E-12 2.46E-12 2.50E-12
-5.5 3.31E-12 2.54E-12 1.62E-12 1.67E-12
-6.0 2.09E-12 .
-6.5 1.31E-12 1.05E-12 6.70E-13 6.88E-13 I i
k 6-37 5
FIGURE 6.2-1 t: FAST NEUTRON Fl.UX (E > I.0 MeV) AS A FUNCTION OF AXIAL POSI110N d ALONG THE O DEGREE TRAVERSE IN THE REACTOR CAVITY ', CYCLE 16 IRRADIATION , Neutron Flux (n/cm2-sec) i V 1.000E + 09 \ i l
' \
1.000E + 08 ' i
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) .
l 0 CYCLE 16 - 0 DEG - l
'i i
6-38 l l l
FIGURE 6.2-2 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION AI.ONG THE IS DEGREE TRAVERSE IN THE REACTOR CAVITY , CYCLE 16 IRRADIATION i. Neutron Flux (n/cm2-sec) 1.000E + 10 _
. -a 1.000E + 09 I k
_ r3 l 1 l 1.000E + 08 l
-8 -6 -4 -2 0 2 4 6 8 l Distance From Core Midplane (ft) . O CYCLE 16 - 15 DEG t
6-39
FIGURE 6.2-3 FASTNEUTRONFi.UX(E>1.0MeV)ASAFUNCTIONOFAXIALPOSITION . ALONG THE 30 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 16 IRRADIATION , Neutron Flux (n/cm2-sec) 1.000E + 10 _ i l I
~
1.000E + 09 _ g g ; m 1.000E + 08 ..
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) .
O CYCLE 16 - 30 DEG - 6-40
FIGURE 6.2-4 , FASTNEUTRONFi.UX(E>1.0MeV)ASAFUNCTIONOFAXIALPOSITION ALONG THE 45 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 16 IRRADIATION i Neutron Flux (n/cm2-sec) 1.000E + 09 _ u m_u m _
., 1.000E + 08 '
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft)
O CYCLE 16 - 45 DEG l 6-41
6.3 - Cycle 17 Results 6.2.1 - Maasured 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 [1]:
~
CAPSULE IDENTIFICATION - AZIMUTH CORE CORE CORE (dearees) TOP MIDPLANE BOTTOM - 0 AA BB CC 15 DD 30 EE 45 FF The contents of each of these irradiation capsules is specified in Reference 1 - and, for completeness, is also included in Appendix B to this report. The irradiation history of the Point Beach Unit 2 reactor during Cycle 17 is also listed in Appendix B. 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 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 17 are provided in Table 6.3-1. Corresponding reaction rate data from the the
- four stainless steel gradient 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,1) reactions, ~
respectively. 6-42
i I l 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 covered measurements for each capsule. Likewise, the U-238 (n,f) reaction rates we"e obtained by averaging 9 the results of the radiometric foil and solid state track recorder data. In addition, the fission rate measurements include corrections for U-235 impurities and the effects of 1,f reactions in the U-238 sensors as well for the effects of 7,f reactions in the Np-237 monitors.
. 6.3.2 - F.esults 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 a priori and adjusted spectra to the measurements are given. Also included in the tabulations are the lo 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 "a priori cale" were obtained by normalizing the neutron spectral data from Table 4.1-1 to the measured le-54 (n.p) reaction rates from each sensor set as discussed in Section 3.0. Thus, the comparisons illustrated in Tables 6.3-5 through 6.3-10 indicate only the degree to which the relative neutron energy spectra matched the measured data before and after adjustment. These data are not meant to provide an absolute comparison of calculation and measurement. Absolute comparisons are discussed 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) and Ni-58 (n p) reaction rate measurements from the gradient chains. 1he 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 6-43
~
r J,
- midplane sensor sets to produce axial distributions of exposure rates in terms af ( (E > 1.0 MeV), p (E > 0.1 MeV), and dpa/sec in the reactor cavity.
The resultant axial distributions of ( (E > 1.0,*ieV), p (E > 0.1 MeV),
- and dpa/sec 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 presentations, the solid symbols represent the explicit results of the FERRET evaluations, while the open symbols depict
- the normalized data from the gradient chains.
e e 9 9 e 6-44
=
i TABLE 6.3-1
SUMMARY
OF REACTION RATES DERIVED FROM MULTIPLE F0ll SENSOR SETS IRRADIATED DURING CYCLE 17 l i ~ REACTION RATE fros/ nucleus) CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE l REACTION BB DD EE FF AA CC
'. Cu63(n,a) 9.04E-19 8.01E-19 6.79E-19 6.97E-19 4.23E-19 3.92E-19 i Ti46(n p) 1.32E-17 1.16E-17 1.00E-i? 9.83E-18 6.56E-18 6.08E-18
- Fe54(n,p) 7.79E-17 6.79E-17 5.6iE-17 ~i.41E-17 3.55E-17 3.62E-17 NiS8(n.p) 1.10E-16 9.61E-17 8.01C r/ 7.65E-17 5.66E-17 5.22E-17 U238(n,f) 3.74E-16 3.09E-16 2.78E-16 2.58E-16 2.12E-16 1.78E-16 {
Np237(n,f) 6.20E-15 5.01E-15 4.41E-15 3.66E-15 2.54E-15 l
*CoS9(n,1) 9.31E-14 1.17E-13 1.05E-13 6.50E-14 3.89E-14 4.25E-14 (
CoS9(n,y) 5.52E-14 6.68E-14 5.76E-14 4.41E-14 2.79E-14 2.79E-14 . *U235(n,f) 7.83E-13 9.11E-13 1.02E-12 4.52E-13 3.32E-13 3.95E-13 ; U235(n,f) 2.63E-13 2.98E-13 2.66E-13 1.97E-13 9.87E-14 7.12E-14 j . i
* - Bare foil, all others were cadmium covered ,
i I i t e 6-45 i
l TABLE 6.3-2 l Fe-54.(n,p) REACTION RATES DERIVED FR0t1 THE STAINLESS STEEL ! GRADIENT CHAINS IRRADIATED DURING CYCLE 17 l
- l l
FEET REACTION RATE (ros/ nucleus) [ FROM f MIDPLANE O DEG 15 DEG 30 DEG 45 DEG
+6.5 2.02E-17 1.52E-17 1.17E-17 1.09E-17 - !
+5.5 5.04E-17 3.84E-17 2.96E-17 2.52E-17 >
+4.5 7.48E-17 5.23E-17 4.49E-17 3.94E-17 - !
+3.5 8.36E-17 6.31E-17 5.28E-17 4.53E-17 i
+2.5 8.80E-17 6.99E-17 4.94E-17 5.04E-17 i
+1.5 8.41E-17 6.80E-17 5.04E-17 4.69E-17
+0.5 7.83E-17 6.41E-17 5.38E-17 4.58E-17 l
. 6 7 7 0 -
8 -
-2.5 6.55E-17 6.21E-17 5.18E-17 5.14E-17 !
-3.5 7.24E-17 6.11E-17 4.86E-17 4.73E-17 - !
-4.5 7.04E-17 6.llE-17 4.61E-17 4.17E-17 !
-5.5 4.99E-17 4.06E-17 2.79E-17 2.90E-17
-6.5 2.07E-17 1.49E-17 1.13E-17 1.27E-17 i
e* l 6-46 l r i
TABLE 6.3-3 Ni-58 (n p) REACTION RATES DERIVED'FROM THE STAINLESS STEEL GRADIENT CHAINS IRRADIATED DURING CYCLE 17 FEET REACTION RATE (ros/ nucleus) FROM MIDPLANE O DEG 15 DEG 30 DEG 45 DEG
. +6.5 3.13E-17 2.33E-17 1.87E-17 'l.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 5.82E-17
+3.5 1.19E-16 9.48E-17 7.20E-17 6.53E-17
+2.5 1.28E-16 1.03E-16 7.56E-17 7.46E-17
+1.5 1.22E-16 9.38E-17 7.89E-17 7.12E-17
+0.5 1.09E-16 9.00E-17 7.30E-17 6.69E-17
-0.5 9.40E-17 8.29E-17 7.95E-17 7.24E-17
. -1.5 9.10E-17 7.99E-17 7.48E-17 7.32E-17
-2.5 1.03E-16 8.35E-17 7.68E-17 7.10E-17
- -3.5 1.05E-16 9.12E-17 7.06E-17 6.83E-17
-4.5 1.08E-16 8.43E-17 6.71E-17 6.05E-17
-5.5 7.22E-17 5.88E-17 4.43E-17 2.06E-17
-6.5 2.95E-17 2.35E-17 1.79E-17 1.86E-17 O
6-47 __ _ -_ ._._ _ _ . _ _ . . _ _ . w
TABLE 6.3-4 i i Co-59 (n,y) REACTION RATES DERIVED FROM THE STAINLESS STEEL , GRADIENT CHAINS IRRADIATED DURING CYCLE 17 ! i
*l FEET REACTION RATE (ros/ nucleus) .
FROM f MIDPLANE O DEG 15 DEG 30 DEG 45 DEG i
+6.5 2.llE-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 5.59E-14
+1.5 6.63E-14 1.15E-13 9.45E-14 5.90E-14 l
+0.5 6.68E-14 1.07E-13 9.45E-14 5.90E-14 !
-0.5 6.63E-14 8.46E-14 7.62E-14 4.92E-14 I
-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 4.63E-14 t
-3.5 6.37E-14 7.73E-14 6.5SE-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 !
l e l 6-48
TABLE 6.3-5 DERIVED EXPOSURE RATES FROM THE CAPSULE BB DOSIMETRY EVALUATION O DEGREE AZIMUTH - CORE MIDPLANE ~ A PRIORI ADJUSTED PARAMETER VALUE__ VALUE UNCERTAINTY p (E > 1.0 Mev) 1.56E+09 1.50E+09 8% 4 (E > 0.1 Mev) 1.25E+10 1.13E+10 16%
. p (E < 0.414 ev) 2.38E+09 1.49E+09 25%
p (Total) 3.40E+10 3.10E+10 15%
. dpa/sec 5.23E-12 4.79E-12 14%
COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES 0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 9.04E-19 1.16E-18 9.39E-19 1.29 1.04 Ti-46 (n,p) 1.32E-17 1.40E-17 1.31E-17 1.06 0.99 Fe-54 (n,p) 7.79E-17 7.70E-17 7.57E-17 0.99 0.97 Ni-58 (n,p) 1.10E-16 1.06E-16 'l.07E-16 0.96 0.97 U-238 (n,f) (Cd) 3.74E-16 4.51E-16 4,30E-16 1.21 1.15
. Np-237 (n,f) (Cd) 6.20E-15 6.68E-15 6.22E-15 1.08 1.00 Cc-59 (n,y) 9.31E-14 1.27E-13 9.33E-14 1.37 1.00 Co-59 (n,7) (Cd) 5.52E-14 6.37E-14 5.53E-14 1.15 1.00 U-235 (n,f) 7.83E-13 1.16E-12 7.87E-13 1.48 1.01 U-235 (n,f) (Cd) 2.63E-13 2.44E-13 2.54E-13 0.93 0.97 t
6-49 J
TABLE 6.3-6 i i DERIVED EXPOSURE RATES FROM THE CAPSULE DD DOSIMETRY EVALUATION : 15 DEGREE AZIMUTH - CORE MIDPLANE i PARAMETER A PRIORI ADJUSTED. VALUE VALUE
'f UNCERTAINTY. [
~
p (E > 1.0 Mev) 1.40E+09 1.25E+09 8% [ p (E > 0.1 Mev) 1.25E+10 9.91E+09 16% i ( (E < 0.414 ev) 2.27E+09 1.87E+09 24% .' ( (Total) 3.05E+10 2.72E+10 13% ; dpa/sec 4.45E-12 3.64E-12 13% . l l COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES l 15 DEGREE AZIMUTH - CORE MIDPLANE ! l REACTION RATE (ros/ nucleus) C/M !
.j A PRIORI ADJUSTED i REACTION MEASURED . CALC m _ CALC. A PRIORI ADJUSTED l Cu-63 (n,a) 8.01E-19 1.02E-18 8.34E-19 1.27 1.04 !
Ti-46 (n,p) 1.16E-17 1.21E-17 1.15E-17 1.05 0.99 ; Fe-54 (n.p) 6.79E-17 6.68E-17 6.57E-17 0.98 0.97 i Ni-58 (n p) 9.61E-17 9.18E-17 9.29E-17 0.96 0.97 l U-238 (n,f) (Cd) 3.09E-16 3.97E-16 3.61E-16 1.29 1.17 l Np-237 (n,f) (Cd) 5.01E-15 6.32E-15 5.17E-15 1.26 1.03 . Co-59 (n,1) 1.17E-13 1.28E-13 1.15E-13 1.10 0.98 , Co-59 (n,1) (Cd) 6.68E-14 6.85E-14 6.76E-14 1.02 1.01 - U-235 (n,f) 9.llE-13 1.10E-12 9.49E-13 1.21 1.04 U-235 (n,f) (Cd) 2.98E-13 2.51E-13 2.84E-13 0.84 0.95 - [ 4 6-50
TABLE 6.3-7 i DERIVED EXPOSURE RATES FROM THE CAPSULE EE DOSIMETRY EVALUATION 30 DEGREE AZIMUTH - CORE MIDPLANE ~ A PRIORI ADJUSTED , PARAMETER VALUE -VALUE UNCERTAINTY d (E > 1.0 Mev) 1.12E+09 1.08E+09 8% ( (E > 0.1 Mev) 9.99E+09 8.74E+09 16% i
. p (E < 0.414 ev) 1.91E+09 1.94E+09 23%
( (Total) 2.51E+10 2.44E+10 13%
. dpa/sec 3.54E-1? 3.17E-12 13%
COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES 30 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED , CALC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 6.79E-19 8.84E-19 7.09E-19 1.30 1.04 Ti-46 (n,p) 1.00E-17 1.04E-17 9.87E-18 1.04 0.99 Fe-54 (n,p) 5.64E-17 5.54E-17 5.50E-17 0.98 0.98 Ni-58 (n,p) 8.01E-17 7.59E-17 7.78E-17 0.95 0.97 U-238 (n,f) (Cd) 2.78E-16 3.20E-16 3.10E-16 1.15 1.11
. Np-237 (n,f) (Cd) 4.41E-15 5.02E-15 4.54E-15 1.14 1.03 Co-59 (n,7) 1.05E-13 1.09E-13 1.07E-13 1.03 1.02 Co-59 (n,7) (Cd) 5.76E-14 5.84E-14 5.72E-14 1.01 0.99 U-235 (n,f) 1.02E-12 9.28E-13 9.73E-13 0.91 0.95 U-235 (n,f) (Cd) 2.66E-13 2.14E-13 2.59E-13 0.80 0.97 9
6-51
-. - _ _ _ __ A
, m _ - _ _ . _. TAILE 6.3-8 l DERIVED EXPOSURE RATES FROM THE CAPSULE FF DOSIMETRY EVALUATION 45 DEGREE AZIMUTH - CORE MIDPLANE
~
A PRIORI ADJUSTED , PARAMETER VALUE VALUE UNCERTAINTY p (E > 1.0 Mev) 1.01E+09 9.55E+08 8% p (E > 0.1 Mev) 8.50E+09 7.04E+09 '12%
~
p (E < 0.414 ev) 1.92E+09 8.03E+08 27% - p (Total) 2.16E+10 1.82E+10 13% ; dpa/sec 3.06E-12 2.62E-12 13% - COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES ! 45 DEGREE AZIMUTH - CORE MIDPLANE > REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED ! Cu-63 (n,a) 6.97E-19 8.87E-19 7.24E-19 1.27 1.04 Ti-46 (n,p) 9.83E-18 1.03E-17 9.74E-18 1.04 0.99 Fe-54 (n,p) 5.41E-17 5.31E-17 5.28E-17 0.98 0.98 Ni-58 (n,p) 7.65E-17 7.24E-17 7.42E-17 0.95 0.97 U-238 (n,f) (Cd) 2.58E-16 2.95E-16 2.84E-16 1.15 1.10 Np-237 (n,f) (Cd) 3.66E-15 4.40E-15 3.79E-15 1.20 1.03 - ! Co-59 (n,7) 6.50E-14 1.02E-13 6.48E-14 1.56 1.00 : Co-59 (n,7) (Cd) 4.41E-14 5.04E-14 4.43E-14 1.14 1.01 ** [ U-235 (n,f) 4.52E-13 9.21E-13 4.65E-13 2.04 1.03 f U-235 (n,f) (Cd) 1.97E-13 1.84E-13 1.89E-13 0.94 0.96 i T D 6-52 i k b
i TABLE 6.3-9 DERIVED EXPOSURE RATES FROM THE CAPSULE AA DOSIMETRV EVALUATION i 0 DEGREE AZIMUTH - CORE TOP , A PRIORI ADJUSTED PARAMETER VALUE VALUE UNCERTAINTY ' p (E > 1.0 Mev) 7.11E+08 7.91E+08 11% ( (E > 0.1 Mev) 5.72E+09 5.88E+09 22% , ( (E < 0.414 ev) 1.09E+09 5.49E+08 25% ( (Total) 1.55E+10 1.45E+10 18% !
. dpa/sec 2.39E-12 2.45E-12 19%
r COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES i 0 DEGREE AZIMUTH - CORE TOP , REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 4.23E-19 5.31E-19 4.40E-19 1.25 1.04 Ti-46 (,n,p) 6.56E-18 6.39E-18 6.41E-18 0.97 0.98 Fe-54 (a,p) 3.55E-17 3.51E-17 3.62E-17 0.99 1.02 . Ni-58 (n,p) 5.66E-17 4.82E-17 5.41E-17 0.85 0.96 ! U-238 (n,f) (Cd) 2.12E-16 2.06E-16 2.24E-16 0.97 1.06 L
. Co-59 (n,1) 3.89E-14 5.81E-14 4.13E-14 1.49 1.06 Co-59 (n,7) (Cd) 2.79E-14 2.91E-14 2.68E-14 1.04 0.96 i . U-235 (n,f) 3.32E-13 5.29E-13 3.02E-13 1.59 0.91 U-235 (n,f) (Cd) 9.87E-14 1.llE-13 1.02E-13 1.13 1.03 I
6-53 i
i TABLE 6.3-10 DERIVED EXPOSURE RATES FROM THE CAPSULE CC 00SIMETRY EVALUATION j 0 DEGREE AZIMUTH - CORE BOTTOM l A PRIORI ADJUSTED PARAMETER VALUE VALUE UNCERTAINTY p (E > 1.0 Mev) 7.24E+08 6.82E+08 8%
~
p (E > 0.1 Mev) 5.83E+09 4.53E+09 16% p (E < 0.414 ev) 1.11E+09 6.85E+08 21% ( (Total) 1.58E+10 3.15E+10 14% dpa/sec 2.43E-12 3.95E-12 13% - COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0 DEGREE AZIMUTH - CORE BOTTOM REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n.o) 3.92E-19 5.41E-19 4.10E-19 1.38 1.05 Ti-46 (n,p) 6.08E-18 6.50E-18 6.00E-18 1.07 0.99 Fe-54 (n,p) 3.62E-17 3.58E-17 3.52E-17 0.99 0.97 Ni-58 (n.p) 5.22E-17 4.91E-17 5.04E-17 0.94 0.96 U-238 (n,f) (Cd) 1.78E-16 2.09E-16 2.00E-16 1.18 1.12 Np-237 (n,f) (Cd) 2.54E-15 3.10E-15 2.60E-15 1.22 1.02 . Co-59 (n,1) 4.25E-14 5.92E-14 4.57E-14 1.39 1.07 Co-59 (n,7) (Cd) 2.79E-14 2.96E-14 2.66E-14 1.06 0.95 - U-235 (n,f) 3.95E-13 5.38E-13 3.48E-13 1.36 0.88 U-235 (n,f) (Cd) 7.12E-14 1.13E-13 7.66E-14 1.59 1.08 - D 6-54
i' i TABLE 6.3-11 ; FAST NEUTRON FLUX.(E > 1.0 MeV) AS A FUNCTION a' ' 0F AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLE 17 IRRADIATION l AZIMUTHAL ANGLE ,
, HEIGHT I 4(ft) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES l
. *6.5 4.39E+08 3.20E+08 2.50E+08- 2.18Es08
+6.0 7.91E+08
. +5.5 1.07E+09 8.15E+08 5.98E+08 4.96E+08 :
+4.5 1.61E+09 1.17E+09 9.12E+08 7.81E+08 l
+3.5 1.74E+09 1.31E+09 1.04E+09 8.86E+08 !
+2.5 1.85E+09 1.44E+09 1.03E+09 1.00E+09
+1.5 1.77E+09 1.36E+09 1.07E+09 9.42E+08 l
+0.5 1.61E+09 1.29E+09 1.06E+09 9.02E+08
, 0.0 1.50E+09 1.25E+09 1.08E+09 9.55E+08 I
-0.5 1.39E+09 1.21E+09 1.10E+09 1.01E+09 l
. -1.5 1.36E+09 1.13E+09 1.04E+09 9.75E+08
-2.5 1.43E+09 1.22E+09 1.07E+09 9.85E+08 l
-3.5 1.52E+09 1.27E+09 9.90E+08 9.26E+08 i
-4.5 1.52E+09 1.22E+09 9.40E+08 8.19E+08 !
-5.5 1.05E+09 8.30E+08 5.95E+08 4.22E+08
-6.0 6.82E+08 j
-6.5 4.31E+08 3.18E+08 2.41E+08 2.51E+08 I r
. i l
~ \
\
- l 6-55 j
TABLE 6.3-12 FAST NEUTRON FLUX (E > 0.1 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLE 17 IRRADIATION AZIMUTHAL ANGLE
~
HEIGHT (ft) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES
+6.5 3.30E+09 2.54E+09 2.02E+09 1.61E+09 - +6.0 5.88E+09 +5.5 8.07E+09 6.46E+09 4.84E+09 3.66E+09 -
1
+4.5 1.21E+10 9.31E+09 7.38E+09 5.75E+09 +3.5 1.31E+10 1.04E+10 8.44E+09 6.53E+09 +2.5 1.39E+10 1.14E+10 8.36E+09 7.37E409 +1.5 1.33E+10 1.08E+10 8.63E+09 6.95E+09 +0.5 1.21E+10 1.02E+10 8.58E+09 6.65E+09 0.0 1.13E+10 9.91E+09 B.74E+09 7.04E+09 - -0.5 1.05E+10 9.E9E+09 8.90E+09 7.43E+09 -1.5 1.02E+10 8.95E+09 8.44E+09 7.19E+09 * -2.5 1.08E+10 9.70E+09 8.63E+09 7.26E+09 -3.5 1.15E+10 1.01E+10 8.01E+09 6.83E+09 -4.5 1.15E+10 9.67E+09 7.61E+09 6.04E+09 -5.5 7.88E+09 6.58E+09 4.81E+09 3.11E+09 -6.0 4.53E+09 -6.5 3.25E+09 2.52E+09 1.95E+09 1.85E+09 I
x l 6-56 i I
TABLE 6.3-13 IRON ATOM DISPLACEMENT RATE (dpa/sec) AS A FUNCTION +~ CF AXIAL POSITION WITHIN THE REACTOR CAVITY CYCLE 17 IRRADIATION AZIMUTHAL ANGLE HEIGHT (ft) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES
. +6.5 1.40E-12 9.31E-13 7.34E-13 5.98E-13
+6,0 2.45E-12
. +5.5 3.42E-12 2.37E-12 1.79E-12 1.36E-12
+4.5 5.14E-12 3.42E-12 7.68E-12 2.14E-12
+3.5 5.55E-12 3.83E-1: 3.06E-12 2.43E-12
+2.5 5.91E-12 4.20E "2 3.03E-12 2.74E-12
+1.5 5.64E-12 3.95E-12 3.13E-12 2.59E-12
+0.5 5.14E-12 3.76E-12 3.11E-12 2.48E-12
. 0.0 4.79E-12 3.64E-12 3.17E-12 2.62E-12
-0.5 4.44E-12 3.52E-12 3.23E-12 2.76E-12
. -1.5 4.33E-12 3.29E-12 3.06E-12 2.68E-12
-2.5 4.58E-12 3.56E-12 3.13E-12 2.70E-12
-3.5 4.85E-12 3.69E-12 2.91E-12 2.54E-12
-4.5 4.86E-12 3.55E-12 2.76E-12 2.25E-12
-5.5 3.34E-12 2.42E-12 1.75E-12 1.16E-12
-6.0 1.95E-12
-6.5 1.38E-12 9.26E-13 7.06E-13 6.87E-13 e
4 6-57 I
FIGURE 6.3-1 l i FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION I
^
ALONG THE O DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 17 IRRADIATION l
~
Neutron Flux (n/cm2-sec) i
.4 C ,
1.000E + 09
/
i
- k -
- r 1.000E + 08 ' ' '
l
-8 -6 -4 -2 0 2 4 6 8 Distance From Core Midplane (ft) -
O CYCLE 17 - O DEG
~
i 6-58 i
e FIGURE 6.3-2 FAST NEUTRON F' LUX (E'> 1.0 MeV) AS A FUNCTION OF AXIAL POSITION '~ ALONG THE 15 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 17 IRRADIATION Neutron Flux (n/cm2-sec)
, 1 4
_ i 1.000E + 09 ,
~
I 1.000E + 08 <
. l
-8 -6 -4 -2 0 2 4 6 8 ;
Distance From Core Midplane (ft)
- O CYCLE 17 - 15 DEG 6-59 I
i
r-l FIGURE 6.3-3 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION . ALONG THE 30 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 17 IRRADIATION . i Neutron Flux (n/cm2-sec) l i 1.000E + 09 C ' ' 2;G G u a m .; _ l l 1.000E + 08 ' ' ' ' '
-8 -6 -4 -2 0 2 4 6 8 .
Distance From Core Midplane (ft) O CYCLE 17 - 30 DEG 6-60
i
~
FIGURE 6.3-4
~
FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 45 DEGREE TRAVERSE IN THE REACTOR CAVITY : CYCLE 17 IRRADIATION Neutron Flux (n/cm2-sec) 1.000E + 10 - _ . 1 r i
. g
- 1.OOOE + 09 _
- - u i
~
t k
. 1.000E + 08
-8 -6 -4 -2 0 2 4 6 8 1 Distance From Core Midplane (ft)
O CYCLE 17 - 45 DEG 6-61
s.
)
i i I f
. I i
t
. e O
P a r t e I 9 L E B
?
ik h 9 I e r 9 A h i 9 P G f I-t e I L O S + 1 I s b I s I i
l SECTION 7.0 i COMPARIS0N OF CALCULATIONS WITH MEASUREMENTS l In order to develop accurate neutron exposure profiles at the inner ~ diameter and through the thickness of the pressure vessel wall, the ! measurement results provided in Sections 5.0 and 6.0 must be combined with ; analytically determined spatial gradients provided in Section 4.0. In i essence, this approach accepts the measurement results as the best ,
. available exposure rate information for the irradiation period in question l and assumes that the analytically determined radial distribution functions
. provide accurate representations of the spatial gradients that exist among ;
the measurement locations and the points of interest within the pressure vessel wall. This approach is analagous to the common practice of I normalizing a cycle specific forward neutron transport calculation to available measurements from either surveillance capsule or reactor cavity dosimetry programs. An indication of the acceptability of this method of exposure . determination can be gained by an absolute comparison of the results of j neutron transport calculations with all measured results applicable to a given reactor. These comparisons quantify the biases that may exist due to the transport methodology, reactor modeling, and/or reactor operating ; characteristics over the respective irradiation periods; and, furthermore, demonstrate the degree of consistency among the measurements obtained from different geometric locations and varying irradiation intervals.
. In this section, comparisons of the measurement results from surveillance l
capsule and reactor cavity dosimetry with corresponding analytical L 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 d (E > 1.0 MeV), p (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 7-1
I I l dat'a from the counting laboratories. It is shown that these two levels of j comparison yield consistent and similar results, indicating that the least 1 squares adjustment methodology is producing accurate exposure results and . that the calculation / measurement comparisons are yielding accurate bias *i factors that can be applied to neutron transport calculations performed ! for the Point Beach Unit 2 reactor.
'[
i A 7.1 Comparison of Least Squares Adjustment Results with Calculation , In Table 7.1-1, comparisons of calculated and measured exposure rates for i the four surveillance capsule dosimetry sets and for the three cycles of
- reactor riidplane dosimetry sets irradiated during Cycles 15,16, and 17 [
are given. In all cases, the calculated values were based on the fuel i cycle specific exposure calculations averaged over the appropriate l irradiation period. That is, the Capsule V values apply to Cycle 1, the ! Capsules T, R, and S values represent averages over Cycles 1 through 3,1 ! through 5, and 1 through 16, respectively; and, the cavity measurements , directly apply to Cycles 15, 16, and 17. t i An examination of Table 7.1-1 indicates that, considering all of the l available core midplane data, the calculated exposure rates underpredicted ; measurements by factors of 0.869, 0.915, and 0.918 for p (E > 1.0 ! MeV), p (E > 0.1 MeV), and dpa/sec, respectively. The standard i deviations associated with each of the 16 sample data sets were 8.3%, j 9.3%, and 8.8%, respectively. l The data comparisons provided in Table 7.1-1 also indicate a bias between - the surveillance capsule and reactor cavity dosimetry comparisons; where ! the agreement between calculation and measurement is best at the cavity ** ! sensor locations and somewhat worse at the surveillance capsules. The l reasons for this apparent bias cannot be acertained at this time, but may
- j be due to the complexities in modeling the Point Beach Unit 2 surveillance I capsules and support structure that exhibit a variable geometry over the *l capsule height. In contrast, the senser holders used in the cavity i i
i 7-2 ; i i
irradiations were designed to provide free field measurements by minimizing perturbations in the neutron field and, thus, simplifying the neutron transport calculations. In any event, the inclusion of both sets of measurements in the data base listed in Table 7.1-1 results in an acceptable determination of calculation to measurement bias factors with standard deviations of better than 10% in all cases. 7.2 Comparisons of Measured and Calculated Sensor Reaction Rates In Table 7.2-1, calculation / measurement 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 standard deviations observed for the six . fast neutron reactions range from 4% to 12% on an individual reaction basis; whereas, the overall average C/M ratio for the entire data set has an associated la standard deviation of approximately 8.5%. Furthermore, the average C/M bias of 0.892 observed in the reaction rate comparisons is in excellent agreement with the values of 0.869, 0.915, and 0.918 observed in the exposure rate comparisons shown in Table 7.1-1. e D 7-3 l
l. 1 TABLE 7.1-1 L b COMPARISON OF MEASURED AND CALCULATED EXPOSURE RATES FROM
~
; SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS
~
FAST NEUTRON FLUX (E > 1.0 MeV)
~
d [n/cm2-sec} MEASURED CALCULATED C/M 13 DEGREE CAPSULE (V) 1.48E+11 1.10E+11 0.743 I 23 DEGREE CAPSULE (T) 8.23E+10 6.61E+10 0.803 ~ 13 DEGREE CAPSJLE (R) 1.42E+11 1.10E+10 0.775
- 33 DEGREE CAfSULE (S) 7.44E+10 5.43E+10 0.730 0 DEGREE CAVITY (CY15) 1.87E+09 1.71E+09 0.914 15 DEGREE CAVITY (CY15) 1.69E+09 1.43E+09 0.846 30 DEGREE CAVITY (CY15) 1.23E+09 1.11E+09 0.902 45 DEGREE CAVITY (CY15) 1.10E+09 9.15E+08 0.832 0 DEGREE CAVITY (CY16) 1.45E+09 1.39E+09 0.960
- 15 DEGREE CAVITY (CY16) 1.28E+09 1.17E+09 0.910 30 DEGREE CAVITY (CY16) 9.73E+08 9.44E+08 0.970 45 DEGREE CAVITY (CY16) 9.63E+08 8.28E+08 0.860 0 DEGREE CAVITY (CY17) 1.50E+09 1.40E+09 0.930 15 DEGREE CAVITY (CY17) 1.25E+09 1.17E+09 0.937 30 DEGREE CAVITY (CY17) 1.08E+09 9.68E+08 0.896 45 DEGREE CAVITY (CY17) 9.55E+08 8.60E+08 0.901 AVERAGE C/M RATIO 0.869 la VARIATION 0.072 -
4 7-4 l
TABLE 7.1-1 (C3ntinued) l COMPARIS0N OF'MEASUPED'AND CALCULATED EXPOSURE RATES FROM - -1
' SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS FAST NEUTRON FLUX (E > 0.1 MeV) j i
p [n/cm2-sec) i MEASURED CALCULATED C/M l 13 DEGREE CAPSULE (V) 5.36E+11 4.21E+11 0.785
~
23 DEGREE CAPSULE (T) 2.84E+11 2.30E+11 0.810 13 DEGREE CAPSULE (R) 5.40E+11 4.19E+11 0.776 33 DEGREE CAPSULE (S) 2.41E+11. 1.90E+11 0.788 l 0 DEGREE CAVITY (CY15) 1.38E+10 1.36E+10 0.986 , 15 DEGREE CAVITY (CY15) 1.45E+10 1.23E+10 0.848 l 30 DEGREE CAVITY (CY15) 1.02E+10 9.73E+09 0.954 l 45 DEGREE CAVITY (CY15) 8.66E+09 7.61E+09 0.879 .. O DEGREE CAVITY (CY16) 1.10E+10 1.10E+10- 1.004 l 15 DEGREE CAVITY (CY16) 1.03E+10 1.00E+10 0.971 l 30 DEGREE CAVITY (CY16) 8.16E+09 8.30E+09 1.017 l 45 DEGREE CAVITY (CY16) 7.65E+09 6.89E+09 0.900 i 0 DEGREE CAVITY (CY17) 1.13E+10 1.11E+10 0.978 l 15 DEGREE CAVITY (CY17) 9.91E+09 1.01E+10 1.009 j 30 DEGREE CAVITY (CY17) 8.74E+09 8.51E+09 0.949 ! 45 DEGREE CAVITY (CY17) 7.04E+09 7.16E+09 0.978 ; AVERAGE C/M RATIO 0.915 l
. la VARIATION 0.085 I L
]
7-5 l f
- , - - . . _ . . , _ . _ . , __ _ .J
m
-{
TABLE?7.1-1 (Continued) COMPARISON OF MEASURED AND CALCULATED EXPOSURE RATES FROM-SURVEILLANCE CAPSULE ^AND CAVITY DOSIMETRY IRRADIATIONS , IRON ATOM DISPLACEMENT RATE . [dpa/sec) ! MEASURED CALCULATED' C/M 13 DEGREE CAPSULE (V) 2.53E-10 2.04E 0.806 I t 23 DEGREE CAPSULE (T) 1.40E-10 1.17E-10 0.836 13 DEGREE CAPSULE-(R) 2.51E-10 2.03E-10 0.809- - l 33 DEGREE CAPSULE (S) 1.22E-10 9.55E 0.783 - 0 DEGREE CAVITY (CY15) 5.87E-12 5.18E-12 0.882 15 DEGREE CAVITY (CY15) 5.21E-12 4.60E-12 0.883 7 30 DEGREE CAVITY (CY15) 3.69E-12 3.64E-12 0.986 l 45 DEGREE CAVITY (CY15) 3.15E-12 2.89E-12 0.917 0 DEGREE CAVITY (CY16) 4.66E-12 4.22E-12 0.905 - 15 DEGREE CAVITY (CY16) 3.76E-12 3.75E-12 0.998 ! 30 DEGREE CAVITY (CY16) 2.95E-12 3.10E-12 1.053 *; 45 DEGREE CAVITY (CY16) 2.78E-12 2.62E-12 0.941 f 0 DEGREE CAVITY (CY17) 4.79E-12 4.23E-12 0.880 j 15 DEGREE CAVITY (CY17) 3.64E-12 3.77E-12 1.031 , 30 DEGREE CAVITY.(CY17) 3.17E-12 3.18E-12 0.979 45 DEGREE CAVITY (CY17) 2.62E-12 2.72E-12 0.998 f AVERAGE C/M RATIO 0.918 i la VARIATION 0.081 - -
= r i
f 7-6 l
TABLE 7.2-1 COMPARIS0N OF MEASURED AND CALCULATED NEUTRON SENSOR REACTION RATES FROM SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS ; L..- [u63(n.0) T146(n.0) Fe54(n.o) NiS8(n.0) U238(n.fi No237(n.f)
- CAPSULES
^
l
. V 0.898 0.898 0.724 0.850 T 0.893 0.873 0.888 0.849 0.830 ;
. R 0.832 0.892 0.864 0.787 0.823 l S 0.797 0.728 0.742 0.835 i CY15 CAVITY 0 DEGREE 0.919 0.920 0.920 0.884 0.979 0.992 15 DEGREE 0.796 0.799 0.812 0.803 0.910 0.862- l 30 DEGREE 0.905 0.893 0.867 0.884 0.982 0.968 l
. 45 DEGREE 0.870 0.851 0.907 0.859 0.887 0.888 CY16 CAVITY
. O DEGREE 0.932 0.955 0.952 0.927 1.034 1.021 15 DEGREE 0.852 0.879 0.887 0.858 0.935 1.004 30 DEGREE 0.854 0.883 0.879 0.863 1.062 1.023 45 DEGREE 0.852 0.887 0.864 0.841 0.940 0.900 CY17 CAVITY 0 DEGREE 0.885 0.919 0.909 0.902 1.109 0.990 15 DEGREE 0.799 0.834 0.827 0.825 1.090 1.058 30 DEGREE 0.844 0.851 0.858 0.845 1.00b 0.991 3
. 45 DEGREE 0.808 0.830 0.835 0.817 0.969 1.016 AVERAGE C/M 0.859 0.875 0.879 0.853 0.938 0,941 la 0.043 0.042 0.036 0.046 0.113 0.080 TOTAL C/M RATIO 0.892 la VARIATION 0.076 7-7
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
*J pressure vessel through the completion of Cycle 17. Based on the continued use of the Cycle 16-17 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 and lower shell forgings and the beltline circumferential weld are highlighted.
8.1 Exposure Distributions Within the Beltline Region In essence, an approach using analytically determined gradient information to extrapolate measurement results to locations of interest within the pressure vessel is based on the assertion that the measured values of exposure rates in the reactor cavity represent the best available neutron flux data for the irradiation period in question and, further, on the assumption that the analytically determined radial distribution functions provide accurate representations of the spatial gradients that exist among
. the measurement locations and points of interest within the pressure vessel wall. This method is analagous to the common practice of
. normalizing a cycle specific forward neutron transport calculation to available measurements from either surveillance capsule or reactor cavity dosimetry programs.
This approach provides accurate assessments of vessel exposure with associated uncertainties for periods of operation during which continuous 8-1
monitoring has occured. In the case of Point Beach Unit 2, the cavity dosimetry program providing a complete spatial mapping of a sector of the beltline region of the pressure vessel was installed at the start of Cycle
- 15. Additional monitoring was limited to the four scheduled surveillance ~
capsule withdrawals described in preceding sections of this report. The dosimetry data from these capsules provide mersurement information at a ' single point within the reactor geometry for the four extended irradiation periods, but cannot be used to establish a verification of the exposure of
- the vessel at azimuthal locations far removed from the measurement point.
~
Therefore, in order to establish a baseline exposure of the pressure - ' vessel applicable to the onset of the reactor cavity measurement program, all available core midplane measured data were combined with fuel cycle - specific transport calculations to provide best estimate exposures for the first 14 cycles of operation. The reactor cavity measurements were then used directly to provide the continuous monitoring capability for Cycles 15 and beyond. 8.1.1 Baseline Exposure at the End of Cycle 14
.(
In Table 7.1-1, comparisons of calculated and measured exposure rates for the four surveillance capsule dosimetry sets and for the twelve cavity dosimetry sets that were located on the core midplane are given. From Table 7.1-1, it was noted that, considering all of the midplane data, the calculated exposure values underpredicted measurement by factors of 0.869, 0.915, and 0.918 for p (E > 1.0 MeV), p (E > 0.1 MeV), and dpa/sec, respectively. The corresponding 10 standard deviations in these averages of the twelve sample data sets were 8.3%, 9.3%, and - 8.8%. In developing the best estimate baseline exposure for the Point Beach Unit 2 reactcr pressure vessel these ratios were employed as bias factors to
- scale the cycle specific neutron transport calculations documented in Section 4.0 of this report. In particular, the following bias factors
- were employed to establish the baselina exposures of the vessel wall:
8-2
l M/C BIAS
+ (E > 1.3 MeV) 1.151
+ (E > 0.1 MeV) 1.093
- dpa 1.089 l
The end of Cycle 14 best estimate exposures at the pressure vessel clad / base metal interface are provided in Tables 8.1-1 through 8.1-4 for 1 4 (E > 1.0 MeV), in Tables 8.1-5 through 8.1-8 for f (E > 0.1 1 MeV), and in Tables 8.1-9 through 8.1-12 for dpa. In these data tables,
". exposures are presented as a function of axial position for four azimuthal locations around the circumference of the vessel. From these tabulations, the locations of maximum exposure of the various materials comprising the )
beltline region can easily be determined. Exposure distributions through the vessel wall can be developed by normalizing the surface exposures from : Tables 8.1-1 through 8.1-12 to the appropriate radial distribution ! functions given in Section 4.0 of this report. l r 8.1.2 Exposure Accrued During Cycles 15, 16, and 17 i To assess the incremental exposure resulting from the Cycles 15, 16, and 17 irradiations, the measured results from the reactor cavity multiple foil sensor sets were directly extrapolated to the vessel clad / base metal interface using the analytically derived gradient data from Section 4.0 of this report. The axial gradient chain measurements were, of course, employed to develop the axial traverses along the vessel wall. The l extrapolated results applicable to the vessel inner surface are also -
. incorporated into Tables 8.1-1 through 8.1-12 to establish the best estimate exposure accrued by the reactor vessel through the end of Cycles
. 15, 16, and 17, respectively. '
Again, as noted above, exposure distributions through the vessel wall, can f be developed using these surface exposures and radial distribution ! functions from Section 4.0. This exposure information, applicable through ' [ the end of Cycle 17, was derived from an extensive set of measurements and 8-3 i
assures that embrittlement gradients can be established with a minimum uncertainty. Further, as the monitoring program continues and additional data become available, the overall plant specific data base for Point Beach Unit 2 will expand resulting in reduced uncertainties and an
- improved accuracy in the assesment of vessel condition.
8.1.3 Projection of Future Vessel Exposure At the end of Cycle 17, the Point Beach Unit 2 reactor had accrued 15.6 I 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 included in Tables 8.1-1 through 8.1-12 in addition to the plant specific exposure assessments through the end of Cycle 17. These temporal extrapolations into the future were based on the assumption that the measured data averaged over the Cycles 16 and 17 irradiations - were representative of all future fuel cycles. That is, that future fuel designs would incorporate the low leakage fuel management concept - including part length hafnium absorbers designed to provide flux reduction measures at the maximum exposure locations along the beltline circumferential weld. Examination of these projected exposure levels establishes the long term effectiveness of flux reduction measures 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 1 l l 8-4
TABLE 8.1-1
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 - 0 DEGREE AZIMUTHAL ANGLE HEIGHT 4 (E > 1.0 MeV) [n/cm2] (ft) E0C 14 E0C 15 E0C 16 E0C 17 32 EFPY 48 EFPY
. 46.0 4.27E+18 4.52E+18 4.79E+18 5.07E+18 1.03E+19 1.55E+19
+5.5 8.32E+18 8.66E+18 9.03E+18 9.41E+18 1.66E+19 2.36E+19
. +4.5 1.22E+19 1.27E+19 1.33E+19 1.39E+19 2.49E+19 3.57E+19
+3.5 1.37E+19 1.42E+19 1.49E+19 1.55E+19 2.75E+19 3.92E+19
+2.5 1.42E+19 1.48E+19 1.55E+19 1.61E+19 2.88E+19 4.12E+19
+1.5 1.44E+19 1.50E+19 1.56E+19 1.62E+19 2.83E+19 4.01E+19
+0.5 1.45E+19 1.52E+19 1.57E+19 1.63E+19 2.72E+19 3.79E+19 0.0 1.45E+19 1.52E+19 1.57E+19 1.62E+19 2.63E+19 3.61E+19
, -0.5 1.45E+19 1.52E+19 1.57E+19 1.61E+19 2.53E+19 3.43E+19
-1.5 1.45E+19 1.52E+19 1.56E+19 1.61E+19 2.51E+19 3.38E+19
. -2.5 1.44E+19 1.51E+19 1.56E+19 1.61E+19 2.57E+19 3.52E+19
-3.5 1.42E+19 1.47E+19 1.53E+19 1.58E+19 2.62E+19 3.64E+19
-4.5 1.30E+19 1.35E+19 1.40E+19 1.46E+19 2.48E+19 3.48E+19
-5.5 9.10E+18 9.45E+18 9.82E+18 1.02E+19 1.73E+19 2.42E+19
-6.0 4.44E+18 4.68E+18 4.93E+18 5.16E+18 9.83E+18 1.44E+19 Note: Height is provided relative to the axial midplane of the active core. !
\
l e 0 m 8-5 1
)
TA8LE 8.1-2 j i
SUMMARY
OF.8EST 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 i HEIGHT *' 4 (E > 1.0 MeV) [n/cm2) ' (ft) EOC 14 EOC 15 EOC 16 EOC 17 32 EFPY 48 EFPY
+6.0 2.61E+18 2.80E+18 2.98E+18 3.15E+18 6.49E+18 9.76E+18 -
f
+5.5 5.08E+18 5.34E+18 5.58E+18 5.81E+18 1.04E+19 1.49E+19
+4.5 7.44E+18 7.83E+18 8.21E+18 8.54E+18 1.54E+19 2.22E+19 -
j
+3.5 8.34E+18 8.77E+18 9.21E+18 9.58E+18 1.73E+19 2.49E+19
]
+2.5 8.62E+18 9.09E+18 9.53E+18 9.94E+18 1.82E+19 2.62E+19
+1.5 8.77E+18 9.23E+18 9.65E+18 1.00E+19 1.78E+19 2.54E+19
+0.5 8.85E+18 9.30E+18 9.69E+18 1.01E+19 1.73E+19 2.44E+19 i
0.0 8.89E+18 9.35E+18 9.71E+18 1.01E+19 1.70E+19 2.38E+19 !
-0.5 8.89E+18 9.36E+18 9.70E+18 1.00E+19 1.66E+19~ 2.31E+19 -!
-1.5 8.87E+18 9.32E+18 9.67E+18 9.98E+18. 1.64E+19 2.27E+19 i
l
-2.5
-3.5 8.80E+18 9.25E+18 9.59E+18 9.'94E+18 1.66E+19 2.31E+19 'f 8.61E+18 9.03E+18 9.43E+18 9.79E+18 1.71E+19 2.42E+19 I
-4.5 7.93E+18 8.32E+18 8.71E+18 9.06E+18 1.62E+19 2.31E+19 l
-5.5 5.55E+18 5.81E+18 6.06E+18 6.29E+18: 1.09E+19 1.55E+19 !
- -6.0 2.72E+18 2.90E+18 3.06E+18 3.21E+18 6.27E+18 9.26E+18 !
i Note: Height is provided relative to the axial ! midplane of the active core. ! I { f i r j l e
'f,
- 8-6 !
l l
]
TABLE 8.1-3 . r
SUMMARY
OF BEST ESTIMATE FAST NEUTRON (E > 1.0 MeV) EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF THE POINT BEACH UNIT 2 l REACTOR PRESSURE VESSEL - 30 DEGREE AZIMUTHAL ANGLE
~
r HEIGHT f (E > 1.0 MeV) [n/cm2] (ft) E0C 14 E0C 15 EOC 16 E0C 17 32 EFPY 48 EFPY
. 46.0 1.99E+18 2.12E+18 2.25E+18 2.38E+18 4.82E+18 7.21E+18
+5.5 3.85E+18 4.04E+18 4.21E+18 4.39E+18 7.73E+18 1.10E+19
. +4.5 5.62E+18 5.90E+18 6.15E+18 6.42E+18 1.14E+19 1.62E+19 !
+3.5 6.29E+18 6.61E+18 6.87E+18 7.17E+18 1.26E+19 1.80E+19 7
+2.5 6.51E+18 6.85E+18 7.15E+18 7.44E+18 1.31E+19 1.87E+19
+1.5 6.61E+18 6.94E+18 7.23E+18 7.53E+18 1.32E+19 1.88E+19
+0.5 6.67E+18 7.01E+18 7.30E+18 7.60E+18 1.33E+19 1.89E+19 ,
0.0 6.69E+18 7.03E+18 7.32E+18 7.63E+18 1.34E+19 1.90E+19
. -0.5 6.69E+18 7.04E+18 7.33E+18 7.64E+18 1.34E+19 1.91E+19
-1.5 6.68E+18 7.02E+18 7.29E+18 7.59E+18 1.31E+19 1.85E+19
. -2.5 6.62E+18 6.95E+18 7.23E+18 7.54E+18 1.32E+19 1.87E+19
-3.5 6.49E+18 6.80E+18 7.06E+18 7.34E+18 1.26E+19 1.77E+19 i
-4.5 5.99E+18 6.27E+18 6.51E+18 6.78E+18 1.17E+19 1.65E+19
-5.5 4.20E+18 4.38E+18 4.54E+18 4.71E+18 7.88E+18 1.10E+19
-6.0 2.07E+18 2.20E+18 2.31E+18 2.42E+18 4.50E+18 6.54E+18 Note: Height is provided relative to the axial midplane of the active core.
l l e 8-7 l
F i i TABLE 8.1-4 i
SUMMARY
OF BEST ESTIMATE FAST NEUTRON (E > 1.0 MeV) EXPOSURE i PROJECTIONS FOR THE BELTLINE REGION OF THE POINT BEACH UNIT 2 ; REACTOR PRESSURE VESSEL - 45 DEGREE AZIMUTHAL ANGLE
~,
HEIGHT !
+ (E > 1.0 MeV) [n/cm2)
(ft) EOC 14 EOC 15 E0C 16 E0C 17 E EFPY 48 EFPY
~ +6.0 1.75E+18 1.87E+18 1.99E+18 2.10E+18 4.33E+18 6.50E+18 - +5.5 3.40E+18 3.57E+18 3.73E+18 3.88E+18 6.92E+18 9.90E+18 +4.5 4.96E+18 5.22E+18 5.47E+18 5.71E+18 1.05E+19 1.52E+19 - +3.5 5.56E+18 5.84E+18 6.12E+18 6.39E+18 1.16E+19 1.67E+19 +2.5 5.75E+18 6.06E+18 6.36E+18 6.66E+18 1.25E+19 1.82E+19 i +1.5 5.84E+18 6.14E+18 6.43E+18 6.72E+18 1.23E+19 1.77E+19 +0.5 5.90E+18 6.19E+18 6.48E+18 6.76E+18 1.22E+19 1.75E+19 0.0 5.92E+18 6.22E+18 6.51E+18 6.81E+18 1.24E+19 1.79E+19 -0.5 5.92E+18 6.23E+18 6.53E+18 6.83E+18 1.27E+19 1.83E+19 ' -1.5 5.91E+18 6.20E+18 6.49E+18 6.79E+18 1.24E+19 1.79E+19 -2.5 5.86E+18 6.15E+18 6.44E+18 6.74E+18 1.24E+19 1.79E+19 ' -3.5 5.74E+18 6.02E+18 6.29E+18 6.58E+18 1.20E+19 1.72E+19 . -4.5 5.29E+18 5.54E+18 5.81E+18 6.06E+18 1.10E+19 1.58E+19 -5.5 3.72E+18 3.88E+18 4.06E+18 4.19E+18 7.12E+18 9.99E+18 ; -6.0 1.84E+18 1.95E+18 2.07E+18 2.15E+18 4.09E+18 5.98E+18 -
Note: Height is provided relative to the axial I midplane of the active core. f i
*l 8-8 ,
e
h i TABLE 8.1 i l
SUMMARY
OF 8EST ESTIMATE FAST NEUTRON (E > 0.1 MeV) EXPOSURE i PROJECTIONS FOR THE BELTLINE REGION OF THE POINT BEACH UNIT 2 . REACTOR PRESSURE VESSEL - 0 DEGREE AZIMUTHAL ANGLE l HEIGHT t (E > 0.1 MeV) [n/cm2] (ft) EOC 14 EOC 15 EOC 16 EOC 17- 32 EFPY 48 EFPY
. +6.0 1.10E+19 1.17E+19 1.23E+19 1.30E+19 2.60E+19 3.86E+19
+5.5 2.15E+19 2.24E+19 2.34E+19 2.43E+19 4.30E+19 6.13E+19
.. +4.5 3.15E+19 3.28E+19 3.43E+19 3.57E+19 .6.45E+19 9.25E+19 ;
+3.5 3.54E+19 3.69E+19 3.85E+19 4.01E+19 7.13E+19 1.02E+20 -
+2.5 3.66E+19 3.82E+19 3.99E+19 4.16E+19 7.45E+19 1.07E+20
+1.5 3.72E+19 3.87E+19 4.04E+19 4.20E+19 7.33E+19 1.04E+20 I
+0.5 3.75E+19 3.91E+19 4.06E+19 4.20E+19 7.04E+19 9.81E+19 j 0.0 3.76E+19 3.92E+19 4.06E+19 4.19E+19. 6.81E+19 9.36E+19 f
. -0.5 3.76E+19 3.93E+19 4.05E+19 4.17E+19 6.56E+]9 8.90E+19 i
-1.5 3.76E+19 3.92E+19 4.04E+19 4.16E+19 6.48E+19 8.75E+19 !
.. -2.5 3.73E+19 3.88E+19 4.01E+19 4.14E+19 6.66E+19 9.12E+19 !
-3.5 3.65E+19 3.79E+1S 3.93E+19 4.07E+19 6.79E+19 9.45E+19
-4.5 3.36E+19 3.dSE+19 3.63E+19 3.77E+19 6.43E+19 9.03E+19 l
-5.5 2.35E+19 2 44E+19 2.53E+19 2.63E+19 4.47E+19 6.26E+19 l
-6.0 1.15E+19 1.21E+19 1.27E+19 1.32E+19 2.44E+19 3.53E+19 I f
Note: Height is provided relative to the axial midplane of the active core. i l
~
, f I
l 8-9
)
i
TABLE 8.1-6
SUMMARY
OF BEST ESTIMATE FAST NEUTRON (E > 0.1 MeV) EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF THE POINT 8EACH UNIT 2 REACTOR PRESSURE VESSEL - 15 DEGREE AZIMUTHAL ANGLE i HEIGHT
+ (E > 0.1 MeV) [n/cm2] i fft) E0C 14 E0C 15 E0C 16 E0C 17 32 EFPY- 48 EFPY
+6.0 7.21E+18 7.76E+18 8.20E+18 8.65E+18 1.72E+19 2.56E+19 -
+5.5 1.41E+19 1.48E+19 1.54E+19 1.61E+19 2.84E+19 4.05E+19
+4.5 2.05E+19 2.17E+19 2.27E+19 2.36E+19 4.24E+19 6.07E+19
- i
+3.5- 2.31E+19 2.43E+19 2.55E+19 2.65E+19 4.75E+19 6.80E+19 !
+2.5 2.39E+19 2.52E+19 2.65E+19 2.75E+19 4.97E+19 7.14E+19 I
+1.5 2.43E+19 2.56E+19 2.68E+19 2.78E+19 4.87E+19 6.92E+19-
+0.5 2.45E+19 2.58E+19 2.68E+19 2.78E+19 4.74E+19 6.65E+19 .
0.0 t 2.46E+19 2.59E+19 2.69E+19 2.79E+19 4.66E+19 6.49E+19 !
-0.5 2.46E+19 2.60E+19 2.69E+19 2.78E+19 4.56E+19 6.30E+19 - '
-1.5 2.45E+19 2.58E+19 2.67E+19 2.76E+19 4.50E+19 6.21E+19
-2.5 2.43E+19 2.43E+19 2.52E+19- 2.62E+19 4.42E+19 6.19E+19 ',
-3.5 2.38E+19 2.50E+19 2.61E+19 2.71E+19 4.68E+19 6.61E+19
-4.5 2.19E+19 2.31E+19 2.41E+19 2.50E+19 4.42E+19 6.29E+19 <
-5.5 1.53E+19 1.61E+19 1.67E+19 1.74E+19 2.99E+19 4.22E+19
-6.0 7.53E+18 8.06E+18 8.49E+18 8.85E+18 1.65E+19 2.39E+19 Note: Height is provided relative to the axial t j
midplane of the active core. [ t t I L 8-10 t
p TABLE 8.1-7
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 - 30 DEGREE AZIMUTHAL ANGLE HEIGHT 4 (E > 0.1 MeV) [n/cm2] (ft) E0C 14 E0C 15 EOC 16 E0C 17 32 EFPY 48 EFPY
. +6.0 5.09E+18 5.45E+18 5.74E+18 6.05E+18 1.19E+19 1.76E+I9
+5.5 9.83E+18 1.03E+19 1.08E+19 1.12E+19 1.96E+19 2.79E+19
- +4.5 1.44E+19 1.52E+19 1.58E+19 1.65E+13 2.90E+19 4.12E+19
+3.5 1.61E+19 1.69E+19 1.76E+19 1.84E+19 3.22E+19 4.56E+19
+2.5 1.67E+19 1.76E+19 1.83E+19 1.91E+19 3.35E+19 4.76E+19
+1.5 1.69E+19 1.78E+19 1.85E+19 1.93E+19 3.36E+19 4.77E+19
+0.5 1.71E+19 1.80E+19 1.87E+19 1.95E+19 3.39E+19 4.81E+19 0.0 1.72E+19 1.81E+19 1.88E+19 1.96E+19 3.42E+19 4.84E+19
- -0.5 1.72E+19 1.81E+19 1.88E+19 1.96E+19 3.43E+19 4.86E+19
-1.5 1.71E+19 1.80E+19 1.87E+19 J.94E+19 3.33E+19 4.70E+19
* -2.5 1.70E+19- 1.79E+19 1.86E+19 1.93E+19 3.35E+19 4.74E+19
-3.5 1.66E+19 1.74E+19 1.81E+19 1.88E+19 3.20E+19 4.49E+19
-4.5 1.53E+19 1.61E+19 1.67E+19 1.74E+19 2.98E+19 4.19E+19
-5.5 1.08E+19 1.13E+19 1.17E+19 1.21E+19 2.01E+19 2.79E+19
-6.0 5.32E+18 5.64E+18 5.90E+18 6.15E+18 1.10E+19 1.57E+19 Note: Height is provided relative to the axial midplane of the active core.
9 4 e 8-11
]
=
TA8LE 8.1-8 - t
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 - 45 DEGREE AZIMUTHAL ANGLE .- HEIGHT 4 (E > 0.1 MeV) [n/cm2] (ft) E0C 14 EOC 15 EOC 16 EOC 17 32 EFPY 48 EFPY
+6.0 4.33E+18 4.64E+18 4.91E+18 5.16E+18 1.02E+19 1.51E+19
+5.5 8.40E+18 8.82E+18 9.22E+18 9.57E+18 1.69E+19 2.40E+19
+4.5 1.23E+19 1.29E+19 1.35E+19 1.41E+19 2.55E+19 3.67E+19
+3.5 1.38E+19 1.45E+19 1.51E+19 1.58E+19 2.83E+19 4.05E+19
+2.5 1.42E+19 1.49E+19 1.57E+19 1.64E+19 3.03E+19 4.39E+19
+1.5 1.44E+19 1.52E+19 1.59E+19 1.66E+19 2.99E+19 4.30E+19
+0.5 1.45E+19 1.53E+19 1.60E+19 1.66E+19 2.97E+19 4.24E+19 0.0 1.46E+19 1.54E+19 1.61E+19 1.68E+19 3.03E+19 4.34E+19
-0.5 1.46E+19 1.54E+19 1.61E+19 1.69E+19 3.08E+19 4.43E+19
-1.5 1.45E+19 1.53E+19 1.60E+19 1.67E+19 3.02E+19 4.33E+19
-2.5 1.44E+19 1.52E+19 1.59E+19 1.66E+19 3.01E+19 4.33E+19
-3.5 1.42E+19 1.48E+19 1.55E+19 1.62E+19 2.90E+19 4.16E+19
-4.5 1.31E+19 1.37E+19 1.44E+19 1.19E+19 2.68E+19 3.84E+19
-5.5 9.17E+18 9.58E+18 1.00E+19 1.05E+19 1.74E+19 2.43E+19
-6.0 4.53E+18 4.81E+18 5.09E+18 5.26E618 9.58E+18 1.38E+19 Note: Height is provided relative to the .*xial midplane of the active core.
e 8-12 m___
. _ .~. _ .. _ . __
TABLE 8.1-9
SUMMARY
OF BEST ESTIMATE IRON ATOM DISPLACEMENT
. PROJECTIONS FOR THE BELTLINE REGION OF THE POINT BEACH UNIT 2 l REACTOR PRESSURE VESSEL - 0 DEGREE AZIMUTHAL ANGLE l
l HEIGHT IRON ATOM DISPLACEMENTS [dpa) l (ft) EOC 14 EOC 15 EOC 16 EOC 17 32 EFPY 48 EFPY [
. +6.0 6.67E-03 7.09E-03 7.52E-03 7.98E-03 1.66E-02 2.50E-02
+5.5 1.30E-02 1.36E-02 1.43E-02 1.49E-02' 2.75E-02 3.98E-02
.- +4.5 1.90E-02 1.99E-02 2.09E-02 2.19E-02 4.12E-02 6.00E-02 l
+3.5 2.14E-02 2.24E-02 2.35E-02 2.45E-02 4.55E-02 6.60E-02 .
+2.5 2.21E-02 2.31E-02 2.43E-02 2.54E-02 4.76E-02 6.93E-02 l
+1.5 2.24E-02 2.35E-02 2.46E-02 2.57E-02 4.68E-02 6.73E l
+0.5 2.26E-02 2.37E-02 2.47E-02 2.57E-02 4.48E-02 6.34E-02 l 0.0 2.27E-02 2.38E-02 2.47E-02 2.56E-02 4.32E-02 6.04E-02 l
- -0.5 2.27E-02 2.38E-02 2.47E-02 2.55E-02 4.15E-02 5.72E-02 .{
-1.5 2.27E-02. 2.38E-02 2.46E-02. 2.54E-02 4.10E-02 5.63E-02 .
-2.5 2.25E-02 2.36E-02 2.45E-02 2.53E-02 4.23E-02 5.89E-02
-3.5 2.21E-02 2.30E-02 2.40E-02 2.49E-02 4.31E-02 6.10E-02 i
-4.5 2.03E-02 2.12E-02 2.21E-02 2.30E-02 4.09E-02 5.84E-02
-5.5 1.42E-02 1.48E-02 1.54E-02 1.61E-02 2.84E-02 4.05E-02
-6.0 6.94E-03 7.34E-03 7.74E-03 8.llE-03 1.56E-02 2.30E-02 :
l Note: Height is provided relative to the axial l midplane of the active core.
\
l
. j i
4 t I c. 8-13 i
-f f
TABLE 8.1-10 t
SUMMARY
OF BEST ESTIMATE IRON ATOM DISPLACEMENT 1 PROJECTIONS FOR THE BELTLINE REGION OF THE POINT-BEACH UNIT 2 j REACTOR PRESSURE VESSEL - 15 DEGREE AZIMUTHAL ANGLE ! HEIGHT IRON ATOM DISPLACEMENTS [dpa] ! (fti EOC-14 EOC 15 EOC 16 EOC 17 32 EFPY 48 EFPY
+6.0 4.20E-03 4.49E-03 4.74E-03 5.00E-03 .9.86E-03 1.46E-02 -
+5.5 8.18E-03 8.60E-03 8.98E-03 9.33E-03 1.64E-02 2.34E-02
+4.5 1.20E-02 1.26E-02 1.32E-02 1.37E-02 2.45E-02 3.50E-02 -
43.5 1.34E-02 1.41E-02 1.48E-02 1.54E-02 -2.75E-02 3.43E-02 ;
+2.5 1.39E-02 1.47E-02 1.54E-02 1.60E-02 2.88E-02 4.13E-02 >
+1.5 1.41E-02 1.49E-02 1.55E-02 1.61E-02 2.81E-02 3.99E-02
+0.5 1.42E-02 1.49E-02 1.56E-02 1.61E-02. 2.74E-02 3.85E-02 0.0 1.43E-02 1.51E-02 1.56E-02 1.62E-02 2.70E-02 .3.75E-02 !
-0.5 1.43E-02 1.51E-02 1.56E-02 1.61E-02 2.64E-02 3.64E-02 -l 4E 4 E- 1 .6 0 3 E -.
-3.5 1.38E-02 1.45E-02 1.51E-02 1.57E-02 2.70E-02 3.81E-02 l
-4.5 1.27E-02 1.34E-02 1.40E-02 1.45E-02 2.56E-02 3.63E-02
-5.5 8.94E-03 9.35E-03 9.74E-03 1.01E-02 1.73E-02 2.44E-02
-6.0 4.38E-03 4.66E-03 4.90E-03 5.11E-03 9.50E-03 1.38E-02 Note: Height is provided relative to the axial midplane of the active core. l e
8-14 1
l TABLE 8.1-11.- , i
SUMMARY
OF BEST ESTIMATE IRON ATOM DISPLACEMENT PROJECTIONS FOR THE BELTLINE REGION OF THE POINT BEACH UNIT 2 ! REACTOR PRESSURE VESSEL --30 DEGREE AZIMUTHAL ANGLE ! l HEIGHT IRON ATOM DISPLACEMENTS [dpa) . l (ft) EOC 14 EOC 15 EOC 16 EOC 17 32 EFPY 48 EFPY
". ' +6.0 3.12E-03 3.32E-03 3.50E-03 3.69E-03 7.18E-03 1.06E-02 1
+5.5 6.05E-03 6.33E-03 6.61E-03 6.86E-03 1.20E-02 1.70E-02 l
+4.5 8.83E-03 9.26E-03 9.66E-03 1.00E-02 1.76E-02 2.50E-02 l t
+3.5 9.90E-03 1.04E-02 1.08E-02 1.12E-02 1.96E-02 2.77E-02 l
+2.5 1.02E-02 1.07E-02 1.12E-02 1.16E-02 2.03E-02 2.89E-02 !
+1.5 1.04E-02 1.09E-02 1.13E-02 1.18E-02 2.05E-02 2.90E-02
+0.5 1.05E-02 1.10E-02 1.14E-02 1.19E-02 2.07E-02 2.92E-02 0.0 1.05E-02 1.10E-02 1.15E-02 1.19E-02 2.07E-02 2.93E-02 ,
'- -0.5 1.05E-02 1.10E-02 1.15E-02 1.19E-02 2.08E-02 2.94E-02 l
-1.5 1.05E-02 1.10E-02 1.14E-02 1.19E-02 2.03E-02 2.85E j
-2.5 1.04E-02 1.09E-02 1.13E-02 1.18E-02 2.04E-02 2.88E-02 l
-3.5 1.02E-02 1.07E-02 1.11E-02 1.15E-02 1.95E-02 2.73E-02 ?
-4.5 9.40E-03 9.84E-03 1.02E-02 1.06E-02 1.81E-02 2.55E-02 l
-5.5 6.60E-03 6.88E-03 7.13E-03 7.39E-03 1.22E-02 1.70E-02 i
-6.0 3.25E-03 3.45E-03 3.60E-03 3.75E-03 6.69E-03 9.56E-03 i
Note: Height is provided relative to the axial j, midplane of the active core. j l l 1 l 8-15 l
TA8LE 8.1-12
SUMMARY
OF 8EST ESTIMATE IRON ATOM DISPLACEMENT PROJECTIONS FOR THE 3ELTLINE REGION OF THE POINT 8EACH UNIT 2 REACT 0F'P." ESSURE VESSEL - 45 DEGREE AZIMUTHAL ANGLE-
~
HEIGHT IRON ATOM DISPLACEMENTS [dpa] (ft) EOC 14 EOC 15 EOC 16 E0C 17 32 EFPY 48 EFPY
~ +6.0 2.72E-03 2.90E-03 3.07E-03 3'23E-03 6.34E-03 9.38E-03 +5.5 5.28E-03 5.53E-03 5.79E-03 6.00E-03 1.06E-02 1.50E-02 +4.5 7.71E-03 8.09E-03 8.49E-03 8.83E-03 1.60E-02 2.30E-02 - +3.5 8.64E-03 9.05E-03 9.48E-03 9.86E-03 1.77E-02 2.53E-02 +2.5 8.92E-03 9.38E-03 9.85E-03 1.03E-02 1.90E-02 2.75E-02 +1.5 9.07E-03 9.52E-03 9.97E-03 1.04E-02 1.87E-02 2.69E-02' +0.5 9.16E-03 9.59E-03 1.01E-02 1.04E-02 1.86E-02 2.66E-02 0.0 9.18E-03 9.64E-03 1.01E-02 1.05E-02 1.89E-02 2.72E-02 -0.5 9.18E-03 9.65E-03 1.01E-02 1.05E-02 1.92E-02 2.77E-02 - -1.5 9.17E-03 9.62E-03 1.01E-02 1.05E-02 1.89E-02 2.72E-02 -2.5 9.10E-03 9.53E-03 9.98E-03 1.04E-02 1.88E-02 2.71E-02 - -3.5 8.91E-03 9.32E-03 9.76E-03 1.02E-02 1.82E-02 2.60E-02 -4.5 8.21E-03 8.59E-03 9.01E-03 9.36E-03 1.68E-02 2.40E-02 -5.5 5.76E-03 6.01E-03 6.29E-03 6.47E-03 1.09E-02 1.52E-02 -6.0 2.85E-03 3.02E-03 3.19E-03 3.30E-03 6.01E-03 8.65E-03 Note: Height is provided realative to the axial midplane of the active core.
4 8-16
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), and a circumferential weld (SA-1484) joining the two ring forgings. The circumferential weld is centered 15.06 inches below the axial 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. Also considered part of the beltline region is the circumferential weld located at the top of the intermediate shell forging 8.44 inches above the active fuel. The corresponding circumferential weld at the bottom of the lower shell forging 39.87 inches below the active fuel experiences neutron radiation levels low enough that the lifetime exposure of the weld will 2
. remain below 1.0E+17 n/cm . Therefore, this weldment is not considered to be part of the beltline region. 8.2-1 Circumferential Weld (SA-1484) The current (End of Cycle 17) 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 expressed in terms of
. t (E > 1.0 MeV), t (E > 0.1 MeV), and dpa.
In developing the exposure profiles for the circumferential weld, it is noted that, although the flux reduction afforded by the Cycle 16-17 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 at the 0 degree azimuth throughout the service life of the unit. However, the magnitude of the projected 8-17
I exposures are significantly lower than would be the case had the flux eduction measures not been implemented. t 8.2-2 Intermediate Shell Forging (123V500) The current and projected maximum exposures of the intermediate shell - forging are given in Table 8.2-2. Again, all three 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 redution in exposure rates in the vicinity of the circumferential weld, the axial location of the maximum exposure at . 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 point 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 . l feet below the active core midplane. However, the absorbers have a j 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. - 8-18 i
i 8.2-4 Upper / Intermediate Shell Circumferential Weld
- The current and and projected exposures for the lower shell forging are ! . listed in Table 8.2-4. As in the case of the intermediate forging, all ;
three exposure parameters are tabulated. Due to its location above the ! top of the reactor core, the inclusion of the part length hafnium absorbers in the fuel cycle design has a negligible impact on the fast I
+
neutron exposure experienced by this weld. ' 0 7 9 i t i l i [ O 6 0 8-19 c
l TABLE 8.2-1 l MAXIMUM FAST NEUTRON EXPOSURE OF-POINT BEACH UNIT'2 8ELTLINE CIRCUMFERENTIAL WELD (SA-1484) *l l 4 (E > 1.0 MeV) In/cm21 AZIMUTHAL EOC 17 .! ANGLE 15.6 EFPY 32.0 EFPY 48.0 EFPY ;
~ i l -0 DEGREES 1.61E+19 2.52E+19 3.39E+19 -
15 DEGREES 9.99E+18 1.65E+19 2.28E+19 30 DEGREES 7.60E+18 1.32E+19 1.87E+19 - 45 DEGREES 6.80E+18 1.25E+19 1.80E+19 1 4 (E > 0.1 MeV) In/cm21 l AZIMUTHAL E0C 17 i ANGLE 15.6 EFPY 32.0 EFPY 48.0 EFPY . O DEGREES 4.16E+19 6.50E+19' 8.79E+19 l 15 DEGREES 2.77E+19 4.52E+19 6.23E+19 I 30 DEGREES 1.95E+19 3.36E+19 4.74E+19 l 45 DEGREES 1.68E+19 3.04E+19 4.36E+19 ! I s IRON ATOM DISPLACEMENTS Idoal l AZIMUTHAL E0C 17 ! ANGLE 15.6 EFPY 32.0 EFPY 48.0 EFPY-0 DEGREES 2.54E-02 4.11E-02 5.65E-02 - 15 DEGREES 1.61E-02 2.62E-02 3.60E-02 30 DEGREES 1.19E-02 2.04E-02 2.87E-02 45 DEGREES 1.05E-02 1.90E-02 2.73E-02 i J 8-20 ;
)
TABLE 8.2-2 MAXIMUM FAST NEUTRON EXPOSURE OF POINT BEACH UNIT 2 ; INTERMEDIATE SHELL FORGING (123V500) i 4 (E > 1.0 MeV) In/cm21
- AZIMUTHAL E0C 11 ANGLE 15.6 EFPY 32.0 EFPY 48.0 EFPY
. O DEGREES 1.63E+19 2.88E+19 4.12E+19 i 15 DEGREES 1.01E+19 1.82E+19 2.62E+19 ,
. 30 DEGREES 7.64E+18 1.34E+19 1.91E+19 l 45 DEGREES 6.83E+18 1.27E+19 1.83E+19 4 (E > 0.1 MeV) In/cm21 AZIMUTHAL E0C 17
. ANGLE 15.6 EFPY 32.0 EFPY 48.0 EFPY 0 DEGREES 4.20E+19 7.45E+19 1.07E+20 . 15 DEGREES 2.79E+19 4.97E+19 7.14E+19 30 DEGREES 1.96E+19 3.43E+19 4.86E+19 45 DEGREES 1.69E+19 3.08E+19 4.43E+19 IRON ATOM DISPLACEMENTS Idoal ! AZIMUTHAL E0C 17 ' ANGLE 15.6 EFPY 32.0 EFPY 48.0 EFPY
, O DEGREES 2.57E-02 4.76E-02 6.93E-02 15 DEGREES 1.62E-02 2.88E-02 4.13E-02 30 DEGREES 1.19E-02 2.08E-02 2.94E-02 45 DEGREES 1.05E-02 1.92E-02 2.77E-02 s
8-21
P TABLE 8.2-3 MAXIMUM FAST NEUTRON EXPOSURE OF POINT BEACH UNIT 2 I LOWER SHELL FORGING (122W195) 4 (E > 1.0 MeV) In/cm21 l AZIMUTHAL E0C 17 ANGLE 15.6 EFPY 32.0 EFPY 48.0 EFPY I
~
O DEGREES 1.61E+19 2.62E+19 3.64E+19 - 15 DEGREES 9.99E+18 1.71E+19 2.42E+19 I 30 DEGREES 7.60E+18 1.32E+19 1.87E+19 - 45 DEGREES 6.80E+18 -1.25E+19 1.80E+19
+ (E > 0.1 MeV) In/cm21 AZIMUTHAL EOC 17 ANGLE 15.6 EFPY 32.0 EFPY 48.0 EFPY -
0 DEGREES 4.16E+19 6.79E+19 9.45E+19 15 DEGREES 2.77E+19 4.68E+19 6.61E+19 - 30 DEGREES 1.95E+19 3.36E+19 4.74E+19 i 45 DEGREES 1.68E+19 3.04E+19 4.36E+19 IRON ATOM DISPLACEMENTS Idoal AZIMUTHAL E0C 17 ANGLE 15.6 EFPY 32.0 EFPY 48.0 EFPY 0 DEGREES 2.54E-02 4.31E-02 6.10E-02 - t 15 DEGREES 1.61E-02 2.70E-02 3.81E-02
~
30 DEGREES 1.19E-02 2.04E-02 2.88E-02 ) 45 DEGREES 1.05E-02 1.90E-02 2.73E-02 ; i 8-22 F t
E F . TABLE 8.2-4 MAXIMUM FAST NEUTRON EXPOSURE OF POINT 8EACH UNIT 2 UPPER / INTERMEDIATE SHELL CIRCUMFERENTIAL WELD 4 (E > 1.0 MeV) In/cm21 AZIMUTHAL E0C 17 ANGLE 15.6 EFPY 32.0 EFPY 48.0 EFPY 0 DEGREES 2.27E+18 3.70E+18 5.00E+18 15 DEGREES 1.51E+18 2.33E+18 3.15E+18 30 DEGREES 1.07E+18 1.73E+18 2.33E+18 45 DEGREES 9.40E+17 1.56E+18 2.10E+18 4 (E > 0.1 MeV) In/cm21 AZIMUTHAL E0C 17
- ANGLE 15.6 EFPY 32.0 EFPY 48.0 EFPY 0 DEGREES 6.17E+18 1.01E+19 1.36E+19 15 DEGREES 4.39E+18 6.78E+18 9.17E+18 30 DEGREES 2.89E+18 4.67E+18 6.29E+18 45 DEGREES 2.44E+18 4.06E+18 5.46E+18 IRON ATOM DISPLACEMENTS Idoal AZIMUTHAL E0C 17 ANGLE 15.6 EFPY 32.0 EFPY 48.0 EFPY . O DEGREES 3.75E-03 6.11E-03 8.25E-03 15 DEGREES 2.58E-03 3.98E-03 5.39E-03 30 DEGREES 1.78E-03 2.87E-03 3.87E-03 ;
45 DEGREES 1.54E-03 2.56E-03 3.44E-03 <, f O 8-23
FIGURE 8.2-1 FA3T NEUTRON FLUENCE (E > 1.0 Mev) AS A FUNCTION OF AZIMUTHAL
~
ANGLE AT THE INNER RADIUS OF THE BELTLINE CIRCUMFERENTIAL WELD i Neutron Fluence (n/cm2)
- 1.OOOE + 20 E
I i O
- a.
+
1.OOOE + 19 _
+
1.000E + 18 - 0 10 20 30 40 50 Azimuthal Angle (Deg)
~^
+ 15.6 EFPY $ 32.0 EFPY O 48.0 EFPY I
8-24 i I
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 Neutron Fluence (n/cm2) ; [ _ i 3 -
~
O : n
+ ,
1.000E + 19 _ l 1 l l 1.000E + 18 l 1 0 10 20 30 40 50
. Azimuthal Angle (Deg) . + 15.6 EFPY $ 32.0 EFPY O 48.0 EFPY j l
8-25 ; l
-er- -
--wnn-.> , ,
FIGURE 8.2-3 IRON ATOM DISPLACEMENTS [dpa) AS A FUNCTION OF AZIMUTHAL ANGLE AT THE INNER RADIUS OF THE BELTLINE CIRCUMFERENTIAL WELD - Displacements (dpa) C
- i 9
~
- D D 1
7 g 0.01 _ 0.001 ' ' O 10 20 30 40 50 - Azimuthal Angle (Deg) 8
+ 15.6 EFPY $ 32.0 EFPY O 48.0 EFPY .
e 8-26 [
l 8.3 Uncertainties in Exposure Projections The overall uncertainties associated with the exposure rates and integrated exposures determined for Point Beach Unit 2 stem from two basic. ; sources; the accuracy of the neutron flux measurements at the sensor set l locations and the accuracy of the radial gradient projections derived from ; the use of the transport code. Based on the least squares adjustment approach used in the FERRET analyses the 1 sigma uncertainties in the measured data were as follows: Jp UNCERTAINTY CAPSULE CAVITY Flux (E > 1.0 MeV) 6% 6% Flux (E > 0.1 MeV) 13% 15% dpa/sec 9% 12% j These values represent uncertainties derived from the reaction rate , measurements and from the least squares fit of the output spectrum to the measured data. As additional data is obtained from the ongoing measurement program, the knowledge of the neutron spectra at the measurement locations will increase and the uncertainties in the measured ' exposure parameters will be reduced somewhat.
- Since the ultimate goal of the cavity measurement program is the evaluation of the exposure of the vessel itself, an additional uncertainty
. associated with the ability to translate results from the measurement locations to the points of interest within the vessel must be included along with the measurement uncertainties listed above. Information -
pertinent to this extrapolation uncertainty has been obtained from benchmarking studies using the Westinghouse neutron transport methodology ; and from several comparisons of power reactor internal surveillance L capsule dosimetry and reactor cavity dosimetry for which the irradiation history of all sensors was the same. i l
~
8-27
I Based on these benchmarking evaluations the uncertainty or bias associated l with the calculated slope through the steel vessel was estimated to be i approximately 5% for all exposure parameters. Thus, the total uncertainty associated with projections at the clad / base metal interface is estimated ~ to be as follows for each exposure parameter of interest. la UNCERTAINTY VESSEL IR Flux (E > 1.0 MeV) 11% . Flux (E > 0.1 MeV) 20% dpa/sec 17% - Use of these values represents the bounding lo uncertainties for vessel exposure, since with penetration into the vessel wall the extrapolation uncertainty lessens until at the outer surface the overall uncertainty reverts simply to the measurement uncertainty. Again, as more , data are accumulated from both reactor cavity and surveillance capsule- - dosimetry sets, the extrapolation uncertainty will also be reduced resulting in higher levels of accuracy in the vessel exposure projections. ' i l e e 4
. l 8-28
SECTION 9.0 ,F REFERENCES
- 1. Anderson, S. L. and Fero, A. H., " Reactor Cavity Neutron Measurement Program for Wisconsin Electric Power Company Point Beach Nuclear Plant Unit I and Unit 2," WCAP-11944, Rev 1, June 1989.
. 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. Soltesz,R. G. , et. al., " Nuclear Rocket Shielding Methods, Modification, Updating, and Input Data Preparation - Volume 5 - Two Dimensional Discrete Ordinates Transport Technique," WANL-PR-(LL)-034, August 1970.
. 4. SAILOR RSIC DATA LIBRARY COLLECTION DLC-76, " Coupled Self-Shielded, 47 Neutron, 20 Gamma Ray, P3, Cross Section Library for Light Water Reactors.
- 5. ASTM Designation E706-87, " Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1989
- 6. 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.,1989.
. 7. ASTM Designation E261-77, " 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.
1989. 0 9-1
- 8. ASTM Designation E262-86, " Standard Method for Measuring Thermal Neutron Flux by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.1989.
- 9. 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.,1989.
~
- 10. ASTM Designation E264-87, " 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.,1989.
- 11. ASTM Designation E481-86, " 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.,
1989.
- 12. ASTM Designation E523-87, " Standard Method for Determining Fast Neutroo -
Flux Density by Radioactivation of Copper," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1989. -
- 13. ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by l Radioactivation of Uranium-238," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1989.
- 14. ASTM Designation E705-84, " 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., .
1989.
- 15. ASTM Designation E1005-84, " 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.,1989. ~: 9-2
I l
- 16. Schmittroth, E. A., " FERRET Data Analysis Code", HEDL-TME-79-40, Hanford Engineering Development Laboratory, Richland, Washington, September 1979.. l I
- 17. 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 Laboratory, Kirkland AFB, NM, July 1967. j e a d Unc r i P c of e4 M P S p Reactor Dosimetry," NUREG/CP-0029, NRC, Washington, D.C., July 1982.
+
- 19. Scherpereel, L. R., " Core Physics Characteristics of the Point Beacn Nuclear Plant - Unit 1, Cycle 1," WCAP-7430, December 1969. [ proprietary]
- 20. Hawrylak, J. P., " Revised Cycle 2 Nuclear Design Characteristics for Point Beach Unit 2," WCAP-8418, Rev 1, November,1974. [ proprietary]
l t i
. 21. Hawrylak, J. P., et. al., "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 3," WCAP-8759, March 1976.
[ proprietary]
- 22. Hawrylak, J. P., et. al., "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 4," WCAP-8934, February 1977. j
[ proprietary] l
- 23. 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] j
- 24. Pilzer, E. H., et. al., "The Nuclear Design - Core Management of the l
Point Beach Unit 2 Nuclear Reactor - Cycle 6," WCAP-9493, April 1979. j [ proprietary] !
- 25. Scherder, W. J., et. al., "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 7," WCAP-9667, February 1980.
t* [ proprietary] g.3 i l i
,. ,__ _. - --- +--- ., ,, , ,_..,. - -
- 26. Smith R. T., et. al., "The Nuclear Design - Core Management of the Point 8each Unit 2 Nuclear Reactor - Cycle 8," WCAP-9846, March 1981. ;
[ proprietary] '
- 27. 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)
- 28. Smith R. T., et. al., "The Nuclear Design - Core Management < ,f the
~
Point Beach Unit 2 Nuclear Reactor - Cycle 10," E'W-10278, Itarch 1983. - [ proprietary). I
?
- 29. Smith R. T., "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 11," WCAP-10583, August 1984.
[ proprietary).
- 30. Smith R. T., "The Nuclear Design - Core Management of the Point Beach
[ Unit 2 Nuclear Reactor - Cycle 12 Rev 1," WCAP-10897, November 1985. 'l [ proprietary). i
~i i
- 31. Smith R. T., "The Nuclear Design - Core Management of the Point Beach !
Unit 2 Nuclear Reactor - Cycle 13 ," WCAP-ll288, November 1986. [ proprietary). l
- 32. Smith R. T., "The Nuclear Design - Core Management of the Point Beach Unit 2 Nuclear Reactor - Cycle 14 ," WCAP-ll571, September 1987.
[ proprietary). I
. I
- 33. Smith R. T., "The Nuclear Design - Core Management of the Point Beach !
Unit 2 Nuclear Reactor - Cycle 15 ," WCAP-Il903, September 1988. ; [ proprietary). ' l
- 34. Smith R. T., "The Nuclear Design - Core Management of the Point Beach
~
Unit 2 Nuclear Reactor - Cycle 16," WCAP-12362, September 1989. [ proprietary). l
'l 9-4
i
- 35. Smith R. T., "The Nuclear Design - Core Management of the Point Beach l Unit 2 Nuclear Reactor - Cycle 17," WCAP-12683, September 1990.
[ proprietary].
- i
- 36. 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. l t
- 37. Yanichko, S. E., et. al., " Analysis of Capsule R from the Wisconsin ,
' i
- Electric Power Company Point Beach Nuclear Plant Unit No. 2 Reactor :
Vessel Radiation Surveillance Program," WCAP-9635, December 1979. t
- 38. Capsule S Report to be determined.
- 39. Perrin, J. S., et. al., " Point Beach Nuclear Plant Unit No. 2 Pressure j Vessel Surveillance Program: Evaluation of Capsule V," Battelle Memorial i Institute Report, June 1975. ,
- 40. Anderson, S. L., et. al., " Adjoint Flux Program for Point Beach Units ;
1 AND 2," WCAP-10638, December 1984. '! i I i k
. i i > b 9-5 I
J b4 +.m44 J d m.4 Wi ne t v Aan , a w 8.dA-5AAm .6 4 5 as 2 4 .a..A- +.& 4-f 4 4e -.-ums+ + -.%~.44=- A-+m J.s. 6- 4 a P m 4 4 . mA & b i o i t
..i i
5 I f 5
'9%
9 l F
}
t t t 5 t O
- l 4
s 9
=
^ b I
i
)
+
-W 5
I e 6 i
.h r
I L i t. L 5 1 s. i e a t L P e g* b r
- I i
-I I
W i i t t I r h t b
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. The irradiation history of capsules withdrawn to date was as follows: CYCLE NO. STARTUP SHUTDOWN -COMMENT I 05/30/72 10/17/74 CAPSULE V WITHDRAWN 2 12/20/74 02/26/76 3 03/26/76 03/04/77 CAPSOLE T WITHDRAWN 4 04/19/77 03/22/78 5 04/17/78 03/23/79 CAPSULE R WITHDRAWN 6 04/19/79 04/11/80 7 05/13/80 04/17/81 8 05/21/81 04/16/82 9 05/26/82 03/25/83 10 07/05/83 09/28/84 11 11/19/84 10/05/85 12 11/24/85 09/27/86 13 12/01/86 10/03/87 14 11/17/87 10/08/88
. 15 11/21/88 09/23/89 16 11/24/89 10/06/90 CAPSULE S WITHDRAWN REF. CORE POWER - 1518 MWt The monthly thermal generation applicable to the Point Beach Unit 2 reactor is 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.
A-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 (W-hr) M (W-hr) M (W-hr) M _{W-hr) 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 885000 7/76 1026604 8/78 1054425 9/80 1047595 .' 6/74 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 12/74 302016 1/77 1102848 2/79 1010047 3/81 1112552 1/75 1113456 2/77 1001354 3/79 746785 4/81 559549 2/75 881295 3/77 100292 4/79 585747 5/81 186873 . 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 6/75 914234 7/77 1072410 8/79 1114720 9/81 988920 7/75 1063799 8/77 973371 9/79 1032935 10/81 1088553 ; 1080909 I 8/75 748416 9/77 1039145 10/79 1117434 11/81 9/75 997380 10/77 1109781 11/79 1045777 12/81 1073535 10/75 989176 11/77 1062668 12/79 1095273 1/82 1106250 11/75 974925 12/77 1106636 1/80 1111502 2/82 1005969 1 12/75 1114475 1/78 1066993 2/80 974189 3/82 1091347 . 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 9 b A-2
MONTHLY THERMAL GENERATION DURING THE FIRST SIXTEEN FUEL CYCLES OF THE POINT 8EACH UNIT 2 REACTOR THERMAL THERMAL THERMAL THERMAL GENERATION GENERATION GENERATION GENERATION 19QHIH (MW-hr) !!QtNITH (MW-hr) tELJTH (MW-hr) fiQNJE (MW-hr) 7/82 1117700 8/84 1126993 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 1066056 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 7/83 883944 8/85 1125504 9/87 1092960 10/89 0 8/83 1075582 9/85 1090774 10/87 70587 11/89 141690 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 3/88 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 S 9 A-3
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A-6
Westinghouse Electric Corporation RapGir Advanced Energ Systems - Analytical Imboratory Malts Mill site Degaeste 14340
'to: S.L. Anderson Energ Center - East (478)
E.hrek Cnergy Center - East (470) Aeosived: 5/1/91
- Reported 6/5/91 IAEEULTB CF Al&INSIS)
Dosiestry: Point Beach Unit #2 Capsule S Originator Lab. Dosiseter (May 20,1991) ID 8e$e6 Meterial !belide 4 sang
- 2 signa FIastat IE3rnt31s U-238 91-1035 U-234 Ca-137 3p-237 1.373+03 +/- 2.3t+01 91-1036 3p-237 Co-137 8.55e+03 +/- 1.23-02 m WIRES Co/A1(Cd) 91-1037 Co/A1 Co-60 1.93e+04 +/- 2.4E+02 Co/A1 91-1038 CW Co 40 4.13E+04 +/- 3.4E+02 Cu 91-1039 Cu Co-60
~ 1.83D*02 +/- 2.2E+00 W MID WIRES Co/A1(Cd) 91-1040 Co/A1 Co 40 1.868+C4 +/- 2.3e+02 Co/A1 91-1041 CWA 1 Co40 3,399+04 +/- 3.2E+02 Cu 91-104? Cu Co-60 1.539+02 +/- 2.12+00 MID WIRES Ce/A1(Cd) 91-1043 Co/A1 Co-60 1.73t+04 +/- 2.2r+02 Co/A1 91-1044 cvAl Co40 3.30B+04 +/- 2.se+02 Ni 91-1045 Mi co-58 2.95E+03 +/- 3.88+01 por MID wInss Co/A1(Cd) 91-1046 Ca/A1 Co40 2.12n+04 +/- 2.2a+02 Co/A1 91-1047 Co/A1 Co 40 4.43E+04 + 3.3t*02 Cu 91-1048 Cu Co40 1.60e+02 - 2.1B+00 30T WIRES Co/A1(Cd) 91-1049 Co/A1 0o-60 1.90D+04 +/- 2.1r+02 CW 91-1050 Co/A1 Co40 4.12B+04 */- 3.2E+02 Cu 91-1051 Cu Co-60 1.91B+C2 +/- 2.73+00 masacks
- Results are in units of # s/(mg of Dosiaster Material).
Referanones negaeste 14340 IAb.nockfes page 169,171. Proce& ares: A-512,A-513,A-524. Analysts w!F, Fpo Appr---i A- *// < A-7 i
i I Since surveillance Capsules T, R, and S were irradiated for multiple fuel ! cycles, the flux adjustment factors, tj, defined in Section 3.0 were employed ! in the reaction rate calculations for the individual sensor sets. , The quantity Cj is defined as the calculated ratio of ( (E > 1.0 MeV) , f, during irradiation period j to the time weighted average ( (E > 1.0 MeV) ' over the entire irradiation period. The values of C d used in the evaluation
- of the Point Beach Unit 2 surveillance capsules were as follows:
C 3 , l CAPSULE T CAPSULE R CAPSULE S CYCLE 1 0.973 1.000 1.081 CYCLE 2 1.030 1.018 1.183 CYCLE 3 1.011 1.009 1.170 CYCLE 4 0.955 1.124 CYCLE 5 1.009 1.121 , CYCLE 6 1.151 l CYCLE 7 0.972 , CYCLE 8 0.930 CYCLE 9 0.941 j CYCLE 10 0.963 CYCLE 11 0.950 CYCLE 12 0.908 CYCLE 13 0.858 CYCLE 14 0.935 CYCLE 15 0.856 CYCLE 16 0.747 l l
*l i
A-8 i
APPENDIX B d MEASURED SPECIFIC ACTIVITY AND IRRADIATION HISTORY OF REACTOR CAVITY SENSOR SETS l 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 15, 16,and 17 are provided. I The irradiation history of the three fuel cycles was as follows: ! l CYCLE 15 CYCLE 16 CYCLE 17 ! CYCLE STARTUP 11/21/88 11/24/89 11/17/90 , CYCLE SHUTDOWN 09/23/89 10/06/90 09/27/91 1 REF. CORE POWER 1518 MWt 1518 MWt 1518 MNt i . CYCLE 15 CYCLE 16 CYCLE 17 THERMAL THERMAL Thermal GENERATION GENERATION GENERATION l MONTH (MW-hr) MONTH (MW-HR) MONTH __ (MW-HR) ! 11/88 2298).2 11/89 141690 11/90 405311 12/88 1113148 12/89 1123322 12/90 1120800 i 1/89 1127540 1/90 1127322 1/91 1123755 l 2/89 1018557 2/90 1018430 2/91 1013700 3/89 1030496 3/90 1118045 3/91 1110255 4/89 1008555 4/90 1090485 4/91 1087195
. 5/89 1117762 5/90 1127775 5/91 1124595 6/89 1089331 6/90 1080806 6/91 1080040 7/89 1126185 7/90 1128256 7/91 1100758 1 8/89 1078103 8/90 1129766 8/91 1087688 9/89 776161 9/90 1120491 9/91 941710 i
10/90 162384 TOTAL 10715717 11368772 11195800 B-1
The irradiation capsule loading diagram and the measured specific activities of the radiometric monitors from the Cycle 15 irradiation are - provided on pages B-3 through B-10. Similar data from the Cycle 16 and . Cycle 17 irradiations are given on pages B-11 through B-17 and B-18 through B-24, respectively. For the multiple foil sensor sets, the : individual foil ID can be correlated with the capsule loading diagrams ; provided in Sections 6.1.1, 6.2.1, and 6.3.1 in order to determine the . location of the foil within the reactor cavity during irradiation.
~
O m b a i e l l B-2 6
I Westirgtecte I;ectric Cer;cratien rmpT AcNar.ced E e gy Systers - Analytical Laberatcry . Walt: Fill Site Eequest* 13910 TO: A.H.Fers (W)McD, Energy Center (East ?-17) Received: 1/19/90 Reported: 2/20/90 [RESULTS OF ANALYSIS} Point Beach Unit 2 Cycle 15 Reactor Cavity Dosimetry Lab Dosimeter (f 2/1/90)
. Foil ID Samplef Material Nuclide dps/ag foil 2 sigma AT 90-233 Ni Co-58 10.5 +/- 0.1 AG 90-243 Ni Co-58 94.9 +/- 1.1 AB 90-253 Ni co-58 245.4 +/- 3.1 AI 90-263 Ni Co-58 93.0 +/- 1.4 r
AJ 90-273 Ni Co-58 215.3 +/- 90-283 2.1 AK Ni co-58 160.3 +/- 1.4 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 AM 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 AK 90-284 Cu Co-60 0.618 +/- 0.011 AL 90-294 Cu Co-60 0.467 +/- 0.010 0.448 +/- 0.007 T 90-235 Ti Sc-46 0.190 +/- 0.011 G 90-245 Ti Sc-46 2.00 +/- 0.05 B 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 Sc-46 Ti 3.30 +/- 0.08 Reentks: AL File: 13910
+
References:
Lab.Bookf41p74. ta435p213-216. Procedures: A-524. Analyst: WIT, WRM, CAB. Approved g. - #-20 -% B-3
l 1 l i I Westir.gtrese Ctetric Ccr;cratier.
- Adianced Enercy Systers - Analytical Lat: oratory elxn haltz Mill Site Eequest8 1391C
'IO: A.R. Fero (W)m2D, Energy Center (East 4-17)
Received: 1/19/90 Reported: 2/20/90 (RESULTS OF ANALYSIS] Point Beach Unit 2 Cycle 15 Reactor Cavity Losimetry
~
Lab Dosimeter (6 2/1/90) - Foil ID Sampled Material Nuclide dps/mg foil 2 signa BA 90-237 AICo Co-60 65.1 +/- 1.1 DA 90-238 AICo Co-60 47.0 +/- 0.9 - BM 90-247 AICo Co-60 178.8 +/- 4.6 DM 90-248 A1Co Co-60 131.5 +/- 1.6 . BL 90-257 A1Co Co-60 525.6 +/- 8.3 DL 90-258 A1Co co-60 310.6 +/- 6.1 ' BK 90-267 AICo 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 AICo Co-60 375.8 +/- 7.0 BJ 90-287 AlCo Co-60 537.8 +/- 8.2 l DJ 90-288 AlCo Co-60 305.9 +/- 6.1 BH 90-297 AICo Co-60 339.9 +/- 6.6 - DH 90-298 Alco 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 *l' CG 90-241 Fe Mn-54 6.72 +/- 0.12 DG 90-242 Fe Mn-54 8.33 +/- 0.09 CH 90-251 Fe Mn-54 20.81 +/- 0.37 DH 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 DJ 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 Remarks: i AL File: 13910
References:
Lab.Bookt41p74. Ia435p213-216. Procedures A-524. g,,g Analyst: WTF, WRM, CAB. Approved: g , , B-4 l r
l l
- est:.rghetse Elec.rie Cer;cratien Advar.ced Energy Systerrs - Analytical Laboratcry
, PrCF'" Waltz Mill Site Pequerte 12910 ! i TO: A.H. Fero (W) m!D, Ihergy Center (East 4-17) Received: 1/19/90 ! Reported: 2/20/90
* (RESULTS OF ANALYSIS) ,
Point Beach Unit 2 Cycle 15 Reactor Cavity Dosimetry l Lab Dosimeter (f 2/1/90) i 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 3r-95 7.74 +/- 0.07 t E 90-259 U-238 Ir-95 19.59 +/- 0.13 I 90-269 U-238 Ir-95 7.90 +/- 0.07 J 90-279 U-238 3r-95 17.76 +/- 0.13 K 90-289 U-238 Er-95 13.32 +/- 0.11 L 90-299 U-238 Ir-95 11.69 +/- 0.08 T 90-239 U-238 Ru-103 0.544 +/- 0.005 G 90-249 U-238 Ru-103 3.780 +/- 0.040 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 K 90-289 U-238 Ru-103 6.082 +/- 0.063
*. L 90-299 U-238 Ru-103 5.451 +/- 0.048 T 90-239 U-238 Cs-137 0.114 +/- 0.003
~ G 90-249 U-238 Cs-137 0.600 +/- 0.019 H 90-259 U-238 Cs-137 1.843 +/- 0.036 I 90-269 U-238 Cs-137 0.677 +/- 0.021 J 90-279 U-238 Cs-137 1.683 +/- 0.036 K 90-289 U-238 Cs-137 1.241 +/- 0.030 L 90-299 U-238 Cs-137 1.102 +/- 0.021 1 90-240 Np Cs-137 2.33 +/- 0.32 2 90-250 hp Cs-137 10.59- +/- 0.53 3 90-260 Np Cs-137 24.07 +/- 0.64 4 90-270 pp 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 Demarks: AL File: 13910
References:
Lab.Bookf41p74. LB035p213-216. Pra m &res: A-524. Analyst: WIT, M m , CAB. Approved: *
# ~#0* E0
= 0 B-5
Westirgriese I* t.~.ric C::;c::::c-Advanced Energy Syste s - Ir.alytical Lt2cratcry F.EPCr Waltz Mill Site Fec,uert# 12910 TO: A.H. Fero (W)lATD, Energy Center (East 4-17) ! Received: 1/19/90 Reported: 2/20/90 ,
. i (REELTS OF ANhLYSIS)
Point Beach Unit 2 Cycle 15 Reactor Cavity Dosimetry , i
~
Bead Chain Tag ID: S-2, 0 degree. i dps/ag of chain 6 2/1/90 Feet [< >] i frcan Lab sti-54 Co-58 Co-60
- Midplane Sanplef dps/ag 2 signa dps/mg 2 signa dps/ag 2 signa l
+6.5 90-301 3.11 +/- 0.11 5.20 +/- 0.12 42.4 +/- 0.3 +5.5 90-302 7.58 +/- 0.40 12.89 +/- 0.53 95.6 +/- 1.1 +4.5 90-303 11.07 +/- 0.53 18.96 +/- 0.63 122.7 +/- 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 90-306 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 - -6.5 90-314 2.78 +/- 0.22 4.90 +/- 0.24 67.4 +/- 0.8 ,
l l NET: For the genma counts, 6 beads wen cut from the chain at the designated points of " Feet frem Midplane" (3 beads on secti side of the point). 6 beads are about 1.4 inches long and weigh about 1.2 grams. i I Remarks: AL File: 13910
References:
Lab.Bookf 41p74.12435p213-216. Procedures: A-524. M' Analyst: WIT, WRM, CAB. Approved: 2'h'9# . B-6
1:estr.gr4*ere E;ectric Cergre.tien Advanced Energy Systers - Analytical Laboratory . TSCF; Kaltz Mill Site Eequest8 1391C
'!Os A.R.Tero (W)NPdD, Energy Center (East 4-17)
Received: 1/19/90 Reported: 2/20/90 ! [RESLTS & ANPLYSIS] Point Beach Unit 2 Cycle 15 Reactor Cavity Dosimetry
- Dead Chain Tag ID: S-2, 15 degree.
reet !< ops /mg of chain e 2/1/90 >) from Lab Ph-54 Co-58 Co-60 >-
, Midplane Samplet ops /ag 2 sigma dps/mg 2 sigma dps/ag 2 sigma ,
l
+6.5 90-315 2.40 +/- 0.20 4.11 +/- 0.23 43.7 +/- 0.5
+5.5 90-316 6.64 +/- 0.51 10.19 +/- 0.53 141.4 +/- 1.3
+4.5 90-317 10.05+/- 0.62 16.52 +/- 0.63 203.6 +/- 1.6 i
+3.5 90-318 11.15 +/- 0.98 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
+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 -
. -4.5 90-328 2.58 +/- 0.21 4.50 +/- 0.22 68.2 +/- 0.5 i i NDIT: For the gamma counts, 6 beads were cut from the chain at the designated ' points of " Feet fr a Midplane" (3 beads on each side of the point). , 6 beads are about 1.4 inches long and weigh about 1.2 grams. ' i
- l Remarks: i AL File: 13910
References:
Lab.Bookl41p74. LB435p213-216. Procedures: A-524. g4 Analyst: WIT, WRM, CAB. Approved g .P l B-7 l
Wertirghetst Electric Cergraticn FHCF; IAanced Er.ergy Syrterns - Analytical Iaboratcry , Kaltz Hill Site Fequest8 12910
'IO: A.R. Fero (W)NMD, Energy Center (East 4-17)
- Received: 1/19/90 l Reported: 2/20/90 :
(REE LTS OF ANRLYSIS] Point Beach Unit 2 Cycle 15 Reactor Cavity Dosiastry need Chain Tag ID: s-2, 30 degree. - ' Feet [< dps/ag of chain f 2/1/90 from Lab Mn-54 ->} Coa-58 Co-60 Midplane Samplet dps/ag 2 signa dps/ag 2 signa dps/ag 2 sigma
+6.5 90-329 2.03+/- 0.11 3.36 +/- 0.11 +5.5 90-330 38.5 0.3 4.95 +/- 0.33 8.58 +/- 0.37 108.2 0.8 +4.5 90-331 7.60 +/- 0.38 12.21 +/- 0.51 +3.5 90-332 154.5 0.9 8.37 +/- 0.67 13.71 +/- 0.54 183.4 1.2 +2.5 90-333 8.93 +/- 0.46 14.84 +/- 0.54 +1.5 90-334 195.0 1.1 8.51 +/- 0.78 14.97 +/- 0.98 204.0 2.0 +0.5 90-335 8.60 +/- 0.84 14.18 +/- 0.93 -0.5 90-336 210.6 2.1 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 -2.5 90-338 1.5
- 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 -4.5 90-340 1.4 7.05 +/- 0.35 11.97 +/- 0.43 138.9 0.9 4 -5.5 90-341 4.45 +/- 0.27 7.94 +/- 0.32 -4.5 90-342 74.5 0.7 -
1.94 +/- 0.20 3.31 +/- 0.26 51.9 0.6 i i e N7FE For the gamma counts, 6 beads wre cut from the chain at the designated ' points of " Feet fran Midplane" (3 beads on each side of the point). 6 beads are about 1.4 inches long and wigh about 1.2 grams. Renarks: ** AL File: 13910 *
References:
Lab.Bookt41p74. LB435p213-216. ys ', CAB. Approwd: dN- R-# f0 . B-8
l l Westzgtecte E*ectric Corg rrtien i Advarced Energy Syrters - Analytical Laboratory Rnc!c Faltz E.ill Site Fequert# 13910 TO: A.H. Fero Of)lAE, Durgy Center (East 4-17)
- Received: 1/19/90 Reported: 2/20/90
[RMATS OF AIRLYSIS) Point Beach Unit 2 Cycle 15 Reactor Cavity Dosinstry Bead Chain Tag ID: S-2, 45 degree. Feet [< dps/mg of chain f 2/1/90 >} frcan Imb Mn-54 Co-58 Co-60 Midplane Samplet dps/ag 2 sigma dps/ag 2 sigma dps/mg 2 sigma 44.5 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
-6.5 90-356 1.83+/- 0.17 3.02 +/- 0.17 48.9 +/- 0.4 101E For the gamuun 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 irches long and weigh about 1.2 grams. Remarks: At. File: 13910
References:
Lab.Bocif 41p74. Ia435p213-216. Pre res: A-524. Analyst: WIT, WRM, CAB. Approved:
'g*
- NOid e
B-9
- I i !
i CONTENTS or AULTIPLE FOIL SENSOR SETS I CAPSULE 10 BARE OR RADIDMETRIC MONITOR ID and CADH!UM SSTR , POSITION SHIELDED [g Ni Q Ti lgt Q y-2}g g-j}} PACKAGE ~ G-1 8 CG -- -- -- -- BM -- -- PB-6B ! G-2 Cd DG AG AG G H DM G -- -- t G-3 Cd -- -- -- -- - -- -- 2 PB-6C , i l H-1 B CH -- -- -- -- BL -- -- PB-168 H-2 Cd DH AH AH H L DL H -- -- ;
. H-3 Cd -- -- -- -- -- -- -- 3 PB-16C I-! 8 C! -- - -- -- BK -- --
PB-7B i 1-2 Cd DI AI Al I K DK 1 -- -- I-3 Cd -- -- -- -- -- - -- 4 PS-7C ! J-l B CJ - -- -- - BI -- - PB-178 l J-2 Cd DJ AJ AJ J ! DI J -- -- l J-3 Cd - -- - -- - -- - 5 P8-17C K-1 8 CK - - - - BJ -- - P8-188 K-2 Cd DK AK AK K J DJ K - - K-3 Cd -- -- - -- - - -- 6 PB-18C L-1 8 CL - - - - SH - -- P8-198 . I L-2 Cd DL AL AL L M DH L -- - L-3 Cd - - - - -- - - 7 PS-19C , IX-1 B CT - - - -- BA - -- PS-1B II-2 Cd DT AT AT T A DA T -- --
- l XI-3 Cd -- - - -- - -- --
1 PB-IC
.i I
'l l
B-10 I l
. ~ _ . --
~
F
>l Westinghouse Advanced Energy Systems l *- RDORT l Analytical Laboratory - Walts Mill Site l Aequest0 14245 o Originator: S.L. Anderson Of) LED, Energy Center (4-36)
Received: 1/14/91 Reported: 3/26/91 , [RESULTS OF ANAL"fSIS) Point Beach Reactor Cavity Dosimetry i
. Lab Dosimeter (6 12/12/90) f
. Foil ID Samplet Material Nuclide dps/ag
- 2 sigma CM 91-254 Fe Mn-54 9.09E+00 +/- 1.3E-01 l DM 91-255 Fe Mn-54 9.83E+00 +/- 1.4E-01 CN 91-263 Fe Kn-54 2.04E+01 +/- 2.0E-01 l 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 i CP 91-281 Fe Mn-54 1.75E+01 +/- 2.0E-01 l DP 91-282 Fe Mn-54 1.66E+01 +/- 1.7E-01 i j CS 91-290 Fe Mn-54 1.47t+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-03 1 AO 91-275 Cu Co-60 2.51E-01 +/- 5.6E-03 AP 91-284 Cu Co-60 5.10E-01 +/- 6.4E-03 As 91-293 Cu Co-60 4.56E-01 +/- 7.5E-03 1 AR 91-302 Cu Co-60 4.49E-01 +/- 7.3E-03 t BG 91-260 A1Co co-60 1.85E+02 +/- 2.9E+00 DG 91-261 A1Co co-60 1.32E+02 +/- 2.4E+0b BF 91-269 A1Co 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 l BD 91-287 AlCo Co-60 5.63E+02 +/- 5.2E+00 DD 91-288 Alco Co-60 3.375+02 +/- 3.3E+00 l BC 91-296 A1Co Co-60 5.03E+02 +/- 4.2E+00 91-297 Co-60 l DC A1Co 2.73E+02 +/- 3.0E+00 BB 91-305 A1Co co-60 3.13E+02 +/- 2.6E+00 DB 91-306 A1Co co-60 2.08E+02 +/- 2.0E+00 l 1
. Assarks:
- Results are in units of dps/(ag of Dosimeter Meterial).
AL File: 14245
References:
Lab.dockt 46 pages 103-104. , j Pramawes: A-524. - g- /
. Anair.t mr, nC Appe m a --
D B-11
.. - ~ - . . . -. .-. - = - .- - .- -- -, -- --
i
*b
- REVIsuD I westinghouse Advenood unergy systems .
REFGW l Analytical Imboratory - Malts Mill Site Ampacett 14245 - j
*I Originator S.L. Anderson Of)MID, Energy Center (4-36) .
Received: 1/14/91 Re,orted: 3/27/91 . (RESULTS OF ANALYSIS] l Point Beach Reactor Cavity Dosimetry
- I Lab Dosimeter (G 12/12/90); . {
Foil ID Sample # Material Nuclide ops /ag
- 2 signa _
M 91-262 U-238 Cs-137 7.63E-01 +/- 2.7E-02 -! N 91-271 U-238 Cs-137 1.61E+00 +/- 8.7E-02 . t O 91-280 U-238 Cs-137 7.30E-01 +/- 3.9E-02 P 91-289 U-238 Cs-137 1.47E+00 +/- 6.2E-02 8 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 0-238 Ru-103 1.26E+01 +/- 2.3E-01 ~ N 91-271 U-238 Ru-103 2.28E+01 +/- 5.2E-01 l 0 91-280 U-238 Ru-103 1.11E+01 +/- 3.2E-01 -l P 91-289 U-238 Ru-103 1.92E+01 +/- 4.2E-01' ! S 91-298 U-238 Ru-103 1.585+01 +/- 2.0E-01 1 R 91-307 U-238 Ru-103 1.48E+01 +/- 1.9E-01 .{ M 91-262 U-238 Ir-95 1.63E+01 +/- 2.1E-01 ! N 91-271 U-238 Ir-95 3.10E+01 +/ 4.1E-01 t 0 91-280 U-238 Ir-95 1.52E+01 +/- 2.3E-01 ' 'l P 91-289- U-238 Ir-95 2.72E+01 +/- 3.8E-01 } 8 91-298 U-238 3r-95 2.24E+01- +/- 2.0E-01 -l R 91-307 U-238 Ir-95 2.07E+01 +/- 1.9E ! I M 91-258 Ti sc-46 3.50E+00 +/- 8.1E-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 5.99E+00 .+/- 8.8E-02 i 8 91-294 Ti sc-46 .5.26t+00 +/ 8.3E-02
R 91-303 Ti Sc-46 5.02E+00 +/- 7.9E-02 t AM 91-256 Ni- Co-58 1.89E+02 +/- 3.7E+00 l AN 91-265 Ni Co-58 3.67E+02 +/- 5.1E+00 l A0 91-274 Ni co-58 1.74E+02 +/- 3.5E+00 l AP 91-283 Ni Co-58 3.17E+02 +/- 4.BE+00
- i AS 91-292 Ni Co-58 2.69E+02 +/- 4.4E+00 l AR 91-301 Ni co-58 2.55E+02 +/- 4.1Et00 '
( l l Rosarks:
- Results are in units of @a/ tag of Desiaster storial). !
- susple 491-280,Er-95 data corrected. ,
I AL File: 14245
References:
Lab.Bocke 49 pages 32-37. . .t Procedure
- ,sts .s .A-524.
.. ,,,r- :
j-- - - -
. l t
B-12 i
Y 1 i l Westinghouse Advanced Energ systees l l REKatt l' Analytical Iaboratory - Maltz Mill Site l Roguestl 14220 ; i o , Originator: S.L. Anderson Oi)NID, Energ Center (4-36) , Received: 1/14/91 > Reported: 3/27/91 [RE!KETS & ANLYSIS] j Point Beach Reactor Cavity Dosimetry Bead Chain Tag ID: 0 dog. Feet [< W ag of chain i 12/12/90 > from Iab . Iti-54 Co-58 Co-60 Midplane Seegdee dpvag 2sipa dpvag 2 sigen dpav'ag 2 sigma
+7.5 91-45-A 1.15E+00 +/- 1.5E-01 3.26E+00 +/- 2.7E-01 3.01E+01+/-2.4E30 46.5 91-45-B 3.65E+00 +/- 1.9E-01 9.89E+00 +/- 4.2E-01 4.31E+01 +/- 2.8E-41
+5.5 91-65-C 8.87E+00 +/- 3.1E-01 2.41E+01 +/- 7.6E-01 9.86E+01 +/- 4.3E-01
+4.5 91-65 4 1.40E+01 +/- 5.6E-01 3.75E+01 +/- 1.2E+00 1.32E+02 +/- 7.0E-01
+3.5 91-65-E 1.56E+01 +/- 8.0E-01 4.0e+01 +/- 1.8E+00 1.53E+02 +/- 1.1E+0C
+2.5 .91-65-F 1.66E+01 +/- 7.2E-01 4.19E+01 +/- 1.8E+00 1.66E+02 +/- 9.9E-01
+1.5 91-6 M 1.58E+01 +/- 7.9E-01 3.895+01 +/- 1.6t+00 1.61E+02 +/- 9.2E-01
+0.5 91-65 4 1.42E+01 +/- 7.2E-01 3.54E+01 +/- 1.7t+00 1.63E+02 +/- 9 M-01
-0.5 91-45-I 1.14E401 +/- 6.5E-01 2.95E+01 +/- 1.5E+00 1.60E+02 +/- 9.7E-01
-1.5 91-6 %T 1.12E+01 +/- 6.6E-01 2.84E+01 +/- 1.4E+00 1.54E+02 +/- 9.6E-01
-2.5 91-65-E 1.26E+01 +/- 7.9E-01 3.12B+01 +/- 1.65+00 1.51E+02 +/- 8.5E-01
- -3.5 91-45-L 1.31E+01 +/- 6.9E-41 3.57E+01 +/- 1.5E+00 1.55E+02 +/- 9.6E-01
-4.5 91-65 +t 1.30E+01 +/- 7.2E-01 3.33E+01 +/- 1.5E+C0 1.27E+02 +/- 7.8E-01
-5.5 91-65 1 8.83E+00 +/- 5.5E-01 2.37E+01 +/- 1.2B+00 9.07E+01 +/- 7.3E-01
-4.5 91-4 5 o 3.36E+00 +/- 5.0E-01 9.73E+00 +/- 9.9E-01 6.68t+01 +/- 5.7E-01 Remarks:
- Romults are in units of W(ag of Dosimeter Raterial).
AL File: 14220
References:
Lab.Bookt 49 pages 32-37. Procedures: A-524. an.1rst: =r, = -M g. i B-13
1 l Westinghouse Advanced thergy Systems l > REPORT l Analytical Laboratory - Waltz Mill Site l Requesti 14220 , Criginator: S.L. Anderson (W)NID, Energy Center (4-36) Received: 1/14/91 - Reported: 3/27/91 [ RESETS T ANhLYSIS] ,
~
Point Beach Reactor Cavity Dosimetry Bead Chain Tag ID: 15 deg. Feet [< dps/mg of chain 6 12/12/90 ; from Lab Mn-54 Co-58 Co-60 Midplane Samplet dps/mg 2 sigma dpe/mg 2 sigma dpsVag 2 sigma
+7.5 91-66-A 8.48E-01 +/- 1.7E-01 2.47E+00 +/- 3.9E-01 2.71E+01 +/- 2.9E-01 +6.5 91-66-B 2.68E+00 +/- 3.5E-01 7.15E+00 +/- 7.2E-01 4.32E+01 +/- 4.1E-01 +5.5 91-66-c 6.66E+00 +/- 5.5E-01 1.88E+01 +/- 1.3E+00 1.42E+02 +/- 9.1E-01 +4.5 91-66-D 1.07E+01 +/- 7.3E-01 2.93E+01 +/- 1.7E+00 2.03E+02 +/- 1.1E+0C +3.5 91-66-E 1.28E+01 +/- 1.4E+00 3.10E+01 +/- 2.6E+00 2.2BE+02 +/- 1.7E+0C +2.5 91-46-F 1.29E+01 +/- 9.8E-01 3.23E+01 +/- 2.3E+00 2.40E+02 +/- 1.5E+00 +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-B 1.11E+01 +/- 9.9E-01 2.86E+01 +/- 1.8E+00 2.17E+02 +/- 1.5E+00 -0.5 91-66-I 9.90E+00 +/- 1.lE+00 2.49E+01 +/- 2.4E+00 2.00E+02 +/- 1.5E+00 -1.5 91-66-J 9.97E+00 +/- 6.5E-01 2.58E+01 +/- 1.6E+00 1.91E+02 +/- 1.lE+00 -2.5 91-66-K 9.75E+00 +/- 9.6E-01 2.60E+01 +/- 2.2E+00 1.86E+02 +/- 1.2E+00 -3.5 91-66-L 1.16E+01 +/- 7.3E-01 2.86E+01 +/- 1.6E+00 1.90E+02 +/- 1.lE+00 -4.5 91-66 41 1.11E+01 +/- 6.8E-01 2.89E+01 +/- 1.6E+00 1.66E+02 +/- 1.0E+00 -5.5 91-66-N 6.95E+00 +/- 5.5E-01 1.84E+01 +/- 1.3E+00 1.22E+02 +/- 8.5E-01 -6.5 91-66-0 2.97E+00 +/- 4.2E-01 7.37E+00 +/- 8.lE-01 6.52E+01 +/- 5.1E-01 Remarks:
- Results are in units of dps/(ag of Dosimeter Material). ,
AL File 14220 . References Lab.Bookt 49 pages 32-37. , Procedures: A-524. ,[ - Analyst: W2F, W. Approved ~C B-14 1
. I Westinghouae Advanced Energy Systems I t RDGtr i Analytical Laboratory - Walts Mill Site l Requeste 14220
- Originator: S.L. Anderson M),'!D, Energy Center (4-36) - l Deported: 3/27/91
[RMILTS T AIRLYSIS] Point Beach Raxtor Cavity Dosisatry i Band Chain Tag ID: 30 dog. - Feet [< dps/mg of chain 0 12/12/90 >J
- frcan Lab Mn-54 -
Co-58 Co-60 Mi? =m sample # % 2 signa dps/ag 2 sips dpa/ag 2 si pa !
+7.5 91-47-A 7.58E-01 +/- 1.1E-01 2.13B+00 +/- 3.1E-01 2.29E+01 +/- 2.1E-01
+6.5 91-67-e 2.25E+00 +/- 2.2E-01 5.96t+00 +/- 5.3E-01 3.74E+01 +/- 2.9E-01 i
+5.5 91-67-c 5.66E+00 +/- 5.3E-01 1.47E+01 +/- 1.0E+00 1.068+02 +/- 7.9E-01
+4.5 91-67-0 8.21E+00 +/- 6.1E-01 2.08E+01 +/- 1.4E+00 1.50E+02 +/- 9.4E-01
+3.5 9147-E 8.59E+00 +/- 9.0E-01 2.268+01 +/- 2.0E+00 1.72E+02 +/- 1.1E+00
+2.5 91-67-F' 9.49E+00 +/- 6.7E-01 2.49E+01 +/- 1.5E+00 1.86E+02 +/- 1.1E+00 :
+1.5
- 91-67-G 9.11B+00 +/- 9.0E-01 2.39E+01 +/- 2.0E+00 1.84E+02 +/- 1.2E+00 t
+0.5 91-67-5 9.54E+00 +/- 6.4E-01 2.40E+01 +/- 1.5E+00 1.88E+02 +/- 1.1E+00 ;
-0.5 91-67-I 9.19E+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 9147-K 9.13E+00 +/- 6.9E-01 2.295+01 +/- 1.6E+00 1.72E+02 +/- 1.0E+00
-3.5 91-67 6 8.27t+00 +/- 6.35-01 2.19E+01 +/- 1.4B+00 1.58E+02 +/- 9.8E-01 r
- -4.5 91-67-M 7.70E+00 +/- 4.0E-01 2.04E+01 +/- 9.4EH01 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 4.975+01 +/- 3.0E-01
)
h 4 l i n
. I l
Remarks:
- Results are in units of dpsV(ag of Dosimeter Material). ,
. 1 AL File: 14220 i
References:
Lab.Booke 49 pages 32-37. ! Prnemaires: A-524.
- Analyst: WTF, 7K Approved: ^-
)
4 l B-15 l
. - - = . _ . . _ - .. _ .- - . ~
f t l
'I.
?
i i I Mastinghouse Advanced Energ Syntams l REPORf I Analytical Laboratory - Waltz Mill Site i Asqueste 14220 t! Originator: S.L. Anderson (W)N2D, Energ Center (4-36) ! Received: 1/14/91 l Baported: 3/27/91 , [REELTS & AlmLYSIS] Point Beach Reactor Cavity Dosimetry ,' , i Bead Chain Tag ID: 45 deg. l Feet [< @sAug of chain 0 12/12/90 i from >} . Lab Mrr-54 Co-58 Co-60 i Mi@ lane Sample 6 % 2 sigua @e/ag 2 signa dps/ag 2 signa [
+7.5 91-68-A 6.41E-01 +/- 9.2E-42 1.78E+00 +/- 1.9E-01 2.0$E+01 +/- 1.3E-01
+6.5 91-68 4 i 2.04E+00 +/- 1.7E-01 5.23E+00 +/- 3.7E4 1 3.20E+01 +/- 2.4E-01 :
+5.5 91-48-C 4.86E+00 +/- 3.4E41 1.24E+01 +/- 7.1E-01 6.48E+01 +/- 3.9E-01
+4.5 91-68-0 7.55E+00 +/- 5.4E-41 1.95E+01 +/- 1.3E+00 {
+3.5 91-68-E 8.975+01 +/- 7.3E41 ;
7.94E+00 +/- 5.9E-01 2.09E+01 +/- 1.3E+00 1.05E+02 +/- 7.9E-41
+2.5 91-68-f 8.91E+00 +/- 4.3E-01 2.24E+01 +/- 9.32-41 }
+1.5 1.13E+02 +/- 5.1E-41 -
9148-G 8.84E+00 +/- 5.9E-01 2.15E+01 +/- 1.2E+00 1.20E+02 +/- 8.5E-01 l
+0.5 91-68 4 8.78E+00 +/- 6.1E-01 2.15E+01 +/- 1.3E+00
-0.5 1.19t+02 +/- 8.4E-01 91-68-I 3.65E+00 +/- 6.7E-01 2.27E441 +/- 1.5E+00 1.15E+02 +/- 7.4E 41
-1.5 91-68-J 8.71E+00 +/- 5.8E-01 2.13E+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 1.07E602+/-7.2E-01
-3.5 91-68-L 8.15E+00 +/- 3.5E-01 2.088+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 1
-5.5 91-48-N 5.12E+00 +/- 1.4E-01 1.35E+01 +/- 3.4E-01 6.22E441 +/- 2.0E-01 '
-4.5 91-68-0 2.03E+00 +/- 2.0E-01 5.78E+00 +/- 4.5E-41 4.63E+01 +/- 2.7E-01 l
Ammarks:
- Assults are in units of dps/(ag of heimmter Material). '
AL File: 14220
References:
Lab.Bookt 49 pages 32-37. . Procedure
.n.1 , .s: A-524.
.. .p, e :.
g- - _ B-16 V -
, .. - - - - . . . . ~
. 1 i
I r k a i I i I CONTENTS OF MULTIPLE FOIL SENSOR SETS i CYCLE 16 IRRADIATION i 1 i i
. CAPSULE ID 8ARE OR RADIOMETRIC MONITOR ID and CADMIUM SSTR POSITION SHIELDED {g gi CM 11 Bh La U-23A PACKAGE I
M-1 8 CM -- -- -- -- BG -- P8-48 M-2 Cd DM AM AM M G DG M -- + 1 M-3 Cd -- -- -- -- -- -- -- PB-4C i N-1 8 CN -- -- -- -- BF -- PB-12B l N-2 Cd DN AN AN N F DF N -- , N-3 Cd -- -- -- -- -- -- -- P8-12C ! 0-1 B C0 -- -- -- -- BE -- PB-5B l 0-2 Cd DO A0 A0 0 E DE O -- 0-3 Cd -- -- -- -- -- -- -- PB-5C l P-1 B CP -- -- -- -- 80 -- PB-13B l P-2 Cd DP AP AP P D 00 P -- I P-3 Cd -- -- -- -- -- -- -- P8-13C J Q-1 B CS -- -- -- -- BC -- PB-148 l l Q-2 Cd DS AS AS S C DC S -- - 0-3 Cd -- -- -- -- -- -- -- P8-14C 1 R-1 B CR -- -- -- -- BB -- P8-15B i
. R-2 Cd DT AT AT T A DA T --
R-3 Cd -- -- -- -- -- -- -- PB-15C l I i I I I l B-17 i i
, _ . _ . _ _ _ _ _ _ , y ,_ . _ . , . ----m ,__ - - - - - - __- .- --- -
l i
............................................... L Westinghouse Advan:ed Energy Systems
- REPORT Analytical Laboratory . Waltz Mill Site Requesti 14477 :
Originator: 5. Anderson (W)NTD, Energy Center Received: 10/18/91 : Reported: 12
................................................................................../12/91.......
(RESULTS OF ANALYSIS) POINT BEACH UNIT 2 CYCLE 17 REACTOR CAVITY 0051 METRY Lab Dosimeter (f 10/24/91) Foil ID Samplef Material Nuclide dps/mg
- 2 sigma
- i BG 91 1822 Fe Mn.54 9.82E+00 +/- 1.lE-01 -
AG 91 1823 Fe Mn.54 1.08E+0! +/. 1.lE 01 . ; BH 91 1831 Fe Mn.54 2.30E+01 +/. 1.6E.01 ' AH 91 1832 Fe Mn.54 2.22E+01 +/. 1.6E.01 Al 91 1840 Fe Mn.54 1.llE+01 +/- 1.2E 01 81 91-1841 Fe Mn.54 9.91E+00 +/. 1.lE.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 i AL 91 1868 Fe Mn-54 1.56E+01 +/- 1.3E-01 j G 91 1824 Ni Co.58 2.83E+02 +/. 1.6E+00 ' H 91 1833 Ni Co.58 5.51E+02 +/- 2.2E+00 1 91 1842 Ni Co.58 2.61E+02 +/. 1.5E+00 J 91-1851 Ni Co.58 2.0E+00 4.80E+02 +/. K 91 1860 N1 Co.58 4.00E+02 +/. 1.5E+00 L 91 1869 Ni Co.58 3.82E+02 +/- 1.8E+00 BG 91 1828 Alto Co-60 1.95E+02 +/. 1.9E+00 AG 91 1829 AlCo Co.60 1.40E+02 +/. 1.6E+00 BH 91 1837 A1Co Co.60 4.67E+02 +/. 3.0E+00 AH 91 1838 AICo Co.60 2.77E+02 +/. 2.3E+00 Al 91 1846 AlCo Co.60 2.13E+02 +/. 2.0E+00 B1 91 1847 AlCo Co.60 1.40E+02 +/. 1.6E+00 BJ 91 1855 AICo Co 60 5.89E+02 +/. 3.3E+00 AJ 91 1856 AICo Co.60 2.6E+00 3.35E+02 +/. 8K 91 1864 AICo Co.60 5.25E+02 +/. 3.2E+00 AK 91 1865 AlCo Co.60 2.89E+02 +/. 2.3E+00 BL 91 1873 AlCo Co.60 3.16E+02 +/. 2.5E+00 At 91 1874 ?.1Co Co.60 2.21E+02 +/. 2.lE+00 Remarks:
- Results are in units of dps/(mg of Dosimeter Material). .
i AL File: 14477 *
References:
Lab Bookf46 pages 246 247 $, Procedures: A.524 ; Analyst: WTF, TRK, MRK Approved. a I , B-18 ,
I
............................................... t Westinghouse Advanced Energy Systems !
REPORT Analytical Laboratory - Waltz Mill Site Requestf 14477
!s Originator: S. Anderson (W)NTD, Energy Center l
Received: 10/18/91 -
....................................................................... ported:
Re 12/12/91 [RESULTSOFANALYSIS) : t POINT BEACH UNIT 2 CYCLE 17 REACTOR CAVITY 00SIMETRY - 1 Lab Dosimeter (# 10/24/91) Foil ID Samplef Material Muelfde dps/mg
- 2 sigma {
. 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 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 ! K 91 1866 U 238 Cs.137 1.15E+00 +/- 2.52E.02 : L 91 1875 U 238 Cr 137 1.01E+00 +/- 2.62E 02 i G 91 1830 U-238 Ru-103 2.61E+01 +/- 1.09E-01 H 91 1839 U 238 Ru 103 k 4.67E+01 +/- 1.52E 01 ; I 91-1848 U 238 Ru2103 2.32E+01 +/- 1.18E 01 J 91 1857 U-238 Ru 103 l 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 i' 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 1 91 1848 U 238 Zr 95 2.31E+01 +/. 1.35E 01 l J 91-1857 U 238 Zr-95 j 4.38E+01 +/. 1.63E 01 ,
K 91 1866 U 238 Zr 95 3.69E+01 +/ 1.50E 01 ' L 91-1875 U 238 Zr 95 3.23E+01 +/- 1.65E 01 i G 91 1826 Ti l Sc-46 4.91E+00 +/- 4.79E-02 ; H 91 1835 Ti Sc 46 s84E+00 +/. 6.94E 02 I 91-1844 T1 Sc 46 4.f. W OO +/- 4.62E-02 ! J 91 1853 Tl Sc-46 8.68 W 9 +/ 6.50E 02 .} K 91 1862 T1 Sc 46 7.50E+00 +/- 6.02E 02 i L 91 1871 Ti Sc 46 7.36E+00 +/. 5.95E-02 Remarks:
- Results are in units of dps/(ag of Dosimeter Material).
~* AL File: 14477 !
References:
Lab Bookf46 pages 246 247 / . Procedures: A-524. , ! Analyst: WTF, TRK. MRK Approved: ,
- f I :
l l l l B-19 I u
k l Westinghouse Advanced Energy Systems (' REPORT Analytical Laboratory . Waltz Mill Site Requesti 14477 Originator: S. Anderson (W)NTD, Energy Center Received: 10/18/91 ,
....................................................................... ported: Re 12/12/91
[RESULTS OF ANALYSIS] POINT BEACH UNIT 2 CYCLE 17 REACTOR CAVITY DOS! METRY . ( Lab Dosimeter (9 10/24/91) Foil ID Sample # Material Nuclide dps/mg
- 2 sigma
- G 91 1825 Cu Co.60 2.87E.01 +/- 3.79E.03 H 91 1834 Cu Co.60 6.14E.01 +/- 5.53E.03 1 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 t p [ a & r i I i I i Remarks:
- Results are in anits of dps/(mg of Dosimeter Material). ,
3
. i h
AL File: 14477
References:
Lab Bookf46 pages 246 247 "'N f Procedures: A.524. /
- l Analyst: WTF, TRK, MRK Approved: fl I .
t B-20 i
,e r - ---? - - -
- L 1
Westinghouse Advanced Energy Systems REPORT Analytical-Laboratory - Waltz Mill Site Requestf 14477 l l Originator: S. Anderson (W)NATD. Energy Center l Radiation Engineering & Analysis o Westinghouse Electric Corporation Received: 10/18/91
........................................................................ Reported: 2/18/92
.................... )
[RESULTS OF ANALYSIS] P0lNT 8EACH UNIT 2 CYCLE 17 REACTOR CAVITY 00SIMETRY 8ead Chain Tag !D: 0 deg.
. Feet from
[< -------.~.--. dps/mg of chain 9 10/24/91 ~.---~~>] Lab ------ Mn.54 -- --- - ----- Co-58 ------- Midplane Samplef --- .-- Co 60 - - - - dps/mg 2 sigma dps/mg 2 sigma
........ ........ .......... dps/mg 2 sigma
. +7.5 91 1818A 1.25E+00 +/- 1.2E-01
+6.5 4.79E+00 +/- 2.9E-01 2.80E+01 +/- 1.6E.01 91-18188 4.13E+00 +/- 2.4E 01 1.55E+01 +/. 5.9E-01
+5.5 91 1818C 4.05E+01 +/- 2.7E-01 1.03E+01 +/ 3.5E 01 3.71E+01 +/. 8.8E-01 7.52E+01 +/. 3.8E.01
+4.5 91-1818D 1.53E+01 +/- 9.9E-01
+3.5 5.65E+01 +/. 2.2E+00 1.02E+02 +/- 1.0E+no l 91-1818E 1.71E+01 +/- 1.lE+00 5.90E+01 +/- 2.5E+00
} +2.5 91 1818F 1.18E+02 +/- 1. lE+00 ' 1.80E+01 +/ 1.lE+00 6.33E+01 +/. 2.6E+00 1.28E+02 +/- 1.lE+00
+1.5 91-1818G 1.72E+01 +/- 1.lE+00
+0.5 6.01E+01 +/. 2.5E+00 1.27E+02 +/- 1.lE+00 91 1818H 1.60E+01 +/- 1.0E+00 5.39E+01 +/. 2.5E+00
-0.5 91-18181 1.28E+02 +/ 1.lE+00 1.38E+01 +/- 1.1E+00 4.65E+01 +/. 2.5E+00 1.27E+02 +/. 1.lE+00 I
-1.5 91 1818J 1.36E+01 +/- 1.1E+00
-2.5 4.50E+01 +/- 2.6E+00 1.23E+02 +/- 1.lE+00 91 1818K 1.34E+01 +/- 9.4E-01
-3.5 5.07E+01 +/. 2.4E+00 1.24E+02 +/- 1.lE+00
, 91-1818L 1.48E+01 +/- 9.8E 01 5.20E+01 +/. 2.3E+00
-4.5 91 1818M 1.22E+02 +/ 1.lE+00 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
-6.5 91-18180 3.57E+01 +/.1.0E+00 6.97E+01 +/- 4.7E-01 4.23E+00 +/- 4.lE 01 1.46E+0! +/. 8.4E 01 6.49E+01 +/- 4.5E-01
\ {
)
Remarks:
- Results are in units of dps/(ag of Dosimeter Material).
AL File: 14477
References:
Lab Bookf46 pages 246 247 - Procedures: A 524 Analyst: WTF, TRK, MRK Approved:! [' Ni ~f C I l l } B-21 __ _ _ - - _ _ - - _ _ - - _ . - - - - _ _ _ . _ _ - _ -- . - - - - - - - . - _. -J
I REPORT l h[stikghous[kd[anc[kn[bySy$teYs Analytical Laboratory - W tz Mill Site l l Request # 14477 *
)
l; Originator: S. Anderson (W) NATO, Energy Center Radiation Engineering & Analysis Received: 10/18/91 Westinghouse Electric Corporation Reported: 2
..................................................................................../18/92 ........ t
[RESULTS OF ANALYSIS) ! POINT BEACH UNIT 2 CYCLE 17 REACTOR CAVITY DOSIMETRY Bead Chain Tag ID: 15 deg. Feet [<-----.--------- dps/mg of chain 9 10/24/91 -..- -.-.--- >] from Lab -----.- Mn-54 ------- - ----- Co-58 ------- ------- Co 60 ------ Midplane Samplef dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma
. l +7.5 91 1819A 1.11E+00 +/- 1.lE 01 4.14E+00 +/- 2.6E-01 3 +6.5 91-1819B 2.56E+01 +/- 1.5E.01 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.lE 01 2.96E+01 +/- 1.8E+00 +4.5 91 18190 1.35E+02 +/- 9.0E-01 i 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 +2.5 91 1819F 2.15E+02 +/. 9.3E-01 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 +0.5 91 1819H 2.20E+02 +/- 1.2E+00 i 1.31E+01 +/ 8.9E 01 4.45E+01 +/- 2.2E+00 2.05E+02 +/. 1.!E+00 0.5 91 18191 1.25E+01 +/- 9.0E 01 4.10E+01 +/- 1.9E+00 i -1.5 91 1819J 1.62E+02 +/- 9.9E 01 1.13E+01 +/- 7.4E 01 3.95E+01 +/- 2.0E+00 1.52E+02 +/. 9.5E-01 -2.5 91-1819K 3.5 91 1819L 1.27E+01 +/- 7.7E-01 1.25E+01 +/- 8.4E 01 4.13E+01 +/. 1.8E+00 4.51E+01 +/- 2.0E+00 1.51E+02 +/ 9.6E-01 *! -4.5 91 1819M 1.48E+02 +/. 9.5E.01 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 l
6.5 91-181C0 9.42E+01 +/- 7.5E-01 ,! 3.04E+00 +/- 2.4E 01 1.16E+01 +/- 6.2E-01 5.03E+0! +/- 3.lE-01 i
?
l
)
i l l I i Reaarks:
- Results are in units of dps/(eg of Dosimeter Material). E AL File: 14477
References:
Lab Bookf46 pages 246 247 -
/ -
Procedures: A 524. . , Analyst: WTF, TRK, MRK Approved: *, / pf, f
/ y i
'I 1
B-22 t i l l
Westin house Advanced Energy Systems o REPORT Analyt cal 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 .....
[RESULTS OF ANALYSIS) POINT BEACH UNIT 2 CYCLE 17 REACTOR CAvlTY DOSIMETRY Jo
, Bead Chain Tag ID: 48 deg.
Feet from [<-------------- dps/mg of chain 9 10/24/91 ------------>] Lab ~ ~--- Mn 54 - --.~ --- -- Co 58 - --- ------- Co 60 - ---.-
. Midplane Samplef dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma
+7.5 91-1820A 8.05E-01 +/- 1.!E-01 3.07E+00 +/. 2.7E 01
+6.5 2.10E+01 +/- 1.4E 01 91-18208 2.39E+00 +/- 1.5E-01 9.26E+00 +/- 4.0E-01
+5.5 91-1820C 3.47E+01 +/- 1.8E 01 6.06E+00 +/- 5.7E-01 2.09E+01 +/- 1.4E+00 9.70E+01 +/ 7.7E-01
+4.5 91-18200 9.17E+00 +/- 7.7E-01 3.21E+01 +/- 1.8E+00
+3.5 91-1820E 1.40E+02 +/- 9.3E Ol 1.08E+01 +/. 8.0E 01 3.56E+01 +/- 1.8E+00 1.64E+02 +/- 9 lE-01
+2.5 91-1820F 1. ole +01 +/- 8.3E-01 3.74E+01 +/- 2.lE+00
+1.5 1.76E+02 +/- 1.0E+00 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.lE+00 0.5 1.81E+02 +/- 1.0E+00 91 18201 1.09E+01 +/ 8.7E 01 3.93E+01 +/- 2.lE+00
-1.5 91-1820J 1.46E+02 +/- 9.4E 01 1.04E+01 +/- 7.2E-01 3.70E+01 +/- 1.8E+00 1.42E+02 +/- 9.ZE 01
-2.5 91 1820K 1.06E+01 +/- 8.2E-01 3.80E+01 +/- 2.0E+00
-3.5 91-1820L 1.37E+02 +/- 9.lE 01 9.93E+00 +/- 7.lE-01 3.49E+01 +/- 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 5.5 91-1820N 1.02E+02 +/- 7.8E 01
. 5.71E+00 +/- 3.2E-01 2.19E+0! +/- 7.0E-01 5.51E+01 +/- 3.2E-01 6.5 91 18200 2.30E+00 +/- 2.lE-01 8.87E+00 +/- 5.lE-01 3.89E+01 +/- 2.7E-01
................................~............ ............................................
Remarks:
- Results are in units of dps/(mg of Dosimeter Material).
Al File: 14477 /
References:
Lab Bookf46 pages 246-247 m / Procedures: A-524. pg7 d Analyst: WTF, TRK, MRK Approved: h'. B-23
I i l i 1
............................................... +'
Westinghouse Advanced Ene my Systems REPORT Analyttcal Laboratory Waltz Mill Site Requestf 14477 Originator: S. Anderson (W)NATD, Energy Center Radiation Engineering & Analysis Westinghouse Electric Corporation Received: 10/18/91
.......................................................................... ported:
Re 2/18/92 [RESULTS OF ANALYSIS) POINT 8EACH UNIT 2 CYCLE 17 REACTOR CAVITY DOSIMETRY . 8ead Chain Tag ID: 45 deg. Feet (<--------- ---- dps/mg of chain 9 10/24/91 ---- --------> from Lab ------- Co-58 -----.. Midplane Samplef
.--.--- Mn 54 ------.
dps/mg 2 sigma
-------Co-60..-.-..) '
dps/mg 2 sigma dps/mg 2 sigma
+7.5 91-1821A 7.41E-01 +/- 9.0E 02
+6.5 2.43E+00 +/- 1.9E-01 2.01E+01 +/- 1.2E-01 91-1821B 2.22E+00 +/- 1.2E-01 8.10E+00 +/- 2.8E 01
+5.5 91-1821C 2.97E+01 +/- 1.5E-01 5.15E+00 +/- 2.9E-01 1.82E+01 +/- 7.CE-01 6.00E+01 +/- 3.4E.01
+4.5 91-1821D 8.05E+00 +/- 3.5E-01
+3.5 2.88E+01 +/- 8.3E-01 8.04E+01 +/- 3.9E-01 91-1821E 9.27E+00 +/ 5.8E-01 3.23E+0! +/- 1.7E+00
+2.5 91 1821F 9.63E+01 +/- 7.6E-01 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
+0.5 3.52E+01 +/- 1.8E+00 1.13E+02 +/- 8.3E.01 91-1821H 9.36E+00 +/- 6.4E 01 ;
-0.5 3.31E+01 +/- 1.7E+00 1.13E+02 +/. 8.2E-01 91 18211 1.08E+01 +/. 6.5E 01 3.58E+01 +/. 1.6E+00 '
-1.5 91-1821J 9.42E+01 +/- 7.5E 01 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
-3.5 3.51E+01 +/- 1.5E+00 8.86E+01 +/- 7.3E-01 91-1821L 9.67E+00 +/- 6.2E 01 3.38E+0! +/- 1.6E+00
-4.5 91-1821M 8.25E+01 +/- 7.0E-01 .;
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
-6.5 1.02E+01 +/- 1.2E+00 2.17E+01 +/- 6.9E 01 91-18210 2.60E+00 +/- 1.5E-01 9,18E+00 +/- 3.8E-01 3.69E+01 +/. 1.9E.01 ;
i Remarks:
- Resul.s are in units of dps/(mg of Dosimeter Material).
AL File: 14477 ,
References:
Lab Bookf46 pages 246 247 Procedures: A-524. , b [' pq ' Analyst: WTF, TRK, MRK Approved: g f t l B-24
1 e i
'j .,
- CONTENTSdFMULTIPLEFOILSENSORSETS i CYCLE 17 IRRADIATION CAPSULE ID BARE OR. RADIOMETRIC MONITOR ID and CADMIUM SSTR POSITION SHIELDED fl Hi G 11 Nh h M-ZH PACKAGE
~
AA-1 B BG -- -- -- -- BG -- PB-27B ; AA-2 Cd AG G GG G AG G -- AA-3 Cd -- -- -- -- -- -- -- PB-27C i BB-1 B BH -- -- -- -- BH -- PB-29B BB-2 Cd AH H HH H AH H -- BB-3 Cd -- -- -- -- -- - -- PB-29C CC-1 B BI -- -- -- -- BI -- PB-288 CC-2 Cd AI I I I I AI I -- CC-3 Cd -- -- -- -- -- -- - PB-28C \ I
~
00-1 B BJ -- -- -- -- BJ -- PB-308 DD-2 Cd AJ J J J J AJ J -- i DD-3 Cd -- -- -- -- -- -- -- PB-30C EE-1 B BK -- -- -- -- BK -- PB-31B , EE-2 Cd AK K K K K AK K -- EE-3 Cd -- -- -- -- -- -- -- PB-31C t FF-1 B BL -- -- -- -- BL -- PB-32B FF-2 Cd AL L L L L AL L -- FF-3 Cd -- -- -- -- -- -- -- PB-32C . 6 h
~
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