ML18065A616

Nonproprietary Consumers Power Co Reactor Vessel Neutron Fluence Measurement Program for Palisades Nuclear Plant - Cycles 1 Through 11.

ML18065A616
Person / Time
Site:	Palisades
Issue date:	03/31/1996
From:	Shaun Anderson, Brassart G, Perock J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML18065A613	List: ML18065A612 ML18065A616
References
WCAP-14557, WCAP-14557-R01, WCAP-14557-R1, NUDOCS 9604110152
Download: ML18065A616 (265)

Text

WCAP-14557 Revision I WESTINGHOUSE NON-PROPRIETARY CLASS 3 Consumers Power Company Reactor Vessel Neutron Fluence Measurement Program for Palisades Nuclear Plant - Cycles 1 through 11 John D. Perock Stanwood L. Anderson

• March 1996 APPROVED:__.(}J f2n~~--:~~~~~~

~sart, Manager Radiation Engineering and Analysis Prepared by Westinghouse for the Consumers Power Company Purchase Order No. XARP93703 Work perfo11Ded under Shop Order No. M2JP-450 WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Business Unit

• P.O. Box 355 Pittsburgh, Pennsylvania 15230

© 1996 Westinghouse Electric Corporation.

All rights reserved.

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Consumers Power Company

. Reactor Vessel Neutron Fluence Measurement Program for Palisades Nuclear Plant - Cycles 1 through 11 .

EXECUTIVE

SUMMARY

Beginning with Fuel Cycle 8, a reactor cavity measurement program was instituted at Palisades to provide a continuous monitoring of the reactor pressure vessel and reactor vessel support structure. This report presents the results of the reactor cavity neutron dosimetry measurements during Fuel Cycles 8, 9, and 10/11 along with the previous A240, W290, W290-9, and WI 10 surveillance capsule measurements. The use of the cavity measurement program coupled with available smveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit For the Palisades Nuclear Plant, the maximum reactor vessel fast neutron exposure is to the circumferential beltline region which has a peak fluence at angles symmetric with 75°. The exposure to the longitudinal welds at 0° and 30° is also important. Based on comparisons of measurements with calculations, it is concluded that the DORT neutron transport calculations, based on plant specific parameters, over-predict the reactor vessel fast neutron fluence (E > 1.0 MeV) by 17%. Taking into account this bias, the best estimate fluence at the core midplane on the pressure vessel at azimuthal locations at 15° intervals over a core quadrant following Cycle 11 are as follows:

<l>(E > 1.0 MeV) [n/cm2 ]

Cycle 11 9.46e+I8 l.32e+l9 9.73e+18 6.17e+18 9.82e+18 1.32e+19 9.47e+18 As further data are accumulated from subsequent irradiations, the neutron environment in the vicinity of the reactor vessel will become better characterized an~ 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~

All of the calculations and dosimetry evaluations presented in this report have been based on the latest available nuclear cross-section data derived from ENDF/B-VI and are intended to be consistent with the requirements of Draft Regulatory Guide DG-1025, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence". As such, the data provided here are intended to supersede prior evaluations documented in References 1 through 7.

Peak azimuthal fluence location.

iii

TABLE OF CONTENTS LIST OF FIGURES TABLE OF CONTENTS Page iv vi LIST OF TABLES x 1.0 OVERVIEW OF THE PROGRAM 1-1

2.0 DESCRIPTION

OF THE MEASUREMENT PROGRAM 2-1 2.1 Description of Reactor Cavity Dosimetry 2-1 2.2 Description of Surveillance Capsule Dosimetry 2-6 3.0 NEUTRON TRANSPORT AND DOSIMETRY EVALUATION 3-1 METHODOLOGIES 3.1 Neutron Transport Analysis Methods 3-1 3.2 Neutron Dosimetry Evaluation Methodology 3-10 3.3 Determination of Best Estimate Reactor Vessel Exposure 3-16 4.0 RESULTS OF NEUTRON TRANSPORT CALCULATIONS 4-1 4.1 Forward Calculation 4-1 5.0 EVALUATIONS OF SURVEILLANCE CAPSULE DOSIMETRY 5-1 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 8 Results 6-1 6.2 Cycle 9 Results 6-28 6.3 Cycles 10/11 Results 6-58 7.0 COMPARISON OF CALCULATIONS WITH MEASUREMENTS 7-1 7.1 Comparison of Least Squares Adjustment Results with Calculation 7-1 7 .2 Comparisons of Measured and Calculated Sensor Reaction Rates 7-2 8.0 BEST ESTIMATE NEUTRON EXPOSURE OF REACTOR VESSEL 8-1 8.1 Exposure Distributions Within the Beltline Region 8-1 8.2 Uncertainties in Exposure Projections 8-6

9.0 REFERENCES

9-1 iv

TABLE OF CONTENTS Apoendices ~

• A Specific Activities and Irradiation History of S.ensors from A-1 Surveillance Capsules A240, W290, W290-9, and WllO B Specific Activities and Irradiation History of Reactor Cavity Sensor B-1 Sets - Cycle 8 C Specific Activities and Irradiation History of Reactor Cavity Sensor C-1 Sets - Cycle 9 D Specific Activities and Irradiation History of Reactor Cavity Sensor D-1 Sets - Cycles 10/11 E Biological Shield Wall Heat Deposition E-1 v

Figure 2.1-1 Title LIST OF FIGURES Azimuthal Location of Sensor Strings Page 2-4 2.1-2 Irradiation Capsule for Cavity Sensor Sets 2-5 2.2-1 Typical Surveillance Capsule 2-7 3.1-1 Reactor Geometry Showing a 90° R,6 Sector for Cycles 1 and 2 3-4 3.1-2 Reactor Geometry Showing a 90° R,6 Sector for Cycles 3-7 and 9 3-5 3.1-3 Reactor Geometry Showing a 90° R,6 Sector for Cycle 8 3-6 3.1-4 Reactor Geometry Showing a 90° R,6 Sector for Cycle 10 3-7 3.1-5 Reactor Geometry Showing a 90° R,6 Sector for Cycle 11 3-8 3.1-6 Internal Surveillance Capsule Geometry 3-9 6.1-1 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position Along the 16° Traverse in the Reactor Cavity 6-21 e Short Gradient Chains - Cycle 8 Irradiation 6.1-2 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-22 Along the 24° Traverse in the Reactor Cavity Short Gradient Chains - Cycle 8 Irradiation 6.1-3 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-23 Along the 26° Traverse in the Reactor Cavity Short Gradient Chains - Cycle 8 Irradiation 6.1-4 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-24 Along the 39° Traverse in the Reactor Cavity Short Gradient Chains - Cycle 8 Irradiation 6.1-5 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-25 Along the 0° Traverse in the Reactor Cavity Long Gradient Chains - Cycle 8 Irradiation 9.

vi

LIST OF FIGURES e Figure Title Page 6.1-6 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-26 Along the 20° Traverse in the Reactor Cavity Long Gradient Chains - Cycle 8 Irradiation 6.1-7 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-27 Along the 30° Traverse in the Reactor Cavity Long Gradient Chains - Cycle 8 Irradiation 6.2-1 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-49 Along the 6° Traverse in the Reactor Cavity Short Gradient Chains - Cycle 9 Irradiation 6.2-2 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-50 Along the 16° Traverse in the Reactor Cavity Short Gradient Chains - Cycle 9 Irradiation 6.2-3 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-51 Along the 24° Traverse in the Reactor Cavity Short Gradient Chains - Cycle 9 Irradiation 6.2-4 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-52 Along the 26° Traverse in the Reactor Cavity Short Gradient Chains - Cycle 9 Irradiation 6.2-5 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-53 Along the 39° Traverse in the Reactor Cavity Short Gradient Chains - Cycle 9 Irradiation 6.2-6 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-54 Along the 0° Traverse in the Reactor Cavity Long Gradient Chains - Cycle 9 Irradiation 6.2-7 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-55 Along the 10° Traverse in the Reactor Cavity Long Gradient Chains - Cycle 9 Irradiation 6.2-8 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-56 Along the 20° Traverse in the Reactor Cavity Long Gradient Chains - Cycle 9 Irradiation vii

LIST OF FIGURES Figure Title Page 6.2-9 Fast Neutron Flux (E > 1.0 Me V) as a Function of Axial Position 6-57 Along the 30° Traverse in the Reactor Cavity Long Gradient Chains - Cycle 9 Irradiation * .

6.3-1 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-81 Along the 6° Traverse in the Reactor Cavity Short Gradient Chains - Cycles 10/11 Irradiation 6.3-2 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-82 Along the 16° Traverse in the Reactor Cavity Short Gradient Chains - Cycles 10/11 Irradiation 6.3-3 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-83 Along the 24° Traverse in the Reactor Cavity Short Gradient Chains - Cycles 10/11 Irradiation 6.3-4 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-84 Along the 26° Traverse in the Reactor Cavity Short Gradient Chains - Cycles 10/11 Irradiation 6.3-5 Fast Neutron Flux (E > 1.0 Me V) as a Function of Axial Position 6-85 Along the 36° Traverse in the Reactor Cavity Short Gradient Chains - Cycles 10/11 Irradiation 6.3-6 Fast Neutron Flux (E > 1.0 Me V) as a Function of Axial Position 6-86 Along the 39° Traverse in the Reactor Cavity Short Gradient Chains - Cycles 10/11 Irradiation_

6.3-7 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-87 Along the 0° Traverse in the Reactor Cavity Long Gradient Chains - Cycles 10/11 Irradiation 6.3-8 Fast Neutron Flux (E > 1.0 Me V) as a Function of Axial Position 6-88 Along the 10° Traverse in the Reactor Cavity Long Gradient Chains - Cycles 10/11 Irradiation 6.3-9 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position 6-89 Along the 20° Traverse in the Reactor Cavity Long Gradient Chains - Cycles 10/11 Irradiation viii

LIST OF FIGURES Figure Title ~

6.3-10 Fast Neutron Flux (E > 1.0 Me V) as a Function of Axial Position 6-90 Along the 30° Traverse in the Reactor Cavity Long Gradient Chains - Cycles 10/11 Irradiation ix

Table 4.1-1 Title LIST OF TABLES Calculated Neutron Energy Spectra at Reactor Cavity Sensor Set

~

4-3 Locations for Cycle 8

\..

4.1-2 Calculated Neutron Energy Spectra at Reactor Cavity Sensor Set 4-5 Locations for Cycle 9 4.1-3 Calculated Neutron Energy Spectra at Reactor Cavity Sensor Set 4-7 Locations for Cycles 10/11 4.1-4 Calculated Neutron Energy Spectra at Surveillance 4-9 Capsule Center 4.1-5 Azimuthal Variation of Neutron Flux and dpa/sec at the 4-11 Reactor Vessel Clad-Base Metal Interface - Cycle 1 4.1-6 Azimuthal Variation of Neutron Flux and dpa/sec at the 4-12 Reactor Vessel Clad-Base Metal Interface - Cycle 2 4.1-7 4.1-8 Azimuthal Variation of Neutron Flux and dpa/sec at the Reactor Vessel Clad-Base Metal Interface - Cycle 3 Azimuthal Variation of Neutron Flux and dpa/sec at the 4-13 4-14 Reactor Vessel Clad-Base Metal Interface - Cycle 4 4.1-9 Azimuthal Variation of Neutron Flux and dpa/sec at the 4-15 Reactor Vessel Clad-Base Metal Interface - Cycle 5 4.1-10 Azimuthal Variation of Neutron Flux and dpa/sec at the 4-16 Reactor Vessel Clad-Base Metal Interface - Cycle 6 4.1-11 Azimuthal Variation of Neutron Flux and dpa/sec at the 4-17 Reactor Vessel Clad-Base Metal Interface - Cycle 7 4.1-12 Azimuthal Variation of Neutron Flux and dpa/sec at the 4-18 Reactor Vessel Clad-Base Metal Interface - Cycle 8 4.1-13 Azimuthal Variation of Neutron Flux and dpa/sec at the 4-19 Reactor Vessel Clad-Base Metal Interface - Cycle 9 x

LIST OF TABLES e Table Title Page 4.1-14 Azimuthal Variation of Neutron Flux and dpa/sec at the 4-20 Reactor Vessel Clad-Base Metal Interface - Cycle 10 4.1-15 Azimuthal Variation of Neutron Flux and dpa/sec at the 4-21 Reactor Vessel Clad-Base Metal Interface - Cycle 11 4.1-16 Relative Radial Distribution of Neutron Flux (E > 1.0 Me V) 4-22 Through the Reactor Vessel Wall - Cycle 11 4.1-17 Relative Radial Distribution of Neutron Flux (E > 0.1 Me V) 4-23 Through the Reactor Vessel Wall - Cycle 11 4.1-18 Relative Radial Distribution of Iron Atom Displacement 4-24 Rate (dpa/sec) Tirrough the Reactor Vessel Wall - Cycle 11 5.1-1 Summary of Reaction Rates Derived from Multiple Foil Sensor 5-3 Sets Withdrawn from Internal Surveillance Capsules

-- 5.2-1 5.2-2 Derived Exposure Rates from Surveillance Capsule A240 Dosimetry Withdrawn at the End of Fuel Cycle 2 Derived Exposure Rates from Surveillance Capsule W290 5-4 5-5 Dosimetry Withdrawn at the End of Fuel Cycle 5 5.2-3 Derived Exposure Rates from Surveillance Capsule W290-9 5-6 Dosimetry Withdrawn at the End of Fuel Cycle 9 5.2-4 Derived Exposure Rates from Surveillance Capsule Wl 10 5-7 Dosimetry Withdrawn at the End of Fuel Cycle 10 6.1-1 Summary of Reaction Rates Derived from Multiple Foil Sensor Sets 6-4 Cycle 8 Irradiation 54 6.1-2 Fe (n,p) Reaction Rates Derived from the Stainless Steel 6-5 Short Gradient Chains - Cycle 8 Irradiation 58 6.1-3 Ni (n,p) Reaction Rates Derived from the Stainless Steel 6-6 Short Gradient Chains - Cycle 8 Irradiation xi

LIST OF TABLES Table Title Page 59 6.1-4 Co (n, y) Reaction Rates Derived from the Stainless Steel 6-7 Short Gradient Chains - Cycle 8 Irradiation 54 6.1-5 Fe (n,p) Reaction Rates Derived from the Stainless Steel 6-8 Long Gradient Chains - Cycle 8 Irradiation 58 6.1-6 Ni (n,p) Reaction Rates Derived from the Stainless Steel 6-9 Long Gradient Chains - Cycle 8 Irradiation 59 6.1-7 Co (n, y) Reaction Rates Derived from the Stainless Steel 6-10 Long Gradient Chains - Cycle 8 Irradiation 6.1-8 Derived Exposure Rates from the Capsule B Dosimetry Evaluation 6-11 16° Azimuth - 280° Reference - Core Midplane 6.1-9 Derived Exposure Rates from the Capsule D Dosimetry Evaluation 6-12 26° Azimuth - 290° Reference - Core Midplane 6.1-10 Derived Exposure Rates from the Capsule E Dosimetry Evaluation 6-13 26° Azimuth - 290° Reference - Core Bottom 6.1-11 Derived Exposure Rates from the Capsule G Dosimetry Evaluation 6-14 39° Azimuth - 315° Reference - Core Midplane 6.1-12 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial 6-15 Position Within the Reactor Cavity Short Gradient Chains - Cycle 8 Irradiation 6.1-13 Fast Neutron Flux (E > 0.1 MeV) as a Function of Axial 6-16 Position Within the Reactor Cavity Short Gradient Chains - Cycle 8 Irradiation 6.1-14 Iron Atom Displacement Rate as a Function of Axial 6-17 Position Within the Reactor Cavity Short Gradient Chains - Cycle 8 Irradiation 6.1-15 Fast Neutron Flux (E > 1.0 Me V) as a Function of Axial 6-18 Position Within the Reactor Cavity Long Gradient Chains - Cycle 8 Irradiation xii

LIST OF TABLES e Table Title Page 6.1-16 Fast Neutron Flux (E > 0.1 MeV) as a Function of Axial 6-19 Position Within the Reactor Cavity Long Gradient Chains - Cycle 8 Irradiation 6.1-17 Iron Atom 'Displacement Rate as a Function of Axial 6-20 Position Within the Reactor Cavity Long Gradient Chains - Cycle 8 Irradiation 6.2-1 Summary of Reaction Rates Derived from Multiple Foil Sensor Sets 6-30 Cycle 9 Irradiation 6.2-2 S4pe (n,p) Reaction Rates Derived from the Stainless Steel 6-31 Short Gradient Chains - Cycle 9 Irradiation 58 6.2-3 Ni (n,p) Reaction Rates Derived from the Stainless Steel 6-32 Short Gradient Chains - Cycle 9 Irradiation

• -* 6.2-4 6.2-5 59 Co (n, y) Reaction Rates Derived from the Stainless Steel Short Gradient Chains - Cycle 9 Irradiation 54 Fe (n,p) Reaction Rates Derived from the Stainless Steel Long Gradient Chains - Cycle 9 Irradiation 6-33 6-34 58 6.2-6 Ni (n,p) Reaction Rates Derived from the Stainless Steel 6-35 Long Gradient Chains - Cycle 9 Irradiation 59 6.2-7 Co (n,y) Reaction Rates Derived froin the Stainless Steel 6-36 Long Gradient Chains - Cycle 9 Irradiation 6.2-8 .. Derived Exposure Rates from the Capsule A Dosimetry Evaluation 6-37 6° Azimuth - 270° Reference - Core Midplane 6.2-9 ' Derived Exposure Rates from the Capsule C Dosimetry Evaluation 6-38 6° Azimuth - 270° Reference - Core Bottom 6.2-10 Derived Exposure Rates from the Capsule J Dosimetry Evaluation 6-39 16° Azimuth - 280° Reference - Core Midplane xiii

LIST OF TABLES Table Title Page 6.2-11 Derived Exposure Rates from the Capsule K Dosimetry Evaluation 6-40 26° Azimuth - 290° Reference - Core Midplane 6.2-12 Derived Exposure Rates from the Capsule L Dosimetry Evaluation 6-41 26° Azimuth - 290° Reference - Core Bottom 6.2-13 Derived Exposure Rates from the Capsule N Dosimetry Evaluation 6-42 39° Azimuth - 315° Reference - Core Midplane 6.2-14 Fast Neutron Flux (E > 1.0 Me V) as a Function of Axial 6-43 Position Within the Reactor Cavity Short Gradient Chains - Cycle 9 Irradiation 6.2-15 Fast Neutron Flux (E > 0.1 Me V) as a Function of Axial 6-44 Position Within the Reactor Cavity Short Gradient Chains - Cycle 9 Irradiation 6.2-16 Iron Atom Displacement Rate as a Function of Axial 6-45 Position Within the Reactor Cavity Short Gradient Chains - Cycle 9 Irradiation 6.2-17 Fast Neutron Flux (E > 1.0 Me V) as a Function of Axial 6-46 Position Within the Reactor Cavity Long Gradient Chains - Cycle 9 Irradiation 6.2-18 Fast Neutron Flux (E > 0.1 Me V) as a Function of Axial 6-47 Position Within the Reactor Cavity Long Gradient Chains - Cycle 9 Irradiation 6.2-19 Iron Atom Displacement Rate ,as a Function of Axial 6-48 Position Within the Reactor Cavity Long Gradient Chains - Cycle 9 Irradiation 6.3-1 Summary of Reaction Rates Derived from Multiple Foil Sensor Sets 6-61 Cycles 10/11 Irradiation 54 6.3-2 Fe (n,p) Reaction Rates Derived from the Stainless Steel 6-62 Short Gradient Chains - Cycles 10/11 Irradiation xiv

LIST OF TABLES Table Title Page

. 6.3-3 58 Ni (n,p) Reaction Rates Derived from the Stainless Steel 6-63 Short Gradient Chains - Cycles 10/11 Irradiation 59 6.3-4 Co (n, y) Reaction Rates Derived from the Stainless Steel 6-64 Short Gradient Chains - Cycles 10/11 Irradiation 54 6.3-5 Fe (n,p) Reaction Rates Derived from the Stainless Steel 6-65 Long Gradient Chains - Cycles 10/11 Irradiation 58 6.3-6 Ni (n,p) Reaction Rates Derived from the Stainless Steel 6-66 Long Gradient Chains - Cycles 10/11 Irradiation 59 6.3-7 Co (n,y) Reaction Rates Derived from the Stainless Steel 6-67 Long Gradient Chains - Cycles 10/11 Irradiation 6.3-8 Derived Exposure Rates from the Capsule 0 Dosimetry Evaluation 6-68 6° Azimuth - 270° Reference - Core Midplane

• 6.3-9 6.3-10 Derived Exposure Rates from the Capsule P Dosimetry Evaluation 16° Azimuth - 280° Reference - Core Midplane Derived Exposure Rates from the Capsule Q Dosimetry Evaluation 6-69 6-70 16° Azimuth - 280° Reference - Core Bottom 6.3-11 Derived Exposure Rates from the Capsule R Dosimetry Evaluation 6-71 26° Azimuth - 290° Reference - Core Midplane 6.3-12 Derived Exposure Rates from the Capsule S Dosimetry Evaluation 6-72 36° Azimuth - 300° Reference - Core Midplane 6.3-13 Derived Exposure Rates from the Capsule T Dosimetry Evaluation 6-73 39° Azimuth - 315° Reference - Core Midplane 6.3-14 Derived Exposure Rates from the Capsule U Dosimetry Evaluation 6-74 24° Azimuth - 330° Reference - Core Midplane

• 6.3-15 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial 6-75

• Position Within the Reactor Cavity Short Gradient Chains - Cycles 10/11 Irradiation xv

LIST OF TABLES Table Title ~

6.3-16 Fast Neutron Aux (E > 0.1 MeV) as a Function of Axial 6-78 Position Within the Reactor Cavity Short Gradient Chains - Cycles 10/11 Irradiation 6.3-17 Iron Atom Displacement Rate as a Function of Axial 6-79 Position Within the Reactor Cavity Short Gradient Chains - Cycles 10/11 Irradiation 6.3-18 Fast Neutron Aux (E > 1.0 MeV) as a Function of Axial 6-80 Position Within the Reactor Cavity Long Gradient Chains - Cycles 10/11 Irradiation 6.3-19 Fast Neutron Aux (E > 0.1 MeV) as a Function of Axial 6-81 Position Within the Reactor Cavity Long Gradient Chains - Cycles 10/11 Irradiation 6.3-20 Iron Atom Displacement Rate as a Function of Axial 6-82 7.1-1 Position Within the Reactor Cavity Long Gradient Chains - Cycles 10/11 Irradiation Comparison of Measured and Calculated Exposure Rates from Surveillance Capsule and Cavity Dosimetry Irradiations 7-3 7.2-1 Comparison of Measured and Calculated Neutron Sensor 7-6 Reaction Rates from Surveillance Capsule and Cavity Dosimetry Irradiations 8.1-1 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure 8-3 Projections for the Beltline Region of the Palisades Reactor Vessel Through Cycle 11

• 8.1-2 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure 8-4 Projections for the Beltline Region of the Palisades Reactor Vessel Through Cycle 11 8.1-3 Summary of Best Estimate Iron Atom Displacement (dpa) Exposure 8-5 Projections for the Beltline Region of the Palisades Reactor Vessel Through Cycle 11

• xvi

LIST OF TABLES Table Title Page 8.2-1 Summary of Tolerance Uncertainty Component for 8-8

'l>(E > 1.0 MeV) Projections at the Vessel Wall xvii

SECTION 1 OVERVIEW OF THE PROGRAM The Reactor Cavity Neutron Measurement Program 11 *3l initiated at Palisades at the start of Fuel Cycle 8 was designed to provide a mechanism for the long term continuous monitoring of the neutron exposure of those portions of the reactor vessel and vessel support structure which may experience radiation induced increases in reference nil ductility transition temperature (RT NDT) over the nuclear power plant lifetime. When used in conjunction with dosimetry from internal surveillance capsules14-7J and with the results of neutron transport calculations, the reactor cavity neutron dosimetry provides an extensive plant specific measurement data base that can be used to determine the best estimate neutron exposure of the reactor vessel and to project embrittlement gradients through the vessel wall with a minimum uncertainty. Minimizing the uncertainty in the neutron exposure projections will, in turn, help to assure that the reactor can be operated in the least restrictive mode possible with respect to:

1- 10CFR50 Appendix G pressure/temperature limit curves for normal heatup and cooldown of the reactor coolant system, 2- Emergency. Operating Procedure (EOP) pressure/temperature limit curves, and 3- Pressurized Thermal Shock (PTS) RTPTS screening criteria.

In addition, an accurate measure of the neutron exposure of the reactor vessel and support structure can provide a sound basis for requalification should operation of the plant beyond the current design and/or licensed lifetime prove to be desirable.

In the assessment of the state of embrittlement of light water reactor vessels, an accurate evaluation of the neutron exposure of the materials comprising the beltline region of the vessel is required. This exposure evaluation must, in general, include assessments not only at locations of maximum exposure at the inner diameter of the vessel, but, also, as a function of axial, azimuthal, and radial location throughout the vessel wall.

In order to satisfy the requirements of 10CFR50 Appendix G for the calculation of pressure/temperature limit curves for normal heatup and cooldown of the reactor coolant system,

• fast neutron exposure levels must be defined at depths within the vessel wall equal to 25 and 75 1-1

percent of the wall thickness for each of the materials comprising the beltline region. These locations are commonly referred to as the 1/4T and 3/4T positions in the vessel wall. The 1/4T exposure levels are also used in the determination of upper shelf fracture toughness as specified in 10CFR50 Appendix G.

In the determination of values of RTPTS for comparison with applicable pressurized thermal shock screening criteria for plates and circumferential welds, maximum neutron exposure levels experienced by each of the beltline materials are required. These maximum levels will, of course, occur at the vessel inner radius.

In the event that a probabalistic fracture mechanics evaluation of the reactor vessel is performed, or if an evaluation of thermal annealing and subsequent material re-embrittlement is undertaken, a complete embrittlement profile is required for the entire volume of the reactor vessel beltline.

The determination of this embrittlement profile would, in turn, necessitate the evaluation of neutron exposure gradients throughout the entire beltline.

The meth<?dology used to provide these required best estimate neutron exposure evaluations for the Palisades reactor vessel is based on the underlying philosophy that, in order to minimize the uncertainties associated with vessel exposure projections, plant specific neutron transport calculations must be supported by benchmarking of the analytical approach, comparison with industry wide power reactor data bases of surveillance capsule and reactor cavity dosimetry, and, ultimately, by validation with plant specific suiveillance capsule and reactor cavity dosimetry data bases.

That is, as a progression is made from the use of a purely analytical approach tied to experimental benchmarks to an approach that makes use of industry and plant specific power reactor measurements to remove potential biases in the analytical method, knowledge regarding the neutron environment applicable to a specific reactor vessel is increased and the uncertainty associated with vessel exposure projections is minimized.

With this overall methodology in mind, the Reactor Cavity Measurement Program was established to meet the following objectives:

1- Provide a measurement data base sufficient to:

a) remove biases that may be present in analytical predictions of neutron exposure;

.and 1-2

b) support the methodology for the projection of exposure gradients through the thickness of the reactor vessel wall.

2- Establish uncertainties in the best estimate fluence projections for the reactor vessel wall.

3- Provide a long term continuous monitoring capability for the beltline region of the reactor vessel.

The reactor cavity dosimetry program has currently been discontinued. It is anticipated that future dosimetry information will be from the in-vessel surveillance capsules.

This report provides the results of neutron dosimetry evaluations performed subsequent to the completion of Fuel Cycle 11. Fast neutron exposure in terms of fast neutron fluence (E > 1.0 MeV) and dpa is established for all measurement locations in the reactor cavity. The analytical formalism describing the relationship among the measurement points and locations within the reactor vessel wall is described and used to project the exposure of the vessel itself.

-e Results of exposure evaluations from surveillance capsule dosimetry withdrawn at the end of Fuel Cycles 2, 5, 9 and 10 as well as reactor cavity dosimetry results from Cycles 8 and 9 are incorporated to provide the integrated exposure of the reactor vessel from plant startup through the end of Cycle 11. Also, uncertainties associated with the derived exposure parameters at the measurement locations and with the projected exposure of the reactor vessel are provided.

All of the calculations and qosimetry evaluations presented in this report are intended to meet the requirements of Draft Regulatory Guide DG-1025, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence"; and, have been based on the latest available nuclear cross-section data derived from ENDF/B-VI. As such, the data provided here are intended to supersede prior evaluations documented in References 1 through 7.

In addition to the dosimetry evaluations, an analysis of the neutron and gamma ray heat deposition in the biological shield wall for Cycle 11 is presented. A detailed description of the analysis and results are provided in Appendix E of this report.

• 1-3

SECTION 2 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) 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 envirorunent over the regions of interest.

Placement of the multiple foil sensor sets was such that spectra evaluations could be made at various 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 reactor vessel. Due to the irregular shape of the reactor core,

• water thickness varies significantly as a function of azimuthal angle. Additional sensor sets were positioned opposite the bottom of the active core at an azimuthal angle of a longitudinal reactor vessel weld (270 degrees) and at a particular azimuthal angle (290 degrees) of an in-vessel surveillance capsule, and at a particular azimuthal angle (280 degrees) of the peak azimuthal flux.

Here the intent was to measure variations in neutron spectra over the core height. At each of the azimuthal locations selected for core midplane spectra measurements, gradient chains extended over the bottom half of the active fuel. Additional gradient chains were placed at symmetric locations in other quadrants to confirm symmetry of the neutron flux distribution. These additional gradient chains extended over nearly the full height of the active fuel (except for the bottom-most nine inches) and extended up past the reactor vessel support structure.

• The sensor set deployment described in the preceding paragraphs is characteristic of the basic long term monitoring program designed to provide fast neutron exposure assessments for materials comprising the beltline region of the reactor vessel. The reactor cavity dosimetry program has currently been discontinued. It is anticipated that future dosimetry informatio.n will be from the in-vessel surveillance capsules.

• 2-1

2.1.1 Sensor Placement in the Reactor Cavity A detailed description of the reactor cavity dosimetry hardware and plant specific installation can be found in References 1 through 3. However, the following information is provided in this report to orient the reader to the plant geometry and the specifics of the sensor sets.

The placement of the individual multiple foil sensor sets and gradient chains within the reactor cavity is illustrated in Figure 2.1-1. In Figure 2.1-1 a plan view of the azimuthal locations of the strings of sensor sets is depicted along with the azimuthal locations of the gradient chains.

The strings were located at azimuthal positions of 270, 280, 290, 300, 315 and 330 degrees relative to the core cardinal axes. The sensor strings were hung in the annular gap between the reactor vessel insulation and the primary biological shield at a nominal radius of 108 inches relative to the core centerline and were hung from an aluminum support bar supported from the reactor cavity seal drip pan. Note that it has been determined that the dosimetry support bar was

-skewed radially (the 270 degree end of the bar was closer to the reactor vessel 16 ~ thus, the bar shifted radial and azimuthal positions of the dosimetry are shown below.

Reference First Octant Bar Dosimeter Azimuth Eguivalent Shifted Angle Radius (in) 270° oo 60 100.7 280° 100 16° 101.6 290° 20° 26° 102.5 300° 30° 36° 103.5 315° 45° 39° 105.3 330° 30° 24° 107.6 The gradient chains, located at azimuthal positions of 30, 90, 150, 210, 260 and 340 degrees, were also located in the annular gap between the reactor vessel insulation and the primary biological shield at a nominal radius of 108 inches relative to the core centerline and were supported from the reactor cavity seal drip pan.

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 2.1-2. Each capsule included three compartments to hold the neutron sensors. The top e

2-2

compartment (position 1) contained the . bare radiometric monitors, whereas, the two remaining compartments (positions 2 and 3) housed the cadmium shielded packages. The separation between positions 1 and 2 was such that cadmium shields inserted into position 2 did not introduce perturbations in the thermal flux in position 1. Aluminum 6061 was selected for the dosimeter capsules in order to minimize neutron flux perturbations at the sensor set locations as well as to limit the radiation levels associated with post-irradiation shipping and handling of the capsules. A summary of the contents of the multiple foil capsules used during each cycle of irradiation is provided in the appendices to this report.

2.1.3 Description of Gradient Chains Along with the multiple foil sensor sets placed at discrete locations within the reactor cavity, gradient chains were employed to obtain axial variations of fast neutron exposure along each of the twelve traverses. Subsequent to irradiation these gradient chains were removed from the reactor cavity and segmented to provide neutron reaction rate measurements at six-inch to 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 gradient chains yielded activation results for the 54Fe(n,p )54Mn, 58Ni(n,p )58Co, and 59Co(n, y ) 60Co 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 a check on the chemical analysis of the Type 304 stainless steel.

• 2-3

F1GURE 2.1-1 AZIMUTHAL LOCATION OF SENSOR STRINGS 260° Inlet Outlet 150° * +

Outlet Inlet Fuel North Transfer TubP. Note: The dosimetry bar has been dete"!'ined to be shifted six dP.grees clockwise relative to the position shown above.

2-4

FIGURE 2.1-2 IRRADIATION CAPSULE FOR REACTOR CAVITY SENSOR SETS

• '!Q .+/-.Cit - - - - - -

1 1.ocd:.* 01

• 1.z.d:..o~

~.so!.01 1.oot..01 1u-.i!Q.*:::,

.~t.01

.'12.:1:.ol "T,P T'f,.

• l't t.01

...................-.~1+/-.01 "T"fP

.cei. !.011--1--t--~ ._..__._ . ~a +/- .os

'T'CP C:il<OUP - 01 2-5

2.2 Description of Surveillance Capsule Dosimetry Over the course of the first 10 fuel cycles at Palisades, three materials surveillance capsules were withdrawn from their positions. One from the outer surface of the core support barrel and the other two from the inner surface of the reactor vessel. The neutron dosimetry contained within these capsules provided a measure of the integral exposure received by each of the capsules during its respective irradiation period; and established a measurement continuity between the initial startup of the reactor and the initiation of the Reactor Cavity Measurement Program. The specific withdrawal dates of these three capsules were as shown below.

Capsule A240 End of Cycle 2 01n8 Capsule W290 End of Cycle 5 08/83 Capsule W 110 End of Cycle 10 06/93 An additional capsule was installed at the W290 location for irradiation in Cycle 9 only. This capsule had the same external configuration as the others with minor design modifications to facilitate remote installation. Figure 2.2-1 shows a typical Palisades in-vessel surveillance capsule assembly geometry.

The type and location of the neutron sensors included in the materials surveillance program are described in some detail in References 4 through 7. Specific information pertinent to the individual sensor sets included in Capsules A240, W290, Wl 10 and W290-9 is provided in the appendices to* this report.

2-6

FIGURE 2.2-1 TYPICAL SURVEILLANCE CAPSULE Lock Assembly

} WedgcCouplingA=mb!Y

} &1ooskm Assembly Tensile-Monitor Compattmem _ _..

Tensile-Monitor Compartmem Tensile-Monitor Companment 2-7

SECTION 3 NEUTRON TRANSPORT AND DOSIMETRY EVALUATION METHODOLOGIES As noted in Section 1 of this report, the best estimate exposure of the reactor vessel was developed using a combination of absolute plant specific neutron transport calculations and plant specific measurements from the reactor cavity and internal surveillance capsules. In this section, the neutron transport and dosimetry evaluation methodologies are discussed in some detail; and the approach used to combine the calculations and measurements to produce the best estimate vessel exposure is presented.

3.1 Neutron Transport Analysis Methods Fast neutron exposure calculations for the reactor and reactor cavity geometry were carried out using forward discrete ordinates transport techniques. By using fuel cycle specific core power distributions these calculations provided the energy distribution of neutrons for use as input to neutron dosimetry evaluations as well as for use in relating measurement results to the actual exposure at key locations in the reactor vessel wall. In addition, they established the means to compute absolute exposure rate values, thus, providing a direct comparison with all dosimetry results obtained over the operating history of the reactor.

Incorporating, the cycle specific data to derive neutron energy spectra distributions from the forward calculation provided the means to:

1- Evaluate neutron dosimetry from reactor cavity and surveillance capsule locations.

2- Enable a direct comparison of analytical prediction with measurement 3- Determine plant specific bias factors to be used in the evaluation of the best estimate exposure of the reactor vessel.

4 - Establish a mechanism for projection of reactor vessel exposure as the design of each new fuel cycle evolves .

• 3-1

A plan view of the reactor geometry at the core midplane *elevation is shown in Figures 3.1-1 through 3.1-5. Figure 3.1-1 shows the model used for Cycles 1 and 2 which includes the

• accelerated 30 degree surveillance capsule location in addition to the 20 degree wall capsule.

The Cycles 3-7 and 9 did not have the accelerated capsule position as shown in Figure 3.1-2.

Cycles 8, 10 and 11 differed in that the fuel region included stainless steel pins in place of actual fuel pins at various locations as depicted in Figures 3.1-3 through 3.1-5, respectively.

The reactor exhibits 1A core symmetry, thus, a 0-90 degree sector is depicted to evaluate the

• maximum flux at the reactor vessel. This 45-90 degree sector does not contain the in-vessel capsules. In addition to the core, reactor internals, reactor vessel, and the primary biological shield, the model also included explicit representations of the surveillance capsules, the reactor vessel cladding, and the mirror insulation located external to the vessel.

The models depicted in Figures 3.1-1 through 3.1-5 were developed using nominal design dimensions for all components. Specified tolerances in the design dimensions are reflected in the overall uncertainty assessments associated with projected neutron exposures. This modeling approach is consistent with the guidelines of DG-1025.

A description of a single surveillance capsule attached to the core support barrel or the reactor vessel cladding is shown in Figure 3.1-6. From a neutronic standpoint, the inclusion of the surveillance capsules and associated support structures in the analytical model is significant.

Since the presence of the capsules and structure has a marked impact on the magnitude of the neutron flux as well as on the relative neutron energy spectra at dosimetry locations within the capsules, a meaningful comparison of measurement and calculation can be made only if these perturbation effects are properly accounted for in the analysis.

In contrast to the relatively massive stainless steel and carbon steel structures associated with the internal surveillance capsules, the small aluminum capsules used in the reactor cavity measurement program were designed to minimize perturbations in the neutron flux and, thus, to

.provide free field data at the measurement locations. Therefore, explicit modeling of these small capsules in the forward transport models was not required.

The forward transport calculations for the reactor model depicted in Figures 3.1-1 through 3.1-5 were carried out in R,6 geometry using the DORT two-dimensional discrete ordinates transport theory code181 and the BUGLE-93 cross-section library 191* The BUGLE-93 library is a 47 neutron and 20 gamma ray energy group, ENDF/B-VI based, data set produced specifically for light water

. reactor applications. In these analyses, anisotropic scattering was treated with a P 3 expansion of 3-2

e the scattering cross-sections and the angular discretization was modeled with an S 16 order of angular quadrature.

The forward calculations were normalized to a core midplane power density and for operation at a thermal power level of 2530 MWt. The spatial core power distributions utilized in the forward calculations were supplied by Consumers Power1101* The power distributions were supplied as pin-by-pin power distributions, initial enrichments, and cycle burnups for each fuel assembly in the quadrant. The neutron source was derived for each fuel pin and for each assembly using burnup dependent values of the fission neutron energy spectrum, neutrons per fission, and energy per fission evaluated at the mean assembly burnup value. The source spectrum was calculated by determining the fraction of fissions occurring in each of the important uranium and plutonium isotopes for the mid-cycle burnup and calculating the resultant average fission spectrum using the ENDF/B-VI fission spectrum for each isotope.

The source from each fuel assembly was spatially located to take into account the varying gaps between fuel assemblies and thus represents the location of source from each pin as accurately as possible. The source was converted from the X-Y pin geometry to the R-0 DORT geometry by distributing the source over a square area equal to the pitch for each pin.

• Details of the fuel assembly locations, core geometry, and other reactor parameters were taken from Reference 6. The core average, by-pass, and inlet coolant temperatures were taken from Reference 10.

The results of the DORT calculations are given in Section 4 of this report.

• 3-3

FIGURE 3.1-1 REACTOR GEOMETRY SHOWING A 90° R,8 SECTOR FOR CYCLES 1 AND 2 EGADS 1.1 X1995/12/12 7975913619452 Palisades RT Cycles 1&2 0

0 0

0 0

0 N

E 0

0 N

0

~

0

~

0 0

r, cm 3-4

FIGURE 3.1-2 REACTOR GEOMETRY SHOWING A 90° R,8 SECTOR FOR CYCLES 3-7 AND 9 EGADS 1.1 X1995/12/12 5745279759137 Palisades RT Cycles 3- 7,9 0

0 0

0 0

0 N

E u

0 0

N 0

~

0

~

0 0 50 100 150 200 250 300 350 400 r, cm 3-5

FIGURE 3.1-3 REACTOR GEOMETRY SHOWING A 90° R,8 SECTOR FOR CYCLE 8 EGADS 1.1 X1995/12/12 2645279759137 Palisades RT Cycle 8 0

0 0

0 0

0 N

E 0

0 N

0

~

0 0

0 50 100 150 200 250 300 350 400 r, cm 3-6

FIGURE 3.1-4 REACTOR GEOMETRY SHOWING A 90° R,8 SECTOR FOR CYCLE 10 EGADS 1.1 X1995/12/12 7591389645279 Palisades RT Cycle 10 0

0 0

0 0

0 N

• E 0

0 N

0

~

0

~

0 0 50 100 150 200 250 300 350 400 r, cm

• 3-7

FIGURE 3.1-5 REACTOR GEOMETRY SHOWING A 90° R,8 SECTOR FOR CYCLE 11

• EGADS 1.1 X1995/12/12 3828452797591 Palisades RT Cycle 11 0

0 0

0 0

0 E

(,.)

0 0

0

~

0

~

0 r, cm 3-8

AGURE 3.1-6 INTERNAL SURVEILLANCE CAPSULE GEOMETRY J.81 CM

3. ~ 8 CM 1

• 91 CM .

CAPSULE LOCATIONS ANGLE - DISTANCE 10 -*215.43 CM 20 - 215.43 CM JO - 196*. 06 CM "'

(A

• • DISTANCE IS.FROM-CORE CENTER TO CAPSULE BLOCK -

CD n

ii:

.CD (0

0 iC

"'WI n

s::

CAPSULE CENT.ER -

WATER GAP HOLDER TUSE 3-9

3.2 Neutron Dosimetry Evaluation Methodology The use of passive 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 the case of internal surveillance capsule A240, the specific activities of the multiple foil sensor set were determined from prior analysis performed by Battelle Columbus Laboratory 141*

In this case, the source of the measured specific activity data was referenced and the remainder

.*of the data evaluation proceeded using the methodology described in this section.

3.2.1 Determination of Sensor Reaction Rates Following irradiation, the multiple foil sensor sets from surveillance capsule and reactor 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.

3-10

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 111 through 211* 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 reactor 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 238U and 237Np fission monitors from internal surveillance capsules, by direct counting preceded by dissolution and chemical separation of cesium from the sensor. For the stainless steel gradient chains used in the reactor cavity irradiations, individual sensors were obtained by cutting the chains into a series of segments to provide data points at six-inch to one-foot intervals. For the long gradient chains, the data points encompass an axial span from 4.5 feet below the core midplane to 8.0 feet above the core midplane. For the short gradient chains, which are attached to the support bar, the data points encompass an axial span from 5.5 feet below the core midplane to 0.5 feet above the core

.midplane.

The irradiation history of the reactor over the first ten operating cycles was obtained from

• NUREG-0020, "Licensed Operating Reactors Status Summary Report" and the irradiation history of Cycle 11 was supplied by Consumers Power1221* In particular, operating data were extracted from these reports on a monthly basis from reactor startup to the end of the current evaluation period. For the sensor sets utilized in surveillance capsule and reactor cavity irradiations, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations.

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 2530 MWt were determined from the following equation:

p -l.I -Al N F Y 0

.EJ _!_ C (1 - e 1 e "

p j ref 3-11

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 Pref (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 <j>(E > 1.0 MeV) during irradiation period j to the time weighted average <j>(E > 1.0 Me V) over the entire irradiation period.

A = decay constant of the product isotope (sec* 1).

~ = 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/Pref accounts for month by month variation of power level

• within a given fuel cycle. The ratio Cj is calculated for each fuel cycle using the forward transport methodology and accounts for the change in sensor reaction rates caused by variations in flux level due to changes in core power spatial distributions from fuel cycle to fuel cycle. For a single cycle irradiation Ci = 1.0. However, for multiple cycle irradiations, particularly those employing low leakage fuel management the additional Cj correction must be utilized.

3.2.1.2 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 23 8u foil measurements to account for the presence of 23 5u impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation. These corrections were location and fluence dependent and were derived from the data obtained from the forward transport calculations ..

In addition to the corrections made for the presence of 235u in the 2 38u fission sensors, corrections 238 237 were also made to both the U and Np sensor reaction rates to account for gamma ray induced fission reactions occurring over the course of the irradiation. These photo-fission 3-12

corrections were, likewise, location dependent and were also .based on the forward transport calculations described in Section 3.1.1.

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 code1231* The FERRET approach used the measured reaction rate data, sensor reaction cross-sections, and a calculated trial spectrum as input and proceeded to adjust the group fluxes from the trial spectrum to produce a best fit (in a least squares sense) to the measured reaction rate data. The "measured" exposure parameters along with the associated uncertainties were then obtained from the adjusted spectrum.

In the FERRET evaluations, a log-normal least squares algorithm weights both the trial values and the measured data in accordance with the assigned uncertainties and correlations. In general, the measured values fare linearly related to the flux cl> by some response matrix A:

/""') = E A <*> cl> (ft.)

I g lg g where i indexes the measured values belonging to a single data set s, g designates the energy group, and a delineates spectra that may be simultaneously adjusted. For example, RI = Eg a.ig cl>g relates a set of measured reaction rates R; to a single spectrum <Pg by the multigroup reaction cross-section a;g- The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with large assigned uncertainties.

In the least squares adjustment, the continuous quantities (i.e., neutron spectra and cross-sections) were approximated in a multi-group format consisting of 53 energy groups. The trial input spectrum was converted to the FERRET 53 group structure using the SAND-II code1241* This procedure was carried out by fust expanding the 47 group calculated spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure in regions where group boundaries do not coincide. The 620 point spectrum was then re-collapsed into the group structure used in FERRET.

3-13

The sensor set reaction cross-sections, obtained from the ENDF/B-VI dosimetry file 1251, were also collapsed into the 53 energy group structure using the SAND-II code. In this instance, the trial

• spectrum, as expanded to 620 groups, was employed as a weighting function in the cross-section collapsing procedure. Reaction cross-section uncertainties in the form of a 53 x 53 covariance matrix for each sensor reaction were also constructed from the information contained on the ENDF/B-VI data files. These matrices included energy group to energy group uncertainty correlations for each of the individual reactions. However, correlations between cross-sections for different sensor reactions were not included. The omission of this additional uncertainty information does not significantly impact the results of the adjustment.

Due to the importance of providing a trial spectrum that exhibits a relative energy distribution close to the actual spectrum at the sensor set locations, the neutron spectrum input to the FERRET evaluation was obtained from the plant specific calculation for each dosimetry location.

While the 53 x 53 group covariance matrices applicable to the sensor reaction cross-sections were developed from the cross-section data files, the covariance matrix for the input trial spectrum was constructed from the following relation:

2 M,=R +RR,P, gg II gggg where Rn specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the set of values. The fractional uncertainties R 8 specify additional random uncertainties for group g that are correlated with a correlation matrix given by:

p

~

I = [1-B] agg I + B e -H

• where:

(g-g')2 H=---

The frrst term in the correlation matrix equation specifies purely random uncertainties, while the second term describes short range correlations over a group range y (8 specifies the strength of the latter term). The value of o is l when g = g' and 0 otherwise. For the trial spectrum used in the current evaluations, a short range correlation of y = 6 groups was used. This choice 3-14

implies that neighboring groups are strongly correlated when 0 is close to L Strong long range correlations (or anti-correlations) were justified based on information presented by R. E.

Maerker 261* Maerker's results are closely duplicated when y = 6. For the integral reaction rate covariances, simple normalization and random uncertainties were combined as deduced from experimental uncertainties.

In performing the least squares adjustment with the FERRET code, the fast neutron flux spectra (E > 1.0 Me V) calculated at the center of the dosimetry location was input to the analyses. The specific assignment of uncertainties in the measured reaction rates and the input (trial) spectra used in the FERRET evaluations was as follows:

REACTION RA TE UNCERTAINTY 5%

FLUX NORMALIZATION UNCERTAINTY 30%

FLUX GROUP UNCERTAINTIES (E > 0.0055 MeV) 30%

(0.68 e V < E < 0.0055 MeV) 58%

• (E < 0.68 eV)

SHORT RANGE CORRELATION (E > 0.0055 Me V) 104%

0.9 (0.68 e V < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 FLUX GROUP CORRELATION RANGE (E > 0.0055 Me V) 6 (0.68 e V < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 It should be noted that the uncertainties listed for the upper energy ranges extend down to the lower range. Thus, the 58% group uncertainty in the second range is made up of a 30%

uncertainty with a 0.9 short range correlation and a range of 6, and a second part of magnitude 50% with a 0.5 correlation and a range of 3 .

• 3-15

These input uncertainty assignments were based on prior experience in using the FERRET least squares adjustment approach in the analysis of neutron dosimetry from surveillance capsule, reactor cavity, and benchmark irradiations. The values are liberal enough to permit adjustment of the input spectrum to fit the measured data for all practical applications.

3.3 Determination of Best Estimate Reactor Vessel Exposure As noted earlier in this report, the best estimate exposure of the reactor vessel was developed using a combination of absolute plant specific transport calculations based on the methodology discussed in Section 3.1 and plant specific measurement data determined using the measurement evaluation techniques described in Section 3.2. In particular, the IJest estimate vessel exposure is obtained from the following relationship:

~ =Kfl>

But &t. Cale.

where: cpBest Est = The best estimate fast neutron exposure at the location of interest.

K = The plant specific measurement/calculation (MIC) bias factor derived from all available surveillance capsule and reactor cavity dosimetry data.

cp Cale. = The absolute calculated fast neutron exposure at the location of interest The approach defined in the above equation is based on the premise that the measurement data represent the most accurate plant specific information available at the locations of the dosimetry; and, further that the use of the measurement data on a plant specific basis essentially removes biases present in the analytical approach and mitigates the uncertainties that would result from the use of analysis alone. That is, at the measurement points the uncertainty in the best estimate exposure is dominated by the uncertainties in the measurement process. At locations within the reactor vessel wall, additional uncertainty is incurred due to the analytically determined relative ratios among the various measurement points and locations within the reactor vessel wall.

3-16

The implementation of this approach acts to remove plant specific biases associated with the definition of the core source, actual vs. assumed reactor dimensions, and operational variations in water density within the reactor. As a result, the overall uncertainty in the best estimate exposure projections within the vessel wall depend on the individual uncertainties in the measurement process, the uncertainty in the dosimetry location, and in the uncertainty in the calculated ratio of the neutron exposure at the point of interest to that at the measurement location. The uncertainties in the measured flux were derived directly from the results of the least squares evaluation of dosimetry data .

3-17

SECTION 4 RESULTS OF NEUTRON TRANSPORT CALCULATIONS 4.1 Forward Calculation As noted in Section 3 of this report, data from the cycle specific forward transport calculations 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 reactor vessel wall. In this section, the key data extracted from the forward calculations are presented and their relevance to the dosimetry evaluations and vessel exposure projections is discussed. This fluence methodology is consistent with the approved method specified in Reference 27.

4.1.1 Reactor Cavity Sensor Set Locations Data from the Cycle 8, Cycle 9, and Cycles 10/11 specific forward calculations pertinent to

• reactor cavity sensor evaluations are provided in Tables 4.1-1 through 4.1-3. Specifically, the calculated neutron energy spectra are listed for the dosimetry locations at 6, 16, 24, 26, and 39 degrees relative to* the core cardinal axes at the axial core midplane and at the radial locations in the reactor cavity corresponding to the dosimetry positions as described in Section 2.

Table 4.1-1 presents the three spectra used for the analysis of the Cycle 8 dosimetry. Table 4.1-2 presents the four spectra used for the analysis of the Cycle 9 dosimetry, and Table 4.1-3 presents the four spectra used for the analysis of the Cycle 10/11 dosimetry. These data represent the trial spectra used as the starting guess in the FERRET least squares adjustment evaluations of the reactor cavity sensor sets. On a relative basis these calculated energy distributions establish a baseline again~t which adjusted spectra may be compared.

4.1.2 Surveillance Capsule Locations Data from the cycle specific forward calculation pertinent to surveillance capsule evaluations are provided in Table 4.1-4.

4-1

In Table 4.1-4, the calculated neutron energy spectra at the geometric center of surveillance capsules located at 30 degrees (symmetric to the actual location of 240 degrees) and radially at

• 196.06 cm and 20 degrees (symmetric to the actual location of 290 degrees) and radially at 215.43 cm are listed.

4.1.3 Reactor Vessel Wall Data from the forward calculations pertinent to the reactor vessel wall are provided in Tables 4.1-5 through 4.1-18.

In Tables 4.1-5 through 4.1-15, the calculated azimuthal distribution of exposure rates in terms of <l>(E > 1.0 MeV), <l>(E > 0.1 MeV), and dpa/sec are listed at approximately five degree intervals over the reactor geometry for Cycles 1 to 11, respectively. These data in these tables are applicable to the reactor vessel clad/base metal interface. Also given in these tables are the exposure rate ratios [(E > 0.1 MeV)]/[<f>(E > 1.0 MeV)] and [dpa/sec]/[<l>(E > 1.0 MeV)] that provide an indication of the variation in neutron spectrum as a function of azimuthal angle at the reactor vessel inner radius.

Radial gradient information for <l>(E > 1.0 MeV), <l>(E > 0.1 MeV), and dpa/sec is given in Tables 4.1-16, 4.1-17, and 4.1-18 for Cycle 11, respectively. These data are presented on a relative basis for each exposure parameter at 15 degree azimuthal intervals of a core quadrant. The cycle to cycle variation is small such that the relative radial distributions for Cycle 11 is similar to the previous cycles and can be used to calculate the exposure rate distributions through the reactor vessel wall. Exposure rate distributions within the vessel wall are obtained by normalizing the

.calculated or best estimate exposure at the vessel inner.radius to the gradient data given in Tables 4.1-16 through 4.1-18.

4-2

TABLE 4.1-1 e CALCULATED NEUTRON ENERGY SPECTRA AT REACTOR CAVITY SENSOR SET LOCATIONS FOR CYCLE 8 Neutron Flux (n/cm2-sec)

Azimuthal Angle Lower Energy Fa = 1.084 Fa= 1.084 Fa= 1.084 CMeV) 16° 26° 39° 1.42e+Ol 3.758e+05 3.143e+05 2.391e+05 1.22e+Ol l.064e+06 8.741e+05 6.504e+05 l.OOe+Ol 4.121e+06 3.311e+06 2.388e+06 8.61e+OO 7.562e+06 6.014e+06 4.287e+06 7.41e+OO l.147e+07 8.981e+06 6.299e+06 6.07e+OO 2.454e+07 1.899e+07 l.321e+07 4.97e+OO 3.334e+07 2.571e+07 1.793e+07 3.68e+OO 5.875e+07 4.540e+07 3.198e+07 3.0le+OO 4.602e+07 3.582e+07 2.557e+07 2.73e+OO 3.634e+07 2.823e+07 2.008e+07 2.47e+OO 4.472e+07 3.512e+07 2.546e+07 2.37e+OO 2.294e+07 l.795e+07 l.294e+07 2.35e+OO 7.037e+06 5.393e+06 3.724e+06 2.23e+OO 3.552e+07 2.743e+07 l.923e+07 l.92e+OO 9.856e+07 7.664e+07 5.432e+07 l.65e+OO l.377e+08 l.072e+08 7.573e+07 l.35e+OO 2.33le+08 l.843e+08 1.334e+08 l.OOe+OO 5.62le+08 4.476e+08 3.251e+08 8.21e-01 5.530e+08 4.440e+08 3.233e+08 7.43e-01 2.548e+08 2.153e+08 l.730e+08 6.08e-01 l.193e+09 9.708e+08 7.199e+08 4.98e-01 l.147e+09 9.547e+08 7.380e+08

• 3.69e-01 1.183e+09 l.007e+09 8.119e+08 NOTE: The upper energy of group 1 is 17.33 MeV .

• 4-3

TABLE 4.1-1 (continued)

CALCULATED NEU1RON ENERGY SPECTRA AT REACTOR CAVITY SENSOR SET LOCATIONS FOR CYCLE 8 Neutron Flux (n/cm2-sec)

Azimuthal Angle Lower Energy Fa= 1.084 Fa= 1.084 Fa= 1.084 CMeV) 16° 26° 39° 2.97e-01 l.773e+09 1.486e+09 1.152e+09 l.83e-Ol 2.188e+09 l.934e+09 l.644e+09

l. lle-01 2.388e+09 2.109e+09 l.786e+09 6.74e-02 l.575e+09 1.420e+09 l.24le+09 4.09e-02 l.172e+09 l.074e+09 9.615e+08 3.18e-02 3.681e+08 3.477e+08 3.244e+08 2.61e-02 l.869e+08 l.800e+08 l.718e+08 2.42e-02 6.482e+08 5.571e+08 4.406e+08 2.19e-02 l.50e-02 7.lOe-03 3.36e-03 4.114e+08 7.520e+08 9.690e+08 l.037e+09 3.595e+08 6.997e+08 9.279e+08 9.898e+08 2.937e+08 6.369e+08 8.804e+08 9.353e+08 l.59e-03 8.718e+08 8.399e+08 8.032e+08

• 4.54e-04 l.329e+09 1.285e+09 l.236e+09 2.14e-04 6.751e+08 6.581e+08 6.389e+08 l.Ole-04 6.994e+08 6.802e+08 6.585e+08 3.73e-05 8.809e+08 8.573e+08 8.306e+08 1.07e-05 l.Olle+09 9.854e+08 9.567e+08 5.04e-06 5.469e+08 5.343e+08 5.202e+08 l.86e-06 6.783e+08 6.643e+08 6.490e+08

• 8.76e-07 4.816e+08 4.729e+08 4.636e+08 4.14e-07 3.865e+08

• 3.817e+08 3.768e+08 1.00e-07 7.955e+08 7.882e+08 7.810e+08 0.00 l.928e+09 l.912e+09 1.896e+09 4-4

TABLE 4.1-2 e CALCULATED NEUTRON ENERGY SPECTRA AT REACTOR CAVITY SENSOR SET LOCATIONS FOR CYCLE 9 Neutron Flux (n/cm2-sec)

Azimuthal Angle Lower Energy Fa= 1.084 Fa= 1.106 Fa= 1.106 Fa= 1.106 CMeV) ~ 16° 26° 39° 1.42e+Ol 3.272e+05 2.958e+05 2.570e+05 1.879e+05 1.22e+Ol 9.067e+05 8.148e+05 7.024e+05 5.038e+05 l.OOe+-01 3.420e+06 3.063e+06 2.61le+06 l.828e+06 8.6le+OO 6.198e+06 5.576e+06 4.724e+06 3.285e+06 7.41e+OO 9.216e+06 8.312e+06 6.990e+06 4.826e+06 6.07e+OO 1.943e+07 l.757e+07 l.470e+07 l.014e+07 4.97e+OO 2.616e+07 2.355e+07 1.969e+07 l.378e+07

... 3.68e+OO 3.0le+-00 2.73e+OO 2.47e+OO 4.599e+07 3.629e+07 2.858e+07 3.561e+07 4.090e+07 3.198e+07 2.512e+07 3.097e+07 3.431e+07 2.695e+07 2.114e+07 2.625e+07 2.458e+07 1.965e+07 l.542e+07 l.951e+07 2.37e+OO l.816e+07 1.583e+07 1.339e+07 9.917e+06 2.35e+OO 5.421e+06 4.802e+06 4.008e+06 2.864e+06 2.23e+OO 2.768e+07 2.432e+07 2.039e+07 l.475e+07 1.92e+OO 7.765e+07 6.775e+07 5.699e+07 4.158e+07 l.65e+OO l.087e+08 9.454e+07 7.955e+07 5.794e+07 l.35e+OO

• l.872e+08 1.607e+08 l.365e+08 l.018e+08 1.00e+-00 4.564e+08 3.881e+08 3.308e+08 2.476e+08 8.21e-01 4.541e+08 3.826e+08 3.273e+08 2.460e+08 7.43e-01 2.207e+08 l.793e+08 l.591e+08 l.306e+08 6.08e-01 9.978e+08 8.284e+08 7.143e+08 5.463e+08 4.98e-01 9.797e+08 8.015e+08 7.024e+08 5.578e+08 3.69e-01 l.036e+09 8.328e+08 7.416e+08 6.llle+-08 NOTE: The upper energy of group 1 is 17.33 MeV .

• 4-5

TABLE 4.1-2 (continued)

CALCULATED NEUTRON ENERGY SPECTRA AT REACTOR CAVITY SENSOR SET LOCATIONS FOR CYCLE 9 Neutron Flux (n/cm2-sec)

Azimuthal Angle Lower Energy Fa= 1.084 Fa= 1.106 Fa= 1.106 Fa= 1.106 CMeV) ~ 16° 26° 39° 2.97e-Ol l.536e+09 l.242e+09 l.092e+09 8.699e+08 l.83e-Ol l.993e+09 l.558e+09 l.424e+09 l.23le+09 l.l le-01 2.178e+09 l.700e+09 l.55le+09 l.337e+09 6.74e-02 l.465e+09 l.129e+09 l.045e+09 9.267e+08 4.09e-02 l.106e+09 8.44le+08 7.907e+08 7.164e+08 3.18e-02 3.573e+08 2.678e+08 2.562e+08 2.410e+08 2.6le-02 l.858e+08 l.370e+08 l.328e+08 l.274e+08 2.42e-02 5.830e+08 4.577e+08 4.076e+08 3.313e+08 2.19e-02 3.741e+08 2.916e+08 2.631e+08 2.202e+08 1.50e-02 7.196e+08 5.432e+08 5.141e+08 4.736e+08 7.IOe-03 9.547e+08 7.078e+08 6.835e+08 6.526e+08 3.36e-03 l.020e+09 7.567e+08 7.291e+08 6.933e+08 l.59e-03 8.629e+08 6.380e+08 6.189e+08 5.948e+08 4.54e-04 l.318e+09 9.735e+08 9.474e+08 9.149e+08 2.14e-04 6.740e+08 4.958e+08 4.852e+08 4.726e+08 l.Ole-04 6.964e+08 5.132e+08 5.015e+08 4.872e+08 3.73e-05 8.772e+08 6.464e+08 6.320e+08 6.145e+08 l.07e-05 l.008e+09. 7.421e+08 7.266e+08 7.077e+08 5.04e-06. 5.454e+08 4.017e+08 3.939e+08 3.848e+08 1.86e-06 6.774e+08 *4.984e+08 4.899e+08 4.799e+08 8.76e-07 4.816e+08 3.542e+08 3.488e+08 3.427e+08 4.14e-07 3.878e+08 2.848e+08 2.816e+08 2.785e+08 l.OOe-07 7.993e+08 5.866e+08 5.816e+08 5.773e+08 0.00 1.939e+09 l.422e+09 l.411e+09 l.402e+09 4-6

TABLE 4.1-3 e CALCULATED NEU1RON ENERGY SPECTRA AT REACTOR CAVITY SENSOR SET LOCATIONS FOR CYCLES 10-11 Neutron Aux (n/cm2-sec)

Azimuthal Angle Lower Energy Fa= 1.108 Fa= 1.108 Fa= 1.108 Fa= 1.108 Fa= 1.108 Fa= 1.108 (MeV} _K_ 16° 24° 26° 36° 39° l.42e+Ol 2.248e+05 2.418e+05 2.228e+05 2.360e+05 2.075e+05 l.981e+05 l.22e+Ol 6.105e+05 6.608e+05 6.056e+05 6.405e+05 5.576e+05 5.303e+05 l.OOe+Ol 2.249e+06 2.460e+06 2.236e+06 2.362e+06 2.030e+06 l.920e+06 8.61e+OO 4.049e+06 4.458e+06 4.035e+06 4.257e+06 3.645e+06 3.437e+06 7.41e+OO 5.943e+06 6.605e+06 5.951e+06 6.262e+06 5.341e+06 5.020e+06 6.07e+OO l.245e+07 1.392e+07 l.250e+07 l.313e+07 l.119e+07 1.050e+07 4.97e+OO l.651e+07 l.857e+07 1.672e+07 1.744e+07 l.494e+07 l.404e+07 3.68e+OO 2.846e+07 3.213e+07 2.905e+07 3.009e+07 2.603e+07 2.452e+07 3.0le+OO 2.232e+07 2.514e+07 2.279e+07 2.354e+07 2.052e+07 l.939e+07

• • 2.73e+OO 2.47e+OO 2.37e+OO 1.746e+07 2.170e+07 1.104e+07 1.972e+07 2.436e+07 1.244e+07 1.786e+07 2.219e+07 1.132e+07 l.843e+07 2.283e+07 1.164e+07 1.606e+07 2.006e+07 l.021e+07 l.517e+07 1.902e+07 9.666e+06 2.35e+OO 3.275e+06 3.748e+06 3.384e+06 3.484e+06 3.007e+06 2.830e+06 *>~

2.23e+OO l.673e+07 1.902e+07 l.715e+07 l.771e+07 l.538e+07 l.448e+07 ..

l.92e+OO 4.694e+07 5.308e+07 4.783e+07 4.950e+07 4.314e+07 4.069e+07 .t*

l.65e+OO 6.544e+07 7.402e+07 6.663e+07 6.901e+07 6.007e+07 5.666e+07 1.35e+OO 1.125e+08 1.262e+08 1.145e+08 1.182e+08 l.038e+08 9.843e+07 l.OOe+OO 2.732e+08 3.052e+08 2.763e+08 2.860e+08 2.515e+08 2.387e+08 8.2le-01 2.707e+08 3.0lle+08 2.729e+08 2.826e+08 2.489e+08 2.365e+08 7.43e-01 l.326e+08 l.433e+08 l.349e+08 1.368e+08 l.251e+08 l.212e+08 6.08e-01 5.932e+08 6.540e+08 5.930e+08 6.155e+08 5.458e+08 5.200e+08 4.98e-01 5.840e+08 6.367e+08 5.870e+08 6.036e+08 5.436e+08 5.224e+08 3.69e-Ol 6.201e+08 6.664e+08 6.216e+08 6.360e+08 5.824e+08 5.634e+08 NOTE: The upper energy of group 1 is 17.33 MeV .

• 7

TABLE 4.1-3 (continued)

CALCULATED NEUTRON ENERGY SPECTRA AT REACTOR CAVITY SENSOR SET LOCATIONS FOR CYCLES 10-11 Neutron Flux (n/cm2-sec)

Azimuthal Angle Lower Energy Fa=l.108 Fa= 1.108 Fa= 1.108 Fa= 1.108 F~ = 1.108 Fa= 1.108 CMeV) _£__ 16° 24° 26° 36° 39° 2.97e-Ol 9.126e+08 9.883e+08 9.065e+08 9.375e+08 8.461e+08 8.123e+08 1.83e-01 1.196e+09 1.260e+09 1.202e+09 1.217e+09 l.138e+09 l.113e+09 i.lle-01 1.305e+09 1.374e+09 1.302e+09 1.325e+09 l.238e+09 1.208e+09 6.74e-02 8.807e+08 9.178e+08 8.809e+08 8.907e+08 8.430e+08 8.274e+08 4.09e-02 6.672e+08 6.898e+08 6.684e+08 6.728e+08 6.432e+08 6.340e+08 3.18e-02 2.166e+08 2.208e+08 2.183e+08 2.175e+08 2.116e+08 2.104e+08 2.61e-02 1.130e+08 1.137e+08 1.125e+08 1.126e+08 1.106e+08 l.102e+08 2.42e-02 3.449e+08 3.666e+08 3.316e+08 3.485e+08 3.177e+08 3.044e+08 2.19e-02 1.50e-02 7.lOe-03 3.36e-03 1.59e-03 2.216e+08 4.330e+08 5.794e+08 6.189e+08 5.252e+08 2.344e+08 4.451e+08 5.860e+08 6.262e+08 5.295e+08 2.172e+08 4.376e+08 5.816e+08 6.171e+08 5.246e+08 2.246e+08 4.367e+08 5.793e+08 6.178e+08 5.240e+08 2.070e+08 4.203e+08 5.676e+08 6.044e+08 5.151e+08 2.005e+08 4.167e+08 5.655e+08 6.006e+08 5.130e+08 4.54e-04 8.036e+08 8.090e+08 8.039e+08 8.017e+08 7.898e+08 7.874e+08 2.14e-04 4.118e+08 4.132e+08 4.118e+08 4.104e+08 4.059e+08 4.053e+08 1.0le-04 4.254e+08 4274e+08 4.253e+08 4241e+08 4.190e+08 4.181e+08 3.73e-05 5.360e+08 5.384e+08 5.357e+08 5.344e+08 5.282e+08 5.269e+08

. 1.07e-05 6.159e+08 6.184e+08 6.160e+08 6.142e+08 6.075e+08 6.062e+08 5.04e-06 3.338e+08 3.350e+08 3.343e+08 3.330e+08 3.297e+08 3.293e+08 1.86e-06 4.148e+08 4.160e+08 4.162e+08 4.140e+08 4.105e+08 4.103e+08 8.76e-07 2.952e+08 2.958e+08 2.967e+08 2.947e+08 2.926e+08 2.927e+08 4.14e-07 2.380e+08 2.382e+08 2.406e+08 2.379e+08 2.368e+08 2.376e+08 l.OOe-07 4.908e+08 4.910e+08 4.981e+08 4.912e+08 4.897e+08 4.920e+08 0~00 1.189e+09 1.190e+09 1.214e+09 1.192e+09 1.189e+09 1.197e+09 4-8

TABLE 4.1-4 (continued)

CALCULATED NEUTRON ENERGY SPECTRA AT SURVEILLANCE CAPSULE CENTER Neutron Flux (n/cm2-sec)

• Azimuthal Angle Lower Energy Fa= 1.149 Fa= 1.149 Fa= 1.106 CMeV) 30° 20° 20°- Cyc 9 2.97e-01 7.273e+10 4.825e+09 2.982e+09 1.83e-01 1.117e+ll 7.697e+09 4.806e+09 1.l le-01 9.796e+10 6.662e+09 4.145e+09 6.74e-02 7.655e+10 5.269e+09 3.288e+09 4.09e-02 6.510e+10 4.531e+09 2.833e+09 3.18e-02 2.594e+10 1.848e+09

• 1.l59e+09 2.61e-02 1.445e+10 9.726e+08 6.112e+08 2.42e-02 1.965e+10 1.359e+09 8.440e+08 2.19e-02 l.50e-02 7.lOe-03 3.36e-03 1.226e+10 3.722e+l0 6.939e+JO 7.430e+l0 8.228e+08 2.737e+09 5.175e+09 5.459e+09 5.112e+08 1.713e+09 3.250e+09 3.428e+09 1.59e-03 7.129e+l0 5.133e+09 3.229e+09 4.54e-04 1.182e+ll 8.518e+09 5.371e+09 2.14e-04 6.763e+l0 4.904e+09 3.097e+09 1.0le-04 7.188e+l0 5.224e+09 3.302e+09 3.73e-05 9.534e+l0 6.876e+09 4.351e+09 1.07e-05 1.185e+ 11 8.543e+09 . 5.414e+09 5.04e-06 6.964e+10 5.034e+09 3.193e+09 1.86e-06 9.435e+l0 6.940e+09 4.410e+09 8.76e-07 *1.198e+10 5.410e+09 3.443e+09 4.14e-07 6.748e+l0 5.198e+09 3.318e+09 1.00e-07 1.519e+ll 1.269e+10 8.143e+09 0.00 4.526e+l 1 6.802e+l0 4.270e+10 4-10

TABLE 4.1-5 e AZIMUTHAL VARIATION OF NEUTRON FLUX AND DPA/SEC AT REACTOR VESSEL CLAD-BASE METAL INTERFACE CYCLE 1 Theta Flux (n/cm2-sec) Ratio E > 0.1 d12alsec (deg.) E > 1.0 E > 0.1 dmVsec E > 1.0 E > 1.0 0 3.694e+10 7.267e+10 5.659e-11 1.97 l.53e-21 5 3.896e+10 7.688e+10 5.957e-11 1.97 1.53e-21 10 4.429e+10 8.749e+l0 6.745e-11 1.98 1.52e-21 15 4.868e+10 9.633e+l0 7.386e-11 1.98 1.52e-21 20 4.221e+l0 8.952e+10 6.431e-11 2.12 l.52e-21 25 3.732e+l0 7.449e+l0 5.705e-11 2.00 1.53e-2l 30 3.579e+10 7.147e+l0 5.460e-11 2.00 1.53e-21 35 3.523e+10 6.999e+10 5.347e-11 1.99 1.52e-21 40 2.696e+10 5.376e+l0 4.130e-11 1.99 1.53e-21 45 2.277e+10 4.531e+10 3.515e-11 l.99 1.54e-21 50 2.697e+10 5.379e+10 4.132e-11 1.99 1.53e-21 55 3.543e+10 7.034e+10 5.379e-11 1.99 1.52e-21 60 3.700e+l0 7.336e+10 5.638e-11 1.98 1.52e-21 65 3.774e+10 7.520e+10 5.768e-11 1.99 1.53e-21 70 4.451e+10 8.873e+10 6.768e-11 1.99 1.52e-21 75 4.899e+10 9.674e+10 7.431e-11 1.97 1.52e-21 80 4.428e+10 8.749e+10 6.745e-11 1.98 1.52e-21 85 3.895e+10 7.685e+10 5.955e-11 1.97 1.53e-21 90 3.693e+10 7.258e+10 5.655e-11 1.97 1.53e-21

• • 4-11

TABLE 4.1-6 AZIMUTHAL VARIATION OF NEU1RON FLUX AND DP NSEC AT REACTOR VESSEL CLAD-BASE METAL INTERFACE CYCLE 2 Theta Flux (n/cm2-sec) Ratio E > 0.1 dpa/sec (deg.) E > 1.0 E > 0.1 dpa/sec E > 1.0 E > 1.0 0 3.358e+10 6.603e+10 5.142e-1 l 1.97 1.53e-21 5 3.513e+10 6.931e+10 5.372e-1 l 1.97 l.53e-21 10 3.932e+10 7.770e+10 5.991e-ll 1.98 l.52e-21 15 4.298e+10 8.512e+10 6.525e-1 l 1.98 1.52e-21 20 3.769e+10 7.998e+10 5.745e-ll 2.12 1.52e-21 25 3.411e+l0 6.803e+10 5.214e-ll 1.99 1.53e-21 30 3.299e+10 6.588e+10 5.032e-1 l 2.00 1.53e-21 35 3.258e+10 6.474e+10 4.946e-1 l 1.99 l.52e-21 40 2.519e+10 5.019e+10 3.858e-11 1.99 1.53e-21 45 2.144e+10 4.261e+10 3.307e-ll 1.99 1.54e-21 50 2.519e+10 5.021e+10 3.859e-11 1.99 1.53e-21 55 3.277e+10 6.506e+10 4.975e-ll 1.99 1.52e-21 60 3.413e+10 6.766e+10 5.200e-1 l 1.98 1.52e-21 65 3.450e+10 6.869e+10 5.272e-11 1.99 1.53e-21 70 3.977e+10 7.930e+10 6.049e-ll 1.99 1.52e-21 75 4.324e+10 8.544e+10 6.561e-ll 1.98 1.52e-21 80 3.932e+10 7.770e+10 5.990e-ll 1.98 1.52e-21 85 3.512e+10 6.928e+10 5.370e-11 1.97 1.53e-21 90 3.357e+10 6.595e+10 5.139e-l 1 1.96 1.53e-21 4-12

TABLE 4.1-7 e AZIMUTHAL VARIATION OF NEU1RON FLUX AND DP NSEC AT REACTOR VESSEL CLAD-BASE METAL INTERFACE CYCLE 3 Theta Flux (n/cm2-sec) Ratio E > 0.1 dgaLsec (deg.) E > 1.0 E > 0.1 dgaLsec E > 1.0 E > 1.0 0 4.461e+l0 8.803e+l0 6.828e-11 1.97 l.53e-21 5 4.702e+10 9.309e+10 7.184e-11 1.98 l.53e-21 10 5.340e+10 l.059e+l 1 8.128e-11 1.98 l.52e-21 15 5.866e+10 l.165e+l 1 8.895e-11 1.99 l.52e-21 20 5.067e+10 l.080e+ll 7.719e-11 2.13 l.52e-21 25 4.498e+10 9.000e+lO 6.870e-11 2.00 l.53e-21 30 4.351e+10 8.660e+10 6.626e-11 1.99 l.52e-21' 35 4.135e+10 8.242e+10 6.275e-11 1.99 l.52e-21

• 40 45 50 55 3.147e+10 2.662e+10 3.146e+l0 4.133e+l0 6.300e+10 5.316e+10 6.298e+10 8.237e+10 4.819e-11 4.106e-11 4.817e-11 6.272e-11 2.00 2.00 2.00 1.99 l.53e-21 l.54e-21 l.53e-21 l.52e-21 60 4.348e+10 8.652e+10 6.621e-11 1.99 l.52e-21 65 4.489e+10 8.977e+10 6.856e-11 2.00 l.53e-2l 70 5.339e+10 l.068e+ll 8.l 14e-11 2.00 l.52e-21 75 5.855e+10 l.162e+ll 8.879e-11 1.99 l.52e-21 80 5.340e+10 l.059e+ll 8.128e-11 1.98 l.52e-21 85 4.706e+10 9.315e+10 7.190e-11 1.98 l.53e-21 90 4.466e+10 8.805e+10 6.833e-11 1.97 l.53e-21 4-13

TABLE 4.1-8 AZIMUTIIAL VARIATION OF NEU1RON FLUX AND DP NSEC AT REACTOR VESSEL CLAD-BASE METAL INTERFACE CYCLE 4 Theta Flux (n/cm2-sec) Ratio E > 0.1 d12a/sec (deg.) E > 1.0 E > 0.1 d12a/sec E > 1.0 E > 1.0 0 4.511e+l0 8.910e+l0 6.903e-1 l 1.98 l.53e-21 5 4.751e+l0 9.415e+l0 7.258e-1 l 1.98 l.53e-21 10 5.394e+10 l.070e+ll 8.209e-ll 1.98 l.52e-21 15 5.946e+10 l.182e+ll 9.018e-11 1.99 1.52e-21 20 5.193e+l0 l.108e+ll 7.912e-1 l 2.13 l.52e-21 25 4.717e+l0 9.448e+10 7.204e-1 l 2.00 l.53e-21 30 4.704e+10 9.370e+l0 7.159e-11 1.99 l.52e-21 35 4.565e+l0 9.102e+l0 6.923e-1 l 1.99 l.52e-21 40 45 50 55 3.491e+l0 2.949e+l0 3.491e+l0 4.566e+10 6.990e+l0 5.893e+10 6.990e+10 9.102e+l0 5.341e-11 4.546e-11 5.341e-1 l 6.924e-ll 2.00 2.00 2.00 1.99 l.53e-21 l.54e-21 l.53e-21 l.52e-21 60 4.704e+10 9.370e+l0 7.160e-11 1.99 l.52e-21 65 4.713e+l0 9.434e+10 7.197e-11 2.00 l.53e-21 70 5.480e+l0 l.098e+l 1 8.328e-1 l 2.00 l.52e-21 75 5~942e+l0 l.181e+ 11 9.0IOe-11 1.99 l.52e-21 80 5.394e+l0 l.070e+ll 8.209e-11 1.98 l.52e-21 85 4.750e+l0 9.412e+l0 7.257e-11 1.98 l.53e-21 90 4.510e+l0 8.900e+l0 6.898e-11 1.97 l.53e-21 4-14

TABLE 4.1-9 e AZIMUTHAL VARIATION OF NEUTRON FLUX AND DP NSEC AT REACTOR VESSEL CLAD-BASE METAL INTERFACE CYCLE 5 Theta Flux (n/cm2-sec) Ratio E > 0.1 dpa/sec (deg.) E > 1.0 E > 0.1 dpa/sec E > 1.0 E > 1.0 0 4.211e+IO 8.317e+IO 6.447e-11 1.97 l.53e-21 5 4.464e+10 8.846e+10 6.823e-11 1.98 l.53e-21 10 5.144e+IO 1.020e+ll 7.829e-11 1.98 l.52e-21 15 5.736e+IO l.140e+ll 8.699e-11 1.99 l.52e-21 20 5.037e+IO 1.074e+ll 7.674e-11 2.13 l.52e-21 25 4.570e+IO 9.144e+IO 6.980e-11 2.00 l.53e-21 30 4.516e+IO 8.988e+IO 6.875e-11 1.99 l.52e-21 35 4.349e+IO 8.665e+IO 6.596e-11 1.99 l.52e-21 40 3.320e+IO 6.644e+IO 5.081e-11 2.00 l.53e-21 45 2.807e+IO 5.605e+IO 4.327e-11 2.00 1.54e-21

. 50 3.320e+IO 6.644e+IO 5.081e-11 2.00 l.53e-21 55 4.349e+IO 8.665e+IO 6.596e-11 1.99 l.52e-21 60 4.516e+IO 8.987e+IO 6.874e-11 1.99 l.52e-21 65 4.566e+IO 9.130e+IO 6.972e-11 2.00 l.53e-21 ,,'

70 5.315e+10 l.064e+ll 8.076e-11 2.00 l.52e-21 75 5.731e+IO l.138e+ll 8.689e-11 1.99 l.52e-21 80 5.142e+IO l.020e+ll 7.828e-11 1.98 l.52e-21 85 4.462e+IO 8.842e+IO

.. - 6.820e-11 1.98 l.53e-21 90 4.209e+IO 8.306e+IO 6.442e-11 1.97 l.53e-21

• 4.;.15

TABLE 4.1-10 AZIMUTHAL VARIATION OF NEUTRON FLUX AND DPA/SEC AT REACTOR VESSEL CLAD-BASE METAL INTERFACE CYCLE 6 .

Theta Flux (n/cm2-sec) Ratio E > 0.1 dpa/sec (deg.) E > 1.0 E > 0.1 dpa/sec E > 1.0 E > 1.0 0 4.560e+l0 8.999e+10 6.978e-1 l 1.97 l.53e-21 5 4.786e+10 9.478e+10 7.313e-11 1.98 l.53e-21 10 5.392e+l0 l.069e+ll 8.207e-1 l 1.98 l.52e-21 15 5.904e+10 l.173e+ll 8.954e-11 1.99 l.52e-21 20 5.128e+10 l.093e+ll 7.813e-ll 2.13 l.52e-21 25 4.619e+l0 9.245e+10 7.055e-1 l 2.00 l.53e-21 30 4.544e+10 9.045e+10 6.917e-ll 1.99 l.52e-21 35 4.360e+10 8.690e+l0 6.613e-ll 1.99 l.52e-21

.40 3.318e+10 6.644e+10 5.079e-1 l 2.00 l.53e-21 45 2.802e+10 5.598e+l0 4.320e-ll 2.00 l.54e-21 50 3.318e+10 6.645e+10 5.079e-1 l 2.00 l.53e-21 55 4.360e+10 . 8.690e+l.O 6.614e-ll 1.99 l.52e-21 60 4.544e+10 9.045e+l0 6.917e-ll 1.99 l.52e-21 65 4.615e+10 9.232e+l0 7.047e-ll 2.00 l.53e-21 70 5.41le+10 l.083e+l 1 8.224e-11 2.00 l.52e-21 75 5.899e+10 l.172e+ll 8.946e-11 1.99 l.52e-21 80 5.392e+10 1.069e+ll 8.208e-ll 1.98 l.52e-21 85 4.786e+IO 9.476e+l0 7.312e-ll 1.98 l.53e-21 90 4.560e+10 8.991e+10 6.974e-ll 1.97 l.53e-21

. 4-16

TABLE 4.1-11 e AZIMUTHAL VARIATION OF NEU1RON FLUX AND DP NSEC AT REACTOR VESSEL CLAD-BASE METAL INTERFACE CYCLE 7 Theta Flux (n/cm2-sec) Ratio E > 0.1 d12aLsec (deg.) E > 1.0 E > 0.1 dpaLsec E > 1.0 E > 1.0 0 4.355e+10 8.594e+10 6.665e-11 1.97 l.53e-21 5 4.570e+10 9.047e+l0 6.983e-11 1.98 l.53e-21 10 5.144e+10 l.020e+ll 7.831e-11 1.98 l.52e-21 15 5.628e+10 l.118e+ll 8.537e-11 1.99 l.52e-21 20 4.892e+10 l.043e+ll 7.454e-11 2.13 l.52e-21 25 4.417e+l0 8.839e+10 6.748e-11 2.00 l.53e-21 30 4.352e+l0 8.663e+l0 6.627e-11 1.99 l.52e-21 35 4.185e+l0 8.339e+l0 6.349e-11 1.99 l.52e-21

• 40 45 50 55 3.201e+l0 2.711e+10 3.201e+l0 4.185e+l0 6.405e+10 5.411e+l0 6.405e+l0 8.339e+10 4.899e-11 4.179e-11 4.899e-11 6.349e-11 2.00 2.00 2.00 1.99 l.53e-21 l.54e-21 l.53e-21 l.52e-21 60 4.352e+l0 8.662e+l0 6.626e-11 1.99 l.52e-21 65 4.414e+l0 8.8i6e+10 6.740e-11 2.00 l.53e-21 70 5.162e+l0 l.033e+ll 7.846e-11 2.00 l.52e-21 75 5.624e+l0 l.117e+ll 8.529e-11 1.99 l.52e-21

80. 5.144e+l0 l.020e+ll 7.831e-11 1.98 l.52e-21 85 4.569e+10 9.045e+10 6.981e-11 1.98 l.53e-21 90 4.354e+l0 8.585e+10 6.661e-11 1.97 l.53e-21

• 4-17

TABLE 4.1-12 AZIMUTHAL VARIATION OF NEUTRON FLUX AND DP NSEC AT REACTOR VESSEL CLAD-BASE METAL INTERFACE CYCLE 8 Theta Flux (n/cm2-sec) Ratio E > 0.1 d12aLsec (deg.) E > 1.0 E > 0.1 d12a/sec E > 1.0 E > 1.0 0 l.940e+10

• 3.853e+10 3.006e-11 1.99 l.55e-21 5 2.404e+10 4.794e+10 3.700e-11 1.99 l.54e-21 10 3.642e+10 7.215e+10 5.540e-11 1.98 l.52e-21 15 4.641e+10 9.178e+10 7.019e-11 1.98 l.51e-21 20 4.026e+10 8.560e+10 6.117e-11 2.13 l.52e-21 25 3.103e+10 6.172e+10 4.739e-11 1.99 l.53e-21 30 2.222e+10 4.406e+10 3.413e-11 1.98 l.54e-21 35 l.760e+10 3.516e+10 2.710e-11 2.00 l.54e-21 40 45 50 55 1.694e+10 l.660e+10 l.700e+10 l.764e+10 3.323e+10 3.235e+10 3.334e+10 3.524e+10 2.607e-11 2.560e-11 2.616e-11 2.716e-11 1.96 1.95 1.96 2.00 l.54e-21 l.54e-21 l.54e-21 l.54e-21 60 2.220e+10 4.403e+10 3.411e-11 1.98 l.54e-21 65 3.097e+10 6.157e+10 4.730e-11 1.99 l.53e-21 70 4.254e+10 8.484e+10 6.447e-11 1.99 l.52e-21 75 4.636e+10 9.163e+10 7.0lOe-11 1.98 l.5 le-21 80 3.642e+10 7.215e+10 5.540e-11 1.98 1.52e-21 85 2.404e+10 4.792e+10 3.699e-11 1.99 1.54e-21 90 1.940e+10 3.849e+10 3.004e-11 1.98 l.55e-21 4-18

TABLE 4.1-13 e AZTh1UTHAL VARIATION OF NEUTRON FLUX AND DP NSEC AT REACTOR VESSEL CLAD-BASE METAL INTERFACE CYCLE 9 Theta Aux {n/cm2-sec) Ratio E > 0.1 d12aLsec (deg.) E > 1.0 E > 0.1 dpa/sec E > 1.0 E > 1.0 0 2.014e+10 3.945e+l0 3.lOOe-11 1.96 1.54e-21 5 2.185e+l0 4.295e+l0 3.354e-11 1.97 1.54e-21 10 2.637e+l0 5.183e+l0 4.025e-11 1.97 l.53e-21 15 3.018e+10 5.941e+10 4.587e-11 1.97 1.52e-21 20 2.656e+10 5.606e+10 4.052e-11 2.11 l.53e-21 25 2.313e+l0 4.566e+l0 3.540e-11 1.97 l.53e-21 30 l.961e+l0 3.859e+l0 3.002e-11 1.97 1.53e-2'1 35 1.640e+10 3.244e+10 2.508e-11 1.98 l.53e-21

• 40 45 50 55 l.253e+l0 l.098e+10 l.253e+10 l.640e+l0 2.479e+10 2.162e+10 2.480e+10 3.245e+10 l.931e-11 1.702e-11 l.931e-11 2.508e-11 1.98 1.97 1.98 1.98 l.54e-21 l.55e-21 l.54e-21 l.53e-21 60 l.961e+10 3.859e+10 3.002e-11 1.97 1.53e-21 65 2.311e+10 4.558e+10 3.535e-11 1.97 l.53e-2L 70 2.809e+10 5.565e+10 4.275e-11 1.98 l.52e-21 75 3.016e+10 5.934e+10 4.584e-11 1.97 l.52e-21 80 2.638e+10 5.186e+10 4.027e-11 1.97 l.53e-21 85 2.186e+10 4.297e+l0 3.356e-11 1.97 l.54e-21 90 2.015e+l0 3.943e+10 3.lOOe-11 1.96 l.54e-21

• 4.-19

TABLE 4.1-14 AZIMUTHAL VARIATION OF NEUTRON FLUX AND DP NSEC AT REACTOR VESSEL CLAD-BASE METAL INTERFACE CYCLE IO Theta Flux (n/cm2-sec) Ratio E > 0.1 d12alsec (deg.) E > 1.0 E > 0.1 d12a/sec E > 1.0 E > 1.0 0 l.471e+IO 2.883e+IO 2.267e-1 l 1.96 l.54e-21 5 l.619e+IO 3.184e+IO 2.489e-11 1.97 1.54e-21 IO 1.988e+IO 3.903e+IO 3.038e-1 l 1.96 1.53e-21 15 2.260e+10 4.443e+IO 3.443e-1 l 1.97 l.52e-21 20 2.060e+10 4.334e+IO 3.150e-ll 2.10 l.53e-21 25 2.003e+10 3.936e+IO 3.068e-11 1.96 1.53e-21 30 1.886e+IO 3.702e+IO 2.885e-1 l 1.96 l.53e-21 35 1.700e+IO 3.354e+IO 2.598e-11 1.97 1.53e-21 40 45 50 55 1.385e+IO 1.252e+IO 1.387e+IO 1.703e+IO 2.727e+10 2.449e+IO 2.729e+IO 3.359e+IO 2.132e-ll l.937e-ll 2.134e-1 l 2.602e-11 1.97 1.96 1.97 1.97 l.54e-21 l.55e-21 1.54e-21 1.53e-21 60 1.890e+IO 3.709e+IO 2.891e-ll 1.96 1.53e-21 65 2.006e+IO 3.940e+IO 3.072e-ll 1.96 1.53e-21 70 2.183e+IO 4.313e+IO 3.331e-ll 1.98 1.53e-21 75 2.261e+IO 4.443e+IO 3.444e-ll 1.96 1.52e-21 80 1.990e+IO 3.906e+IO 3.040e-11 1.96 1.53e-21 85 1.620e+IO 3.184e+IO 2.489e-1 l 1.97 1.54e-21 90 l.470e+IO 2.880e+IO . 2.266e-11

• 1.96 l.54e-21 4-20

TABLE 4.1-15 e AZIMUTHAL VARIATION OF NEUTRON FLUX AND DPA/SEC AT REACTOR VESSEL CLAD-BASE METAL INTERFACE CYCLE 11 Theta Flux (n/cm2-sec) Ratio E > 0.1 ctn alsec (deg.) E > 1.0 E > 0.1 dnalsec E > 1.0 E > 1.0 0 1.348e+10 2.636e+10 2.077e-11 1.95 l.54e-21 5 l.460e+10 2.~62e+10 2.243e-11 1.96 l.54e-21 10 l.726e+10 3.382e+10 2.638e-11 1.96 l.53e-21 15 l.891e+10 3.714e+10 2.882e-11 1.96 l.52e-21 20 l.655e+10 3.481e+10 2.533e-11 2.10 l.53e-21 25 l.584e+10 3.113e+10 2.429e-11 1.97 l.53e-21 30 l.506e+10 2.952e+10 2.304e-11 1.96 l.53e-21 35 1.340e+10 2.645e+10 2.049e-11 1.97 l.53e-21

• 40.

45 50 55 l.094e+10 9.946e+o9 l.103e+10

• 1.363e+10 2.153e+10 l.944e+10 2.171e+10 2.692e+10 l.685e-11 l.540e-11 1.699e-11 2.086e-11 1.97 1.95 1.97 1.97 l.54e-21 l.55e-21 l.54e-21 l.53e-21 60 l.556e+10 3.050e+10 2.381e-11 1.96 l.53e-21 65 l.677e+10 3.297e+10 2.571e-11 1.97 l.53e-21 70 l.914e+10 3.780e+10

• 2.922e-11 1.98 l.53e-21 75 2.089e+10 4.099e+10 3.181e-11 1.96 l.52e-21 80 l.875e+10 3.675e+10 .2.864e-11 1.96 l.53e-21 85 l.531e+10 3.005e+10 2.353e-11 1.96 l.54e-21 90 l.388e+10 2.715e+10 2.140e-11 1.96 l.54e-21

• • 4-21

TABLE 4.1-16 RELATIVE RADIAL DISTRIBUTION OF NEUTRON FLUX (E > 1.0 Me V)

THROUGH THE REACTOR VESSEL WALL CYCLE 11 Radius Vessel Azimuthal Angle J£mL Fraction JE. _u:_ 30° 45° 219.060 (a) 1.0198 1.0204 1.0201 1.0193 219.380 0.000 1.0000 1.0000 1.0000 1.0000 219.910 0.024 0.9673 0.96,63 0.9667 0.9681 220.975 0.071 0.8770 0.8744 0.8759 0.8796 221.613 0.100 0.8193 0.8144 0.8167 0.8214 222.730 0.150 0.7214 0.7148 0.7176 0.7239 223.846 0.200 0.6330 ' 0.6239 0.6278 0.6348 224.963 0.250 0.5515 0.5420 0.5458 0.5541 226.079 0.300 0.4791 0.4703 0.4736 0.4828 227.196 0.350 0.4165 0.4039 0.4094 0.4170 228.312 0.400 0.3587 0.3477 0.3517 0.3605 229.429 0.450 0.3088 0.2983 0.3024 0.3109 230.546 0.500 0.2658 0.2558 0.2596 0.2679 231.662 0.550 0.2285 0.2189 0.2226 0.2305 232.779 0.600 0.1959 0.1869 0.1903 0.1979 233.895 0.650 0.1678 0.1593 0.1626 0.1698 235.012 0.700 0.1434 0.1357 0.1386 0.1455

. 236.128 0.750 0.1222 0.1151 0.1179 0.1245 237.245 0.800 0.1047 0.0970 0.1003 0.1060 238.361 0.850 0.0883 0.0817 0.0843 0.0904 239.478 0.900 0.0743 0.0681 . 0.0706 0.0767 240.594. 0.950 0.0617 0.0558 0.0583 0.0646 241.180 0.976 0.0554 0.0495 0.0521 0.0587 241.711 1.000 0.0522 0.0461 0.0489 0.0557 Note: Base metal inner radius= 219.380 cm; outer radius = 241. 711 cm (a) Center of vessel clad 4-22

TABLE 4.1-16 (continued)

RELATIVE RADIAL DISTRIBUTION OF NEUTRON FLUX (E > 1.0 MeV)

THROUGH THE REACTOR VESSEL WALL CYCLE 11 Radius Vessel Azimuthal Angle (cm) Fraction 60° 75° 90° 219.060 (a) 1.0201 1.0202 1.0197 219.380 0.000 1.0000 1.0000 1.0000 219.910 0.024 0.9667 0.9665 0.9673 220.975 0.071 0.8760 0.8750 0.8772 221.613 0.100 0.8168 0.8148 0.8196 222.730 0.150 0.7177 0.7148 0.7218 223.846 0.200 0.6280 0.6240 0.6335 224.963 0.250 0.5461 0.5421 0.5521 226.079 0.300 0.4739 0.4706 0.4797

• 227.196 228.312 229.429 230.546 0.350 0.400 0.450 0.500 0.4097 0.3520 0.3026 0.2599 0.4042 0.3477 0.2985 0.2559 0.4171 0.3593 0.3094 0.2664 231.662 0.550 0.2229 0.2191 0.2291 232.779 0.600 0.1906 0.1870 0.1965 233.895 0.650 0.1629 0.1595 0.1684 235.012 0.700 0.1389 0.1358 0.1439 236.128 0.750 0.1181 0.1152 0.1227 237.245 0.800 0.1005 0.0969 0.1052 238.361 0.850 0.0845 0.0817 0.0887 239.478 0.900 0.0708 0.0680 0.0747 240.594 0.950 0.0584 . 0.0556 0.0622 241.180 0.976 0.0522 0.0492 0.0559 241.711 1.000 0.0490 0.0458 0.0527 Note: Base metal inner radius = 219 .380 cm; outer radius= 241.711 cm (a) Center of vessel clad 4-23

TABLE 4.1-17 RELATIVE RADIAL DISTRIBUTION OF NEU1RON FLUX (E > 0.1 MeV)

THROUGH THE REACTOR VESSEL WALL CYCLE 11 Radius Vessel Azimuthal Angle (cm) Fraction ~ ~ 30° 45° 219.060 (a) 0.9917 0.9941 0.9933 0.9909 219.380 0.000 1.0000 1.0000 1.0000 1.0000 219.910 0.024 1.0137 1.0097 1.0111 1.0151 220.975 0.071 1.0030 0.9935 0.9972 1.0084 221.613 0.100 0.9907 0.9763 0.9819 0.9966 222.730 0.150 0.9552 0.9350 0.9425 0.9631 223.846 0.200 0.9154 0.8888 0.8990 0.9241 224.963 0.250 0.8705 0.8411 0.8518 0.8822 226.079 0.300 0.8255 0.7921 0.8044 0.8380 227.196 228.312 229.429 230.546 0.350 0.400 0.450 0.500 0.7817 0.7348 0.6886

. 0.6434 0.7435 0.6953 0.6475 0.6011 0.7581 0.7100 0.6630 0.6172 0.7940 0.7490 0.7040 0.6599 231.662 0.550 0.5989 0.5558 0.5724 0.6163 232.779 0.600 0.5549 0.5115 0.5284 0.5734 233.895 0.650 0.5119 0.4686 0.4857 0.5314 235.012 0.700 0.4694 0.4269 0.4438 0.4903 236.128 0.750 0.4286 0.3860 0.4035 0.4502 237.245 0.800 *o.3898 0.3464 0.3649 0.4111 238.361 0.850 0.3482 0.3071 0.3246 0.3722 239.478 0.900 0.3085 0.2682 0.2860 0.3341 240.594 0.950 0.2677 0.2281 0.2464 0.2957 241.180 0.976 0.2456 0.2061 0.2250 0.2752 241.711 1.000 0.2335 0.1938 0.2131 0.2644 Note: Base metal inner radius= 219.380 cm; outer radius = 241. 711 cm (a) Center of vessel clad

.4-24

TABLE 4.1-17 (continued)

RELATIVE RADIAL DISTRIBUTION OF NEU1RON FLUX (E > 0.1 Me V)

THROUGH THE REACTOR VESSEL WALL CYCLE 11 Radius Vessel Azimuthal Angle (cm) Fraction 60° 75° 90° 219.060 (a) 0.9932 0.9942 0.9916 219.380 0.000 1.0000 1.0000 1.0000 219.910 0.024 1.0112 1.0095 1.0140 220.975 0.071 0.9976 0.9932 1.0036 221.613 0.100 0.9823 0.9758 0.9915 222.730 0.150 0.9431 0.9336 0.9563 223.846 0.200 0.8998 0.8873 0.9168 224.963 0.250 0.8527 0.8393 0.8722 226.079 0.300 0.8054 0.7901 0.8273

• 227.196 228.312 229.429 230.546 0.350 0.400 0.450 0.500 0.7592 0.7110 0.6641 0.6183 0.7414 0.6928 0.6450

.0.5985 0.7836 0.7369 0.6908 0.6456 231.662 0.550 0.5734 0.5532 0.6012 232.779 0.600 0.5294 0.5088 0.5572 233.895 0.650 0.4866 0.4659 0.5142 235.012 0.700 0.4447 0.4240 0.4716 236.128 0.750 0.4042 0.3831 0.4309 237.245 . 0.800. 0.3655 0.3433 0.3920 238.361 0.850 0.3251 0.3039 0.3505 239.478 . . 0.900 0.2863 0.2649 0.3108 240.594 0.950 0.2465 0.2245 0.2699

.241.180 0.976 0.2249 0.2023 0.2477 241.711 1.000 0.2129 0.1898 0.2356.

Note: Base metal inner radius= 219.380 cm; outer radius = 241. 711 cm (a) Center of vessel clad 4-25

TABLE 4.1-18 RELATIVE RADIAL DISTRIBUTION OF IRON ATOM DISPLACEMENT RATE (dpa/sec) IBROUGH IBE REACTOR VESSEL WALL CYCLE 11 Radius Vessel Azimuthal Angle (cm) Fraction ~ ~ 30° 45° 219.060 (a) 1.0164 1.0174 1.0170 1.0159 219.380 0.000 1.0000 1.0000 1.0000 1.0000 219.910 0.024 0.9728 0.9713 0.9718 0.9737 220.975 0.071 0.8992 0.8951 0.8969 0.9019 221.613 0.100 0.8526 0.8459 0.8487 0.8552 222.730 0.150 0.7733 0.7640 0.7675 0.7767 223.846 0.200 0.7009 0.6885 0.6934 0.7042 224.963 0.250 0.6331 0.6195 0.6245 0.6377 226.079 0.300 0.5720 0.5576 0.5629 0.5777 227.196 228.312 229.429 230.546 0.350 0.400 0.450 0.500 0.5179 0.4665 0.4206 0.3795 0.4995 0.4486 0.4023 0.3610 0.5069

. 0.4552 0.4093 0.3680 0.5216 0.4718 0.4266 0.3859 231.662 0.550 0.3423 0.3235 0.3307 0.3490 232.779 0.600 0.3081 0.2894 0.2967 0.3153 233.895 0.650 0.2770 0.2584 0.2658 0.2847 235.012 0.700 0.2482 0.2302 . 0.2374 0.2566 236.128 0.750 0.2220 0.2041 0.2114 0.2308 237.245 0.800 0.1985 0.1797 0.1879 0.2068 238.361 . 0.850 0.1748 0.1572 0.1647 0.1845 239:;478 0.900 0.1531 0.1358 0.1435 0.1637 240.594 0.950 0.1320 0.1148 0.1229 0.1440 241.180 0.976 0.1209 . 0.1036 *0.1120 0.1339 241.711 1.000 0.1151 0.0976 0.1062 0.1287 Note: Base metal inner radius= 219.380 cm; outer radius = 241. 711 cm (a) Center of vessel clad 4-26

TABLE 4.1-18 (continued)

RELATIVE RADIAL DISTRIBUTION OF IRON ATOM DISPLACEMENT RATE (dpa/sec) THROUGH THE REACTOR VESSEL WALL CYCLE 11 Radius Vessel Azimuthal Angle (cm) Fraction 60° 75° 90° 219.060 (a) 1.0170 1.0173 1.0163 219.380 0.000 1.0000 1.0000 1.0000 219.910 0.024 0.9719 0.9713 0.9730 220.975 0.071 0.8971 0.8953 0.8995 221.613 0.100 0.8489 0.8458 0.8530 222.730 0.150 0.7679 0.7634 0.7738 223.846 0.200 0.6937 0.6878 0.7017 224.963 0.250 0.6250 0.6187 0.6340 226.079 0.300 0.5633 0.5569 0.5729

• 227.196 228.312 229.429 230.546 0.350 0.400 0.450 0.500 0.5074 0.4557 0.4098 0.3686 0.4987 0.4475 0.4014 0.3599 0.5189 0.4676 0.4217 0.3806

• I 231.662 0.550 0.3313 0.3224 0.3434 232.779 0.600 0.2972 0.2882 0.3092 233.895 0.650 0.2663 0.2573 0.2781 235.012 0.700 0.2378 0.2290 0.2493 236.128 0.750 0.2118 0.2028 0.2231 237.245 0.800 0.1882 0.1784 0.1996 238.361 0.850 0.1650 0.1559 0.1759 239.478 0;900 0.1437 . 0.1344 0.1542 240.594 0.950 0.1229 0.1132 0.1331 241.180 0:976 *0.1120 0.1019 0.1220 241.711 1.000 0.1061 0.0957 0.1161 Note: Base metal inner radius= 219.380 cm; outer radius = 241. 711 cm (a) Center of vessel clad 4-27

SECTION 5 EVALUATIONS OF SURVEILLANCE CAPSULE DOSIMETRY In this section, the results of the evaluations of the four neutron sensor sets withdrawn from the Palisades reactor vessel are presented. The capsule designation, location within the reactor, and time of withdrawal of each of these dosimetry sets were as listed below.

Azimuthal Withdrawal Irradiation Capsule ID Location Time Time CEFPS)

A240 30° EOC2 7.166e+07 W290 20° EOC5 l.642e+08 1

W290-9 20° EOC9 2.579e+07 WllO 20° EOC IO 3.139e+08 EOC - End of Cycle

• 5.1 Measured Reaction Rates With the exception of Capsule A240, radiometric counting of each of these capsule dosimetry data sets was accomplished by Westinghouse using the procedures discussed in Section 3 of this report. The measured specific activities are included in Appendix A to this report Radiometric counting of the sensors from Capsule A240, on the other hand, was carried out by the Battelle Memorial lnstitute 141* However, in this case, the measured specific activities were not reported.

The irradiation history of the Palisades reactor during the first 10 fuel cycles is also listed in

• Appendix A. The irradiation history was obtained from NUREG-0020, "Licensed Operating Reactors Status Summary Report" for the applicable operating periods. In addition to the reactor power history, for the multiple cycle irradiations Capsules A240, W290, and Wl 10, the flux level adjustment factors for each cycle are also tabulated in Appendi.X A. These adjustment factors

.were determined from the fuel cycle specific forward calculations described in Section 4.1 of this report

• Irradiated during Cycle 9 only.

5-1

Based on the irradiation history, the individual sensor characteristics, capsule gradient corrections, and the measured specific activities, reaction rates averaged over the appropriate irradiation periods and referenced to a core power level of 2530 MWt were computed for the sensor sets removed from Capsules W290, W290-9, and Wl 10. In the case of Capsule A240, reaction rates were developed directly from the derived neutron flux and spectrum averaged reaction cross-sections reported in Reference 4. The computed reaction rates for the multiple foil sensor sets from each of the four internal surveillance capsules are provided in Table 5.1-1.

In regard to the data listed in Table 5.1-1, the fission rate measurements for the 23 8u sensors 235 include corrections for U impurities, the build-in of plutonium isotopes during the long irradiations, and for the effects of photofission reactions. Likewise, the fission rate measurements 237 for the Np sensors include adjustments for photofission reactions occurring over the course of the respective irradiation periods.

It should be noted that the changes in the measured data, as compared to the previous analysis in Reference 6, are due to the elimination of conservatism and the explicit source modelling of each cycle which is specific to this Palisades analysis. This included photofission reaction corrections, as noted above, and use of the long half-life 23 8u (n,t) mes and 237Np (n,t) mes reactions.

5.2 Results of the Least Squares Adjustment Procedure The results of the application of the least .squares adjustment procedure to the four sets of surveillance capsule dosimetry are provided in Table 5.2-1 through 5.2-4. In these tables, the derived exposure experienced by the capsule along with data illustrating the fit of both the trial and adjusted spectra to the measurements are given. Also included in the tabulations are the 1 o 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 "trial calc" represent the absolute calculated neutron flux (E > 1.0 MeV) from Table 4.1-4 averaged over the applicable irradiation periods (Cycles 1 and 2 for Capsule A240, Cycles 1 through 5 for Capsule W290, Cycle 9 for Capsule W290-9, and Cycles 1 through 10 for Capsule WllO) as discussed in Section 3. Thus, the comparisons illustrated in Tables 5.2-1 through 5.2-4 indicate the degree to which the calculated neutron energy spectra matched the measured sensor data before and after adjustment. Absolute comparisons are discussed further in Section 7 of this report.

5-2

TABLE 5.1-1

SUMMARY

OF REACTION RATES DERIVED FROM MULTIPLE FOIL SENSOR SETS WITHDRAWN FROM 1NTERNAL SURVEILLANCE CAPSULES Reaction Rate (rps/nucleus)

Reactiori A240 W290 W290-9 W110 63 Cu (n,<X) 60Co Cd 4.97e-16 9.50e-17 5.72e-17 8.85e-17 54 54 Fe (n,p) Mn 5.58e-14 7.70e-15 4.28e-15 7.0le-15 58 58 Ni (n,p) Co Cd 6.99e-14 l.Ole-14 5.76e-15 9.08e-15 4

6Ti (n,p) 46Sc 9.65e-15 l.5 le-15 9.44e-16 l.43e-15 238 U (n,f) 137Cs Cd 2.53e-l4 l.48e-14 237 Np (n,f) 137Cs Cd 6.26e-14 59 Co (n,y) 60Co l.70e-12 59 Co (n, y) 60Co Cd 2.38e-13

• Note: Cd indicates that the sensor was cadmium covered

• 5-3

TABLE 5.2-1 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE A240 DOSIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 2 Trial Adjusted la Value Value Uncertainn'.

cf>(E > 1.0 MeV) 6.29e+l 1 5.36e+ 11 11%

cf>(E > 0.1 MeV) 1.38e+12 1.18e+ 12 19%

cf>(E < 0.414 eV) 6.62e+l 1 6.33e+ 11 86%

dpa/sec 9.06e-10 7.83e-10 12%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE A240 Reaction Rate (rps/nucleus)

Trial Adjusted MIC MIC Measured Cale. Cale. Trial Adjusted 63 Cu (n,a) Cd 4.97e-16 5.06e-16 4.99e-16 0.98 1.00 54 Fe (n,p) 5.58e-14 6.20e-14 *s.63e-14 0.90 0.99 58 Ni (n,p) Cd 6.99e-14 8.lOe-14 7.13e-14 .. 0.86 0.98 4

6Ti (n,p) 9.65e-15 9.03e~15 9.35e-15 1.07 1.03 5-4

TABLE 5.2-2 DERIVED EXPOSURE RA TES FROM SURVEILLANCE CAPSULE W290 DOSIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 5 Trial Adjusted la Value Value Uncertainty

<f>(E > 1.0 MeV) 6.69e+10 5.63e+10 9%

<f>(E > 0.1 MeV) 1.24e+l 1 1.06e+ll 16%

<f>(E < 0.414 eV) 1.0le+ll 9.56e+10 87%

dpa/sec 9.63e-11 8.25e-11 9%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RA TES:

SURVEILLANCE CAPSULE W290

• Reaction Rate (rps/nucleus)

Trial Adjusted MIC MIC Measured Cale. Cale. Trial Adjusted 63 Cu (n,a) Cd 9.50e-17 9.70e-17 9.41e-17 0.98 1.01 54 Fe (n,p) 7.70e-15 9.00e-15 7.87e-15 0.86 0.98 58 Ni (n,p) Cd l.Ole-14 1.15e-14 l.Ole-14 0.88 1.00

~i (n,p)

• l.51e-15 1.57e-15 1.49e-15 0.96 1.01 238 U (n,f) Cd 2.53e-14 2.95e-14 2.52e-14 0.86 1.00

• 5-5

TABLE 5.2-3 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE W290-9 DOSIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 9 Trial Adjusted la Value Value Uncertainty cp(E > 1.0 Me V) 3.82e+10 3.12e+l0 7%

cp(E > 0.1 MeV) 6.99e+10 5.76e+l0 13%

cp(E < 0.414 eV) 5.61e+10 5.96e+l0 14%

dpa/sec 5.50e-11 4.59e-11 7%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE W290-9 Reaction Rate (rps/nucleus)

Trial Adjusted MIC MIC Measured Cale. Cale. Trial Adjusted 63 Cu (n,a) Cd 5.72e-17 5.85e-17 5.67e-17 0.98 1.01 54 Fe (n,p) 4.28e-15 5.26e-15 4.48e-15 0.81 0.96 58 Ni (n,p) Cd 5.76e-15 6.70e-15

• 5.78e-15 0.86 1.00 46-y'i (n,p) 9.44e-16 9.36e-16 9.16e-16 1.01 1.03 z3su (n,f) Cd l.48e-14 1:70e-14 l.42e-14 0.87 1.04 237 Np (n,f) Cd 6.26e-14 7.66e-14 6.26e-14 0.82 1.00 59 Co (n,y) l.70e-12 l.71e-12 l.70e-12 0.99 1.00 59 Co (n,y) Cd 2.38e-13 3.Sle-13 2.39e-13 0.68 1.00 5-6

TABLE 5.2-4 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE WllO DOSIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 10 Trial Adjusted la Value Value Uncertain!)'.

<f>(E > 1.0 MeV) 6.12e+10 5.06e+10 11%

<f>(E > 0.1 MeV) 1.14e+ll 9.51e+10 18%

<f>(E < 0.414 eV) 9.22e+10 8.75e+10 87%

dpa/sec 8.8le-11 7.44e-11 11%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES SURVEILLANCE CAPSULE Wl 10

• Reaction Rate (rps/nucleus)

Trial Adjusted MIC MIC Measured Gile. Cale. Trial Adjusted ..

63 Cu (n, a) Cd 8.85e-17 8.88e-17 8.77e-17 1.00 1.01 54 Fe (n,p) 7.0le-15 8.23e-15 7.19e-15 0.85 0.97 58 Ni (n,p) Cd 9.08e-15 1.05e-14 9.14e-15 0.86 0.99 46-y'j (n,p). 1.43e-15 1.44e-15 l.40e-15 0.99 1.02

• 5-7

SECTION 6 EVALUATIONS OF REACTOR CAVITY DOSIME1RY 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 Palisades the program was initiated prior to the startup of Cycle 8; and, to date, has included measurement evaluations at the conclusion of Cycles 8, 9, and 11. The evaluation of each of these sets of measured data was accomplished using a consistent approach based on the methodology discussed in Section 3, resulting in an accurate data base to be used in defining the best estimate neutron exposure of the reactor vessel wall.

6.1 Cycle 8 Results 6.1.1 Measured Reaction Rates

• During the Cycle 8 irradiation, four multiple foil sensor sets, and ten stainless steel gradient chains were deployed in the reactor cavity. The capsule identifications associated with each of the multiple foil sensor sets mounted from the dosimetry support bar are listed below.

Reference Bar Capsule Identification Azimuth FOE Shifted Angle Core Midplane Core Bottom 270° oo 60 Not Recovered Not Recovered 280° 100 16° B 290° 20° 26° D E

• 300° 30° 36° Not Recovered

• 315° 45° 39° G 330° 30° 24° No Capsule The contents of each of these irradiation capsules is specified in Appendix B to this report.

• . 6-1

The irradiation history of the Palisades reactor during Cycle 8 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 2530 MWt were computed for each multiple foil sensor and gradient chain segment.

The computed reaction rates for the radiometric foil sensor sets irradiated during Cycle 8 are provided in Table 6.1-1. Corresponding reaction rate data from the ten stainless steel gradient 54

. chains are recorded in Tables 6.1-2 through 6.1-7 for the Fe(n,p ), 58Ni(n,p ), and 59 Co(n, y) reactions, respectively.

In regard to the data listed in Table 6.1-1, the 54Fe(n,p) reaction rates represent an average of the bare and cadmium covered measurements for each capsule. In addition, the fission rate measurements include corrections for 23 5u impurities in the 238 U sensors as well as corrections 23 23 for photofission reactions in both the 8u and Np sensors. It should be noted that the changes in the measured data, as compared to the previous analysis in Reference 6, are due to the elimination of conservatism which is specific to this Palisades analysis. This included photofission reaction corrections, as noted above, and use of the long half-life 237 Np (n,f) 137 Cs reactions.

23 8u (n,f) 137 Cs and 6.1.2 Results of the Least Squares Adjustment Procedure The results of the application of the least squares adjustment procedure to the four sets of multiple foil measurements obtained from the Cycle 8 irradiation are provided in Tables 6.1-8 through 6.1-11. In these tables, the derived exposure experienced at each sensor set location along with data illustrating the fit of both the trial and adjusted spectra to the measurements are given. Also included in -the tabulations are the 1 o uncertainties associated with each of the derived exposure rates.

In regard to the comparisons listed in Tables 6.1-8 through 6.1-11, it should be noted that the columns labeled "trial calc" represent the absolute calculated neutron flux (E > 1.0 Me V) from Table 4.1-1 averaged over the Cycle 8 irradiation period as discussed in Section 3. Thus, the comparisons illustrated in Tables 6.1-8 through 6.1-11 indicate the degree to which the calculated neutron energy spectra matched the measured data before and after adjustment Absolute comparisons of calculation and measurement are discussed further in Section 7 of this report.

6-2

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 54pe(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 cf>(E > 1.0 Me V), cf>(E > 0.1 Me V), and dpa/sec in the reactor cavity.

The resultant axial distributions of cf>(E > 1.0 Me V), cf>(E > 0.1 Me V), and dpa/sec from the gradient chain measurements are given in Tables 6.1-12 through 6.1-14 for the short chains and Tables 6.1-15 through 6.1-17 for the long chains, respectively. The axial distributions of fast neutron flux cf>(E > 1.0 Me V) are depicted graphically in Figures 6.1-1 through 6.1-7. In these graphical presentations, results for axial locations of 0.0 feet relative to the core midplane represent the explicit results of the FERRET evaluations summarized in Tables 6.1-8 through 6.1-11, while results at the remaining axial locations depict the normalized data from the gradient chains .

• 6-3

TABLE 6.1-1

SUMMARY

OF REACTION RATES DERIVED FROM MULTIPLE FOIL SENSOR SETS CYCLE 8 IRRADIATION Reaction Rate (rps/nucleus)

Reaction Cansule B Cansule D Cansule E Cansule G

- 63 Cu (n,<X) Cd 9.76e-19 7.5le-19 1.99e-19 5.53e-19 4

6Ti (n,p) Cd 1.45e-17 l.lle-17 3.34e-18 7.89e-18 54 Fe (n,p) Cd 7.42e-17 5.60e-17 1.75e-17 4.00e-17 58Ni (n,p) Cd l.04e-16 7.8le-17 2.57e-17 5.51e-17 238U (n,f) Cd 3.42e-16 2.52e-16 1.92e-16 z3su (n,f) Cd 9.70e-14 9.15e-14 8.97e-14 23su (n,f) Cdl 4.0le-16 3.04e-16 8.93e-17 1.92e-16 235 1 U (n,f) Cd 8.36e-14 8.22e-14 4.77e-14 9.46e-14 z31Np (n,f) Cd 6.67e-15 4.99e-15 1.65e-15 3.55e-15 59Co (n,y) 4.63e-14 4.34e-14 2.59e-14 4.82e-14 59 Co (n,y) Cd 3.08e-14 2.99e-14 1.75e-14 2.90e-14 Note: Cd indicates that the sensor was cadmium covered.

Paired Uranium Dosimeter (PUD) measurement

TABLE 6.1-2 54 Fe (n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL SHORT GRADIENT CHAINS - CYCLE 8 IRRADIATION Feet from Reaction Rate (rps/nucleus)

Mid12lane ~ 24° 26° 39°

+0.5 4.08e-18 3.19e-18 3;09e-18 2.32e-18 0.0 3.56e-18 3.15e-18

-0.5 4.15e-18 3.38e-18 3.2le-18 2.28e-18

-1.0 4.15e-18 3.42e-18 3.15e-18 2.28e-18

-1.5 4.04e-18 3.03e-18 3.07e-18 2.03e-18

-2.0 3.90e-18 2.88e-18 2.90e-18 2.14e-18

-2.5 3.29e-18 2.70e-18 2.76e-18 1.89e-18

-3.0 3.16e-18 2.46e-18 2.68e-18 1.85e-18

• -3.5

-4.0

-4.5

-5.0 3.02e-18 2.2le-18 l.82e-18 l.23e-18 2.03e-18 1.74e-18 l.24e-18 8.97e-19 2.27e-18 1.9le-18 1.27e-18 1.65e-18 1.44e-18 1.32e-18 9.38e-19

-5.5 8.12e-19 5.36e-19 6.94e-19

• 6-5

TABLE 6.1-3 58 Ni (n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL SHORT GRADIENT CHAINS - CYCLE 8 IRRADIATION Feet from Reaction Rate (rps/nucleus)

Midplane ~ 24° 26° 39°

+0.5 7.69e-17 5.70e-17 5.56e-17 4.3le-17 0.0 5.43e-17

-0.5 7.97e-17 5.88e-17 5.37e-17 3.83e-17

-1.0 7.53e-17 5.58e-l 7 5.4le-17 3.7le-17

-1.5 7.17e-17 5.24e-17 5.60e-17 3.9le-17

-2.0 6.93e-17 5.40e-17 5.56e-17 3.83e-17

-2.5 6.89e-17 4.90e-17 5.33e-17 3.34e-17

-3.0 5.97e-17 4.38e-17 4.9le-17 3.34e-17

-3.5

-4.0

-4.5

-5.0 5.73e-17 4.76e-17 3.56e-17 2.37e-17 3.65e-17 3.05e-17 2.04e-17 l.62e-17 4.68e-17 4.4le-17 3.42e-17 2.30e-17 3.04e-17 2.90e-17 2.4le-17 1.80e-17

-5.5 1.70e-17 l.18e-17 1.48e-17 6-6

TABLE 6.1-4 59 Co (n,y) REACTION RATES DERIVED FROM THE STAINLESS STEEL SHORT GRADIENT CHAINS - CYCLE 8 IRRADIATION Feet from Reaction Rate (rps/nucleus)

Midplane ~ 24° 26° 39°

+o.5 4.82e-14 5.84e-14 4.3~e-14 5.23e-14 0.0 5.84e-14

-0.5 4.82e-14 5.84e-14 4.38e-14 5.28e-14

-1.0 4.82e-14 5.69e-14 4.39e-14 5.18e-14

-1.5 4.73e-14 5.5.9e-14 4.33e-14 5.18e-14

-2.0 4.59e-14 5.38e-14 4.30e-14 5.03e-14 4.46e-14 5.lle-14 4.14e-14 4.89e-14

-2.5

-3.0 4.27e-14 4.8le-14 4.lOe-14 4.57e-14

-3.5 4.12e-14 4.48e-14 3.94e-14 4.44e-14

-4.0 3.85e-14 4.12e-14 3.73e-14 4.06e-14

-4.5 3.28e-14 3.31e-14 3.49e-14 3.78e-14

-5.0 2.80e-14 2.83e-14 2.80e-14 3.03e-14

-5.5 2.57e-14 2.60e-14 2.57e-14

• 6-7

TABLE 6.1-5 54 Fe (n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL LONG GRADIENT CHAINS - CYCLE 8 IRRADIATION Reaction Rate (rps/nucleus)

Feet from Ref. 90° Ref. 260° Ref. 340° Ref. 30° Ref. 150° Ref. 210° Average Mi!!I!lane FOE 0° FOE10° FOE 20° FOE 30° FOE 30° FOE 30° FOE 30°

+8.0 2.39e-19 2.47e-19 2.89e-19 1.69e-19 2.32e-19 1.96e-19 1.99e-19

+7.5 3.lle-19 3.22e-19 3.02e-19 2.25e-19 3.52e-19 2.90e-19 2.89e-19

+7.0 5.18e-19 5.00e-19 5.02e-19 4.35e-19 6.70e-19 3.95e-19 5.00e-19

+6.5 7.83e-19 6.65e-19 7.57e-19 5.41e-19 6.43e-19 5.58e-19 5.81e-19

+6.0 1.16e-18 9.89e-19 9.13e-19 9.46e-19 9.41e-19 9.34e-19

+5.5 1.45e-18 1.47e-18 1.15e-18 1.44e-18 1.29e-18 1.29e-18

+5.0 1.77e-18 1.82e-18 1.49e-18 1.71e-18 1.64e-18 1.61e-18

+4.5 2.14e-18 2.09e-18 1.85e-18 1.89e-18 1.98e-18 1.91e-18

+4.0 2.36e-18 2.44e-18 1.98e-18 2.18e-18 2.17e-18 2.lle-18

+3.5 2.60e-18 2.66e-18 2.34e-18 2.61e-18 2.41e-18 2.45e-18

+3.0 2.70e-18 2.91e-18 2.46e-18 2.55e-18 2.47e-18 2.49e-18

+2.5 2.80e-18 2.98e-18 2.60e-18 2.66e-18 2.76e-18 2.67e-18

+2.0 2.85e-18 3.03e-18 2.58e-18 2.77e-18 2.57e,18 2.64e-18

+1.5 2.77e-18 3.02e-18 2.53e-18 2.74e-18 2.64e-18 2.64e-18

+1.0 2.67e-18 2.82e-18 2.58e-18 2.88e-18 2.60e-18 2.69e-18

+0.5 2.82e-18 2.91e-18 2.65e-18 2.78e-18 2.65e-18 2.70e-18 0.0 3.14e-18 3.06e-18 2.70e-18 2.90e-18 2.51e-18 2.71e-18

-0.5 2.97e-18 2.95e-18 2.68e-18 2.75e-18 2.77e-18 2.73e-18

-1.0 2.70e-18 2.94e-18 2.70e-18 2.79e-18 2.62e-18 2.70e-18

-1.5 2.73e-18 2.94e-18 2.55e-18

• 2.73e-18 2.63e-18 .2.63e-18

-2.0 2.62e-18 3.03e-18 2.72e-18 2.66e-18 2.72e-18 2.70e-18

-2.5 2.62e-18 2.99e-18 2.60e~18 2.63e-18

• 2.47e-18 2.57e-18

-3.0 2.24e-18 2.70e-18 2.48e-18 2.35e-18 2.35e-18 2.39e-18

-3.5 2.48e-18 2.56e-18 2.38e-18 2.33e-18 2.38e-18 2.36e-18

-4.0 2.08e-18 2.27e-18 2.14e-18 2.15e-18 1.95e-18 2.08e-18

-4.5 1.70e-18 1.72e-18 1.81e-18 1.60e-18 1.71e-18 6-8

TABLE 6.1-6 58 Ni (n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL LONG GRADIENT CHAINS - CYCLE 8 IRRADIATION Reaction Rate (rps/nucleus)

Feet from Ref. 90° Ref. 260° Ref. 340° Ref. 30° Ref. 150° Ref. 210° Average Midplane FOE 0° FOE10° FOE 20° FOE 30° FOE 30° FOE 30° FOE 30°

+8.0 4.65e-18 5.33e-18 4.76e-18 3.89e-18 3.76e-18 3.78e-18 3.81e-18

+7.5 6.40e-18 6.19e-18 8.70e-18 4.40e-18 6.8le-18

• 5.51e-18 5.57e-18

+7 ..0 1.15e-17 9.17e-18 9.56e-18 7.83e-18 1.25e-17 7.87e-18 9.41e-18

+6.5 1.49e-17 1.04e-17 1.32e-17 9.69e-18 1.27e-17 1.22e-17 1.15e-17

+6.0 2.0le-17 2.08e-17 1.58e-17 1.76e-17 1.9le-17 1.75e-17

+5.5 2.66e-17 2.72e-17 2.24e-17 2.47e-17 2.41e-17 2.38e-17

+5.0 3.24e-17 3.16e-17 2.67e-17 3.09e-17 3.17e-17 2.97e-17

+4.5 3.76e-17 3.82e-17 3.38e-17 3.97e-17 3.83e-17 3.73e-17

+4.0 4.15e-17 4.0le-17 3.57e-17 4.2le-17 3.96e-17 3.9le-17

+3.5 4.88e-17 4.31e-17 4.12e-17 4.40e-17 4.12e-17 4.21e-17

• • +3.0

+2.5

+2.0

+1.5 5.04e-17 5.08e-17 5.50e-17 4.88e-17 4.65e-17 4.95e-17 5.02e-17 5.02e-17 4.24e-17 4.28e-17 4.20e-17 4.32e-17 4.71e-17 4.63e-17 5.02e-17 4.71e-17 4.6le-17 4.36e-17 4.85e-17 4.77e-17 4.52e-17 4.42e-17 4.69e-17 4.60e-17

+1.0 5.27e-17 5.06e-17 4.52e-17 4.59e-17 4.57e-17 4.56e-17

+0.5 5.08e-17 5.lOe-17 4.40e-17 4.71e-17 4.98e-17 4.70e-17 0.0 4.84e-17 4.83e-17 4.52e-17 5.14e-17 4.77e-17 4.8le-17

-0.5 5.04e-17 5.2le-17 4.60e-17 4.94e-17 4.85e-17 4.80e-17

-1.0 4.92e-17 5.25e-17 4.60e-17 4.44e-17 4.81e-17 4.62e-17

-1.5 4.57e-17 5.43e-17 5.00e-17 4.59e-17 5.18e-17 4.93e-17

-2.0 4.65e-17 5.32e-17 4.72e-17 4.75e-17 4.85e-17 4.77e-17

-2.5 4.88e-17 5.lOe-17 4.76e-17 *4.48e-17 4.44e-17 4.56e-17

-3.0 4.38e-17 4.68e-17 4.44e-17 4.63e-17 4.24e-17 4.44e-17

-3.5 4.15e-17 4.09e-17 4.02e-17 4.2le-17 4.07e-17 4.lOe-17

-4.0 3.45e-17 3.94e-17 3.74e-17 3.73e-17 3.85e-17 3.77e-17

-4.5 2.83e-17 1.21e-17 3.32e-17 3.29e-17 3.29e-17 3.30e-17

• . 6-9

TABLE 6.1-7 59 Co (n, y) REACTION RA TES DERIVED FROM TIIE STAINLESS STEEL LONG GRADIENT CHAINS - CYCLE 8 IRRADIATION Reaction Rate (rps/nucleus)

Feet from Ref. 90° Ref. 260° Ref. 340° Ref. 30° Ref. 150° Ref. 210° Average Mignlane FOE 0° FOE10° FOE 20° FOE 30° FOE 30° FOE 30° FOE 30°

+8.0 2.05e-14 1.75e-14 1.92e-14 2.03e-14 2.lOe-14 l.97e-14 2.04e-14

+7.5 2.24e-14 2.0le-14 2.lOe-14 2.2le-14 2.43e-14 2.27e-14 2.30e-14

+7.0 2.67e-14 2.18e-14 2.43e-14 2.62e-14 3.09e-14 2.62e-14 2.77e-14

+6.5 2.87e-14 2.29e-14 2.54e-14 2.71e-14 3.09e-14 2.92e-14 2.91e-14

+6.0 3.20e-14 2.95e-14 3.lOe-14 3.36e-14 3.15e-14 3.20e-14

+5.5 3.42e-14 3.08e-14 3.18e-14 3.57e-14 3.39e-14 3.38e-14

+5.0 3.89e-14 3.42e-14 3.67e-14 3.92e-14 3.67e-14 3.75e-14

+4.5 4.13e-14 3.70e-14 3.94e-14 4.20e-14 4.02e-14 4.05e-14

+4.0 4.33e-14 3.97e-14 3.90e-14 4.45e-14 4.25e-14 4.20e-14

+3.5 4.72e-14 4.07e-14 4.36e-14 4.74e-14 4.55e-14 4.55e-14

+3.0 4.85e-14 4.40e-14 4.54e-14 5.00e-14 4.68e-14 4.74e-14

+2.5 5.26e-14 4.51e-14 4.65e-14 5.18e-14 4.88e-14 4.90e-14

+2.0 5.66e-14 4.84e-14 4.91e-14 5.37e-14 5.03e-14 5.lOe-14

+1.5 5.30e-14 5.02e-14 5.06e-14 5.56e-14 5.26e-14 5.29e-14

+LO 5.84e-14 5.06e-14 5.08e-14 5.70e-14 5.35e-14 5.38e-14

+0.5 5.75e-14 5.33e-14 5.29e-14 5.81e-14 5.45e-14 5.52e-14 0.0 5.93e-14 5.42e-14 5.35e-14 5.88e-14 5.68e-14 5.64e-14

-0.5 6.02e-14 5.38e-14 5.30e-14 5.90e-14 5.64e-14 5.6le-14

-1.0 5.93e-14 5.Sle-14 5.46e-14 5.86e-14 5.64e-14 5.65e-14

-1.5 5.89e-14 5.33e-14 5.29e-14 5.76e-14 5.54e-14 5.53e-14

-2.0 5.71e-14 5.38e-14 5.32e-14 5.70e-14 5.45e-14 5.49e-14

-2.5 5.48e-14 5.24e-14 5.22e-14 5.51e-14 5.21e-14 5.32e-14

-3.0 5.08e-14 4.93e-14 4.94e-14 5.41e-14 4.93e-14 5.09e-14

-3.5 4.90e-14 4.84e-14 4.78e-14 5.06e-14 4.70e-14 4~85e-14

-4.0 4.45e-14 4.44e-14 4.47e-14 4.81e-14 4.36e-14 4.54e-14

-4.5 4.lle-14 l.46e-14 3.99e-14 4.36e-14 3.95e-14 4.lOe-14 6-10

TABLE 6.1-8 DERIVED EXPOSURE RA TES FROM CAPSULE B DOSIME1RY EV ALDATION 16° AZIMUTH - 280° REFERENCE - CORE MIDPLANE Trial Adjusted la Value Value Uncertainty cf>(E > 1.0 Me V) 1.52e+o9 1.34e+09 7%

cf>(E > 0.1 MeV) 1.34e+10 1.18e+l0 16%

cf>(E < 0.414 eV) 2.92e+o9 6.26e+08 27%

dpa/sec 4.70e-12 4.09e-12 12%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 16° AZIMUTH - 280° REFERENCE - CORE MIDPLANE

• 63 Cu (n, a) Cd Reaction Rate (rps/nucleus)

Measured 9.76e-19 Trial Cale.

1.08e-18 Adjusted Cale.

9.71e-19 MIC Trial 0.90 MIC Adjusted 1.01

~i (n,p) Cd 1.45e-17 l.53e-17 1.42e-17 0.95 1.02 54 Fe (n,p) Cd 7.42e-17 8.69e-17 7.5le-17 0.85 0.99 58 Ni (n,p) Cd 1.04e-16 l.22e-16 l.05e-16 0.85 0.99 238 U (n,f) Cd 3.42e-16 4.46e-16 3.84e-16 0.77 0.89 23su (n,f) Cd 9.70e-14 2.44e-13 9.58e-14 0.40 1.01 23su (n,f) Cd 1 4.0le-16 4.46e-16 3.84e-16 0.90 1.04.

235 1 U (n,f) cd 8.36e-14 2.44e-13 9.58e-14 0.34 0.87 237 Np (n,f) Cd 6.67e-15 6.16e-15 6.02e-15 1.08 1.11 59 Co (n,y) 4.63e-14 1.47e-13 4.68e-14 0.31 0.99.

59 Co (n,y) Cd 3.08e-14 6.84e-14 3.04e-14 0.45 1.01

• Paired Uranium Dosimeter (PUD) measurement 6-11

TABLE 6.1-9 DERIVED EXPOSURE RATES FROM CAPSULED DOSIMETRY EVALUATION 26° AZIMUTH - 290° REFERENCE - CORE MIDPLANE Trial Adjusted la Value Value Uncertainty

<f>(E > 1.0 MeV) l.19e+09 9.97e+08 7%

<f>(E > 0.1 MeV) l.14e+10 9.48e+09 16%

<f>(E < 0.414 eV) 2.90e+09 5.55e+08 28%

dpa/sec 3.93e-12 3.25e-12 13%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 26° AZIMUTH - 290° REFERENCE - CORE MIDPLANE

.Reaction Rate (rps/nucleus) 63 4

Cu (n,cx) Cd 6Ti (n,p) Cd .

Measured 7~51e-19 l:lle~17 Trial Cale.

8.47e-19 l.19e-17 Adjusted Cale.

7.48e-19 l.08e-17 MIC Trial 0.89 0.93 MIC Adjusted 1.00 1.03 54 Fe (n,p) Cd 5.60e-17 6.75e-17 5.66e-17 0.83 0.99 58 .* 0.82 Ni (9,p) Cd 7.8le-17 9.47e-17 7.88e-17 0.99 238 U (n,t) Cd 2.52e-16 3.49e-16 2.87e-16 0.72 0.88 235 U (n,t) Cd

• 9.15e-14 -2.36e-13 9.16e-14 0.39 1.00

. 23su (n,t) Cd 1 3.04e-16 3.49e-16 2.87e-16 0.87 1.06 235 U (n,t) Cd 1 8.22e-14 2.36e-13 9.16e-14 0.35 c0.90 z31Np (n,t) Cd 4.99e-15 5.03e-15 4;56e-15 . 0.99 1.09 59 Co (n,y). 4.34e-14 l.44e-13 4.40e-14 0.30 0.99 59 Co (n, y) Cd 2.99e-14 6.66e-14

• 2.95e-14 0.45 1.01 Paired Uranium Dosimeter (PUD) measurement 6-12

TABLE 6.1-10 DERIVED EXPOSURE RA TES FROM CAPSULE E DOSIMETRY EVALUATION 26° AZIMUTH - 290° REFERENCE - CORE BOTTOM Trial Adjusted 1(J Value Value Uncertaintt cl>(E > 1.0 MeV) l.19e+09 3.45e+08 8%

cl>(E > 0.1 MeV) 1.14e+IO 3.76e+09 17%

cl>(E < 0.414 eV) 2.90e+09 3.40e+08 28%

dpa/sec 3.93e-12 1.25e-12 13%

COMPARIS ON OF MEASURED AND CALCULATED SENSOR REACTION RA TES 26° AZIMUTH - 290° REFERENCE - CORE BOTTOM Reaction Rate (rps/nucleus)

Trial 1 Adjusted MIC MIC Measured Cale. Cale. Trial Adjusted 63 Cu (n, a) Cd 1.99e-19 8.47e-19 2.05e-19 0.23 0.97 46-y'i (n,p) Cd 3.34e-18 1.19e-17 3.25e-18 0.28 1.03 54 Fe (n,p) Cd 1.75e-17 6.75e-17 1.76e-17 0.26 0.99.

58 Ni (n,p) Cd 2.57e-17 9.47e-17 2.55e-17 0.27 1.01 238 U (n,f) Cd 8.93e-17 3.49e-16 9.56e-17 0.26 0.93 235 U (n,f) Cd 4.77e-14 2.36e-13 5.16e-14 0.20 0.92 237 Np (n,f) Cd 1.65e-15 5.03e-15 1.60e-15 0.33 1.03 59 Co (n,y) 2.59e-14 l.44e-13 2.62e-14 0.18 0.99 59 Co (n, y) Cd l.75e-14 6.66e-14 l.73e-14 0.26 1.01

• Calculated using midplane reference spectrum without axial adjustment 6-13

TABLE 6.1-11 DERIVED EXPOSURE RATES FROM CAPSULE G DOSIMETRY EVALUATION 39° AZIMUTH - 315° REFERENCE - CORE MIDPLANE Trial Adjusted 1CJ Value Value Uncertainty cj>(E > 1.0 MeV) 8.60e+08 6.94e+08 7%

cj>(E > 0.1 MeV) 9.08e+09 7.52e+09 17%

cj>(E < 0.414 eV) 2.88e+09 7.60e+08 25%

dpa/sec 3.09e-12 2.52e-12 13%.

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 39° AZIMUTII - 315° REFERENCE - CORE MIDPLANE Reaction Rate (rps/nucleus) 63 Cu (n, a) Cd

~i (n,p) Cd Measured 5.53e-19 7.89e-18 Trial Cale.

6.00e-19 8.39e-18 Adjusted Cale.

5.50e-19 7.74e-18 MIC Trial 0.92 0.94 MIC Adjusted 1.01 1.02 54 Fe (n,p) Cd 4.00e-17 4.76e-17 4.02e-17 0.84 1.00 58 Ni (n,p) Cd 5.51e-17 6.70e-17 5.56e-17 0.82 0.99 238 U (n,f) Cd l.92e-16 2.49e-16 2.00e-16 0.77 0.96 235 U (n,f) Cd 8.97e-14 2.26e-13 9.60e-14 0.40 0.93 23su (n,f) Cd 1 l.92e-16 2.49e-16 2.00e-16 0.77 0.96 23su (n,f) Cd1 9.46e-14 2.26e-13 9.60e-14 0.42 0.99 237Np (n,f) Cd 3.55e-15 3.80e-15 3.29e-15 0.93 1.08 59 Co (n,y) 4.82e-14 1.42e-13 4.86e-14 0.34 0.99 59 Co (n, y) Cd 2.90e-14 6.45e-14 2.88e-14 0.45 1.01 Paired Uranium Dosimeter (PUD) measurement 6-14

TABLE 6.1-12 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSffiON WITHIN THE REACTOR CAVITY SHORTGRADIENT CHAINS - CYCLE 8 IRRADIATION Feet from Neutron Flux (n/cm2-sec)

Midplane .lQ:. 24° 26° 39°

+o.5 1.33e+o9 8.94e+o8 9.87e+o8 7.0le+o8 0.0 9.97e+o8

-0.5 1.35e+o9 9.45e+o8 l.Ole+o9 6.87e+o8

-1.0 1.35e+o9 9.59e+-08 l.03e+-09 6.87e+o8

-1.5 l.31e+o9 8.48e+-08 1.0le+-09 6.13e+-08

-2.0 1.27e+o9 8.07e+-08 9.83e+-08 6.46e+-08

-2.5 l.07e+o9 7.55e+-08 9.26e+o8 5.69e+-08

• -3.0

-3.5

-4.0

-4.5 l.03e+-09 9.8le+-08 7.17e+-08 5.93e+o8 6.87e+-08 5.68e+-08 4.88e+-08 3.47e+-08 8.81e+-08 8.57e+-08 7.26e+-08 6.1 le+-08 5.58e+o8 4.98e+o8 4.34e+o8 3.99e+-08

-5.0 3.99e+o8 2.51e+-08 4.07e+-08 2.83e+-08

-5.5 2.64e+o8 l.50e+-08 2.09e+-08

• 6-15

TABLE 6.1-13 FAST NEUTRON FLUX (E > 0.1 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 8 IRRADIATION Feet from Neutron Flux (n/cm2-sec)

Midplane ~ 24° 26° 39°

+0.5 l.17e+10 8.50e+09 9.40e+o9 7.50e+09 0.0 9.48e+09

-0.5 l.19e+10 8.99e+o9 9.57e+o9 7.36e+09

-1.0 l.19e+10 9.12e+09 9.75e+o9 7.36e+o9

-1.5 l.16e+ 10 8.07e+09 9.57e+o9 6.57e+o9

-2.0 l.12e+l0 7.68e+09 9.35e+o9 6.92e+09

-2.5

-3.0

-3.5

-4.0 9.41e+09 9.06e+09 8.65e+09

. 6.32e+09 7.18e+09 6.54e+o9 5.40e+09 4.64e+09 8.8le+o9 8.38e+o9 8.16e+o9 6.9le+o9 6.10e+09 5.97e+o9 5.33e+o9 4.65e+09

-4.5 5.23e+09 3.30e+09 5.8le+o9 4.28e+09

-5.0 3.51e+09 2.39e+09 3.88e+o9 3.03e+09

-5.5 2.33e+09 1.43e+09 2.24e+09 6-16

TABLE 6.1-14 IRON ATOM DISPLACEMENT RATE AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 8 IRRADIATION Feet from Displacement Rate (dpa/sec)

Midl'!lane ~ 24° 26° 39°

+0.5 4.06e-12 2.92e-12 3.22e-12 2.55e-12 0.0 3.25e-12

-0.5 4.13e-12 3.08e-12 3.28e-12 2.50e-12

-1.0 4.13e-12 3.13e-12 3.35e-12 2.50e-12

-1.5 4.02e-12 2.77e-12 3.28e-12 2.23e-12

-2.0 3.88e-12 2.63e-12 3.21e-12 2.35e-12

-2.5 3.27e-12 2.46e-12 3.02e-12 2.07e-12

• -3.0

-3.5

-4.0

-4.5 3.15e-12 3.0le-12 2.20e-12 l.82e-12 2.24e-12 l.85e-12 1.59e-12 l.13e-12 2.87e-12 2.80e-12 2.37e-12 l.99e-12 2.03e-12 l.81e-12 l.58e-12 l.45e-12

-5.0 1.22e-12 8.19e-13 L33e-12 l.03e-12

-5.5 8.08e-13 4.90e-13 7.62e-13 6-17

TABLE 6.1-15 FAST NEUTRON fLUX (E > 1.0 MeV) AS A FUNCTION e

OF AXIAL POSITION WITHIN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 8 IRRADIATION Feet from Neutron Flux (n/cm2-sec)

Midplane ~ ~ 20° 30°

+8.0 l.02e+08 l.26e+08 7.33e+07

+7.5 1.32e+08 l.32e+08 1.07e+08

+7.0 2.20e+08 2.19e+08 l.84e+08

+6.5 3.33e+08 3.31e+08 2.14e+08

+6.0 4.95e+08 4.32e+08 3.44e+08

+5.5 6.19e+08 6.42e+08 4.76e+08

+5.0 7.52e+08 7.93e+08 5.95e+08

+4.5 9.10e+08 9.lle+08 7.03e+08

+4.0 1.00e+09 l.07e+09 7.78e+08

+3.5 . 1.lle+09 l.16e+09 9.04e+08

+3.0 1.15e+09 1.27e+09 9.19e+08

+2.5 l.19e+09 1.30e+09 9.85e+08

+2.0 1.21e+09 l.32e+09 9.72e+08

+1.5 1.18e+09 1.32e+09 9.71e+08

+1.0 1.13e+09 l.23e+09 9.91e+08

+0.5 1.20e+09 l.27e+09 9.93e+08 0.0 1.34e+09 1.34e+09 9.97e+08

-0.5 l.26e+09 1.29e+09 l.Ole+09

-1.0 1.15e+09 1.28e+09 9.97e+08

-1.5 l.16e+09 l.28e+09 9.70e+08

. -2.0 1.lle+09 1.33e+09 9.94e+08

-2.5 1.12e+09 1.30e+09 9.46e+08

-3.0 9.53e+08 l.18e+09 8.82e+08

-3.5 1.05e+09 1.12e+09 8.69e+08

-4.0 8.85e+08 9.90e+08 7.66e+08

-4.5 7.25e+08 6.29e+08 6-18

TABLE 6.1-16 e FAST NEUTRON FLUX (E > 0.1 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 8 IRRADIATION Feet from Neutron Flux (n/cm2-sec)

Midplane ~ ....ill:. 20° 30°

+8.0 8.97e+08 l.lle+09 6.98e+08

+7.5 l.17e+09 l.16e+09 l.Ole+09

+7.0 l.94e+09 l.93e+09 l.75e+09

+6.5 2.93e+09 2.91e+09 2.04e+09

+6.0 4.36e+09 3.8le+09 3.27e+09

+5.5 5.45e+09 5.66e+09 4.53e+09

+5.0 6.62e+09 6.99e+09 5.66e+09

+4.5 8.02e+09 8.03e+09 6.69e+09

+4.0 8.83e+09 9.39e+09 7.40e+09

• +3.5

+3.0

+2.5

+2.0 9.74e+09 I.Ole+ IO l.05e+10 l.07e+10 l.02e+10 l.12e+10 l.15e+10 l.17e+10 8.60e+09 8.75e+09 9.38e+09 9.24e+09

+1.5 l.04e+10 l.16e+10 9.24e+09

+LO 9.99e+09 l.09e+10 9.43e+09

+0.5 l.06e+10 l.12e+10 9.45e+09 0.0 l.18e+10 l.18e+10 9.48e+09

-0.5 . l.lle+IO l.14e+10 9.58e+09

-1.0 I.Ole+ IO l.13e+10 9.48e+09

-1.5 l.02e+10 l.13e+10 9.23e+09

-2.0 9.8le+09 l.17e+10 9.45e+09

-2.5 9.84e+09 l.15e+10 9.00e+09

-3.0 8.40e+09 l.04e+10 8.40e+09

-3.5 9.28e+09 9.86e+09 8.27e+09

-4.0 7.80e+09 8.72e+09 7.29e+09

-4.5 6.39e+09 5.99e+09 6-19

TABLE 6.1-17 IRON ATOM DISPLACEMENT RA TE AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 8 IRRADIATION Feet from Displacement Rate (dpa/sec)

Midplane ...Q:. 100 20° 30°

+8.0 3.12e-13 3.86e-13 2.39e-13

+7.5 4.05e-13 4.04e-13 3.48e-13

+7.0 6.75e-13 6.72e-13 6.0le-13

+6.5 l.02e-12 l.Ole-12 6.98e-13

+6.0 l.52e-12 l.32e-12 l.12e-12

+5.5 l.89e-12 l.97e-12 l.55e-12

+5.0 2.30e-12 2.43e-12 l.94e-12

+4.5 2.79e-12 2.79e-12 2.29e-12

+4.0 3.07e-12 3.27e-12 2.54e-12

+3.5 3.39e-12 3.56e-12 2.95e-12

+3.0 3.52e-12 3.89e-12 3.00e-12

+2.5 3.65e-12 3.99e-12 3.22e-12

+2.0 3.72e-12 4.05e-12 3.17e-12

+1.5 3.61e-12 4.04e-12 3.17e-12

+LO 3.47e-12 3.78e-12 3.23e-12

+0.5 3.67e-12 3.90e-12 3.24e-12 0.0 4.09e-12 4.09e-12 3.25e-12

-0.5 3.87e-12 3.95e-12 3.28e-12

-1.0 3.52e-12 3.93e-12 3.25e-12

-1.5 3.56e-12

• 3.93e-12 3.17e-12

.. -2.0 3.41e-12 4.06e-12 3.24e-l2

-2.5 3.42e-12 4.00e-12 3.09e-12

-3.0 2.92e-12 3.61e-12 2.88e-12

-3.5 3.23e-12 3.43e-12 2.84e-12

-4.0 2.71e-12 3.03e-12 2.50e-12 *

-4.5 2.22e-12 2.05e-12 6-20

FIGURE 6.1-1 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 16.0 DEGREE TRAVERSE IN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 8 IRRADIATION

-C.I

~ a - - -- -- *-

fl}

I 10

~

e

-=

C.I

=

~

.~

~ ./

ei::

...=

-=

z Q

~

10 R

7 10 I I I I I I I

-6 -5 -4 -3 -2 -1 0 Distance from Core Midplane (ft) 6-21

FIGURE 6.1-2 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 24.0 DEGREE TRAVERSE IN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 8 IRRADIATION 0

~

_., ./'

_,,, ,) r"'"

R 7

10 I I I I I I I

-6 -5 -4 -3 -2 -1 0 Distance from Core Midplane (ft) 6-22

FIGURE 6.1-3 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 26.0 DEGREE TRAVERSE IN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 8 IRRADIATION 1010 Q

• R 7

10 I I I I I I I

-6 -5 -4 -3 -2 -1 0 Distance from Core Midplane (ft) 6-23

FIGURE 6.1-4 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 39.0 DEGREE TRAVERSE IN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 8 IRRADIATION 0

.-~ - - -

~

~*

2 7

10 I I I I I I I

-6 -5 .-4 -3 -2 -1 0 Distance from Core Midplane (ft) 6-24

FIGURE 6.1-5 FAST NEUTRON FLUX (E >LO MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 0.0 DEGREE TRAVERSE IN Tiffi REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 8 IRRADIATION 10 Q

- \..

• • "\ I 10 SI 7

10 I I I I I I I I I I I I I I

-5 -4 -3 -2 -1 0 2 3 4 5 6 7 8 9 Distance from Core Midplane (ft) 6-25

FIGURE 6.1-6 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 20.0 DEGREE TRAVERSE IN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 8 IRRADIATION 0 ~.

" 111

'\ I R ~-

7 10 I I I I I I I I I I I I I I

-5 -4 -3 -2 -1 0 2 3 4 5 6 7 8 9 Distance from Core Midplane (ft) 6-26

FIGURE 6.1-7 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 30.0 DEGREE TRAVERSE IN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 8 IRRADIATION

- <:.I

~

fl}

I IO 0

'E<:.I - ...._

-r;::=

= ~

WI..

'\a.

= '\

....e R

=

~

IO z

7 IO I I I I I I I I

-5 3 -2 -1 0 2 3 4 5 6 7 8 9 Distance from Core Midplane (ft) 6-27

6.2 6.2.1 Cycle 9 Results Measured Reaction Rates During the Cycle 9 irradiation, six multiple foil sensor sets, and ten stainless steel gradient chains were deployed in the reactor cavity. The capsule identifications associated with each of the multiple foil sensor sets mounted from the dosimetry support bar are listed below.

Reference Bar Capsule Identification Azimuth FOE Shifted Angle Core Mid,glane Core Bottom 270° oo 60 A1 c1 280° 100 16° J 290° 20° 26° K L 2

300° 30° 36° Mispositioned 315° 45° 39° N 330° 30° . 24° No Capsule The contents of each of these irradiation capsules is specified in Appendix C to this report.

• The irradiation history of the Palisades reactor during Cycle 9 is also listed in Appendix C. The irradiation history was obtained from NUREG-0020, "Licensed Operating Reactors Status Sununary Report" for the applicable operating period. Based on this reactor operating history, the individual sensor characteristics, and the measured specific activities given in Appendix C, cycle average reaction rates referenced to a core power level of 2530 MWt were computed for each multiple foil* sensor and gradient chain segment.

The computed reaction rates for the radiometric foil sensor sets irradiated during Cycle 9 are .

provided in Table 6.2-1. Corresponding reaction rate *data from the ten stainless steel gradient 54 chains are recorded in Tables 6.2-2 through 6.2-7 for the Fe(n,p), 58Ni(n,p), and 59 Co(n,y) reactions, respectively.

Dosimetry capsules A and C irradiated during Cycles 8 and 9, Reference 6 2

Mispositioned prior to Cycle 9 irradiation, Reference 6.

6-28

In regard to the data listed in Table 6.2-1, the 54Fe(n,p) reaction rates represent an average of the e bare and cadmium covered measurements for each capsule. In addition, the fission rate measurements include corrections for 235u impurities in the 238U sensors as well as corrections for photofission reactions in both the 238u and 237Np sensors. It should be noted that the changes in the measured data, as compared to the previous analysis in Reference 6, are due to the elimination of conservatism which is specific to this Palisades analysis. This included photofission reaction corrections, as noted above, and use of the long half-life 238u (n,f) 137Cs and 237 Np (n,f) 137Cs reactions.

6.1.2 - 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 9 irradiation are provided in Tables 6.2-8 through 6.2-13. In these tables, the derived exposure experienced at each sensor set location along with data illustrating the fit of both the trial and adjusted spectra to the measurements are given. Also included in the tabulations. are the 1a uncertainties associated with each of the derived exposure rates.

• In regard to the comparisons listed in Tables 6.2-8 through 6.2-13, it should be noted that the columns labeled "trial calc" represent the absolute calculated neutron flux (E > 1.0 MeV) from Table 4.1-2 averaged over the Cycle 9 irradiation period as discussed in Section 3. Thus, the comparisons illustrated in Tables 6.2-8 through 6.2-13 indicate the degree to which the calculated neutron energy spectra matched the measured data before and after adjustment Absolute comparisons of calculation and measurement are discussed further in Section 7 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 5Fe(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 ct>CE > 1.0 MeV), ct>CE > 0.1 MeV), and dpa/sec in the reactor cavity.

The resultant axial distributions of ct>CE > 1.0 MeV), ct>CE > 0.1 Me V), and dpa/sec from the gradient chain measurements are given in Tables 6.2-14 through 6.2-16 for the short chains and

• Tables 6.2-17 through 6.2-19 for the long chains, respectively. The distributions of ct>CE > 1.0 6-29

• TABLE 6.2-1

SUMMARY

OF REACTION RATES DERIVED FROM MULTIPLE FOIL SENSOR SETS CYCLE 9 IRRADIATION Reaction Rate (rps/nucleus)

Reaction Ca12sule A 1 Ca12sule C 1 Ca12sule J Ca12sule K Cansule L Ca12sule N 63 Cu (n, a) Cd 7.45e-19 1.13e-19 7.06e-19 6.05e-19 1.69e-19 4.19e-19 4

6Ti (n,p) Cd 1.07e-17 1.81e-18 1.02e-17 8.65e-18 2.58e-18 5.89e-18 54 Fe (n,p) Cd 5.44e-17 9.88e-18 5.13e-17 4.28e-17 1.35e-17 2.96e-17 58 Ni (n,p) Cd 7.43e-17 1.42e-17 7.25e-17 6.12e-17 1.99e-17 4.19e-17 238 U (n,f) Cd 2.44e-16 5.62e-17 2.38e-16 2.33e-16 6.46e-17 1.38e-16 .

mu (n,f) Cd 8.47e-14 4.33e-14 6.13e-14 5.00e-14 3.07e-14 6.42e-14 238 2 U (n,f) Cd 3.23e-16 5.55e-17 mu (n,f) Cd2 7.24e-14 4.05e-14 237 Np (n,f) Cd 4.38e-15 1.lOe-15 3.94e-15 3.73e-15 1.16e-15 2.2le-15 59 Co (n,y) 4.32e-14 2.17e-14 3.57e-14 3.19e-14 1.9le-14 3.55e-14 59 Co (n,y) Cd 2.72e-14 1.51e-14 2.36e-14 2.28e-14 1.30e-14 2.23e-14 Note: Cd indicates that the sensor was cadmium covered.

Dosimetry capsules A and C irradiated during Cycles 8 and 9, Reference 6 2

Paired Uranium Dosimeter (PUD) measurement 6-31

TABLE 6.2-2 54 Fe (n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL SHORT GRADIENT CHAINS - CYCLE 9 IRRADIATION Feet from Reaction Rate (rps/nucleus)

Midplane ~ ~ 24° 26° 39°

+0.5 2.85e-18 2.69e-18 2.26e-18 2.39e-18 l.62e-18 0.0 2.30e-18

-0.38 2.8le-18

-0.5 2.78e-18 2.72e-18 2.21e-18 2.44e-18 l.64e-18

-1.0 2.73e-18 2.71e-18 2.26e-18 2.30e-18 l.60e-18

-1.5 2.66e-18 2.59e-18 2.19e-i8 2.40e-18 l.42e-18

-2.0 2.27e-18 2.39e-18 2.14e-18 2.22e-18 l.38e-18

-2.5 2.12e-18 2.26e-18 l.98e-18 2.lOe-18 l.30e-18

-3.0

-3.5

-4.0

-4.5 l.71e-18 l.62e-18 l.55e-18 9.50e-19 2.02e-18 l.88e-18 l.55e-18 1.13e-18 1.85e-18 l.66e-18 l.43e-18 1.14e-18 1.93e-18 l.81e-18 l.69e-18 1.38e-18 l.29e-18 l.15e-18 9.70e-19 8.29e-19

-5.0 6.67e-19 8.lOe-19 7.64e-19 9.63e-19 6.35e-19

-5.5 6.20e-19 6.02e-19 4.88e-19 Dosimetry capsules A and C irradiated during Cycles 8 and 9, Reference 6 6-32

TABLE 6.2-3 58 Ni (n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL SHORT GRADIENT CHAINS - CYCLE 9 IRRADIATION Feet from Reaction Rate (rps/nucleus)

MidI!lane ~ 16° 24° 26° 39°

+-0.5 4.78e-17 4.68e-17 3.67e-17 3.73e-17 2.59e-17 0.0 3.67e-17

-0.38 4.69e-17

-0.5 4.67e-17 4.6le-17 3.61e-17 3.79e-17 2.58e-17

-1.0 4.54e-17 4.48e-17 3.58e-17 3.77e-17 2.53e-17

-1.5 4.16e-17 4.36e-17 3.57e-17 3.69e-17 2.41e-17

-2.0 4.lle-17 4.13e-17 3.47e-17 3.56e-17 2.34e-17

-2.5 3.62e-17 3.87e-17 3.28e-17 3.44e-17 2.15e-17

• -3.0

-3.5

-4.0

-4.5 2.99e-17 2.66e-17 2.43e-17 1.68e-17 3.58e-17 3.30e-17 2.74e-17 2.13e-17 3.07e-17 2.81e-17 2.44e-17 2.0le-17 3.24e-17 3.06e-17 2.70e-17 2.31e-17 2.04e-17 1.89e-17 l.63e-17 l.42e-l 7

-5.0 l.20e-17 l.48e-17 l.33e-17 l.56e-17 l.03e-l 7

-5.5 l.09e-17 9.73e-18 7.98e-18

• Dosimetry capsules A and C irradiated during Cycles 8 and 9, Reference 6 6-33

TABLE 6.2-4 59 Co (n,y) REACilON RATES DERIVED FROM 1HE STAINLESS STEEL SHORT GRADIENT CHAINS - CYCLE 9 IRRADIATION Feet from Reaction Rate (rps/nucleus)

• Midplane ~ ~ 24° 26° 39°

+o.5 4.38e-14 3.29e-14 3.73e-14 3.08e-14 3.36e-14 0.0 3.68e-14

-0.38 3.31e-14

-0.5 4.38e-14 3.30e-14 3.70e-14 3.lle-14 3.39e-14

-1.0 4.30e-14 3.26e-14 3.67e-14 3.12e-14 3.35e-14

-1.5 4.18e-14 3.21e-14 3.61e-14 3.07e-14 3.31e-14

-2.0 4.13e-14. 3.13e-14 3.53e-14 3.03e-14 3.23e-14

-2.5 3.79e-14 3.06e-14 3.41e-14 2.94e-14 3.13e-14

-3.0

-3.5

-4.0

-4.5 3.84e-14 3.59e-14 3.27e-14 2.67e-14 2.94e-14 2.79e-14 2.64e-14 2.29e-14 3.26e-14 3.09e-14 2.87e-14 2.64e-14 2.89e-14 2.77e-14 2.64e-14 2.47e-14 3.00e-14 2.86e-14 2.67e-14 2.46e-14.

-5.0 2.37e-14 l.93e-14 2.03e-14 2.00e-14 l.89e-14

-5.5 l.83e-14 L81e-14 l.65e-14 Dosimetry capsules A and C irradiated during Cycles 8 and 9, Reference 6 6-34

TABLE 6.2-5 54 Fe (n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL LONG GRADIENT CHAINS - CYCLE 9 IRRADIATION Reaction Rate (rps/nucleus)

Feet from Ref. 90° Ref. 260° Ref. 340° Ref. 30° Ref. 150° Ref. 210° Average MiQ]2lane FOE 0° FOE 10° FOE 20° FOE 30° FOE 30° FOE 30° FOE 30°

. +8.0 l.61e-19 l.46e-19 1.62e-19 1.32e-19 1.35e-19 1.33e-19

+7.5 2.50e-19 2.17e-19 2.31e-19 1.84e-19 2.07e-19 1.95e-19

+7.0 4.17e-19 3.57e-19 3.15e-19 2.96e-19 3.35e-19 3.15e-19

+6.5 6.30e-19 5.61e-19 5.55e-19 4.53e-19 5.05e-19 4.79e-19

+6.0 9.06e-19 7.75e-19 8.08e-19 6.65e-19 6.61e-19 6.63e-19

+5.5 1.18e-18 l.08e-18 l.09e-18 8.65e-19 8.98e-19 8.81e-19

+5.0 l.44e-18 l.35e-18 1.44e-18 1.1 le-18 1.12e-18 1.l2e-18

+4.5 l.72e-18 1.60e-18 l.70e-18 l.27e-18 l.40e-18 l.33e-18

+4.0 l.84e-18 l.81e-18 l.93e-18 l.48e-18 l.50e-18 1.49e-18

+3.5 2.07e-18 1.97e-18 2.06e-18 1.59e-18 l.63e-18 l.61e-18

+3.0 2.17e-18 2.15e-18 2.15e-18 1.67e-18 l.72e-18 l.70e-18

+2.5 2.24e-18 2.23e-18 2.17e-18 l.74e-18 1.88e-18 l.8le-18

+2.0 2.24e-18 2.18e-18 2.27e-18 l.71e-18 l.90e-18 1.80e-18

+1.5 2.27e-18 2.20e-18 2.27e-18 l.77e-18 2.00e-18 l.89e-18

+1.0 2.34e-18 2.31e-18 2.3le-18 l.78e-18 1.87e-18 1.83e-18

+0.5 2.32e-18 2.36e-18 2.28e-18 l.82e-18 l.92e-18 l.87e-18 0.0 2.35e-18 2.35e-18 2.39e-18 l.84e-18 l.95e-18 1.89e-18

-0.5 2.33e-18 2.35e-18 2.42e-18 l.85e-18 2.02e-18 1.93e-18

-1.0 2.34e-18 2.36e-18 2.45e-18 l.77e-18 2.02e-18 l.89e-18

-1.5 2.08e-18 2.40e-18 2.43e-18 l.83e-18 l.88e-18 1.85e-18

-2.0 2.03e-18 2.33e-18 2.36e-18 l.79e-18 l.91e-18 1.85e-18

-2.5 1.87e-18 2.18e-18 2.23e-18 l.76e-18 l.82e-18 1.79e-18

-3.0 1.87e-18 2.06e-18 2.lOe-18 l.74e-18 l.75e-18 l.74e-18

-3.5 l.75e-18 l.95e-18 1.96e-18 l.51e-18 1.68e-18 l.60e-18

-4.0 l.55e-18 l.67e-18 l.68e-18 l.34e-18 l.45e-18 l.39e-18

-4.5 1.33e-18 l.41e-18 l.52e-18 1.l8e-18 l.32e-18 l.25e-18

• 6-35

TABLE 6.2-6 58 Ni (n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL LONG GRADIENT CHAINS - CYCLE 9 IRRADIATION Reaction Rate (rps/nucleus)

Feet from Ref. 90° Ref. 260° Ref. 340° Ref. 30° Ref. 150° Ref. 210° Average Mi!!ulane FOE 0° FOE 100 FOE 20° FOE 30° FOE 30° FOE 30° FOE 30°

+8.0 3.16e-18 2.8le-18 2.97e-18 2.52.e-18 2.52.e-18 2.52.e-18

+7.5 5.18e-18 4.29e-18 4.43e-18 3.66e-18 3.57e-18 3.62.e-18

+7.0 7.41e-18 6.54e-18 6.60e-18 5.59e-18 6.0le-18 5.80e-18

+6.5 1.07e-17 9.85e-18 9.84e-18 8.24e-18 8.62.e-18 8.43e-18

+6.0 1.51e-17 1.40e-17 1.41e-17 1.12.e-17 1.20e-17 1.16e-17

+5.5 2.0le-17 1.87e-17 1.85e-17 1.59e-17 1.66e-17 1.63e-17

+5.0 2.52.e-17 2.47e-17 2.39e-17 1.92.e-17 2.05e-17 1.98e-17

+4.5 2.98e-17 2.88e-17 2.84e-17 2.28e-17 2.44e-17 2.36e-17

+4.0 3.30e-17 3.22.e-17 3.15e-17 2.52.e-17 2.61e-17 2.56e-17

+3.5 3.56e-17 3.52.e-17 3.42.e-17 2.75e-17 3.00e-17 2.87e-17

+3.0 3.74e-17 3.70e-17 3.59e-17 2.82.e-17 3.16e-17 2.99e-17

+2.5 3.77e-17 3.80e-17 3.68e-17 2.97e-17 3.27e-17 3.12.e-17

+2.0 3.77e-17 3.83e-17 3.76e-17 *3.04e-17 3.29e-17 3.17e-17

+1.5 3.86e-17 3.89e-17 3.82e-17 3.04e-17 3.32.e-17 3.18e-17

+1.0 3.91e-17 3.92.e-17 3.81e-17 3.09e-17 3.32.e-17 3.20e-17

+0.5 3.90e-17 3.95e-17 3.82.e-17 3.12.e-17 3.40e-17 3.26e-17 0.0 3.84e-17 3.98e-17 3.89e-17 3.04e-17 3.42.e-17 3.23e-17

-0,5 3.83e-17 4.0le-17 3.83e-17 3.lOe-17 3.43e-17 3.27e-17

-1.0 3.80e-17 4.06e-17 3.89e-17 3.13e-17 3.37e-17 3.25e-17

-1.5 3.6le-17 4.06e-17 3.85e-17 3.13e-17 3.29e-17 3.21e-17

-2.0 3.49e-17 3.91e-17 3.85e-17 3.13e-17 3.24e-17 3.19e-17

-2.5 3.35e-17 3.73e-17 3.72.e-17 3.07e-17 3.20e-17 3.14e-17

-3.0 3.14e-17 3.59e-17 3.51e-17 2.90e-17 3.07e-17 2;98e-17

-3.5 2.94e-17 3.33e-17 3.25e-17 2.65e-17 2.93e-17 2.79e-17

-4.0 2.66e-17. 3.02.e-17 2.87e-17 2.43e-17 2.62.e-17 2.53e-17

-4.5 2.27e-17 2.56e-17 2.44e-17 2.05e-17 2.21e-17 2.13e-17 6-36

TABLE 6.2-7 59 Co (n,y) REACTION RATES DERIVED FROM THE STAINLESS STEEL LONG GRADIENT CHAINS - CYCLE 9 IRRADIATION Reaction Rate (rps/nucleus)

Feet from Ref. 90° Ref. 260° Ref. 340° Ref. 30° Ref. 150° Ref. 210° Average Midnlane FOE 0° FOE 10° FOE 20° FOE 30° FOE 30° FOE 30° FOE 30°

+8.0 1.50e-14 1.52e-14 1.46e-14 1.42e-14 1.35e-14 1.39e-14

+7.5 1.75e-14 1.70e-14 1.64e-14 1.62e-14 l.58e-14 1.60e-14

+7.0 1.94e-14 1.88e-14 1.81e-14 l.79e-14 1.83e-14 1.8le-14

+6.5 2.14e-14 2.07e-14 2.0le-14 l.94e-14 2.0le-14 l.97e-14

+6.0 2.32e-14 2.24e-14 2.2le-14 2.09e-14 2.17e-14 2.13e-14

+5.5 2.55e-14 2.42e-14 2.43e-14 2.28e-14 2.35e-14 2.31e-14

+5.0 2.76e-14 2.59e-14 2.60e-14 2.47e-14 2.52e-14 2.49e-14

+4.5 2.98e-14 2.77e-14 2.79e-14 2.63e-14 2.74e-14 2.69e-14

+4.0 3.19e-14 2.90e-14 2.96e-14 2.78e-14 2.93e-14 2.85e-14

+3.5 3.37e-14 3.03e-14 3.13e-14 2.91e-14 3.lOe-14 3.0le-14

• +3.0

+2.5

+2.0

+1.5

+1.0 3.58e-14 3.72e-14 3.73e-14 4.04e-14 4.18e-14 3.17e-14 3.28e-14 3.39e-14 3.47e-14 3.57e-14 3.27e-14 3.44e-14 3.58e-14 3.69e-14 3.79e-14 3.04e-14 3.14e-14 3.24e-14 3.36e-14 3.43e-14 3.26e-14 3.37e-14 3.50e-14 .

3.62e-14 3.72e-14 3.15e-14 3.26e-14 3.37e-14 3.49e-14 3.57e-14

+0.5 4.24e-14 3.63e-14 3.89e-14 3.53e-14 3.77e-14 3.65e-14 0.0 4.30e-14 3.64e-14 3.92e-14 3.55e-14 3.83e-14 3.69e-14

-0.5 4.30e-14 3.69e-14 3.97e-14 3.60e-14 3.87e-14 3.73e-14

-1.0 4.32e-14 3.67e-14 3.95e-14

• 3.58e-14 3.81e-14 3.70e-14

-1.5 4.27e-14 3.62e-14 3.93e-14 3.58e-14 3.75e-14 3.66e-14

-2.0 4.18e-14 3.55e-14 3.82e-14 3.55e-14 3.80e-14 3.68e-14

-2.5 3.99e-14 3.47e-14 3.72e-14 3.47e-14 3.65e-14 3.56e-14

-3.0 3.83e-14 3.33e-14

• 3.59e-14 3.35e-14 3.57e-14 3.46e-14

-3.5 3.61e-14 3.19e-14 3.41e-14 3.18e-14 3.38e-14 3.28e-14

-4.0 3.40e-14 3.00e-14 3.21e-14 2.99e-14 3.18e-14 3.09e-14

-4.5 3.03e-14 2.74e-14 2.92e-14 2.74e-14 . 2.93e-14 2.84e-14

• 6-37

TABLE 6.2-8 DERIVED EXPOSURE RATES FROM CAPSULE A1 DOSIME1RY EVALUATION 6° AZIMU1H - 270° REFERENCE - CORE MIDPLANE Trial Adjusted 1a Value Value Uncertainty

<f>(E > 1.0 Me V) 1.lle+09 9.57e+08 7%

<f>(E > 0.1 MeV) l.07e+10 8.58e+09 16%

<f>(E < 0.414 eV) 2.68e+09 6.34e+08 26%

dpa/sec 3.68e-12 2.99e-12 12%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 6° AZIMU1H - 270° REFERENCE - CORE MIDPLANE 63 Cu (n,a) Cd

. Reaction Rate (rps/nucleus)

Measured 7.45e-19 Trial Cale.

7.94e-19 Adjusted Cale.

7.38e-19 MIC Trial 0.94 MIC Adjusted 1.01

'*Ti (n,p) Cd 1.07e-17 l.lle-17 l.05e-17 0.96 1.02 54 Fe (n,p) Cd 5.44e-17 6.27e-17 5.5le-17 0.87 0.99 58 Ni (n,p) Cd 7.43e-17 8.80e-17 7.57e-17 0.84 0.98 238 U (n,f) Cd 2.44e-16 3.24e-.16 2.8le-16 0.75 0.87 23su (n,f) Cd 8.47e-14 2.20e-13 8.31e-14 0.39 1.02 23su (n,f) Cd2 3.23e-16 3.24e-16 2.8le-16 1.00 1.15.

235 U (n,f) Cd2 7.24e-14 2.20e-13 8.31e-14 0.33 *0.87 z31Np (n,f) Cd 4.38e-15 4.70e-15 4.13e-15 0.93 1.06 59 Co (n,y) 4.32e-14 1.34.e-13 . 4.36e-14 0.32 0.99 59 Co (n,y) Cd 2.72e-14 6:22e-14 2.69e-14 0.44 1.01 Dosimetry capsule A irradiated. during Cycles 8 and 9, Reforence 6 2

Paired Uranium Dosimeter (PUD) measurement 6-38

TABLE 6.2-9 DERIVED EXPOSURE RA TES FROM CAPSULE c1 DOSIMETRY EVALUATION 6° AZIMUTH - 270° REFERENCE - CORE BOTTOM Trial Adjusted la Value Value Uncertainty

<J>(E > 1.0 MeV) l.lle+09 2.13e+08 8%

<J>(E > 0.1 MeV) l.07e+IO 2.70e+09 17%

<J>(E < 0.414 eV) 2.68e+09 2.72e+08 29%

dpa/sec 3.68e-12 8.73e-13 14%

COMPARJSON OF MEASURED AND CALCULATED SENSOR REACTION RA TES 6° AZIMUTH - 270° REFERENCE - CORE BOTTOM Reaction Rate (rps/nucleus)

• 63 Cu (n, a) Cd 4

6Ti (n,p) Cd Measured l.13e-19 l.81e-18 Trial2 Cale.

7.94e-19 l.lle-17 Adjusted Cale.

l.16e-19 l.78e-18 MIC Trial 0.14 0.16 MIC Adjusted 0.97 1.02 54 Fe (n,p) Cd 9.88e-18 6.27e-17 9.93e-18 0.16 0.99 58Ni (n,p) Cd l.42e-17 8.80e-17 1.42e-17 0.16 1.00 238U (n,t) Cd 5.62e-17 3.24e-16 5.70e-17 0.17 0.99 235U (n,t) Cd 4.33e-14 2.20e-13 4.32e-14 0.20 1.00 238 U (n,t) Cd 3 5.55e-17 3.24e-16 5.70e-17 0.17 0.97 z3su (n,t) Cd3 4.05e-14 2.20e-13 4.32e-14 0.18 0.94 z31Np (n,t) Cd 1.lOe-15 4.70e-15 1.07e-15 0.23 1.03 59Co (n,y) 2.18e-14 l.34e-13 2.21e-14 0.16 0.99 59Co (n,y) Cd l.5le-14 6.22e-14 l.49e-14 0.24 1.01 Dosimetry capsule C irradiated during Cycles 8 and 9, Reference 6 2

Calculated using midplane reference spectrum without axial adjustment 3

Paired Uranium Dosimeter (PUD) measurement 6-39

TABLE 6.2-10 DERIVED EXPOSURE RATES FROM CAPSULE J DOSIMETRY EVALUATION 16° AZIMUTH - 280° REFERENCE - CORE MIDPLANE Trial Adjusted la Value Value Uncertainty

<f>(E > 1.0 Me V) 1.07e+09 8.56e+08 8%

<f>(E > 0.1 MeV) 9.63e+09 7.36e+09 16%

<f>(E < 0.414 eV) 2.20e+09 4.84e+08 27%

dpa/sec 3.37e-12 2.59e-12 13%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 16° AZIMUTH - 280° REFERENCE - CORE MIDPLANE 63 Cu (n,a) Cd Reaction Rate (rps/nucleus)

Measured 7.06e-19 Trial Cale.

8.00e-19 Adjusted Cale.

7.03e-19 MIC Trial 0.88 MIC Adjusted 1.00

~i (n,p) Cd l .02e-17 l.12e-17 l.OOe-17 0.91 1.02 54 Fe (n,p) Cd 5.13e-17 6.25e-17 5.19e-17 0.82 0.99 58 Ni (n,p) Cd 7.25e-17 8.73e-17 7.25e-17 0.83 1.00 238 U (n,f) Cd 2.38e-16 3.16e-16 2.53e-16 0.75 0.94 235 U (n,f) Cd 6.13e-14

• l.82e-13 6.89e-14 0.34 0.89 237 Np (n,f) Cd 3.94e-15 4.38e-15 3.66e-15 0:90 1.08 59 Co (n,y) 3.57e-14 l.lOe-13 3.61e-14 0.32 .0.99 59 Co (n,y) Cd 2.36e-14 5.12e-14 2.33e-14 0.46 1.01 6-40

TABLE 6.2-11 DERIVED EXPOSURE RATES FROM CAPSULE K DOSIMETRY EVALUATION 26° AZIMUTH - 290° REFERENCE - CORE MIDPLANE Trial Adjusted la Value Value Uncertainty

<f>(E > 1.0 Me V) 9.06e+08 7.83e+08 8%

<f>(E > 0.1 MeV) 8.55e+09 6.86e+09 16%

<f>(E < 0.414 eV) 2.18e+09 3.76e+08 29%

dpa/sec 2.97e-12 2.40e-12 12%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 26° AZIMUTH - 290° REFERENCE - CORE MIDPLANE

• 63 Cu (n,a) Cd Reaction Rate (rps/nucleus)

Measured 6.05e-19 Trial Cale.

6.75e-19

• Adjusted Cale.

5.98e-19 MIC Trial 0.90 MIC Adjusted 1.01

'*Ti (n,p) Cd 8.65e-18 9.41e-18 8.47e-18 0.92 1.02 54

• Fe (n,p) Cd *4.28e-17 5.25e-17 4.41e-17 0.82 0.97 58Ni (n,p) Cd 6.12e-17 7.34e-17 6.18e-17 0.83 0.99 238U (n,f) Cd 2.33e-16 2.67e-16 2.29e-16 0.87 1.02 235U (n,f) Cd 5.00e-14 l.77e-13 5.85e-14 0.28 0.85 z31Np (n,f) Cd 3.73e-15 3.80e-15 -

3.44e-15 -0.98 1.08 59Co (n, y). 3.20e-14 l.09e-13 3.25e-14 0.29 0.98 9

~ Co (n, y) Cd 2.28e-14 5.0le-14 2.24e-14 0.46 1.02

• 6-41

TABLE 6.2-12 DERIVED EXPOSURE RA TES FROM CAPSULE L DOSIMETRY EVALUATION 26° AZIMUTH - 290° REFERENCE - CORE BOTTOM Trial Adjusted la Value Value Uncertaint)::

<f>(E > 1.0 MeV) 9.06e+08 2.48e+08 8%

<f>(E > 0.1 MeV) 8.55e+09 2.60e+09 17%

<f>(E < 0.414 eV) 2.18e+09 2.48e+08 27%

dpa/sec 2.97e-12 8.76e-13 13%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 26° AZIMUTH - 290° REFERENCE - CORE BOTTOM Reaction Rate (rps/nucleus) 63 4

Cu (n, a) Cd 6Ti (n,p) Cd

.Measured l.69e-19 2.58e-18 Trial 1 Cale.

6.75e-19 9.41e-18

• Adjusted Cale.

l.72e-19 2.54e-18 MIC Trial 0.25 0:27 MIC Adjusted 0.98 1.02 54 Fe (n,p) Cd l.35e-17 5.25e-17 1.36e-17 0.26 0.99 58 Ni (n,p) Cd l.99e-17 7.34e-17 l.96e-17 0.27 1.02 238 U (n,f) Cd 6.46e-17 2.67e-16 7.02e-17 0.24 0.92 235 U (n,f) Cd 3.07e-14 l.77e-13 3.41e-14 0.17 0.90 z31Np (n,f) Cd l.16e-15 3.80e-15 l.12e-15 0.31 1.04 59 Co (n,y) L92e-14 l.09e-13 l.94e-14 0.18 0.99 59 Co (n,y) Cd l.30e-14 5.0le-14 1.28e-14 0.26. L02 Calculated using midplane reference spectrum without axial adjustment 6-42

TABLE 6.2-13 DERIVED EXPOSURE RA TES FROM CAPSULE N DOSIMETRY EVALUATION 39° AZIMUTH - 315° REFERENCE - CORE MIDPLANE Trial Adjusted la Value Value Uncertainty

<J>(E > 1.0 Me V) 6.70e+08 4.87e+08 8%

<J>(E > 0.1 MeV) 6.98e+09 4.88e+09 17%

<J>(E < 0.414 eV) 2.17e+09 5.26e+08 26%

dpa/sec 2.38e-12 1.67e-12 13%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 39° AZIMUTH - 315° REFERENCE - CORE MIDPLANE Reaction Rate (rps/nucleus)

Trial Adjusted MIC MIC Measured Cale. Cale. Trial Adjusted 63 Cu (n, a) Cd 4.19e-19 4.70e-19 4.16e-19 0.89 1.01

~i (n,p) Cd 5.89e-18 6.57e-18 5.80e-18 0.90 1.02 *'

54

- Fe (n,p) Cd 2.96e-17 3.73e-17 3.00e-17 0.79 0.99 58 Ni (n,p) Cd 4.19e-17 5.25e-17 4.19e-17 0.80 1.00 238 U (n,t) Cd l.38e-16 l.95e-16 1.45e-16 0.71 0.95 235U (n,t) Cd 6.42e-14 1.71e-13 6.95e-14 0.38 0.92 z31Np (n,t) Cd 2.21e-15 2.94e-15 2.13e-15 0.75 1.04 59Co (n,y) 3.55e-14 1.07e-13 3.58e-14 0.33 0.99 59Co (n, y) Cd 2.23e-14 4.87e-14 2.21e-14 0.46 1.01

• 6-43

TABLE 6.2-14 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 9 IRRADIATION Feet from Neutron Flux (n/cm2-sec)

Midplane ~ ~ 24° 26° 39°

+-0.5 9.69e+-08 8.5le+-08 7.69e+-08 7.74e+-08 4.84e+-08 0.0 7.83e+-08

-0.38 8.90e+-08

-0.5 9.44e+-08 8.6le+-08 7.53e+-08 7.92e+-08 4.90e+-08

-1.0 9.28e+-08 8.59e+-08 7.69e+-08 7.45e+-08 4.79e+-08

-1.5 9.03e+-08 8.18e+-08 7.46e+-08 7.79e+-08 4.23e+-08

-2.0 7.7le+-08 7.56e+-08 7.29e+-08 7.21e+-08 4.lle+-08

-2.5 7.22e+-08 7.16e+-08 6.72e+-08 6.82e+-08 3.88e+-08

-3.0

-3.5

-4.0

-4.5 5.82e+-08 5.50e+-08 5.27e+-08 3.23e+-08 6.39e+-08 5.96e+-08 4.9le+-08 3.59e+-08 6.30e+-08 5.65e+-08 4.86e+-08 3.89e+-08 6.26e+-08 5.86e+-08 5.47e+-08 4.47e+-08 3.84e+-08 3.42e+-08 2.90e+-08 2.47e+-08

-5.0 2.27e+-08 2.56e+-08 2.60e+-08 3.12e+-08 l.90e+-08

-5.5 l.96e+-08 2.05e+-08 1.46e+-08 Dosimetry capsules A and C irradiated during Cycles 8 and 9, Reference 6 6-44

TABLE 6.2-15 FAST NEUTRON FLUX (E > 0.1 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 9 IRRADIATION Feet from Neutron Flux (n/cm2-sec)

Midplane ~ ..lE 24° 26° 39°

+0.5 8.70e+09 7.32e+-09 6.74e+-09 6.78e+09 4.85e+09 0.0 6.86e+09

-0.38 7.65e+-09

-0.5 8.47e+09 7.40e+-09 6.60e+09 6.94e+09 4.90e+09

-1.0 8.33e+09 7.38e+-09 6.74e+-09 6.53e+09 4.79e+09 "

-1.5 8.lle+09 7.04e+-09 6.54e+09 6.82e+09 4.23e+09

• -2.0

-2.5

-3.0

-3.5 6.92e+09 6.48e+09 5.22e+09 4.94e+09 6.50e+-09 6.15e+-09 5.49e+-09 5.12e+-09 6.39e+09 5.89e+09 5.52e+09 4.95e+09 6.32e+09 5.98e+09 5.49e+09 5.14e+09 4.lle+09 3.88e+09 3.84e+09 3.43e+09

-4.0 4.73e+09 4.22e+-09 4.26e+-09 4.80e+09 2.90e+09

-4.5 2.90e+09 3.08e+-09 3.41e+09 3.92e+09 2.48e+09 "

-5.0 2.03e+09 2.20e+-09 2.28e+-09 2.74e+09 l.90e+09

-5.5 l.69e+-09 l.79e+09 l.46e+09

• Dosimetry capsules A and C irradiated during Cycles 8 and 9, Reference 6

'6-45

TABLE 6.2-16 IRON ATOM DISPLACEMENT RA TE AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 9 IRRADIATION Feet from Displacement Rate (dpa/sec)

Midplane ~ ~ 24° 26° 39°

+0.5 3.03e-12 2.58e-12 2.35e-12 2.37e-12 l.66e-12 0.0 2.40e-12

-0.38 2.69e-12

-0.5 2.95e-12 2.60e-12 2.31e-12 2.43e-12 l.68e-12

-1.0 2.90e-12 2.60e-12 2.36e-12 2.28e-12 l.64e-12

-1.5 2.82e-12 2.48e-12 2.29e-12 2.38e-12 l.45e-12

-2.0 2.41e-12 2.29e-12 2.23e-12 2.21e-12 l.41e-12

-2.5

-3.0

-3.5

-4.0 2.26e-12 l.82e-12 1.72e-12 l.65e-12 2.16e-12 l.93e-12 l.80e-12 l.48e-12 2.06e-12 l.93e-12 l.73e-12 l.49e-12 2.09e-12 l.92e-12 l.80e-12 l.68e-12 l.33e-12 l.32e-12 l.18e-12 9.94e-13

-4.5 l.Ole-12 1.08e-12

• l.19e-12

• i.37e-12 8.50e-13

-5.0 7.08e-13 7.76e-13 7.97e-13 9.56e-13 6.51e-13 .

-5.5 5.93e-13 6.27e-13. 5.00e-13 Dosimetry capsules A and C irradiated during Cycles 8 and 9, Reference 6 6-46

TABLE 6.2-17 e r FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 9 IRRADIATION Feet from Neutron Flux (n/cm2-sec)

Midplane ~ 100 20° 30°

+8.0 6.56e+o7 5.96e+o7 5.79e+o7 5.51e+o7

+7.5 l.02e+o8 8.84e+o7 8.26e+o7 8.07e+o7

+7.0 l.70e+o8 1.45e+o8 l.13e+o8 l.30e+o8

+6.5 2.57e+o8 2.28e+o8 1.99e+o8 l.98e+o8

+6.0 3.70e+o8 3.15e+o8 2.89e+o8 2.74e+o8

+5.5 4.82e+o8 4.40e+o8 3.90e+o8 3.64e+o8

+5.0 5.86e+o8 5.50e+o8 5.14e+o8 4.62e+o8

+4.5 7.02e+o8 6.51e+o8 6.09e+o8 5.51e+o8

+4.0 7.52e+o8 7.38e+o8 6.9le+o8 6.16e+o8

• +3.5

+3.0

+2.5

+2.0 8.45e+o8 8.83e+o8 9.13e+o8 9.15e+o8 8.0le+o8 8.73e+o8 9.07e+o8 8.87e+o8 7.38e+o8 7.7le+o8 7.77e+o8 8.12e+o8 6.66e+o8 7.0le+o8 7.48e+o8 7.45e+o8 7.80e+o8

+1.5 9.26e+o8 8.96e+o8 8.13e+o8

+l.O 9.55e+o8 9.40e+o8 8.28e+o8 7.55e+o8

+o.5 9.46e+o8 9.62e+o8 8.17e+o8 7.74e+o8 0.0 9.57e+o8 9.57e+o8 8.56e+o8 7.83e+o8

-0.5 9.52e+o8 9.57e+o8 8.66e+o8 7.99e+o8

-1.0 9.53e+o8 9.60e+o8 8.79e+o8 7.83e+o8

-1.5 8.47e+o8 9.76e+o8 8.70e+o8 7.67e+o8

-2.0 8.27e+o8 9.47e+o8 8.43e+o8 7.65e+o8

-2.5 7.6le+o8 8.89e+o8 8.00e+o8 7.40e+o8

-3.0 7.63e+o8 8.39e+o8 7.51e+o8 7.21e+o8

-3.5 7.14e+o8. 7.95e+o8 7.0le+o8 6.60e+o8

-4.0 6.33e+o8 6.8le+o8 6.02e+o8 5.77e+o8

-4.5 5.43e+o8 5.76e+o8 5.46e+o8 5.17e+o8

• 6-47

TABLE 6.2-18 FAST NEUTRON FLUX (E > 0.1 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 9 IRRADIATION Feet from Neutron Flux (n/cm2-sec)

Midplane ~ --1!t... 20° 30°

+8.0 5.89e+08 5.35e+08 4.97e+08 4.83e+08

+7.5 9.13e+08 7.93e+08 7.10e+08 7.07e+08

+7.0 1.53e+09 1.30e+09 9.69e+08 l.14e+09

+6.5 2.30e+09 2.05e+09 1.71e+09 1.73e+09

+6.0 3.32e+09 2.83e+09 2.49e+09 2.40e+09

+5.5 4.33e+09 3.95e+09 3.35e+09 3.19e+09

+5.0 5.26e+09 4.93e+09 4.42e+09 4.05e+09

+4.5 6.30e+09 5.84e+09 5.24e+09 4.83e+09

+4.0 6.75e+09 6.62e+09 5.94e+09 5.40e+09

+3.5

+3.0

+2.5

• 2.0 7.58e+09 7.92e+09 8.19e+09 8.21e+09 7.19e+09 7.84e+09 8.14e+09 7.96e+09 6.35e+09 6.63e+09 6.67e+09 6.98e+09 5.84e+09 6.14e+09 6.55e+09 6.53e+09

+1.5 8.31e+09 8.04e+09 6.99e+09 6.83e+09

+1.0 8.57e+09 8.44e+09 7.12e+09 6.61e+09

+0.5 8.49e+09 8.63e+09 7.02e+09 6.78e+09 0.0 8.59e+09 8.59e+09 7.36e+09 6.86e+09

-0.5 . 8.54e+09 8.59e+09 7.44e+09 7.00e+09

-1.0 8.55e+09 8.62e+09 7.55e+09 6.86e+09

-1.5 7.60e+09 8.76e+09 7.48e+09 6.72e+09

-2.0 7.42e+09 8.50e+09 7.25e+09 6.70e+09

-2.5 6.82e+09

• 7.97e+09 6.87e+09 6.48e+09

-3.0 6.85e+09 7.53e+09 6.45e+09 6.32e+09

-3.5 6.41e+09 7.13e+09 6.03e+09 5.79e+09

-4.0 5.68e+09 6.l le+09 5.17e+09 5.05e+09

-4.5 4.87e+09 5.17e+09 4.69e+09 4.53e+09.

6-48

TABLE 6.2-19 IRON ATOM DISPLACEMENT RA TE AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 9 IRRADIATION Feet from Displacement Rate (dpa/sec)

Midplane JE. 100 20° 30°

+8.0 2.05e-13 l.86e-13 l.75e-13 l.69e-13

+7.5 3.18e-13 2.76e-13 2.50e-13 2.47e-13

+7.0 5.31e-13 4.54e-13 3.41e-13 3.99e-13

+6.5 8.02e-13 7.14e-13 6.02e-13 6.06e-13

+6.0 l.15e-12 9.85e-13 8.75e-13 8.39e-13

+5.5 l.51e-12 l.37e-12 l.18e-12 l.12e-12

+5.0 l.83e-12 l.72e-12 l.55e-12 l.42e-12

+4.5 2.19e-12 2.03e-12 l.84e-12 l.69e-12

+4.0 2.35e-12 2.30e-12 2.09e-12 l.89e-12

• +3.5

+3.0

+2.5

+2.0 2.64e-12 2.76e-12 2.85e-12 2.86e-12 2.50e-12 2.73e-12 2.83e-12 2.77e-12 2.23e-12 2.33e-12 2.35e-12 2.45e-12 2.04e-12 2.15e-12 2.29e-12 2.28e-12

+1.5 2.89e-12 2.80e-12 2.46e-12 2.39e-12

+LO 2.99e-12 2.94e-12 2.5le-12 2.31e-12

+o.5 2.96e-12 3.00e-12 2.47e-12 .237e-12 0.0 2.99e-12 2.99e-12 2.59e-12 2.40e-12

-0.5 2.97e-12 2.99e-12 2.62e-12 2.45e-12

-1.0 2.98e-12. 3.00e-12 2.66e-12 2.40e-12

-1.5 2.65e-12 3.05e-12 2.63e-12 2.35e-12

-2.0 2.58e-12 2.96e-12 2.55e-12 2.34e-12

-2.5 2.38e-12 2.78e-12 2.42e-12 2.27e-12

-3.0 2.38e-12 2.62e-12 2.27e-12 2.2le-12

-3.5 2.23e-12 2.48e-12 2.12e-12 2.02e-12

-4.0 l.98e-12 2.13e-12 l.82e-12 l.77e-12

-4.5 l.70e-12 l.80e-12 l.65e-12 l.58e-12

• 6-49

FIGURE 6.2-1 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 6.0 DEGREE TRAVERSE IN 1HE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLES 8 AND 9 IRRADIATION 1010

'CJ!>>. 0

'fl 10 M

s

~

-~ .v

~

/

c

-ze= 10 R

7 10 I I I I I I I

-6 -5 -4 -3 -2 -1 0 Distance from Core Midplane (ft) 6-50

FIGURE 6.2-2 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 16.0 DEGREE TRAVERSE IN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 9 IRRADIATION

~

C.I Cl.I (I}

10 a

e -

=

C.I

~-

~

~

~

i

&'.: _.......A ......

=

=

-""': i Cl.I

z 10 R

7 10 I I I I I I I

-6 -5 -4 -3 -2 -1 0

. Distance from Core Midplane (ft) 6-51

FIGURE 6.2-3 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 24.0 DEGREE TRAVERSE IN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 9 IRRADIATION 0

~

_,....-4,.

R 7

10 I I I I I I I

-6 -5 -4 -3 -2 -1 0 Distance from Core Midplane (ft) 6-52

FIGURE 6.2-4 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 26.0 DEGREE TRAVERSE IN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 9 IRRADIATION 1010 0

~

• • R I../"

7 10 I I I I I I I

-6 -5 -4 -3 -2 -1 0 Distance from Core Midplane (ft) 6-53

FIGURE 6.2-5 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 39.0 DEGREE TRAVERSE IN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 9 IRRADIATION 0

~ -----

R

.7 IO I I I I I I I

-6 -5 -4 -3 -2 -1 0 Distance from Core Midplane (ft) 6-54

FIGURE 6.2-6 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG TIIE 0.0 DEGREE TRAVERSE IN TIIE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 9 IRRADIATION

~

~

I 10 0

N s

-=

~

=

~

fi:

=

Q z=

~

10 l! ' \.. '

7 10 I I I I I I I I I I I I I I

-5 -4 -3 -2 -1 0 2 3 4 5 *6 7 8 9 Distance from Core Midplane (ft)

• '6-55

FIGURE 6.2-7 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 10.0 DEGREE TRAVERSE IN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 9 IRRADIATION a

lt

'\

R \

7 10 I I I I I I I I I I I I I I

-5 -4 -3 -2 - -1 -0 2-3 4-5 6 7 8 9 Distance from Core Midplane (ft) 6-56

FIGURE 6.2-8 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 20.0 DEGREE TRAVERSE IN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 9 IRRADIATION

- <:,,/

GI

<I) 10 0

I M

e -

<:,,/

=

~

• i£ z

~

I

=

.......:I Q

GI 10 R

' ~

'\

~

10 I I I I I I I I I I I I I I

-5 -4 ~3 -2 -1 0 2 3 4 5 6 7 8 9 Distance from Core Midplane (ft) 6-57

FIGURE 6.2-9 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 30.0 DEGREE TRAVERSE IN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 9 IRRADIATION 1010 a

10 - -

l l 1~

R ~~

10 7

10 I I I I I I I I I I I . I I

-5 -4 -3 -2 -1 0 2 3 4 5 6 7 .* 8. 9

. Distance from Core Midplane (ft) 6-58

6.3 Cycle 10/11 Results 6.3.1 Measured Reaction rates During the Cycle 10/11 irradiation, seven multiple foil sensor sets, and twelve stainless steel gradient chains were deployed in the reactor cavity. The capsule identifications associated with each of the multiple foil sensor sets mounted from the dosimetry support bar were as follows:

Reference Bar Capsule Identification Azimuth FOE Shifted Angle Core Midplane Core Bottom 270° oo 60 0 280° 100 16° p Q 290° 20° 26° R 300° 30° 36° s 315° 45° 39° T 330° 30° 24° u

• The contents of each of these irradiation capsules is specified in Appendix D to this report.

The shipment of irradiation capsules and gradient chains was examined upon receipt at our counting facilities at Waltz Mill. All dosimetry was found to be in good order. As the individual capsules were opened to remove the sensors, the sensor IDs were cross checked against the as-built documentation 131

• A capsule by capsule description of the contents of the irradiation capsules follows. .

• Capsule 0 (270 degrees - Core Midplane): All of the dosimetry IDs matched the data in Table 2-3 in WCAP-13552. All of the metal foils were bright and shiny with the exception of the copper foil which exhibited some discoloration.

Capsule P (280 degrees - Core Midplane): All of the dosimetry IDs matched the data in Table 2-3 in WCAP-13552. All of the metal foils were bright and shiny with the exception of the copper foil which exhibited some discoloration.

6-59 .

Capsule 0 (280 degrees - Bottom of Core): All of the dosimetry IDs matched the data in Table 2-3 in WCAP-13552. The foils in this set had discolorations that were indicative of high temperatures. In particular, the iron foils had a "blue-black" tempered look.

Capsule R (290 degrees - Core Midplane): All of the dosimetry IDs matched the data in Table 2-3 in WCAP-13552. All of the metal foils were bright and shiny with the exception of the copper foil which exhibited some discoloration.

Capsule S (300 degrees - Core Midplane): All of the dosimetry IDs matched the data in Table 2-3 in WCAP-13552 All of the metal foils were bright and shiny with a couple of exceptions.

The copper foil exhibited a blackened discoloration on the edge which covered approximately 15% of the foil surface.

Capsule T (315 degrees - Core Midplane): All of the dosimetry IDs matched the data in Table 2-3 in WCAP-13552. All of the metal foils were bright *and shiny with the exception of the copper foil which exhibited some discoloration.

Capsule U (330 degrees - Core Midplane): All of the dosimetry IDs matched the data in Table 2-3 in WCAP-13552. All of the metal foils were bright and shiny.

In general, the copper foils tended to exhibited discoloration, whereas, the other foils were bright .

and shiny.

The irradiation history of the Palisades reactor during Cycle 10/11 is also listed in Appendix D.

The irradiation history was obtained from NUREG-0020, "Licensed Operating Reactors Status Summary Report" for the applicable operating period. Based on this reactor operating history, the individual sensor characteristics, and the measured specific activities given in Appendix D, cycle average reaction rates referenced to a core power level of 2530 MWt were computed for each multiple foil sensor and gradient chain segment.

The computed reaction rates for the radiometric foil sensor sets irradiated during Cycle 10/11 are provided in Table 6.3-1. Corresponding reaction rate data from the twelve stainless steel gradient chains are recorded in Tables 6.3-2 through 6.3-7 for the 54Fe(n,p), 58Ni(n,p), and 59Co(n,y) reactions, respective!y.

In regard to the data listed in Table 6.3-1, the 54Fe(n,p) reaction rates represent an average of the bare and cadmium covered measurements for each capsule. In addition, the fission rate 6-60

measurements include corrections for 235U impurities in the 238u sensors as well as corrections for photofission reactions in both the 238u and 237Np sensors. It should be noted that the changes in the measured data, as compared to the previous analysis in Reference 6, are due to the elimination of conservatism which is specific to this Palisades analysis. This included photofission reaction corrections, as noted above, and use of the long half-life 238u (n,f) mes and 237 Np (n,f) .mes reactions.

6.3.2 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 10/11 irradiation are provided in Tables 6.3-8 through 6.3-14. In these tables, the derived exposure experienced at each sensor set location along with data illustrating the fit of both the trial and adjusted spectra to the measurements are given. Also included in the tabulations are the 1 a uncertainties associated with each of the derived exposure rates.

In regard to the comparisons listed in Tables 6.3-8 through 6.3-14, it should be noted that the

• columns labeled "trial calc" represent the absolute calculated neutron flux (E > 1.0 MeV) from Table 4.1-3 averaged over the Cycle 10/11 irradiation period as discussed in Section 3. Thus, the comparisons illustrated in Tables 6.3-8 through 6.3-14 indicate the degree to which the calculated neutron energy spectra matched the measured data before and after adjustment Absolute comparisons of calculation and measurement are discussed further in Section 7 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 54pe(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 <f>(E > .1.0 MeV), <f>(E > 0.1 MeV), and dpa/sec in the reactor cavity.

The resultant axial distributions of <f>(E > 1.0 Me V), <f>(E > 0.1 Me V), and dpa/sec from the gradient chain measurements are given in Tables 6.3-15 through 6.3-17 for the short chains and Tables 6.3-18 through 6.3-20 for the long chains, respectively. The distributions of <f>(E > 1.0

• MeV) are depicted graphically in Figures 6.3-1 through 6.3-10. In these graphical presentations, 6-61

• TABLE 6.3-1

SUMMARY

OF REACTION RATES DERIVED FROM MULTIPLE FOIL SENSOR SETS CYCLE 10/11 IRRADIATION Reaction Rate (rps/nucleus)

Reaction Ca12sule 0 Ca12sule P Cii12sule 0 Ca12sule R Ca12sule S Ca12sule T Ca12sule U 63 Cu (n,cx) Cd 5.20e-19 5.42e-19 1.31e-19 4.95e-19 4.15e-19 3.70e-19 4.48e-19 4

6Ti (n,p) Cd 7.02e-18 7.52e-18 1.94e-18 7 .lle-18 5.54e-18 5.5le-18 6.40e-18 54 Fe (n,p) Cd 3.62e-17 3.83e-17 9.9le-18 3.53e-17 2.80e-17 2.73e-17 3.20e-17 58 Ni (n,p) Cd 5.03e-17 5.48e-17 1.47e-17 4.86e-17 7.70e-17 3.83e-17 4.54e-17 238 U (n,f) Cd 2.16e-16 1.97e-16 4.73e-17 1.85e-16 1.52e-16 1.40e-16 1.86e-16 237 Np (n,f) Cd 2.57e-15 2.73e-15 2.61e-15 2.09e-15 1.99e-15 2.04e-15 59 Co (n,y) 2.84e-14 2.74e-14 1.54e-14 2.58e-14 2.54e-14 2.94e-14 3.08e-14 59 Co (n,y) Cd 1.84e-14 1.82e-14 1.03e-14 1.78e-14 1.81e-14 1.86e-14 1.92e-14 Note: Cd indicates that the sensor was cadmium covered.

6-63

TABLE 6.3-2 54 Fe (n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL SHORT GRADIENT CHAINS - CYCLE 10/11 IRRADIATION Feet from Reaction Rate (rps/nucleus)

Midplane ..&:. ..lQ:. 24° 26° 36° 39°

+0.5 2.03e-18 2.16e-18 1.64e-18 1.83e-18 1.59e-18 l.50e-18

-0.5 2.00e-18. 2.06e-18 1.75e-18 1.89e-18 1.56e-18 1.53e-18

-1.0

-1.5 1.73e-18 2.09e-18 1.63e-18 l.91e-18 1.51e-18 l.38e-18

-2.0

-2.5 1.49e-18 1.79e-18 1.65e-18 1.72e-18 1.54e-18 l.16e-18

-3.0

-3.5 1.13e-18 1.58e-18 1.40e-18 l.50e-18 l.26e-18 l.03e-18

-4.0

-4.5 7.14e-19 l.02e-18 l.03e-18 l.15e-18 1.00e-18 8.84e-19

-5.0

-5.5 3.52e-19 4.07e-19 4.94e-19 6.23e-19 5.83e-19 4.33e-19

TABLE 6.3-3 58 Ni (n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL SHORT GRADIENT CHAINS - CYCLE 10/11 IRRADIATION Feet from Reaction Rate (rps/nucleus)

Midplane .£.. ~ 24° 26° 36° 39°

+o.5 3.32e-17 3.65e-17 2.78e-17 3.lOe-17 2.70e-17 2.50e-17

-0.5 3.23e-17 3.62e-17 2.87e-17 3.13e-17 2.68e-17 2.37e-17

-1.0

-1.5 2.88e-17 3.43e-17 2.83e-17 3.15e-17 2.60e-17 2.35e-17

-2.0

-2.5 2.69e-17 2.99e-17 2.87e-17 2.87e-17 2.38e-17 2.18e-17

-3.0

-3.5 2.02e-17 2.74e-17 2.54e-17 2.6le-17 2.21e-17 l.91e-17

-4.0

-4.5 l.24e-17 l.88e-17 l.88e-17 2.07e-17 1.84e-17 l.56e-17

-5.0

-5.5 6.52e-18 7.44e-18 9.40e-18 l.14e-17 l.OOe-17 8.69e-18 6-65.

TABLE 6.3-4 59Co (n,y) REACTION RATES DERIVED FROM THE STAINLESS STEEL SHORT GRADIENT CHAINS - CYCLE 10/11 IRRADIATION Feet from Reaction Rate (rps/nucleus)

Midplane .£.. ~ 24° 26° 36° ..12:.

+o.5 2.66e-14 2.62e-14 2.74e-14 2.30e-14 2.3le-14 2.59e-14

-0.5 2.58e-14 2.59e-14 2.76e-14 2.32e-14 2.32e-14 2.59e-14

-1.0

-1.5 2.54e-14 2.56e-14 2.71e-14 2.28e-14 2.25e-14 2.50e-14

-2.0

-2.5 2.36e-14 2.41e-14 2.59e-14 2.2le-14 2.16e-14 2.38e-14

-3.0

-3.5 2.12e-14 2.24e-14 2.32e-14 2.07e-14 2.0le-14 2.16e-14

-4.0

-4.5 . 1.60e-14 1.86e-14 1.97e-14 1.85e-14 1.79e-14 1.86e-14

-5.0

-5.5 1.36e-14 1.43e-14 1.40e-14 1.39e-14 l.31e-14 1.27e-14

TABLE 6.3-5 54 Fe (n,p) REACTION RATES DERIVED FROM THE STAINLESS STEEL LONG GRADIENT CHAINS - CYCLE 10/11 IRRADIATION Reaction Rate (rps/nucleus)

Feet from Ref. 90° Ref. 260° Ref. 340° Ref. 30° Ref. 150° Ref. 210° Average MidDlane FOE0° FOE 10° FOE 20° FOE 30° FOE 30° FOE 30° FOE 30°

+8.0 l.31e-19 1.15e-19 l.50e-19 l.02e-19 l.27e-19 1.12e-19 l.14e-19

+7.5 l.46e-19 l.87e-19 2.13e-19 l.76e-19 l.87e-19 l.93e-19 1.85e-19

+7.0

+6.5 4.48e-19 4.24e-19 4.18e-19 4.49e-19 4.56e-19 3.90e-19 4.31e-19

+6.0

+5.5 8.33e-19 8.19e-19 7.70e-19 8.48e-19 8.94e-19 8.62e-19 8.68e-19

+5.0

+4.5 l.32e-18 l.24e-18 l.16e-18 l.22e-18 l.32e-18 l.29e-18 1.27e-18

+4.0

+3.5 l.56e-18 l.53e-18 l.45e-18 l.4le-18 l.51e-18 l.54e-18 l.49e-18

• +3.0

+2.5

+2.0

+1.5

+1.0 l.71e-18 l.81e-18 l.57e-18 l.57e-18 l.58e-18 l.58e-18 l.59e-18 l.58e-18 l.74e-18 l.64e-18 l.53e-18 l.67e-18 1.62e-18 1.63e-18

+0.5 l.67e-18 l.65e-18 l.67e-18 l.52e-18 l.74e-18 l.68e-18 1.65e-18 0.0 l.72e-18 l.59e-18 l.72e-18 l.53e-18 l.70e-18 l.83e-18 1.69e-18

-0.5 l.67e-18 l.65e-18 l.68e-18 l.63e-18 l.66e-18 l.67e-18 1.65e-18

-1.0

-1.5 l.65e-18 l.68e-18 l.65e-18 l.58e-18 1.63e-18 l.69e-18 1.63e-18

-2.0

-2.5 l.54e-18 l.49e-18 l.59e-18 l.48e-18 l.59e-18 . l.58e-18 l.55e-18

-3.0

-3.5 1.37e-18 1.32e-18 1.42e-18 l.52e-18 1.42e-18 1.33e-18 1.43e-18

-4.0

-4.5 1.07e-18 9.31e-19 1.19e-18 1.08e-18 l.06e-18 l.05e-18 1.06e-18

• 6-67

TABLE 6.3-6 58 Ni (n,p) REACTION RATES DERJVED FROM THE STAINLESS STEEL LONG GRADIENT CHAINS - CYCLE 10/11 IRRADIATION Reaction Rate (rps/nucleus)

Feet from Ref. 90° Ref. 260° Ref. 340° Ref. 30° Ref. 150° Ref. 210° Average Midnlane FOE 0° FOE 10° FOE 20° FOE 30° FOE 30° FOE 30° FOE 30°

+8.0 2.14e-18 2.59e-18 2.53e-18 2;12e-18 2.48e-18 1.94e-18 2.18e-18

+7.5 3.33e-18 3.48e-18 4.30e-18 3.47e-18 3.64e-18 3.72e-18 3.61e-18

+7.0

+6.5 7.32e-18 7.95e-18 8.41e-18 7.61e-18 8.56e-18 7.77e-18 7.98e-18

+6.0

+5.5 1.54e-17 1.43e-17 1.38e-17 1.50e-17 1.60e-17 1.58e-17 1.56e-17

+5.0

+4.5 2.32e-17 2.21e-17 2.14e-17 2.14e-17 2.23e-17 2.45e-17 2.28e-17

+4.0

+3.5 2.57e-17 2.69e-17 2.55e-17 2.55e-17 2.82e-17 2.74e-17 2.70e-17

+3.0

+2.5 3.03e-17 2.80e-17 2.79e-17 .2.57e-17 2.82e-17 2.82e-17 2.74e-17

+2.0

+1.5 3.lle-17 2.83e-17 2.88e-17 2.68e-17 2.88e-17 2.85e-17 2.80e-17

+1.0

+0.5 2.90e-17 2.69e-17 2.88e-17 2.62e-17 2.85e-17 2.96e-17 2.81e-17 0.0 2.84e-17 2.78e-17 2.85e-17 2.68e-17 2.88e-17 2.94e-17 2.83e-17

-0.5 2.82e-17 2.80e-17 2.93e-17 2.73e-17 2.72e-17 2.91e~11 2. 79e-17

-1.0

. -1.5 2.84e-17 2.88e-17 2.79e-17 2.59e-17

• 2.71e-17 2.85e-17 2.72e-17

-2.0

-2.5 2.64e-17 2.62e-17 2.74e-17 2.72e-17 2.82e-17 2.80e-17 2.78e-17

-3.0

-3.5 2.49e-17 2.30e-17 2.46e-17 2.5ie-17 2.45e-17 2.35e-17 2.44e-17

-4.0

-4.5 1.79e-17 1.79e-17 2.16e-17 1.87e-17 1.89e-17 1.93e-17 1.90e-17 6-68

TABLE 6.3-7 59 Co (n,y) REACTION RATES DERIVED FROM 1HE STAINLESS STEEL LONG GRADIENT CHAINS - CYCLE 10/11 IRRADIATION Reaction Rate (rps/nucleus)

Feet from Ref. 90° Ref. 260° Ref. 340° Ref. 30° Ref. 150° Ref. 210° Average Mi~lane FOE 0° FOE 10° FOE 20° FOE 30° FOE 30° FOE 30° FOE 30°

+8.0 l.08e-14 l.06e-14 l.Ole-14 l.03e-14 l.09e-14 l.04e-14 1.05e-14

+7.5 l.26e-14 l.19e-14 1.13e-14 l.19e-14 l.24e-14 1.20e-14 1.21e-14

+7.0

+6.5 l.55e-14 1.44e-14 l.43e-14 1.44e-14 1.57e-14 1.53e-14 l.51e-14

+6.0

+5.5 1.82e-14 l.68e-14 l.72e-14 1.69e-14 1.83e-14 1.79e-14 1.77e-14

+5.0

+4.5 2.1 le-14 l.89e-14 2.0le-14 1.95e-14 2.13e-14 2.12e-14

• 2.07e-14

+4.0

+3.5 2.37e-14 2.lOe-14 2.26e-14 2.19e-14 2.42e-14 2.35e-14 2.32e-14

• +3.0

+2.5

+2.0

+1.5

+1.0 2.63e-14 2.85e-14 2.27e-14 2.38e-14 2.47e-14 2.69e-14 2.35e-14 2.47e-14 2.62e-14 2.79e-14 2.57e-14 2.74e-14 2.51e-14 2.67e-14

+0.5 2.97e-14 2.45e-14 2.84e-14 2.60e-14 2.91e-14 2.84e-14 2.78e-14 0.0 2.97e-14 2.47e-14 2.88e-14 2.64e-14 2.94e-14 2.86e-14 2.81e-14 ~~

-0.5 3.02e-14 2.52e-14 2.91e-14 2.64e-14 2.96e-14 2.89e-14 2.83e-14

-1.0

-1.5 3.04e-14 2.45e-14 2.88e-14 2.67e-14 2.94e-14 2.84e-14 2.8le-14

-2.0

-2.5 2.89e-14 2.36e-14 2.76e-14 2.55e-14 2.84e-14 2.72e-14 2.70e-14

-3.0

-3.5 2.61e-14 2.16e-14 2.55e-14 2.37e-14 2.57e-14 2.47e-14 2.47e-14

-4.0

-4.5 2.03e-14 1.84e-14 2.19e-14 2.03e-14 2.24e-14 2.06e-14 2.lle-14

• 6-69

TABLE 6.3-8 DERIVED EXPOSURE RATES FROM CAPSULE 0 DOSIMETRY EVALUATION 6° AZIMUTII - 270° REFERENCE - CORE MIDPLANE Trial Adjusted la Value Value Uncertain~

<f>(E > 1.0 Me V) 6.97e+08 6.43e+08 8%

<f>(E > 0.1 MeV) 6.66e+09 5.90e+09 17%

<f>(E < 0.414 eV) l.72e+09 4.22e+08 28%

dpa/sec 2.3le-12 2.05e-12 13%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 6° AZIMUTII - 270° REFERENCE - CORE MIDPLANE 63Cu (n, a) Cd Reaction Rate (rps/nucleus)

Measured 5.20e-19 Trial Cale.

5.36e-19 Adjusted Cale.

5.13e-19 MIC Trial 0.97 MIC Adjusted 1.01 46-yi (n,p) Cd 7.02e-18 7.42e-18 6.97e-18 0.95 1.01 54 Fe (n,p) Cd 3.62e-17 4.09e-17 3.7le-l 7 0.89 0.98 58 Ni (n,p) Cd 5.03e-17 5.7le-17 5.12e-17 0.88 0.98 238 U (n,f) Cd 2.16e-16 2.06e-16 1.93e-16 1.05 1.12 231Np (n,f) Cd 2.57e-15 2.94e-15 2.6le-15 0.87 0.98 59 Co (n,y) 2.84e-14 8.56e-14 2:88e-14 0.33. 0.99 59 Co (n,y) Cd 1.84e-14 3.96e-14 1.83e-14 0.46 1.01 6-70

TABLE 6.3-9 DERIVED EXPOSURE RATES FROM CAPSULE P DOSIMETRY EVALUATION 16° AZIMUTH - 280° REFERENCE - CORE MIDPLANE Trial Adjusted lo Value Value Uncertainty

<f>(E > 1.0 MeV) 7.65e+08 6.5le+08 8%

<f>(E > 0.1 Me V) 7.0le+09 5.97e+09 17%

<f>(E < 0.414 eV) 1.68e+09 3.9le+08 29%

dpa/sec 2.45e-12 2.07e-12 13%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 16° AZIMUTH - 280° REFERENCE - CORE MIDPLANE

• 63 Cu (n, a) Cd Reaction Rate (rps/nucleus)

Measured 5.42e-19 Trial Cale.

5.80e-19 Adjusted Cale.

5.38e-19 MIC Trial 0.93 MIC Adjusted 1.01

'*Ti (n,p) Cd 7.52e-18 8.09e-18 7.45e-18 0.93 1.01 54 Fe (n,p) Cd 3.83e-17 4.49e-17 3.9le-17 0.85 0.98 58 Ni (n,p) Cd 5.48e-17 6.27e-17 5.48e-17 0.87 1.00 238 U (n,t) Cd l.97e-16 2.27e-16 1.95e-16 0.87 1.01 z31Np (n,t) Cd 2.73e-15 3.16e-15 2.69e-15 0.86 1.01 59 Co (n,y) 2.74e-14 8.39e-14 2.78e-14 0.33 0.99 59

,,. Co (n, y) Cd l.82e-14 3.89e-14 1.81e-14 0.47 1.01

• 6-71

TABLE 6.3-10 DERIVED EXPOSURE RA TES FROM CAPSULE Q DOSIMETRY EVALUATION 16° AZIMUTH - 280° REFERENCE - CORE BOTTOM Trial Adjusted la Value Value Uncertain!Y cf>(E > 1.0 MeV) 7.65e+o8 l.79e+08 11%

cf>(E > 0.1 MeV) 7.0le+o9 2.05e+09 25%

cf>(E < 0.414 eV) l.68e+o9 2.10e+08 29%

dpa/sec 2.45e-12 6.80e-13 21%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 16° AZIMUTH - 280° REFERENCE - CORE BOTTOM Reaction Rate (rps/nucleus) 63Cu (n,a) Cd

~i (n,p) Cd Measured l.31e-19 l.98e-18 Trial 1 Cale.

5.80e-19 8.09e-18 Adjusted Cale.

1.33e-19 l.92e-18 MIC Trial 0.23 0.24 MIC Adjusted 0.98 1.03 54 Fe (n,p) Cd 9.9le-18 4.49e-17 l.OOe-17 0.22 0.99 58 Ni (n,p) Cd 1.47e-17 6.27e-17 l.45e-17 0.23 1.01 238 U (n,f) Cd 4.73e-17 2.27e-16 5.lle-17 0.21 0.93 59 Co (n,y) 1.54e-14 8.39e-14 1.56e-14 0.18 0.99 59 Co (n,y) Cd 1.04e-14 3.89e-14 1.03e-14 0.27 1.01 Calculated using midplane reference spectrum without axial adjustment 6-72

TABLE 6.3-11 DERIVED EXPOSURE RA TES FROM CAPSULE R DOSIMETRY EVALUATION 26° AZIMUTH - 290° REFERENCE - CORE MIDPLANE Trial Adjusted la Value Value U ncertaint)'.

cj>(E > 1.0 MeV) 7.06e+o8 6.05e+08 8%

cj>(E > 0.1 MeV) 6.59e+o9 5.73e+09 17%

cj>(E < 0.414 eV) l.66e+o9 3.45e+08 30%

dpa/sec 2.29e-12 1.97e-12 13%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 26° AZIMUTH - 290° REFERENCE - CORE MIDPLANE

• 63 Cu (n, a) Cd Reaction Rate (rps/nucleus)

Measured 4.95e-19 Trial Cale.

5.44e-19 Adjusted Cale.

4.93e-19 MIC Trial 0.91 MIC Adjusted 1.00  !~.

~i (n,p) Cd 7.lle-18 7.54e-18 6.98e-18 0.94 1.02 "(

54 Fe (n,p) Cd 3.53e-17 4.16e-17 3.59e-17 0.85 0.98 .l 58 Ni (n,p) Cd 4.87e-17 5.81e-17 4.94e-17 0.84 0.99 238U (n,f) Cd l.85e-16 2.09e-16 l.80e-16 0.89 1.03 z31Np (n,f) Cd 2.6le-15 2.94e-15 2.56e-15 0.89 1.02 59Co (n,y) 2.58e-14 8.25e-14 2.63e-14 0.31 0.98 59Co (n, y) Cd L78e-14 3.80e-14 l.77e-14 0.47 1.01

• 6-73

TABLE 6.3-12 DERNED EXPOSURE RATES FROM CAPSULES DOSIMETRY EVALUATION 36° AZIMUTH - 300° REFERENCE - CORE MIDPLANE Trial Adjusted Ia Value Value Uncertainty cf>(E > 1.0 MeV) 6.16e+o8 4.89e+08 8%

cf>(E > 0.1 MeV) 6.00e+o9 4.82e+09 17%

cf>(E < 0.414 eV) 1.65e+o9 3.15e+08 31%

dpa/sec 2.07e-12 l.65e-12 14%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 36° AZIMUTH - 300° REFERENCE - CORE MIDPLANE 63 Cu (n, a) Cd Reaction Rate (rps/nucleus)

Measured 4J5e-19 Trial Cale.

4.65e-19 Adjusted Cale.

4.09e-19 MIC Trial 0.89 MIC Adjusted 1.01 4

6Ti (n,p) Cd 5.54e-18 6.45e-18 5.50e-18 0.86 1.01 54 Fe (n,p) Cd 2.80e-17 3.58e-17 2.88e-17 0.78 0.97 58 Ni (n,p) Cd 238 U (n,f) Cd l.52e-16 1.82e-16 1.46e-16 0.84 1.04 z31Np (n,f) Cd 2.09e-15 2.62e-15 2.08e-15 0.80 1.00 59 Co (n,y) 2.52e-14 8.19e-14 2.57e-14 0.31 0.98 59 Co (n, y) Cd l.80e-14 3.75e-14 1.78e-14 0.48 1.01

.6-74

TABLE 6.3-13 DERIVED EXPOSURE RA TES FROM CAPSULE T DOSIMETRY EV ALU ATl ON 39° AZIMUTH - 315° REFERENCE - CORE MIDPLANE Trial Adjusted la Value Value Uncertain~

cl>(E > 1.0 Me V) 5.84e+-08 4.64e+08 8%

cl>(E > 0.1 MeV) 5.80e+-09 4.60e+09 17%

cl>(E < 0.414 eV) l.66e+-09 4.48e+08 28%

dpa/sec 2.00e-12 l.58e-12 14%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 39° AZIMUTH - 315° REFERENCE - CORE MIDPLANE

• 63 Cu (n, a) Cd Reaction Rate (rps/nucleus)

Measured 3.70e-19 Trial Cale.

4.38e-19 Adjusted Cale.

3.72e-19 MIC Trial 0.84 MIC Adjusted 0.99 4

6Ti (n,p) Cd 5.51e-18 6.07e-18 5.39e-18 0.91 1.02 CJ 54 Fe (n,p) Cd 2.73e-17 3.38e-17 2.77e-17 0.81 0.99 58 Ni (n,p) Cd . 3.83e-17 4.73e-17 3.85e-17 0.81 0.99 238U (n,f) Cd l.40e-16 l.72e-16 1.38e-16 0.81 1.01 z31Np (n,f) Cd 1.99e-15 2.51e-15 l.98e-15 0.79 1.01 59 Co (n,y) 2.94e-14 8.22e-14 2.98e-14 0.36 0.99 59 Co (n, y) Cd 1.86e-14 3.75e-14 l.85e-14 0.50 1.01

• 6-75

TABLE 6.3-14 DERIVED EXPOSURE RA TES FROM CAPSULE U DOSIMETRY EVALUATION 24° AZIMUTH - 330° REFERENCE - CORE MIDPLANE Trial Adjusted la Value Value Uncertaintt cp(E > 1.0 MeV) 6.84e+08 5.46e+08 8%

cp(E > 0.1 MeV) 6.45e+09 4.82e+09 17%

cp(E < 0.414 eV) l.69e+09 4.79e+08 27%

dpa/sec 2.24e-12 l.70e-12 13%

COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 24° AZIMUTH - 330° REFERENCE - CORE MIDPLANE Reaction Rate (rps/nucleus)

Trial Adjusted MIC MIC Measured Cale. Cale. Trial Adjusted 63 Cu (n, a) Cd 4.48e-19 5.19e-19 4.46e-19 0.86 1.00 4

6Ti (n,p) Cd 6.40e-18 7.22e-18 6.30e-18 0.89 1.02 54 Fe (n,p) Cd 3.20e-17 4.0le-17 3.29e-17 0.80 0.97 58 Ni (n,p) Cd 4.54e-17 5.60e-17 4.58e-17 0.81 0.99 238 U (n,f) Cd 1.86e-16 2.03e-16 l.68e-16 0.92 1.11 z31Np (n,f) Cd 2.04e-15 2.86e-15 2.12e-15 0.71 0.96 59 Co (n,y) 3.08e-14 8.38e-14 3.lle-14 0.37 0.99 59 Co (n, y) Cd 1.92e-14 3.83e-14 l.9le-14 0.50 1.01 6-76

• TABLE 6.3-15 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 10/11 IRRADIATION Feet from Neutron Flux (n/cm2-sec)

Midplane ~ ..lQ:. 24° 26° 36° 39°

+0.5 6.48e+08 6.66e+o8 5.28e+o8 5.94e+o8 4.95e+08 4.59e+08

-0.5 6.38e+08 6.37e+o8 5.64e+08 6.15e+08 4.83e+08 4.69e+08

-1.0

-1.5 5.52e+08 6.44e+08 5.25e+08 6.22e+08 4.68e+08 4.24e+08

-2.0

-2.5 4.76e+08 5.52e+o8 5.33e+08 5.59e+08 4.77e+08 3.55e+08

-3.0

-3.5 3.63e+08 4.87e+o8 4.52e+08 4.88e+08 3.90e+08 3.16e+08

-4.0

-4.5 2.28e+08 3.15e+08 3.33e+08 3.75e+08 3.lle+08 2.71e+08

-5.0

-5.5 l.13e+08 1.25e+08 1.59e+08 2.03e+08 l.8le+08 l.33e+08 6-77

TABLE 6.3-16 FAST NEUTRON FLUX (E > 0.1 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 10/11 IRRADIATION Feet from Neutron Flux (n/cm2-sec)

Midylane .£.. 16° 24° 26° 36° 39°

+0.5 5.94e+09 6.10e+09 4.67e+09 5.63e+09 4.88e+09 4.55e+09

-0.5 5.85e+09

• 5.83e+09 4.98e+09 5.83e+09 4.76e+09 4.65e+o9

-1.0

-1.5 5.06e+09 5.90e+09 4.63e+09 5.89e+09 4.62e+09 4.2le+09

-2.0

-2.5 4.36e+09 5.05e+09 4.71e+o9 5.29e+09 4.7le+09 3.52e+o9

-3.0

-3.5 3.33e+09 4.46e+09 3.99e+09 4.62e+09 3.85e+09 3.14e+09

-4.0

-4.5 2.09e+09 2.89e+09 2.94e+09 3.55e+09 3.07e+09 2.69e+09

-5.0

-5.5 l.03e+09 l.15e+09 l.41e+09 1.92e+09 l.78e+09 1.32e+09

TABLE 6.3-17 IRON ATOM DISPLACEMENT RA TE AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY

• SHORT GRADIENT CHAINS - CYCLE 10/11 IRRADIATION Feet from . Displacement Rate (dpa/sec)

Midulane ~ ~ 24° 26° 36° 39°

+0.5 2.06e-12 2.12e-12 1.64e-12 1.94e-12 1.67e-12 1.56e-12

-0.5 2.03e-12 2.03e-12 1.75e-12 2.0le-12 1.63e-12 1.60e-12

-1.0

-1.5

• 1.76e-12 2.05e-12 1.63e-12 2.03e-12 1.58e-12 1.44e-12

-2.0

-2.5 1.52e-12 1.76e-12 1.66e-12 1.83e-12 1.62e-12 1.21e-12

-3.0

-3.5 1.16e-12 1.55e-12 1.4le-12 1.59e-12 1.32e-12 1.08e-12

-4.0

-4.5 7.28e-13 l.OOe-12 1.04e-12 1.22e-12 1.05e-12 9.22e-13

-5.0

-5.5 3.59e-13 3.99e-13 4.96e-13 6.62e-13 6.12e-13 4.52e-13 6-79

TABLE 6.3-18 FAST NEUTRON FLUX (E > 1.0 Me V) AS A FUNCTION e

OF AXIAL POSITION WITHIN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 10/11 IRRADIATION Feet from Neutron Flux (n/cm2-sec)

Midplane oo --1.{E_ 20° 30°

+8.0 4.89e+07 4.63e+07 5.67e+07 4.07e+07

+7.5 5.45e+07 7.57e+07 8.05e+07 6.63e+07

+7.0

+6.5 1.67e+08 l.72e+08 1.58e+08 l.55e+08

+6.0

+5.5 3.lle+08 3.31e+08 2.9le+08 3.l le+08

+5.0

+4.5 4.92e+08 5.0le+08 4.37e+08 4.57e+08

+4.0

+3.5

+3.0

+2.5

+2.0 5.84e+08 6.38e+08 6.19e+08 6.34e+08 5.49e+08 5.96e+08 5.32e+08

(

5.81e+08 *

+1.5 6.75e+08 6.36e+08 5.96e+08 5.84e+08

+1.0

+0.5 6.22e+08 6.68e+08 6.31e+08 5.91e+08 0.0 6.43e+08 6.43e+08 6.5le+08 6.05e+08

-0.5 6.23e+08 6.68e+08 6.36e+08 5.93e+08

-LO

-1.5 6.17e+08 6.80e+08 6.24e+08 5.84e+08

-2.0

-2.5 5.75e+08 6.03e+08 6.00e+08 5.55e+08

-3.0

-3.5 5.12e+08 5.33e+08 5.36e+08 5.lle+08

-4.0

-4.5 4.00e+08 3.77e+08 4.51e+08 3.81e+08 6-80

TABLE 6.3-19 e FAST NEUTRON FLUX (E > 0.1 MeV) AS A FUNCTION OF AXIAL POSIDON WITHIN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 10/11 IRRADIATION Feet from Neutron Flux (n/cm2-sec)

Midplane JE_ ...l(L_ 20° 30°

+8.0 4.48e+08 4.25e+08 5.19e+08 3.85e+08

+7.5 5.00e+08 6.94e+08 7.37e+08 6.28e+08

+7.0

+6.5 1.53e+09 l.57e+09 1.45e+09 l.46e+09

+6.0

+5.5 2.85e+09 3.04e+09 . 2.66e+09 2.94e+09

+5.0

+4.5 4.5le+09 4.59e+09 4.00e+09 4.33e+o9

+4.0

• +3.5

+3.0

+2.5

+2.0 5.35e+09 5.85e+09 5.68e+09 5.8le+09 5.03e+o9 5.46e+o9 5.04e+09 5.50e+09

+1.5 6.19e+09 5.83e+09 5.46e+09 5.53e+09

+LO

+o.5 5.71e+09 6.13e+09 5.78e+09 5.60e+09 0.0 5.90e+09 5.90e+09 5.97e+09 5.73e+o9

-0.5 5.71e+09 6.12e+09 5.83e+09 5.6le+09

-1.0

-1.5 5.65e+09 6.24e+09 5.7le+09 5.53e+09

-2.0

-2.5 5.27e+09 5.52e+09 5.50e+o9 5.26e+09

-3.0

-3.5 4.69e+09 4.88e+09 4.9le+09 4.84e+09

-4.0

-4.5 3.67e+09 3.45e+09 4.13e+09 3.60e+09

• 6-81

TABLE 6.3-20 IRON ATOM DISPLACEMENT RA TE AS A FUNCTION OF AXIAL POSITION WITHIN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 10/11 IRRADIATION Feet from Displacement Rate (dpa/sec)

Midplane ~ _l!E_ 20° 30°

+8.0 1.56e-13 1.48e-13 l.80e-13 l.33e-13

+7.5 l.74e-13 2.41e-13 2.56e-13 2.17e-13

+7.0

+6.5 5.32e-13 5.47e-13 5.03e-13 5.05e-13

+6.0

+5.5 9.90e-13 l.06e-12 9.26e-13 l.02e-12

+5.0

+4.5 l.57e-12 l.60e-12 l.39e-12 l.49e-12

+4.0

+3.5 .

+3.0

+2.5

+2.0 l.86e-12 2.03e-12 l.97e-12 2.02e-12 l.75e-12 l.90e-12 l.74e-12 l.90e-12

+1.5 2.15e-12 2.03e-12 l.90e-12 l.91e-12

+LO

+o.5 l.98e-12 2.13e-12 2.0le-12 l.93e-12 0.0 2.05e-12 2.05e-12 2.07e-12 1.97e-12

-0.5 l.99e-12 2.l3e-12 2.03e-12 l.94e-12

-1.0

-1.5 1.96e-12 2.17e-12 1.99e-12 l.91e-12

-2.0

-2.5 l.83e-12 l.92e-12 l.91e-12 l.8le-12

-3.0

-3.5 1.63e-12 l.70e-12 l.71e-12 l.67e-12

-4.0

-4.5 l.27e-12 l.20e-12 l.43e-12 l.24e-12

. 6-82

FIGURE 6.3-1 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 6.0 DEGREE TRAVERSE IN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 10/11 IRRADIATION 1010 N

<:.I

~

Ill I 10 a

=

5<:.I I

• =

~

r;::=

Q

z:=

~

10 ll ~

v-

~

7 10 I I I I I I I

-6 -5 -4 -3 -2 -1 0 Distance from Core Midplane (ft) 6-83

FIGURE 6.3-2 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 16.0 DEGREE TRAVERSE IN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 10/11 IRRADIATION 1010 a

~-

I/ ----

~

/

7 10 I I I I I I I

-6 -5 -4 -3 -2 -1 0 Distance from Core Midplane (ft)

'\

6-84

FIGURE 6.3-3 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 24.0 DEGREE TRAVERSE IN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 10/11 IRRADIATION 1010

<:.I

~ 0

<ll

~

10 E

<:.I - - - ..

== _.._

~ ..--

~

.. v

= ..,,-

....=

... R z

=

~

10 7

10 I I I I I I I

-6 -5 -4 -3 -2 -1 0 Distance from Core Midplane (ft)

• 6-85

FIGURE 6.3-4 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 26.0 DEGREE TRAVERSE IN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 10/11 IRRADIATION a

10 c.-- -

-~

R 10 7

10 I I I I I I I

-6 -5 -4 -3 -2 -1 0 Distance from Core Midplane (ft) 6-86

FIGURE 6.3-5 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 36.0 DEGREE TRAVERSE IN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 10/11 IRRADIATION Q

~--

R 7

IO I I I I I I I

-6 -5 -4 -3 -2 -1 0 Distance from Core Midplane (ft)

• 6-87

FIGURE 6.3-6 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 39.0 DEGREE TRAVERSE IN THE REACTOR CAVITY SHORT GRADIENT CHAINS - CYCLE 10/11 IRRADIATION 0

10

~

l1

~

10 7

10 I I I I I I I

-6 -5 -4 -3 -2 -1 0 Distance from Core Midplane (ft) 6-88

FIGURE 6.3-7 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 0.0 DEGREE TRAVERSE IN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 10/11 IRRADIATION N

Col

~

~

I 10 Q

s - -

=

Col

• ~

~

=

=

Q

=

~

z 10 R

\ ..

7 10 I I I I I I I I I I I I I I

-5 -4 -3 . -2 -1 0 2 3 4 5 6 7 8 9 Distance from Core Midplane (ft)

• 6-89

.FIGURE 6.3-8 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 10.0 DEGREE TRAVERSE IN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 10/11 IRRADIATION 0

10

~~

10 R

7 10 I I I I I I I I I I I I I I

-5 -4 -3 -2 -1 0 2 3 4 5 6 7 8 9 Distance from Core Midplane (ft) 6-90

FIGURE 6.3-9 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 20.0 DEGREE TRAVERSE IN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 10/11 IRRADIATION N

C.I QI

~

I 10 Q

= -

-=

C.I

=

~

• ~

ri: ~

=

~

-=

Q QI

z

10 R

7 10 I I I I I I I I I I I I I I

-5 -4 -3 -2 -1 0 2 3 4 5 6 7 8 9 r- Distance from Core Midplane (ft)

• 6-91

FIGURE 6.3-10 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 30.0 DEGREE TRAVERSE IN THE REACTOR CAVITY LONG GRADIENT CHAINS - CYCLE 10/11 IRRADIATION 1010 Q

10

~.....- ~

'\

10 l! ~

.7 10 I I I I I I I I I I I I

-5 -4 -3 -2 -1 0 2 3 4 5 6 7 8 9 Distance from Core Midplane (ft)

. 6-92

SECTION 7 COMPARISON OF CALCULATIONS WITH MEASUREMENTS As described in Section 3.3, *the best estimate neutron exposure projections for the Palisades reactor vessel were based on a combination of plant specific neutron transport calculations and plant specific measurements. Direct comparisons of the transport calculations with the Palisades measurement data base were used to quantify the biases that may exist due to the transport methodology, reactor modeling, and/or reactor operating characteristics over the respective irradiation periods.

In this section, comparisons of the measurement results from surveillance capsule and reactor cavity dosimetry with corresponding analytical predictions at the measurement locations are presented. These comparisons are provided on two levels. In the first instance, predictions of fast neutron expos~e rates in terms of cl>(E > 1.0 MeV), cl>(E > 0.1 MeV), and dpa/sec are compared with the results of the FERRET least squares adjustment procedure; while, in the second case, calculations of individual sensor reaction rates are compared directly with the measured data from the counting laboratories. It is shown that these two levels of comparison

• yield consistent and similar results, indicating that the least squares adjustment methodology is producing accurate exposure results and that the measurement/calculation (M/C) comparisons yield an accurate plant specific -bias factor that can be applied to neutron transport calculations performed for the Palisades reactor to produce "best estimate" exposure projections for the reactor vessel wall.

7 .1 Comparison of Least Squares Adjustment Results with Calculation In Table 7 .1-1, comparisons of measured and calculated exposure rates for the four surveillance capsule dosimetry sets withdrawn to date as well as for the three cycles of reactor cavity rnidplane dosimetry sets irradiated during Cycles 8, 9, and 10/11 are given. In all cases, the calculated values were based on the .fuel cycle specific exposure calculations averaged over the appropriate irradiation period.

An examination of Table 7 .1-1 indicates that, considering all of the available core rnidplane data, the measured exposure rates were less than the calculated values by factors of 0.831, 0.819, and 0.820 for cl>(E > 1.0 MeV), cl>(E > 0.1 MeV), and dpa/sec, respectively. The standard deviations 7-1

associated with each of the 17 sample data sets were +/-0.067 (8.1 %), +/-0.136 (16.6% ), and +/-0.100 (12.2%), respectively.

7.2 Comparisons of Measured and Calculated Sensor Reaction Rates In Table 7 .2-1, measurement/calculation (M/C) ratios for each fast neutron sensor reaction rate from the surveillance capsule and reactor cavity irradiations are listed. This tabulation, provides a direct comparison, on an absolute basis, of calculation and measurement prior to the application of the least squares adjustment procedure as represented in the FERRET evaluations.

An examination of Table 7 .2-1 shows consistent behavior for all reactions and all measurement points. The overall average M/C ratio for the entire data set has an associated 1cr standard deviation of +/-0.072 (8.1 %). Furthermore, the average M/C bias of 0.879 observed in the reaction rate comparisons is in excellent agreement with the values of 0.831, 0.819, and 0.820 observed in the exposure rate comparisons shown in. Table 7 .1-1.

• 7-2

TABLE 7.1-1 COMPARISON OF MEASURED AND CALCULATED EXPOSURE RATES FROM SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS cj>(E > 1.0 MeV) [n/cm 2-sec]

Calculated Measured MIC Internal Cal!sules A240 (30°) 6.29e+ll 5.36e+ll 0.852 W290 (20°) 6.69e+10 5.63e+10 0.842 W290-9 (20°) 3.82e+10 3.12e+10 0.818 WllO (20°) 6.12e+10 5.06e+10 0.826 6° Cavity Cycle 8 Cycle 9 1.lle+09 9.57e+08 0.863 Cycle 10/11 6.97e+08 6.43e+08 0.922 16° Cavity Cycle 8 1.52e+09 1.34e+09 0.883 Cycle 9 1.07e+09 8.56e+08 0.801 Cycle 10/11 7.65e+08 6.51e+08 0.851 24° Cavity Cycle 8 Cycle 9 Cycle 10/11 6.84e+08 5.46e+08 0.798 26° Cavitv Cycle 8 l.19e+09 9.97e+08 0.835 Cycle 9 9.06e+08 7.83e+08 0.864 Cycle 10/11 7.06e+08 6.05e+08 0.856 36° Cavity Cycle 8 Cycle 9 Cycle 10/11 6.16e+08 4.89e+08 0.794 39° Cavity.

Cycle 8 8.60e+08 6.94e+08 0.807 Cycle 9 6.70e+08 4.87e+08 0.727 Cycle 10/11 5.84e+08 4.64e+08 0.794 Average Bias Factor (K) 0.831 Standard Deviation ( 1o) +/-0.067

• 7-3

TABLE 7.1-1 (continued)

COMPARISON OF MEASURED AND CALCULATED EXPOSURE RATES FROM SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS

<f>(E > 0.1 MeV) [n/cm 2-sec]

Calculated Measured _M&.

Internal Cai!§ules A240 (30°) 1.38e+12 l.18e+12 0.855 W290 (20°) 1.24e+ll l.06e+ll 0.852 W290-9 (20°) 6.99e+10 5.76e+10 0.824 WllO (20°) 1.14e+ll 9.5le+10 0.836 6° Cavity Cycle 8 Cycle 9 1.07e+10 8.58e+09 0.804 Cycle 10/11 6.66e+09 5.90e+09 0.886 16° Cavity Cycle 8 1.34e+10 1.18e+10 0.878 Cycle 9 9.63e+09 7.36e+09 0.764 Cycle 10/11 7.0le+09 5.97e+09 0.851 24° Cavity Cycle 8 Cycle 9 Cycle 10/11 6.45e+09 4.82e+09 0.748 26° Cavity Cycle 8 l.14e+10 9.48e+09 0.834 Cycle 9 8.55e+09 6.86e+09 0.803 Cycle 10/11 6.59e+09 5.73e+09 0.870 36° Cavity Cycle 8 Cycle 9 Cycle 10/11 6.00e+09 4.82e+09 0.804 39° Cavity Cycle 8 9.08e+09 7.52e+09 0.828 Cycle 9 6.98e+09 4.88e+09 0.698 Cycle 10/11 5.80e+09 4.60e+09 0.793 Average Bias Factor (K) 0.819 Standard Deviation (lo) +/-0.136 7-4

TABLE 7.1-1 (continued)

COMPARISON OF MEASURED AND CALCULATED EXPOSURE RATES FROM SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS Iron Displacements (dpa/sec)

Calculated Measured MIC Internal Causules A240 (30°) 9.06e-10 7.83e-10 0.865 W290 (20°) 9.63e-11 8.25e-11 0.857 W290-9 (20°) 5.50e-11 4.59e-11 0.834 WllO (20°) 8.81e-11 7.44e-11 0.844 6° Cavity Cycle 8 Cycle 9 3.68e-12 2.99e-12 0.811 Cycle 10/11 2.3le-12 2.05e-12 0.888 16° Cavity Cycle 8 4.70e-12 4.09e-12 0.872 Cycle 9 3.37e-12 2.59e-12 0.769

• Cycle 10/11 24° Cavity Cycle 8 Cycle 9 Cycle 10/11 2.45e-12 2.24e-12 2.07e-12 1.70e-12 0.848 0.757 26° Cavity Cycle 8 3.93e-12 3.25e-12 0.827 Cycle 9 2.97e-12 2.40e-12 0.808 Cycle 10/11 2.29e-12 1.97e-12 0.861 36° Cavity Cycle 8 Cycle 9 Cycle 10/11 2.07e-12 1.65e-12 0.798 39° Cavity Cycle 8 3.09e-12 2.52e-12 0.817 Cycle 9 2.38e-12 1.67e-12 0.702 Cycle 10/11 2.00e-12 1.58e-12 0.790 Average Bias Factor (K) 0.820 Standard Deviation (lo) +/-0.100

• 7-5

TABLE 7.2-1 COMPARISON OF MEASURED AND CALCULATED NEUTRON SENSOR REACTION RA TES FROM SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS Cu63Cn.a) Ti46(n,p) Fe54(n,p) Ni58(n,p) U238(n.O Np237(n,0 Internal A240 (30°) 0.982 1.069 0.900 0.863 W290 (20°) 0.979 0.962 0.856 0.878 0.858 W290-9 (20°) 0.978 1.009 0.814 0.860 0.871 0.817 WllO (20°) 0.997 0.993 0.852 0.865 6° Cavin:

Cycle 8 Cycle 9 0.938 0.964 0.868 *o.844 0.875 0.932 Cycle 10/11 0.970 0.946 0.885 0.881 1.049 0.874 16° Cavi~

Cycle 8 0.904 0.948 0.854 0.852 0.833 1.083 Cycle 9 0.883 0.911 0.821 0.830 0.753 0.900 Cycle 10/11 24° Cavi~

Cycle 8 Cycle 9 Cycle 10/11 0.934 0.863 0.930 0.886 0.853 0.798 0.874 0.811 0.868 0.916 0.864 0.713 26° Cavi~

Cycle 8 0.887 0.933 0.830 0.825 0.797 0.992 Cycle 9 0.896 0.919 0.815 0.834 0.873 0.982 Cycle 10/11 0.910 0.943 0.849 0.838 0.885 0.888 36° Cavi~

cycle 8 Cycle 9 Cycle 10/11 0.892 0.859 0.782 0.835 0.798 39° Cavi~

Cycle 8 0.922 0.940 0.840 0.822 0.771 0.934 Cycle 9 0.891 0.896 0.794 0.798 0.708 0.752 Cycle 10/11 0.845 0.908 0.808 0.810 0.814 0.793 Average 0.922 0.942 0.836 0.843 0.847 0.880 Std. Dev. (lo) 0.046 0.049 0.033 0.026 0.079 0.101 Average Bias Factor (K)

Standard Deviation (1 a) 0.879

+/-0.072 e 7-6

SECTION 8 BEST ESTIMATE NEUTRON EXPOSURE OF THE REACTOR VESSEL In this section the measurement results provided in Sections 5 and 6 are combined with the results of the neutron transport calculations described in Section 4 to establish a mapping of the best estimate neutron exposure of the beltline region of the Palisades reactor vessel for Cycle 11.

8.1 Exposure Distributions Within the Beltline Region As described in Section 3.3 of this report, the best estimate vessel exposure was determined from the following relationship:

~D_.,._ = K ~ Cale.

~

where: 'l>BestEst = The best estimate fast neutron exposure at the location of interest.

K = The plant specific measurement/calculation (MIC) bias factor derived from all available surveillance capsule and reactor cavity dosimetry data.

<PCalc. = The absolute calculated fast neutron exposure at the location of interest.

From the data provided in Table 7 .1-1, the plant specific bias factors (K) to be applied to the calculated exposure values given in Section 4 and the axial variations in Section 6 were as follows:

'I> (E > 1.0 Me V) 0.831 +/- 0.067 (8.1 % )

'l>(E > 0.1 MeV) 0.819 +/- 0.136 (16.6%)

• dpa 0.820 +/- 0.100 (12.2%)

8-1

These bias factors were based on the results of the continuous monitoring program at Palisades that has provide measured data from four internal surveillance capsules and thirteen reactor cavity sensor sets through the first 11 cycles of operation.

The uncertainties listed with the individual bias factors are at the 1 a level and are given on an absolute.

8.1.1 Exposure Accrued During Cycles 1 through 11 To assess the incremental exposure resulting from irradiation during Cycles 1 through 11, the bias factors listed in Section 8.1 were applied directly to the calculated values from Section 4.1 including the axial peaking factor for the vessel clad/base metal interlace to produce best estimate fluence levels characteristic of the midplane of the reactor core. The axial gradient chain measurements were then employed to develop the complete axial traverse along the vessel wall.

The best estimate results applicable to the vessel inner surface are incorporated into Tables 8.1-1 through 8.1-3 to establish the exposure accrued by the reactor vessel through the end of Cycle

11. The azimuthal locations presented in the tables represent 15° intervals of a quadrant, where the 75° azimuthal location is the maximum azimuthal flux position on the reactor vessel wall .

Exposure distri~utions through the vessel wall, can be developed using these surface exposures and radial distribution functions from Section 4. This exposure information, applicable through the end of Cycle 11, was derived from an extensive set of measurements and assures that embrittlement gradients can be established with a minimum uncertainty. Further, as the monitoring program continues and additional data become available, the overall plant specific data base for Palisades will expand resulting in reduced uncertainties and an improved accuracy in the assessment of vessel condition.

8-2

TABLE 8.1-1 e

SUMMARY

OF BEST ESTIMATE FAST NEUTRON (E > 1.0 MeV) EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF 1HE PALISADES REACTOR VESSEL THROUGH CYCLE 11 4>(E > 1.0 MeV) [n/cm2]

z (ft) ..Q:. it'._ 30° 45° 60° 7501 90°

+8.0 6.96e+17 9.70e+17 7.16e+17 4.54e+17 7.22e+17 9.75e+17 6.97e+l7

+7.5 9.8le+17 1.37e+18 1.0le+18 6.40e+17 l.02e+18 l.37e+18 9.82e+17

+7.0 1.53e+18 2.13e+18 1.57e+18 9.99e+17 l.59e+18 2.14e+18 l.53e+l8

+6.5 2.35e+18 3.27e+18 2.42e+18 l.53e+18 2.44e+l8 3.29e+18 2.35e+18

+6.0 3.29e+18 4.58e+18 3.38e+18 2.14e+I8 3.41e+18 ,4.60e+18 3.29e+18

+5.5 4.53e+18 6.32e+18 4.66e+18 2.96e+18 4.70e+l8 6.35e+18 4.54e+18

+5.0 5.57e+18 7.76e+18 5.73e+18 3.63e+18 5.78e+18 7.80e+18 5.58e+18

+4.5 6.78e+18 9.44e+18 6.97e+18 4.42e+18 7.03e+18 9.49e+18 6.79e+18

+4.0 7.40e+18 1.03e+19 7.61e+18 4.83e+18 7.67e+18 l.04e+19 7.41e+18

+3.5 8.30e+18 1.16e+19 8.53e+18 5.41e+18 8.61e+18 l.16e+19 8.31e+l8

+3.0 8.61e+18 1.20e+19 8.86e+18 5.62e+18 8.93e+18 l.21e+19 8.62e+1*8

+2.5 9.02e+18 1.26e+19 9.27e+18 5.88e+18 9.35e+18 l.26e+19 9.03e+l8

+2.0 8.98e+18 l.25e+19 9.23e+18 5.85e+18 9.31e+18 l.26e+19 8.99e+18

+1.5 9.12e+18 1.27e+19 9.38e+18 5.95e+18 9.46e+18 l.28e+19 9.13e+18

+1.0 9.06e+18 1.26e+19 9.32e+18 5.91e+18 9.40e+18 l.27e+19 9.07e+18

+0.5 9.26e+18 1.29e+19 9.53e+18 6.04e+18 9.61e+18 l.30e+19 9.28e+18  :

0.0 9.46e+18 1.32e+19 9.73e+18 6.17e+18 9.82e+18 l.32e+19 9.47e+18

-0.5 9.40e+18 1.31e+19 9.67e+18 6.13e+18 9.75e+18 l.32e+19 9.41e+18

-1.0 9.28e+18 1.29e+19 9.55e+18 6.06e+18 9.63e+18 . 1.30e+19 9.30e+18

-1.5 9.20e+18 1.28e+19 9.47e+18 6.00e+18 9.55e+18 l.29e+19 9.22e+18

-2.0 9.02e+18 1.26e+19 9.27e+18 5.88e+18 9.35e+18 l.26e+19 9.03e+18

-2.5 8.64e+18 1.20e+19 8.89e+18 5.64e+18 8.96e+18 l.21e+19 8.65e+18

-3.0 8.07e+18 1.12e+19 8.30e+18 *

• 5.26e+18 8.37e+18 l.13e+19 8.08e+18

-3.5 7.77e+18 l.08e+19 7.99e+18 '5.07e+18 8.06e+18 l.09e+19 7.78e+18

-4.0 6.74e+18 9.38e+18 6.93e+18 4.39e+18 6.99e+18 9.43e+18 6.75e+18

-4.5 5.87e+18 8.17e+18 6.04e+18 3.83e+18 6.09e+18 8.22e+18 5.88e+18

• Peak azimuthal fluence location.

8-3

TABLE 8.1-2

SUMMARY

OF BEST ESTIMATE FAST NEUTRON (E > 0.1 MeV) EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF TIIE PALISADES REACTOR VESSEL THROUGH CYCLE 11 4>(E > 0.1 MeV) [n/cm 2]

z (ft) JE.. ~ 30° 45° 60° 7501 90°

+8.0 l.35e+18 1.89e+18 1.40e+18 8.89e+17 l.41e+18 l.90e+18 l.35e+18

+7.5 1.90e+18 2.67e+18 1.98e+18 1.25e+18 l.99e+18 2.68e+18 l.91e+18

+7.0 2.97e+18 4.16e+18 3.08e+18 1.95e+18 3.lle+18 4.18e+18 2.98e+18

+6.5 4.57e+18 6.39e+18 4.74e+18 3.00e+18 4.77e+18 6.42e+18 4.57e+18

+6.0 6.38e+18 8.94e+18 6.62e+18 4.19e+18 6.66e+18 8.97e+18 6.38e+18

+5.5 8.81e+18 1.23e+19 9.13e+18 5.79e+18 9.20e+18 l.24e+19 8.8le+18

+5.0 l.08e+19 1.51e+19 1.12e+19 7.lle+18 1.13e+19 l.52e+19 l.08e+19

+4.5 l.32e+19 l.84e+19

• 1.37e+19 8.65e+18 l.37e+19 l.85e+19 l.32e+19

+4.0 l.44e+19 2.0le+19 1.49e+19 9.45e+18 l.50e+19 2.02e+19 1.44e+19

+3.5 l.61e+19 2.26e+19 1.67e+19 1.06e+19 l.68e+19 2.27e+19 l.61e+19

+3.0 1.67e+19 2.34e+19 1.73e+19 1.10e+19 l.75e+19 2.35e+19 l.67e+19

+2.5 1.75e+19 2.'45e+19 l.82e+19 1.15e+19 l.83e+19 2.46e+19 1.75e+19

+2.0 1.74e+19 2.44e+19 1.81e+19 1.15e+19 1.82e+19 2.45e+19 1.74e+19

+1.5 1.77e+19 2.48e+19 1.84e+19 1.16e+19 l.85e+19 2.49e+19 1.77e+19

+1.0 1.76e+19 2.46e+19 1.83e+19 1.16e+19

• l.84e+19 2.48e+19 1.76e+19

+0.5 1.80e+19 2.52e+19 1.87e+19 1.18e+19 l.88e+19 2.53e+19 1.80e+19 0.0 l.84e+19 2.57e+19 1.91e+19 1.21e+19 l.92e+19 2.59e+19 1.84e+19

-0.5 l.83e+19 2.56e+19 1.89e+19 1.20e+19 l.9le+19 2.57e+19 l.83e+19

-1.0 l.80e+19 2.53e+19 1.87e+19 1.19e+19 l.88e+19 2.54e.f.19 1.80e+19

-1.5 l.79e+19 2.50e+19 1.85e+19 1.18e+19 l.87e+19 2.5le+19 1.79e+19

-2.0 l.75e+19 2.45e+19 1.82e+19 1.15e+19 l.83e+19 2.46e+19 1.75e+19

-2.5 l.68e+19 2.35e+19 1.74e+19 1.10e+19 l.75e+19 2.36e+19 1.68e+19

-3.0 l.57e+19 2.20e+19 1.63e+19 l.03e+19 l.64e+19 2.20e+19 1.57e+19

-3.5 1.51e+19 2.lle+19 1.56e+19 9.92e+18 l.58e+19

• 2.12e+19 1.51e+19

-4.0 l.31e+19 l.83e+19 1.36e+19 8.60e+18 l.37e+19 l.84e+19 1.31e+19

-4.5 l.14e+19 1.60e+l9 1.18e+19 7.49e+l8 l.19e+19 1.60e+l9 1.14e+19 Peak azimuthal fluence location.

8-4

TABLE 8.1-3

SUMMARY

OF BEST ESTIMATE IRON ATOM DISPLACEMENT [dpa] EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF THE PALISADES REACTOR VESSEL THROUGH CYCLE 11 Displacements [dpa]

z (ft} JE_ .ll'.:... 30° 45° 60° 7501 90°

+8.0 l.05e-03 l.45e-03 l.08e-03 6.9le-04 l.09e-03 l.46e-03 l.05e-03

+7.5 I.48e-03 2.04e-03 l.52e-03 9.74e-04 l.53e-03 2.06e-03 l.48e-03

+7.0 2.32e-03 3.19e-03 2.37e-03 l.52e-03 2.39e-03 3.2le-03 2.32e-03

+6.5 3.56e-03 4.90e-03 3.64e-03 2.33e-03 3.67e-03 4.93e-03 3.56e-03

+6.0 4.97e-03 6.85e-03 5.08e-03 3.26e-03 5.13e-03 6.89e-03 4.97e-03

+5.5 6.86e-03 9.46e-03 7.02e-03 4.50e-03 7.08e-03 9.50e-03 6.87e-03

+5.0 8.42e-03 l.16e-02 8.62e-03 5.53e-03 8.69e-03 1. l 7e-02 8.43e-03

+4.5 l.03e-02 l.4le-02 l.05e-02 6.73e-03 l.06e-02 l.42e-02 l.03e-02

+4.0 l.12e-02 l.54e-02 l.14e-02 7.35e-03 l.15e-02 l.55e-02 l.12e-02

+3.5 l.26e-02 l.73e-02 l.28e-02 8.24e-03 l.30e-02 l.74e-02 l.26e-02

+3.0 l.30e-02 l.80e-02 l.33e-02 8.55e-03 l.34e-02 l.80e-02 l.30e-02

+2.5 l.36e-02 l.88e-02 l.40e-02 8.95e-03 l.4le-02 l.89e-02 l.37e-02

+2.0 l.36e-02 l.87e-02 l.39e-02 8.9le-03 l.40e-02 l.88e-02 l.36e-02

+1.5 l.38e-02 l.90e-02 l.4le-02 9.06e-03 l.42e-02 l.9le-02 l.38e-02

+I.O l.37e-02 l.89e-02 l.40e-02 9.00e-03 l.4le-02 l.90e-02 l.37e-02

+0.5 l.40e-02 l.93e-02 l.43e-02 9.20e-03 l.45e-02 l.94e-02 l.40e-02 -~~

0.0 l.43e-02 l.97e-02 l.46e-02 9.40e-03 l.48e-02 l.98e-02 l.43e-02

-0.5 l.42e-02 l.96e-02 l.46e-02 9.34e-03 l.47e-02 l.97e-02 l.42e-02

-1.0 l.40e-02 l.94e-02 I.44e-02 9.22e-03 l.45e-02 l.95e-02 l.4le-02

-1.5 l.39e-02 l.92e-02 l.42e-02 9.14e-03 l.44e-02 I.93e-02 l.39e-02

-2.0 l.36e-02 l.88e-02 l.40e-02 8.96e-03 l.4le-02 l.89e-02 l.37e-02

-2.5 l.3le-02 l.80e-02 l.34e-02 8.58e-03 l.35e-02 l.8le-02 l.3le-02

-3.0 l.22e-02 l.68e-02 l.25e-02 *8.02e-03 l.26e-02 l.69e-02 l.22e-02

-3.5 l.18e-02 l.62e-02 l.20e-02 7.72e-03 l.2le-02 l.63e-02 l.18e-02

-4.0 l.02e-02 l.40e-02 l.04e-02 . 6.69e-03 l.05e-02 l.4le-02 l.02e-02

-4.5 8.88e-03 l.22e-02 9.08e-03 5.83e-03 9.16e-03 l.23e-02 8.89e-03

• Peak azimuthal dpa location.

8-5

8.2 Uncertainties in Exposure Projections The overall uncertainty in the best estimate exposure projections within the pressure vessel wall stem primarily from two sources; a) the uncertainty in the bias factor (K) derived from the plant specific measurement data base and b) the analytical uncertainty associated with relating the results at the measurement locations to the desired results within the pressure vessel wall.

Uncertainty in the bias factor derives directly from the individual uncertainties in the measurement process, and in the least squares adjustment procedure, and in the location of the surveillance capsule and cavity dosimetry sensor sets. The analytical uncertainty in the

. relationship between the exposure of the pressure vessel and the exposure at the measurement locations are based on the vessel thickness tolerance and on the location of the dosimetry sensor

• sets for the cavity* data; and on downcomer water density variations, vessel inner radius tolerance, cladding thickness tolerance, and dosimetry sensor set locations for the surveillance capsule data.

The 1a uncertainties associated with the bias factors applicable to <l>(E > 1.0 MeV), <l>(E > 0.1 Me V), and dpa are as follows:

<l>(E > 1.0 MeV)

<l>(E > 0.1 MeV) dpa 8.1%

16.6%

12.2%

• The additional information pertinent to the required analytical uncertainty for vessel locations has been obtained from an analytical sensitivity study of the parameters important to the Palisades fluence evaluation. Based on these sensitivity studies the additional components to the overall uncertainty associated with the tolerances in dosimetry positioning, vessel thickness, vessel inner radius, cladding thickness and downcomer temperature are summarized in Table 8.2-1 for <l>(E

> 1.0* MeV), <l>(E > 0.1 MeV), and dpa, respectively. In addition to the uncertainties derived from the analytical sensitivity analyses, a 5% uncertainty component was included to account for small factors not explicitly addressed.

Combining the uncertainties from the bias factor determination and the sensitivity studies results in the following total uncertainties associated with the neutron exposure projections at the pressure vessel clad/base metal interface:

8-6

4>(E > 1.0 MeV) 14.5%

4>(E > 0.1 MeV) 20.6%

dpa 17.3%

These uncertainty values are well within the 20% 1 o uncertainty required by the PTS rule for 4>(E > 1.0 MeV) projections for the reactor pressure vessel.

Additionally, as part of the uncertainty examinations, gradient measurements for the 30°, 150°,

and 210° azimuths were compared to determine if any effect of core centering offset was observable. The average gradient measurements for the three angles agree within+/- 3%. Since the measurements, themselves, have a+/- 5% measurement uncertainty, the impact of core offset is not evident.

• 8-7

TABLE 8.2-1

SUMMARY

OF TOLERANCE UNCERTAINTY COMPONENT FOR

<l>(E > 1.0 MeV) PROJECTIONS AT THE VESSEL WALL Cavity Components Dosimeter Position 4.6%

Vessel Thickness 8.3%

Miscellaneous Factors 5.0%

Wall Capsule Component Dosimeter Position 3.8%

Vessel Inner Radius 2.6%

Cladding Thickness 1.0%

Downcomer Water Temperature 1.0%

Miscellaneous Factors 5.0%

Accelerated Capsule Component Dosimeter Position 3.5%

Vessel Inner Radius 2.6%

Cladding Thickness 1.0%

Downcomer Water Temperature 2.5%

Miscellaneous Factors 5.0%

8-8

SECTION 9 REFERENCES

1. Fero, A. H., "Reactor Cavity Neutron Measurement Program for Consumers Power Company Palisades Nuclear Plant", WCAP-11911, December 1988.

2. Fero, A. H., "Reactor Cavity Neutron Measurement Program for Consumers Power Company Palisades Nuclear Plant", WCAP-12847, January 1991.

3. Fero, A. H., "Reactor Cavity Neutron Measurement Program for Consumers Power Company Palisades Nuclear Plant (As-Installed Description for Cycle 10)", WCAP-13552, November 1992.

4. Perrin, J. S., et. al., "Palisades Nuclear Plant Reactor Pressure Vessel Surveillance Program: Capsule A-240", BCL-585-12, March 13, 1979.

• 5. Kunka, M. K., Cheney, C. A., "Analysis of Capsules T-330 and W-290 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program",

WCAP-10637, September 1984.

6. Lippincott, E. P., et~ al., "Palisades Nuclear Plant Reactor Vessel Neutron Fluence Measurement Program for Consumers Power Company, Results to End of Cycle 9",

WCAP-13534, Rev. 2, March 1993.

7. Peter, P.A., et. al., "Analysis of Capsule W-110 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program", WCAP-14014, May 1994.

8. RSIC Computer Code Co]J.ection CCC-543, "TORT-DORT Two- and Three-Dimensional Discrete Ordinates Transport, Version 2.8.14", January 1994.

9. RSIC Data Library Collection DLC-175, "BUGLE-93, Production and Testing of the VIT AMIN-B6 Fine Group and the BUGLE-93 Broad Group Neutron/Photon Cross-Section Libraries Derived from ENDF/B-VI Nuclear Data", April 1994.

9-1

10. Snuggerud, R. D., "Cycle Averaged Source and Temperature Information for Cycles 1 through 11 ", from EA-DOR 95-01 and EOC 11 Update, transmitted by letter and

• electronic file August 2, 1995.

11. ASTM Designation E706-87 (Reapproved 1994), "Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1995

12. 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, 1995.

13. ASTM Designation E261-90, "Standard Practice for Determining Neutron flux, Fluence, and Spectra by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA 1995.

14. ASTM Designation E262-86 (Reapproved 1991), "Standard Method for Measuring Thermal Neutron Reaction and Fluence Rates by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA 1995.

15. ASTM Designation E263-93, "Standard Test Method for Measuring Fast Neutron Reaction Rates by Radioactivation of Iron," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1995.

16. ASTM Designation E264-92, "Standard Test Method for Measuring Fast Neutron Reaction Rates by Radioactivation of Nickel," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1995.

17. ASTM Designation E481-86 (Reapproved 1991), "Standard Test Method for Measuring Neutron Fluence Rate by Radioactivation of Cobalt and Silver," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1995.

18. ASTM Designation E523-92, "Standard Test Method for Determining Fast Neutron Reaction Rate by Radioactivation of Copper," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1995.

9-2

19. ASTM Designation E704-90, "Standard Test Method for Measuring Reaction Rates by Radioactivation of Uraniwn-238," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1995.

20. ASTM Designation E705-90, "Standard Test Method for Measuring Reaction Rate by Radioactivation of Neptunium-237," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1995.

21. ASTM Designation El005-84 (Reapproved 1991), "Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1995.

22. Snuggerud, R. D., "Palisades Operating Cycle 11 Irradiation History", personal communication via fax, August 17, 1995.

23. Schmittroth, E. A., "FERRET Data Analysis Code", HEDL-TME-79-40, Hanford Engineering Development Laboratory, Richland, Washington, September 1979.

• 24. 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.

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

26. Maerker, R. E. as reported by Stallman, F. W., "Workshop on Adjustment Codes and Uncertainties - Proc. of the 4th ASTMJEURATOM Symposium on Reactor Dosimetry,"

NUREG/CP-0029, NRC, Washington, D.C., July 1982.

27. Andrachek, J. D., et. al.,* "Methodology Used to Develop Cold Overpressure and Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,"

WCAP-14040-NP-A, January 1996.

• 9-3

APPENDIX A SPECIFIC ACTIVITIES AND IRRADIATION HISTORY OF SENSORS FROM SURVEILLANCE CAPSULES A240, W290, W290-9 AND Wl 10 In this appendix, the irradiation history as extracted from NUREG-0020 and the measured specific activities of radiometric monitors irradiated in surveillance capsules A240, W290, W290-9, and WllO are provided.

The irradiation history of capsules withdrawn to date is as follows:

Cycle Startyp Shutdown Comment 1 12/31/71 12/20/75 2 05/08/76 01/06/78 Capsule A240 Withdrawn 3 04/20/78 09/08/79 4 05/24/80 08/29/81 5 12/25/81 08/12/83 Capsule W290 Withdrawn 6 07/29/84 11/30/85 7 03/02/86 08/08/88 8 11/01/88 09/15/90 9 03/10/91 02/06/92 Caps. W290-9 Installed/Withdrawn 10 04/18/92 06/05/93 Capsule WllO Withdrawn 11 11/06/93 05/22/95 Reference Core Power = 2530 MWt The monthly thermal generation applicable to the palisades reactor is provided in addition to the specific activities of the sensors on the following pages.

Since the in-vessel surveillance capsules were irradiated for multiple fuel cycles, the flux adjustment factors, C;, defined in section 3.0 were employed in the reaction rate calculations for the individual sensor sets.

The quantity C; is defined .as the calculated ratio of cf>(E > 1.0 MeV) during the irradiation period j to the time weighted average cf>(E > 1.0 MeV) over the entire irradiation period. The e values of C; used in the evaluation of the Palisades surveillance capsules were as follows:

A-1

Cvcle A 240 Flux Adjustment Factor Ci W290 w 290-9 w 110 e

1 0.918 1.004 2 0.819 0.895 3 1.100 1.202 4 1.126 1.231 5 1.093 1.194 6 1.216 7 1.159 8 0.963 9 1.00 0.630 10 0.485 e

e A-2

e TABLE A-1 IRRADIATION HISTORY OF PALISADES INTERNAL SURVEILLANCE CAPSULES Thermal Thermal Thermal Generation Generation Generation Month (MW-hr} Month (MW-hr) Month (MW-hr)

Dec-71 625 Apr-75 967872 Aug-78 1049064 Jan-72 155642 May'-75 1334640 Sep-78 556008 Feb-72 16679 Jun-75 873360 Oct-78 1172520 Mar-72 247284 Jul-75 1116216 Nov-78 1683216 Apr-72 519993 Aug-75 749376 Dec-78 849192 May-72 0 Sep-75 977856 Jan-79 1801656 Jun-72 684662 Oct-75 1135152 Feb-79 1604952 Jul-72 669575 Nov-75 1212960 Mar-79 1785288 Aug-72 792640 Dec-75 537672 Apr-79 1371072 Sep-72 490476 Jan-76 0 May-79 590664 Oct-72 731045 Feb-76 0 Jun-79 1297416 Nov-72 552165 Mar-76 0 Jul-79 1702776 Dec-72 1071439 Apr-76 0 Aug-79 1570656 Jan-73 667608 May-76 569280 Sep-79 322368 Feb-73 0 Jun-76 1520760 Oct-79 0 Mar-73 1059289 Jul-76 1052469 Nov-79 0 Apr-73 1549797 Aug-76 1260240 Dec-79 0 May-73 983014 Sep-76 1449288 Jan-80 0 Jun-73 1578251 Oct-76 1207248 Feb-80 0 Jul-73 1534211 Nov-76 1080384 Mar-80 0 Aug-73 476077 Dec-76 1531608 Apr-80 0 Sep-73 0 Jan-77 1426488 May-80 161088 Oct-73 0 Feb-77 1428888 Jun-80 1600296 Nov-73 0 Mar-77 1507152 Jul-80 1182912 Dec-73 0 Apr-77 1454856 Aug-80 1335552 Jan-74 0 May-77 1024776 Sep-80 1328640 Feb-74 0 Jun-77 1596000 Oct-80 1663008 Mar-74 0 Jul-77 1554528 Nov-80 0 Apr-74 0 Aug-77 1122840 Dec-80 920760 May-74 0 Sep-77 1431480 Jan-81 1777944 Jun-74 0 Oct-77 1630296 Feb-81 1684176 Jul-74 0 Nov-77 1457736 Mar-81 1867008 Aug-74 0 Dec-77 1703640 Apr-81 1750200 Sep-74 0 Jan-78 270336 May-81 1641384 Oct-74 387048 Feb-78 0 Jun-81 1531584 Nov-74 8400 Mar-78 0 Jul-81 604440 Dec-74 0 Apr-78 381600 Aug-81 845688 Jan-75 0 May-78 947376 Sep-81 0 Feb-75 0 Jun-78 1245312 Oct-81 0 Mar-75 0 Jul-78 1288344 Nov-81 O*

A-3

TABLE A-1 e

IRRADIATION HISTORY OF PALISADES INTERNAL SURVEILLANCE CAPSULES Thermal Thermal Thermal Generation Generation Generation Month (MW-hr) Month (MW-hr) Month (MW-hr)

Dec-81 1104 Apr-85 1622592 Aug-88 444768 Jan-82 947952 May-85 1841352 Sep-88 0 Feb-82 168384 Jun-85 1708032 Oct-88 0 Mar-82 682224 Jul-85 1823376 Nov-88 29640 Apr-82 0 Aug-85 640848 Dec-88 454344 May-82 362304 Sep-85 1372872 Jan-89 1657920 Jun-82 1614336 Oct-85 1557216 Feb-89 0 Jul-82 581544 Nov-85 1744224 Mar-89 1248144 Aug-82 0 Dec-85 0 Apr-89 1392120 Sep-82 1558992 Jan-86 0 May-89 1499736 Oct-82 1669680 Feb-86 0 Jun-89 1457664 Nov-82 1802688 Mar-86 331392 Jul-89 1510872 Dec-82 1841424 Apr-86 1321872 Aug-89 1341864 Jan-83 1742448 May-86 1107336 Sep-89 1453344 Feb-83 1675200 Jun-86 0 Oct-89 504 Mar-83 1862568 Jul-86 0 Nov-89 0 Apr-83 May-83 Jun-83 1713816 1688184 1761720 Aug-86 Sep-86 Oct-86 0

0 0

Dec-89 Jan-90 Feb-90 502200 1372848 1352352 e

Jul-83 1735776 Nov-86 0 Mar-90 1378920 Aug-83 543720 Dec-86 0 Apr-90 741096 Sep-83 0 Jan-87 0 May-90 536208 Oct-83 0 Feb-87 0 Jun-90 1047984 Nov-83 0 Mar-87 0 Jul-90 1501584 Dec-83 0 Apr-87 951309 Aug-90 1501896 Jan-84 0 May-87 1454016 Sep-90 704184 Feb-84 0 Jun-87 1387536 Oct-90 0 Mar-84 0 Jul-87 875304 Nov-90 0 Apr-84 0 Aug-87 1410336 Dec-90 0 May~84 0 Sep-87 1566648 Jan-91 0 Jun-84 0 Oct-87 14832 Feb-91 0 Jul-84 9816 Nov-87 968160 Mar-9.1 480456 Aug-84 166704 Dec-87 197136 Apr-91 1809167 Sep-84 222792 Jan-88 204576 May-91 1885464 Oct-84 0 Feb-88 1484904 Jun-91 1818648 Nov-84 485160 Mar-88 1878312 Jul-91 1143408 Dec-84 1838256 Apr-88 1522344 Aug-91 1837560 Jan-85 1802520 May-88 1731336 Sep-91 1818984 Feb-85 1562424 Jun-88 1818696 Oct-91 1882521 Mar-85 1843632 Jul-88 1794168 Nov-91 . 1712592 A-4

TABLE A-1 IRRADIATION HISTORY OF PALISADES INTERNAL SURVEILLANCE CAPSULES Thermal Generation Month (MW-hrl Dec-91 1513368 Jan-92 1867224 Feb-92 357888 Mar-92 0 Apr-92 620112 May-92 1878432 Jun-92 1819464 Jul-92 1392552 Aug-92 1459272 Sep-92 1260672 Oct-92 1779079 Nov-92 1326168 Dec-92 1880496 Jan-93 1879536 Feb-93 1698408 Mar-93 1880544 Apr-93 1688919 May-93 862632 Jun-93 237864 A-5

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I A-8

e RSAG rAL { { ')

Westinghouse Electric Corporation Advanced Programs - Analytical Laboratory .

REPORT Waltz Hill Site Request# 14629 TO: A.H.Fero (W)~nergy Center - East {4-17)

Radiation Engineering and Analysis Received: 4/1/92 Westinghouse Electric Corporation Reported: 4/15/92

---

~-------------------------------------------*

[RESULTS OF ANALYSIS]

Block# 1A7F Palisades In-vessel Dosimetry Originator Lab Dosimeter {@ 3/4/92)

ID Sample# Material Nuclide dps/mg

• 2 sigma Co-Al (Cd-V) 92-1148 Co-Al Co-60 2.48E+03 +/- 3.2E+Ol u (Cd-V) 92-1149 u Zr-95 l.36E+03 +/- l.SE+Ol u (Cd-V) 92-1149 u Ru-103 1.47E+03 +/- 1. 9E+Ol u {Cd-V) 92-1149 u Cs-137 4.16E+Ol +/- 4.6E+OO Ti (Bare) 92-1150 Ti Sc-46 6.78E+02 +/- l.OE+Ol Fe (Bare) 92-1151 Fe Mn-54 1. 24E+03 +/- 8.9E+OO Co-Al (Bare) 92-1152 Co-Al Co-60 I. 75E+04 +/- l.4E+Oi

• • u u

u Ni

{V)

(V)

(V)

{Cd-V) 92-1153 92-1153 92-1153 92-1154 u

u u

Ni Zr-95 Ru-103 Cs-137 Co-58 1.26E+03 +/- l. lE+Ol l.40E+03 +/- l.2E+Ol 4.48E+Ol +/- 3.2E+OO 2.84E+04 +/- l.4E+02 T

Cu (Cd-V) 92-1155 Cu Co-60 3.83E+Ol +/- 7.OE-01'.

Np (Cd-V) 92-1156 Np Zr-95 5.83[+03 +/- 8.8E+Ol Np (Cd-V) 92-1156 Np Ru-103 4.98E+03 +/* 7.2E+Ol Np (Cd-V) 92-1156 Np Cs-137 l.90E+Ol +/- 2.6E-Ol Remarks:

• Results are in units of dps/(mg of Dosimeter Material).

1) All dosimeter material was analyzed after removal from Cadmium and Vanadium encapsulation AL File: 14629

~proved:~~

References:

Lab.Book# 56 page 16.

Procedures: A-524.

Analyst: WTF,FRC,MRK.

A-9

RSAC PAL 775 Westinghouse Electric Co~poration e

. Advanced Programs - Analytical Laboratory REPORT Waltz Mill Site Request# 14629 TO: A.H.Fero (W)Energy Center - East (4-17)

Radiation Engineering and Analysis Received: 4/1/92 Westinghouse Electric Corporation Reported: 4/15/92

[RESULTS OF ANALYSIS]

Block# IA4F Palisades In-vessel Dosimetry Originator Lab Dosimeter {@ 3/4/92)

ID Sample# Material Nuclide dps/mg

• 2 sigma Co*A 1 {Cd-V) 92-1139 Co-Al Co-60 ----------

2.52E+03 +/- 3.2E+Ol u (Cd-V) 92-1140 u Zr-95 1.3SE+03 +!- 1. 5E+Ol u (Cd-V) 92-1140 u Ru-103 1.38E+03 +/- 1.2E+Ol u (Cd-V) 92-1140 u Cs-137 4.79E+Ol +/- 4.6E+OO Ti {Bare) 92-1141 Ti Sc-46 6.97E+02 +/- 5.lE+OO Fe (Bare) 92-1142 Fe Hn-54 1.28E+03 +/- 9. lE+OO Co-Al (Bare) 92-1143 Co-Al Co-60 1. 77E+04 +/- 2.4E+02

_ J (V) 92-1144 u Zr-95 l.25E+03 +/- l.3E+Ol u (V) 92-1144 u Ru-103 l.28E+03 +/- 1. lE+Ol u (V) 92-1144 u Cs-137 4.31E+Ol +/- 3.9E+OO Ni {Cd-V) 92-1145 Ni Co-58 2.74E+04 +/- 1.2E+02 Cu (Cd-V) 92-1146 Cu Co-60 3.88E+Ol +/- 5.5E-01 Np {Cd-V) 92-1147 Np Zr-95 S.80E+03 +/- 8.7E+Ol Np (Cd-V) 92-1147 Np Ru-103 S.27E+OO +/- 9.4E+Ol Np {Cd-V) 92-1147 Np Cs-137 l.59E+02 +/- 2.3E+Ol Remarks: *Results are in units of dps/(mg of Dosimeter Material).

1) All dosimeter material was analyzed after removal from Cadmium and Vanadium encapsulation AL File: 14629

References:

lab.Book# 56 page 16.

Procedures: A-524.

Analyst: WTF,FRC,HRK.

- A-10

e RSAC PAL 775 J

>J Westinghouse Electric Corporation Advanced Programs - Analytical Laboratory REPORT Waltz Hill Site Request# 14629 TO: A.H.Fero (W)Energy Center - East (4-17}

Radiation Engineering and Analysis Received: 4/1/92 Westinghouse Electric Corporation Reported: 4/15/92

[RESULTS OF ANALYSIS]

Block# lAlF Palisades In-vessel Dosimetry Originator Lab Dosimeter (@ 3/4/92}

ID Sample# Material Nuclide dps/mg

• 2 sigma Co-Al (Cd-V} 92-1130 Co-Al Co-60 2.48E+03 +/-

2.3E+Ol u (Cd-V} 92-1131 u Zr-95 1.20E+03 +/- 2.0E+Ol u (Cd-V} 92~1131 u Ru-103 l.31E+03 +/- l.3E+Ol u (Cd-V} 92-1131 u Cs-137 3.39E+Ol +/- 3.3E+OO Ti (Bare} 92-1132 Ti Sc-46 6.28E+02 +/- 6.7E+OO Fe (Bare} 92-1133 Fe Mn-54 1.12E+03 +/- 4.4E+OO

• Co-Al (Bare} 92-1134 u

u u

Ni (V}

(V}

(V}

(Cd-V}

92-1135 92-1135 92-1135 92-1136 Co-Al u

u u

N1 Co-60 Zr-95 Ru-103 Cs-137 Co-58 l.82E+04 +/-

l.19E+03 +/-

l.38E+03 +/-

4.34E+Ol +/-

2.71E+04 +/-

l.9E+02 l.7E+Ol 1.4E+Ol 4.2E+OO 1.2E+02 Cu (Cd-V) 92-1137 Cu Co-60 3.59E+Ol +/- 6.0E-01 Np (Cd-V) 92-1138 Np Zr-95 5.59E+03 +/- 7.4E+Ol Np (Cd-V) 92-1138 Np Ru-103 4.92£+03 +/- 4.SE+Ol Np (Cd-V) 92-1138 Np Cs-137 1.77£+02 +/- l.7E+Ol Remarks: *Results are 1n units of dps/(mg of Dosimeter Material).

1) All dosimeter material was analyzed after removal from Cadmium and Vanadium encapsulation Al File: 14629 Approved:~

References:

Lab.Book# 56 page 16.

Procedures: A-524.

Analyst: WTF,FRC,MRK *

• A-11

Westinghouse Electric Corporation Chemistry &Materials Technology - Analytical L:boratory REPORT Waltz Hill Site Request# 15214 Originator: Ed Terek (W)NTD Structural Reliability &Plant Life Optimization Received: 10/11/93 Westinghouse Electric Corporation Reported: 1/11/94

---

~-------------------------------------------------------------

[RESULTS OF ANALYSIS]

Dosimetry: Pallisades In-vessel Dosimetry (October 12, 1993)

Originator Lab Dosimeter ID Sample # Material Nuclide dps/mg +/- 2 sigma

---

~---~-----------

11414 TOP U-238(1 ring) 93-4192 U-238 Zr-95 7.09E+02 +/- 3.0E+OO Ru-103 3.26E+02 +I- 2.9E+OO Cs-137 l.23E+03 +/- 2.4E+OO Ti (2 rings) 93-4193 Ti Sc-46 1.98E+02 +/- 2.7E+OO Fe (4 rings) 93-4194 Fe Mn-54 l.34E+03 +/- 1.2E+Ol U-238 (Cd) (5 rings) 93-4195 U-238 Zr-95 6.9SE+Ol +/- 1.SE+OO Ru-103 3.46E+Ol +/- 1.3E+OO Cs-137 1.30E+02 +/- l.OE+OO Ni (Cd) (6 rings) 93-4196 Ni Co-58 7.0lE+03 +/- 6.6E+Ol Cu (Cd) (7 rings) 93-4197 Cu Co-60 2.22E+02 +I- 2.7E+OO Glass Sulphur 93-4198 s NA NA Fe (1 ring)t 93-4199 Fe Mn-54 1.31E+03 +/- 1.SE+Ol Fe (2 rings)t 93-4200 Fe Mn-54 l.31E+03 +/- l.SE+Ol Fe (3 rings)t 93-4201 Fe Mn-54 l.28E+03 +/- 1.SE+Ol Fe (4 rings)t 93-4202 Fe Mn-54 1.39E+03 +/- 1.6E+Ol Fe (5 rings)t 93-4203 Fe Mn-54 l.22E+03 +/- 1. 4E+Ol Remarks: *Results are in units of dps/(mg of Dosimeter Material).

Combustion Engineering Capsule Design .

_. *,~,-'l.*~~.- (~. :*

1

• t* denotes thin rings j Procedures:A-512,A-513,A-524 ,v~v'7/f..._v~'-

Ana lyst:WTF, FRC,MRK Approved: _ _ _ _1 : . - - - + - - - - - -

Page 1 A-12

Westinghouse Electric Corporation Chemistry &Materials Technology - Analytical Laboratory REPORT Waltz Hill Site Request# 15214 Originator: Ed Terek (W)NTD Structural Reliability & Plant Life Optimization Received: 10/11/93 Westinghouse Electric Corporation Reported: 1/11/94

[RESULTS OF ANALYSIS]

Dosimetry: Pallisades In-vessel Dosimetry "" - lt 0 (October 12, 1993}

Originator Lab Dosimeter ID Sample I Material Nuclide* dps/mg +/- 2 sigma 11441 MI i)

U-238(1 ring} 93-4204 U-238 Zr-95 6.69E+02 +/- 7.0E+OO Ru-103 3.06E+02 +/- 7.6E+OO*

Cs-137 l.30E+03 +/- S.9E+OO Ti (2 rings) 93-4205 Ti Sc-46 1.99£+02 +/- 2.SE+OO Fe (3 rings) 93-4206 Fe Mn-54 1.41£+03 +/- l.2E+Ol U-238 (Cd) (5 rings) 93-4208 U-238 Zr-95 7.49E+Ol +/- 5.6E-Ol Ru-103 3.49E+Ol +/- 5.6E-01 Cs-137 1.60£+02 +/- 4.6E-01 Ni (Cd) (6 rings) 93-4209 Ni Co-58 7.11E+03 +/- 6.7E+Ol Cu (Cd) (7 rings) 93-4210 Cu Co-60 2.58£+02 +/- 3.lE+OO Glass Sulphur 93-4208 s NA NA Fe (1 ring)t 93-4211 Fe Mn-54 l.42E+03 +/- l.SE+Ol Fe (2 rings)t 93-4212 Fe Mn-54 l.42E+03 +/- l.SE+Ol Fe (3 rings)t 93-4213 Fe Mn-54 l.40E+03 +/- l.SE+Ol .

Fe (4 rings)t 93-4214 Fe Mn-54 1.51E+03 +/- 1.6E+Ol Remarks: *Results are in units of dps/(mg of Dosimeter Material).

Combustion Engineering Capsule Design

• t* denotes thin rings Procedures:A-512,A-513,A-524 Analyst:WTF,FRC,MRK Approved:

Page 2 '

• A-13

. Westinghouse Electric Corporation Chemistry & Materials Technology - Analytical Laboratory REPORT Waltz Hill Site . Request# 15214 Originator: Ed Terek (W)NTD Structural Reliability & Plant Life Optimization Received: 10/11/93 Westinghouse Electric Corporation Reported: 1/11/94

[RESULTS OF ANALYSIS]

Dosimetry: Pallisades In-vessel Dosimetry (October 12, 1993)

Originator Lab Dosimeter ID Sample I Material Nuclide dps/mg +/- 2 sigma 11473 .iS ~7 U-238(1 ring) 93-4215 ..

U-238 '-'l Zr-95 7.53E+o2 +/- 1.2E+Ol Ru-103 3.55E+o2 +/- 1.lE+Ol Cs-137 1.41E+o3 +/- 8.2E+OO Ti (2 rings) 93-4216 Ti Sc-46 1.9BE+o2 +/- 2.7E+OO Fe (3 rings) 93-4217 Fe Mn-54 1.37E+o3 +/- 1.2E+Ol U-238 (Cd) (5 rings) 93-4218 U-238 :1, Zr-95 Ru-103 7.97E+ol 3.69E+ol

+/- 2.lE+OO l .8E+OO

+/-

Cs-137 l.65E+o2 +/- 1.4E+OO Ni (Cd) (6 rings) 93-4219 Ni Co-58 7.02E+o3 +/- 6.6E+Ol Cu (Cd) (7 rings) 93-4220 Cu Co-60 2.50E+o2 +/- 3.lE+OO Glass Sulphur 93-4221 s NA NA Fe (1 ring)t 93-4222 Fe Mn-54 l.42E+o3 +/- 1.5E+Ol Fe (2 rings)t 93-4223 Fe Mn-54 l.41E+o3 +/- 1.5E+Ol Fe (3 rings)t 93-4224 Fe Mn-54 l.38E+o3 +/- l.5E+Ol Fe (4 rings)t 93-4225 Fe Mn-54 l.53E+o3 +/- 1.6E+Ol Fe (5 rings)t 93-4226 Fe Mn-54 1.31E+o3 +/- l.5E+Ol Remarks:

• Results are in units of dps/(mg of Dosimeter Material).

Combustion Engineering Capsule Design .

• t* denotes thin rings

--\

Procedures:A-512,A-513,A-524 Analyst:WTF,FRC,MRK Approved:

• 1~

Page 3 A-14

APPENDIX B SPECIFIC ACTIVITIES AND IRRADIATION HISTORY OF REACTOR CAVITY SENSOR SETS - CYCLE 8 In this appendix, the irradiation history as extracted from NUREG-0020 and the measured specific activities of radiometric monitors irradiated in the reactor cavity during Cycle 8 are provided.

The irradiation history of Cycle 8 was as follows:

Cycle Startup Shutdown Comment 8 11/01/88 09/15/90 Reference Core Power = 2530 MWt The monthly thermal generation applicable to the Palisades reactor is provided in addition to

• the specific activities of the sensors on the following pages .

• .B-1

TABLE B-1 IRRADIATION HISTORY OF REACTOR CAVITY SENSOR SETS Cycle 8 Thermal Generation Date MW-hr Sep-88 0 Oct-88 0 Nov-88 29640 Dec-88 454344 Jan-89 1657920 Feb-89 0 Mar-89 1248144 Apr-89 1392120 May-89 1499736 Jun-89 Jul-89 Aug-89 Sep-89 1457664 1510872 1341864 1453344 Oct-89 504 Nov-89 0 Dec-89 502200 Jan-90 1372848 Feb-90 1352352 Mar-90 1378920 Apr-9b 741096 May-90 536208 Jun 1047984 Jul-90 1501584 -

Aug-90. 1501896 Sep-90 704184 B-2

TABLE B-2 e CONTENTS OF MULTIPLE FOIL SENSOR SETS CYCLE 8 IRRADIATION Radiometric Monitor Foil ID Capsule ID/ Bare or Cd 23su 23su NBS (nat) 231Np Position Shielded Fe Ni Cu Ti Nb Co ll!W PUD A-1 B A - - - - A A-2 Cd K A BA AK BA K AA BG A-3 Cd - - - - - - - - llNl 8 B-1 B B - - - - B B-2 Cd L B BB AL BB L AB BH B-3 Cd - - - - - - - - 12N2 9 C-1 B c - - - - c C-2 Cd M c BC AM BC M AC BI C-3 Cd - - - - - - - - 13N3 10 D-1 B D - - - - D D-2 Cd N D BD AN BD N AD BJ D-3 Cd - - - - - - - - 14N4 11 E-1 B E - - - - E E-2 Cd 0 E BE AO BE 0 AE AL E-3 Cd - - - - - - - - 15N5 12 F-1 B F - - - - F F-2 Cd p *F BF AP BF p AF AM F-3 Cd - - - - - - - - 16N6 13 G-1 B G - - - - G G-2 Cd R G BG AR BG R AG u G-3 Cd - - - - - - - - 17N7 14

• B-3

REOOR!'

Adv!lllced RSAC P. AL .6 7: 1 Westinghouse Electric Corporation Energy Systems - Analytical Laboratory Waltz Mill Site 7

Requesti 14175 TO: A.B.Fero (W)Energy Center - F.ast (4-17)

Received: 9/27/90 Reported: 12/21/90

[RESULTS OF ANALYSIS]

Palisades Cycle 8 Reactor Cavity Dosimetry Lab Dosimeter (@ 12/12/90)

Foil ID Samplet Material Nuclide dps/mg

• 2 sigma B

90-1857 Fe Mn-54 17.43 +/- ----------0.19 L 90-1858 Fe Mn-54 17.23 +/- 0.19 D 90-1870 Fe Mn-54 13.20 +/- 0.16 N 90-1871 Fe Mn-54 12.97 +/- 0.14 E 90-1883 Fe Mn-54 4.28 +/- 0.07 0 90-1884 Fe Mn-54 3.89 +/- 0.05 G 90-1896 Fe Mn-54 9.36 +/- 0.07 R 90-1897 Fe Mn-54 9.35 +/- 0.07 B 90-1859 Ni Co-58 201.10 +/- 1.43 D 90-1872 Ni Co-58 151.20 +/- 1.76 E 90-1885 Ni Co-58 49.72 +/- 0.55 G 90-1898 Ni Co-58 106.70 +/- 1.49 BB 90-1860 Cu Co-60 0.745 +/- 0.010 BD 90-1873 Cu Co-60 0.573 +/- 0.003 BE 90-1886 Cu Co-60 0.152 +/- 0.003 BG 90-1899 Cu Co-60 0.422 +/- 0.010 AL 90-1861 Ti Sc-46 4.57 +/- 0.07 AN 90-1874 Ti Sc-46 3.50 +/- 0.03 AO 90-1887 Ti Sc-46 I.OS +/- 0.01 AR 90-1900 Ti Sc-46 2.48 +/- 0.02 B 90;..1863 Al Co Co-60 270. 7 . +/- 4.1 L 90-1864 Al Co Co-60 179.7 +/- 3.2 D 90-1876 Al Co Co-60 253.4 +/- 4.2 N 90-1877 Al Co Co-60 174.7 +/- 3.4 E 90-1889 Al Co Co-60 151.S +/- 3.3 0 90-1890 Al Co Co-60 101.9 +/- 2.0 G 90-1902 Al Co Co-60 281.4 +I- 2.8 R 90-1903 Al Co Co-60 169.6 +/- 2.3 Remarks:

• Results are in units of dps/(mg of DosiJieter K:iterial).

AL File: 14175

References:

Lab.Bookt 35 page 298 - 300 Procedures: A-524.

Analyst: W1'F, CAB. Approved: ea.~""- lf).*"-1-'io 8-4

e RSAC Pt~L 671 5' Westinghouse Electric Corporation Advanced Energy System; - Analytical Laboratory REroRl' Waltz Mill Site Requestt 14175 TO: A.H. Fero (W)Energy Center - F.ast (4-17)

Beceived: 9i27/90 Reported: 12/20/90 (RESULTS OF ANALYSIS]

Palisades Cycle 8 Reactor Cavity Dosimetry Lab Dosimeter (@ 12/12/90)

Foil ID Samplet Material Nuclide dps/mg

• 2 sigma AB 90-1865 ---------

0(2)

Zr-95 ---------

41.60 +/- ----------

0.07 BH 90-1866 0(2) Zr-95 12.40 +/- 0.10 12 90-1867 0(1) Zr-95 12.17 +/- 0.05 N2 90-1868 0(2) Zr-95 40.21 +/- 0.07 AD 90-1878 0(2) Zr-95 36.71 +/- 0.19 BJ 90-1879 0(2) Zr-95 9.75 +/- 0.10 14 90-1880 0(1) Zr-95 9.75 +/- 0.05 N4 90-1881 0(2) Zr-95 36.62 +/- 0.21 AE 90-1891 0(2) Sample lost during inspection AL 90-1892 0(2) Zr-95 3.78 +/- 0.05 15 90-1893 O(l) Zr-95 3.22 +/- 0.04

• • NS AG 17 N7 0

90-1894 90-1904 90-1905 90-1906 90-1907 0(2) 0(2) 0(2) 0(1) 0(2)

Zr-95 Zr-95 Zr-95 Zr-95 Zr-95 19.02 +/-

34.40 +/-

7.48 +/-

6.19 +/-

35.87 +/-

0.10 0.17 0.06 0.05 0.18 AB 90-1865 0(2) Ru-103 17.49 +/- 0.04 BB 90-1866 0 (2) Ru-103 8.64 +/- 0.07 12 90-1867 0(1) Ru-103 8.88 +/- 0.04 N2 90-1868 0(2) Ru-103 16.92 +/- 0.04 AD 90-1878 0(2) Ru-103 14.80 +/- 0.11 BJ 90-1879 0(2) Ru-103 6.78 +/- 0.06 14 90-1880 0(1) Ru-103 6.95 +/- 0.03 N4 90-1881 0(2) Ru-103 14.82 +/- 0.13 AE 90-1891 0(2) Sample lost during inspection AL 90-1892 0(2) Ru-103 2.46 +/- 0.03 .

15 90-1893 0(1) Ru-103 2.29 +/- 0.02 NS 90-1894 0(2) Ru-103 6.97 +/- 0.06 AG 90-1904 0(2) Ru-103 13.03 +/~ 0.09 u 90-1905 0(2)

• Ru-103 4.86 +/- 0.04 17 90-1906 O(l) Ru-103 4.39 *+/- *0.03 N7 90-1907 0(2) Ru-103 13.45 +/- 0.13 Remarks:

• The U foils turned to a powder during irradiation. Oxide form unknown. .

U(l) For the RJD depleted foils (tl2,U4,US,U7), the foils were counted unopened, and the results calculated using the Net 0308 weight supplied by A.Fero.

O(2) For all other .0 foils, the foils were opened and the reoovered powder weighed.

The results were calculated using the recovered powder weight.

AL File: 14175.

References:

Lab.Book# 35 page 298 - 300 Procedures: A-524.

Analyst: Wl'F, CAB

• Approved: fl.ti.~ /~-~o-'lo

• 8-5

REroRl' Westinghouse Electric Corporation RSACPi4 L 67.1 Advanced Energy Systems - Analytical Laboratory Waltz Mill Site 7

Requestt 14175 v

TO: A.H. Fero (W) Energy Center - F.ast (4-17)

Beceived: 9/27/90 Beported: 12/20/90 Palisades Cycle 8 Reactor Cavity Dosimetry Lab D.osimeter (@ 12/12/90)

Foil ID Samplet Material Nuclide dps/mg

• 2 sigma AB 90-1865 0(2)

Cs-137 3.700 +/- ----------0.021 BH 90-1866 0(2) Cs-137 1.300 +/- 0.031 12 90-1867 0 (1) Cs-137 1.272 +/- 0.019 N2 90-1868 0(2) Cs-137 3.572 +/- 0.021 AD 90-1878 0(2) Cs-137 3.237 +/- 0.060 BJ 90-1879 0 (2.) Cs-137 0.975 +/- 0.027 14 90-1880 0 (1) Cs-137 0.981 +/- 0.017 N4 90-1881 0 (2) Cs-137 3.192 +/- 0.059 AE 90-1891 0(2) Sample lost during inspection AL 90-1892 , 0 (2) Cs-137 . 0.346 +/- 0.014 15 90-1893 0 (1) Cs-137 0.303 +/- 0.013 NS 90-1894 0(2) Cs-137 1.537 +/- 0.029 AG 90-1904 0(2) Cs-137 2.980 +/- 0.053 0 90-1905 0(2) Cs-137 0.764 +/- 0.018 17 90-1906 0(1) Cs-137 0.651 +/- 0.016 N7 90-1907 0(2) Cs-137 3.105 +/- 0.072 Renarks:

• The U foils turned to a powder during irradiation. Oxide form unknown.

U(l) For the PUD depleted foils (tl2,t14,t15,t17), *the foils were counted unopened, and the results calculated using the Net 0308 weight supplied by A.Fero.

U(2) For all other U foils, the foils were opened and the recovered powder weighed.

• The results were calculated using the recovered powder weight.

AL File: 14175.

References:

Lab.Book# 35 page 298 - 300 .

Procedures: A-524.

Analyst: WrF, G.B.

ga /t I! I J

• Approved:<..". *f~JMM.- l~-"3.o-'io B-6

e RSACPA L 671 10

(,..

Westinghouse Electric Corporation Advanced Energy Systems - Analytical Laboratory REroRl' Waltz Mill Site Request# 14175 TO: A.H. Fero (W)Energy Center - F.ast (4-17)

Received: 9/27/90 Reported: 12/20/90

[RESULTS OF ANALYSIS]

Palisades Cycle 8 Reactor Cavity Dosimetry Lab Dosimeter (@ 12/12/90)

Foil ID Sample~ Material Nuclide dps/mg

• 2 sigma 9

90-1869 Np-237 Zr-95 238.8 +/- ---------

---

1.4 11 90-1882 Np-237 Zr-95 175.0 +/- 1.5 12 90-1895 Np-237 Zr-95 60.3 +/- 1.1 14 90-1908 Np-237 Zr-95 129.6 +/- 1.3 9 90-1869 Np-237 Ru-103 144.2 +/- 0.9 11 90-1882 Np-237 Ru-103 106.4 +/- 1.4 12 90-1895 Np-237 Ru-103

• 39.5 +/- 0.8 14 90-1908 Np-237 Ru-103 79.8 +/- 1.2 e 11 12 9 90-1869 90-1882 90-1895 Np-237 Np-237 Np-237 Cs-137 Cs-137 Cs-137 24.01 17.97 5.94

+/-

+/-

+/-

0.47 0.77 0.40 14 90-1908 Np-237 Cs-137 12.82 +/- 0.61 Remarks: *Results are in units of dps/(trr:J of Dosimeter Material).

AL File: 14175.

References:

Lab.Book# 35 page 298 - 300 Procedures: A-524. A A I ff Analyst: ~m', CAB. Approved:~(}.~ Ja-~-'fo

• B-7

RSAC PAL 671 g7 (.

• Westinghouse Electric Corporation Advanced Energy Systems - Analytical Laboratory REroRI' Waltz Mill Site RequesU 14175 TO: E.P.Lippincott (W)Energy Center - East (4-17)

Received: 9/27/90 Reported: 4/3/91

[RFSJLTS OF ANALYSIS]

(10 'l>~G.)

Palisades Cycle 8 Reactor cavity Dosinetry Azi.l!uth: 280 deg.

Bead Oiclin Tag ID: 280 deg.

Feet [< dps/mg of chain @ 12/12/90 >

fran Lab Mn-54 Co-58 Co-60 Midplane Scmplet dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma

-+o.s 90-1915A l.14E+ol +/- 7. OE-01 l.92E+ol +/- l.OE+oO 1.0lE+o2 +/- 9.9E-Ol

-0.5 90-1915'8 l.16E+Ol +/- 5.9E-Ol l.99E+ol +/- l.3E+o0 l.OlE+o2 +/- 7.7E-Ol

-1.0

-l.5

-2.0

-2.5

-3.0

-3.5

-4.0

-4.5 90-1915C 90-1915D 90-lSlSE 90-1915F 90-lSlSG 90-1915B 90-1915I 90-1915.J l.16E+ol +/-

l.l3E+ol +/-

l.09E+ol +/-

9.19E+o0 +/-

8.84E+OO +/-

8.44E+o0 +/-

6.17E+o0 +/-

S.lOE+oO +/-

7. OE-01 6.0E-01 6.SE-01 5.SE-01 .

S.OE-01 6.SE-01 3.9E-01 3.SE-01 1.88E+ol +/- 9.lE-01 l.79E+ol +/- l.3E+o0 l.73E+ol +/- 8.6E-01 l.72E+ol +/- l.2E+o0 l.49E+ol +/- 7.0E-01 l.43E+ol +/- l.lE+oO l.l9E+ol +/- 7.0E-01 8.90E+o0 +/- 7.3E-Ol l.OlE+o2 +/-

9.90E+ol +/..;.

9.62E+ol +/-

9.34E+ol +/-

8.9SE+ol +/-

8.63E+ol +/-

8.07E+ol +/-

6.87E+ol +/-

9.lE-01

7. 7E-01 8.9E-Ol 7.4E-Ol 7.2E-Ol 7.lE-01 6.8E-01 4.lE-01

-s.o 90-1915K 3.43E+o0 +/- 3.4E-01 S.92E+o0 +/- S.2E-Ol S.86E+ol +/- S.BE-01

-5.5 90-1915L 2.27E+o0 +/- 1.9E-01 4.24E+o0 +/- 4.3E-01 S.39E+ol +/- 2.7E-Ol Remarks: *Results are in units of dps/(mg of .Dosimeter Material).

JlJ..

-*-~

File: 14175

References:

Lab.Eookli 49 paoes 3-20. /,/'. /

/Z //* e 8-8

• RSAC pr.

;:*. '- E71 Westinghouse Electric Corporation

.Advanced Energy Systems - Analytical Laboratory Waltz Mill Site Requestt 14175 TO: .E.P.Lippinc:ott (W)Energy Center - F.ast (4-17)

Received: 9/27/90 Reported: 4/3/91

[RFSJLTS OF AN.l\LYSIS)

(ao t>ec;.)

Palisades Cycle 8 Reactor cavity Dosimetry Azilmth: 290 deg.

Bead Chain Tag ID: 290 deg.

Feet *c< dps/mg of chain @ l2/l2/90 >]

fran Lab Mn-54 Co-58 Co-60 Midplane Sanplei dps/mg 2 sigma dps/nr;i . 2 sigma dps/nr;i 2 sigma

+o.5 90-1916A 8.36E+o0 +/- S.2E-Ol l.45E+ol +/- 7.4E-Ol 9.00E+ol +/- 7~3E-01

-o.s 90-1916B 8.52E+o0 +/- 3.lE-01 l.40E+Ol +/- 6.7E-Ol 9.06E+ol +/- 4.lE-01

-1.0 90-1916C 8.68E+o0 +/- S.SE-01 l.41E+ol +/- 6.6E-Ol 9.09E+ol +/- 7.2E-Ol

-1.s 90-19160 8.52E+o0 +/- 3.2E-Ol l.46E+Ol +/- 7.4E-01 B.96E+ol +/- 4.2E-Ol

• -2.0

-2.5

-3.0

-3.5

-4.0

-4.5

-s.o 90-1916E 90-l916F 90-1916G 90-1916H 90-19161 90-1916J 90-1916K 8.32E+o0 +/-

7.84E+o0 +/-

7.46E+o0 +/-

7.26E+o0 +/-

6.lSE+OO +/-

S.l7E+o0 +/-

3.45E+o0 +/-

S.OE-01 S.3E-Ol 3.0E-01 S.SE-01 4.9E-Ol 2.BE-01 2.2E-01 l.45E+ol +/- 7.3E-Ol l.39E+ol +/- l.lE+oO l.28E+ol +/- 4.SE-01 l.22E+ol +/- l.2E+o0 l.lSE+ol +/- l.lE+oO 8.91E+o0 +/- 4.0E-01 S.99E+o0 +/- 2.9E-01 8.90E+ol +/- 7.lE-01 B.57E+ol +/- 7.2E-Ol 8.49E+ol +/- 4.9E-01 8.16E+ol +/- 7.0E-Ol 7.72E+ol +/- 6.BE-01 7.23E+ol +/- 4.6E-Ol 5.79E+ol +/- 3.3E-01 '

Remarks: *Results are in units of dps/(mg of Dosimeter Material).

AL File: 14175

References:

Lab.Book# 49 pages 3-20.

Procedures: A-52,.

M.i-'°J \-~t : T,7I'F, TK.

B-9

. Westinghouse Electric CorporaRcS Advanced Energy Systems - Analytical Laboratory AC~ "_ L ( 71 Waltz Mill Site Request#. 14175 TO: E.P.Lippinoott (W)Energy Center - F.ast (4-17)

Received: 9/27/90 Reported: 4/3/91

[RmJLTS OF AIW.l'SISJ

( '/S t>eG.)

Palisades Cycle 8 Reactor Cavity Dosilietry Atim.Jth: 315 deg.

Bead Chain Tag ID: 315 deg.

Feet [< dps/mg of chain @ 12/12/90 >]

fran Lab Mn-54 Co-58 Co-60 Midplane Sanplei dps/mg 2 sigma dps/Jm3 2 sigma dps,'Jll3 2 sigma

+o.s *90-1918A 6.l9E+o0 +/- S.6£-01 l.06E+ol +/- l.lE+OO l.02E+02 +/- 7.SE-01

-o.s 9o-1918B 6.07E+o0 +/- S.4E-Ol 9.43E+o0 +/- l.lE+oO l.03E+o2 +/- 7.7E-Ol

-1.0 90-l.SlSC 6.* 07E+o0 +/- 6.lE-01 9.12E+o0 +/- l.2E+OO l.OlE+o2 +/- 8.7E-Ol

-1.5 90-19180 *s.42E+o0 +/- 4.SE-01 9.61E+o0 +/- l.lE+oO l.01E+o2 +/- 7.SE-01

-2.0 90-1918E S.71E+o0 +/- 3.2E-Ol 9.42E+OO +/- 4.7E-Ol 9.SlE+ol +/- S.3E-01

-2.5 90-1918F 8.21E+o0 +/- l.lE+oO 9.S4E+ol +/- 7.SE-01 S.03E+o0 +/- S.2E-Ol

-3.0 90-1918G 4.93E+o0 +/- 4.lE-01 8.23E+o0 +/- 7.9E-Ol 8.91E+ol +/- S.JE-01

-3.S 90-19168 4.40E+o0 +/- 2.9E-Ol 7.49E+o0 +/- 4.2E-Ol S.-66E+ol +/- 5.0E-01

,_ -4.0 90-1916! 3.84E+o0 +/- 3.JE-01 7.13E+oO +/- 7.JE-01 7.92E+ol +/- 4.4E-01

-4.5 90-1916J 3.53E+00 +/- 2.9£-01 5.94E+o0 +/- 3.6E-01 7.38E+ol +/- 4.6E-01

-5.0 90-l.Sl6K 2.SOE+oO +/- 2.6E-01 4.44E+o0 +/- 3.2E-Ol S.92E+ol +/- 4.lE-01

-5.5 90-1916L l.8SE+o0 +/- 1.9£--01 3.64E+OO +/- S.OE-01 S.02E+ol +/- 3.lE-01 Remarks: *Results are in units of dps/(mg of Dos.il!leter Material).

AL File: 14175

References:

Lab.Book# 49 pages 3-20.

Procedures: A-524.

Analyst: WI'F,TK.

B-10

• RSAC 671 f! v

. Westinghouse Electric Corporation Advanced Energy Systems - Analytical Laboratory waltz Mill Site Requestl 14175 TO: E.P.Lippincott (W)Energy Center - F.ast (4-17)

Received: 9/27/90 Reported: 3/13/91

[RESJLTS CE MW.YSIS)

(JS l>EG.)

Palisades Cycle 8 Reactor cavity Dosimetry Azi.mJth: 330 deg.

Bead Chain ~g m: 330 deg.

Feet [< dps/mg of chain @12/12/90 >]

fran Lab Mn-54 Co-58 Co-60 Midplane Smlplet dps/mg 2 sigma dps/mg 2 sigma . dps/mg 2 sigma

+o.5 90-1919A 8.69E+o0 +/- S.7E-Ol l.Sl>>t-01 +/- l.2E+OO l.14E+o2 +/- 8~2E-Ol o.o 9o-1919B 9.69E+OO +/- 6.6E-Ol l.44E+ol +/- l.OE+OO l.l4E+o2 +/- 8.3E-Ol

-0.5 90-1919C 9.19E+OO +/- S.SE-01 l.56E+ol +/- l.lE+OO l.14E+o2 +/- 8.2E-Ol

-1.0 90-19190 9.32E+OO +/- 6.lE-01 l.48E+Ol +/- l.OE+OO l.11E+02 +/- 8.l.E-01

-1.5 90-1919E 8.24E+o0 +/- S.3E-01 l.39E+Ol +/- l.lE+OO l.09E+02 +/- 8.0E-Ol

-2.0 90-1919F 7.SSE+OO +/- S.4E-ol l.43E+Ol +/- l.OE+OO l.OSE+02 +/- 7.9E-Ol

-2.5 90-1919G 7.34E+OO +/- 3.3E-01 l.30E+Ol +/- 6.SE-01 9.96E+Ol +/- 4.9E-Ol

-3.0 90-1919B 6.68E+OO +/- 3.SE-01 l.16E+Ol +/- 6.9E-Ol 9.38E+ol +/- S.SE-01

-3.5 90-19191 S.52E+OO +/- 3.3E-01 9.69E+OO +/- 7.0E-Ol 8.7SE+Ol +/- S.2E-Ol

-4.0 90-1919J 4.74E+o0 +/- 3.9E-Ol 8.09E+OO +/- S.9E-Ol 8.03E+Ol +/- S.OE-01

-4.S 90-1919K 3.37E+OO +/- 2. 7E-Ol . S.40E+OO +/- S.OE-Ol 6.45E+ol +/- 4.SE-01

-s.o 90-1919L 2.44E+o0 +/- 2.6E-Ol 4.30E+o0 +/- S.OE-01 S.52E+Ol +/- 4.l.E-01 '

-s.s 90-1919M l.46E+OO +/- 2.SE-ol 3.12E+OO +/- 4.2E-Ol S.OSE+Ol +/- 4.0E-Ol Remarks:

• Results are in units of 'dps/ (mg of Dosimeter Material).

AL File: 14175

• Apprcwed:~c ~

References:

Lab.Book# 49 pages 3-20.

Procedur~!i:_A-524

• Analyst: WIT, TR.

B-11

Westinghouse Electric Corporation Advanced Energy Systems - Analytical Laboratory Waltz Mill Site Requestl.14175 TO: E.P.Lippincott (W)Energy Center - East (4-17)

Received: 9/27/90 Reported: 4/2/91 (RE9JLTS CE ANALYSIS]

Palisades Cycle 8 Reactor cavity Dosimetry Azinuth: 30 deg.

Bead Chain Tag ID: (30).

Feet (< dps/mg of chain @ 12/12/90 >J fran Lab Mn-54 Co-58 Co-60 Midplane Sanplet dps/mg 2 sigma dps/nr; 2 sigma dps/mg 2 sigma

+a.o 90-1909A 4.32E-Ol +/- l.2E-Ol 9.65E-Ol +/- 3.0E-01 3.0SE+ol +/- 2.3E-Ol

+7.5 90-1909B S.75E-01 +/- l.lE-01 l.09E+OO +/- 2.4E-Ol 3.31E+ol +/- l.7E-Ol

+7.0 '9D-1909C 1.llE+oO +/- l.SE-01 l.94E+o0 +/- 3.SE-01 3.92E+Ol +/- 2.6E-01

+6.5 90-19090 l.38E+o0 +/- l.4E-Ol 2.40E+o0 +/- 2.9E-Ol 4.06E+Ol +/- l.BE-01

+6.0 90-1909E 2.33E+o0 +/- 2.0E-01 3. 92E+o0 +/- 3.9E-Ol 4.65E+Ol +/- 2.9E-01

+5.5 90-1909F 2.94E+o0 +/- l.6E-01 5.S6E+o0 +/- 3.SE-01 4.77E+ol +/- 2.0E-01

+5.o 90-1909G 3.SlE+oO +/- 2.2E-Cl 6.61E+OO +/- 4.SE-01 S.SOE+ol +/- 3.2E-Ol *

+4.5 90-1909H .4.72E+o0 +/- 2.3E-Ol 8.38E+o0 +/- 4.9E-Ol S.90E+Ol +/- 3.3E-Ol

+4.0 90-1909! 5.04E+OO +/- 2.0E-01 8.84E+o0 +/- 3.9E-Ol S.SSE+ol +/- 2.2E-Ol

+3.5 90-1909J S.98E+o0 +/- 2.9E-01 l.02E+ol +/- S.6E-Ol 6.53E+ol +/- 3.4E-Ol

+3.0 90-1909K 6.28E+o0 +/- 3.2E-Ol l.OSE+Ol +/- S.2E-Ol 6.SOE+ol +/- 3.SE-01

+2.5 90-19091. 6.64E+o0 +/- 3.5E-01 l.06E+ol +/- 8.0E-01 6.97E+Ol +/- 4.0E-01 '

+2.0 90-1909M 6.57E+o0 +/- 3.lE-01 l.04E+ol +/- S.6E-Ol 7.36E+ol +/- 3.7E-Ol

+l.5 90-l909N 6.46E+o0 +/- 3.lE-01 l.07E+ol +/- S.3E-Ol 7.SSE+ol +/- 3.7E-Ol

+l.O 90-19090 6.59E+o0 +/- 3.6E-Ol l.l2E+ol +/- 7.6E-Ol 7.6lE+Ol +/- 4.lE-01

+o.5 90-1909P 6.77E+o0 +/- 2.SE-01 l.09E+Ol +/- S.SE-01 7.93E+ol +/- 3.SE-01 o.o 90-19090 6.89E+o0 +/- 2.SE-01 l.12E+ol +/- 6.0E-01 8.02E+Ol +/- 3.BE-01

-0.5 90-l909R 6.84E+o0 +/- 3.7E-Ol l.14E+Ol +/- 9.0E-01 7.94E+ol +/- 4.SE-01

-l.O 90-l909S 6.90E+o0 +/- 2.7E-Ol l.14E+Ol +/- S.9E-Dl 8.lSE+Ol +/- 3..BE-Dl

-1.5 90-19091' 6.SOE+OO +/- 3.SE-01 l.24E+ol +/- 8.4E-Ol 7.93E+ol +/- 4.3E-01

-2.0 90-19090 6.93E+o0 +/- 2.SE-01 l.17E+ol +/- S.SE-01 7.98E+Ol +/- 3.BE-01

-2.5 90-l909V 6.64E+o0 +/- 2.BE-01 l.lSE+ol +/- 6.4E-Ol 7.83E+ol +/- 3.BE-01

-3.0 90-l90!:M 6.33E+o0 +/- 3. 7E-Ol l.lOE+ol +/- 7.SE-01 7.40E+Ol +/- 4.JE-01

-3.5 . *90-1909X 6.06E+00 +/- 2.7E-Ol 9.96E+o0 +/- S.3E-Ol 7.17E+ol +/- 3.6E-Ol

-4.0 90-l909Y 5.47E+00 +/- 2.6E-01 9.27E+o0 +/- 4.7E-Ol 6.70E+Ol +/- 3.5E-Ol

. -4.5 90-1909Z 4.38E+o0 +/- 3.2E-Ol 8.24E+o0 +/- 7.7E-Ol 5.98E+Ol +/- 3.7E-01 Remarks:

• Results are in units of dps/(mg of Dosilreter f.laterial).

AL File: 14175

References:

Lab.Book# 49 pages 3-20.

Procedures: A-524.

Analyst: Wl'F ,TK. Approved:~ ~

B-12

Rv(' r\....:

Westinghouse Electric Corporation AC!vanced Energy Systems - Analytical Laboratory Waltz Mill Site RequesU *14175 TO: E.P.Lippincott (W)Energy Center - East (4-17)

Received: 9/27/90 Reported: 4/2/91 (RESULTS OF AW\LYSIS]

(o 'PEG.)

Palisades Cycle 8 Reactor cavity Dosilretry Azi.nuth: 90 deg.

Bead Chain Tag m: (90).

Feet [< dps/ng of chain @ 12/12/90 >J fran Lab Hn-54 Ccr.58 Co-60 Midplane Sairplet dps/ng 2 sigma dps/ng 2 sigma dps/ng 2 sigma

+8.0 90-l910A 6.42E-Ol +/- 7.6E-02 l.20E+o0 +/- l.4E-Ol 4.55E+ol +/- l.9E-Ol

+7.5 90-l910B 8.34E-01 +/- 8.2E-02 l.6SE+o0 +/- l.BE-01 4.99E+ol +/- l*2E-Ol

+7.0 90-l910C l.39E+o0 +/- 2.lE-01 2.98E+o0 +/- 4.lE-01 S.93E+ol +/- 3~3E-Ol

+6.5 90-19100 2.lOE+oO +/- 2. 7E-01 3.83E+o0 +/- 5.4E-01 6.38E+ol +/- 3.9E-Ol

+6.0 90-1910E 3.12E+o0 +/- 2.4E-Ol S.18E+o0 +/- 4.3E-Ol 7.13E+Ol +/- 3.3E-Ol

+5.5 90-1910F 3.90E+o0 +/- 3.4E-Ol 6.85E+o0 +/- 6.8E-Ol 7.62E+ol +/- 4.2E-Ol

+5.o 90-19100 4.74E+00 +/- 2.4E-01 8.35E+o0 +/- 4.9E-01 8.65E+ol +/- 5.lE-01

+4.5 90-l910B 5.74E+o0 +/- 2.5E-01 9.69E+o0 +/- 5.3E-01 9.18E+ol +/- 3.BE-01

+4.0 90-19101 6.32E+o0 +/- 3. 7E-01 l.07E+ol +/- 7.7E-Ol 9.63E+ol +/- 4.7E-Ol

+3.5 90-l91QJ 6.97E+o0 +/- 4.9E-01 l.26E+ol +/- l.OE+oO l.OSE+o2 +/- 7.2E-Ol

+3.0 90-l910K 7.25E+o0 +/- 4. lE-01 l.30E+ol +/- 8.9E-Ol l.08E+o2 +/- 5.0E-01

+2.5 90-1910!. 7.52E+o0 +/- 5.2E-Ol l.31E+ol +/- l.lE+oO l.17E+o2 +/- 7.SE-01 *

+2.0 .90-191()1 7.6SE+o0 +/- S.6E-01 l.42E+ol +/- l.4E+o0 l.26E+o2 +/- 7.8E-Ol

+l.S 90-l910N 7.44E+o0 +/- 4.3E-01 l.26E+ol +/- 9.0E-01 l.18E+o2 +/- S.3E-01 ,.

+1.0 90-19100 7.lSE+oO +/- 5.SE-01 l.36E+ol +/- l.OE+oO l.30E+o2 +/- 7.9E-01 .~J.,

+o.5 90-l910P 7.SSE+oO +/- 4.SE-01 l.31E+ol +/- 9.3E-01 l.28E+o2 +/- 5~4E-01 "

o.o 90-191QJ 8.43E+o0 +/- 6.0E-01 l.25E+ol +/- l.lE+OO l.32E+o2 +/- 8.0E-01

-o.s 90-l910R 7.96E+o0 +/- S.SE-01 l.30E+ol +/- l.2E+o0 l.34E+o2 +/- 8.lE-01

-1.0 90-1910$ 7.24E+o0 +/- 4.6E-Ol l.27E+Ol +/- l.lE+oO l.32E+o2 +/- 8.lE-01

-1.5 90-1910T 7.33E+o0 +/- S.4E-01 l.lBE+ol +/- l.lE+oO l.31E+o2 +/- 8.0E-01

-2.0 90-19100 7.02E+o0 +/- S.4E-Ol l.20E+ol +/- l.2E+OO l.27E+o2 +/- 7.9E-Cl

-2.5 90-l9lOV 7.04E+o0 +/- S.3E-01 l.26E+ol +/- l.3E+o0 l.22E+o2 +/- 7.7E-Ol

-3.0 90-191~ 6.0lE+oO +/- 3.SE-01 l.13E+ol +/- 9.9E-01 l.13E+o2 +/- S.2E-Ol

-3.5 90-1910X 6.64E+o0 +/- 5.SE-01 l.07E+ol +/- 9.9E-01 l.09E+o2 +/- 7.2E-01

-4.0 90-1910Y 5.58E+o0 +/- 4.lE-01 8.90E+o0 +/- 8.4E-Ol 9.91E+Ol +/- 4.9E-Ol

-4.5 90-l910Z 4.57E+00 +/- 4.SE-01 7.3lE+oO +/- 9.SE-01 9.lSE+Ol +/- 6.7E-Ol Renarks:

• Results are in units of dps/(1b3 of Dosilreter Material).

AL File: 14175

References:

Lab.Book# 49 pages 3-20.

Procedures: A-524

• Analyst: mT,'ll\. Approved~~

8-13

(

/_ 71

'- j Westinghouse Electric Corporation Advanced Energy Systems - Analytical Laboratory Waltz Mill Site RequesU 14175 TO: E.P.Lippinex>tt (W)Energy Center - F.ast (4-17)

Received: 9/27/90 Reported: 3/13/91

[RESJL'l'S OF ANALYSIS]

(s~ })FG.)

Palisades Cycle 8 Reactor Cavity Dosimetry Azillllth: 150 deg.

Bead Chain Tag m: (RH:).

Feet [< dps/mg of chain @12/12/90 >:

fran Lab Mn-54 Ccr58 Co-60 Midplane Smlplet dps/mg . 2 sigma dps/mg 2 sigma dps/1113 2 sigma

+8.o 90-l9llA 6.13E-Ol +/~ l.3E-Ol 9.66E-Ol +/- 2.3E-Ol 3.28E+ol +/- 2.2E-Ol

/ +7.5 90-l9llB ..idQE=:Ol +/- 1. SE-01 l.7SE+<>O +/- 2.SE-01 3.78E+Ol +/- 2.4E-Ol

+7.0 90-l911C 1. 77E+OO '+/- 1. SE-01 3.22E+OO +/- 3.lE-01 4.SlE+ol +/- 2. 7E-Ol .

+6.5 90-l9llD -r.70E:+oo +/- 1.7£-ol 3.27E+OO +/- 2.SE-01 4.81E+Ol +/- 2.7E-Ol

+6.0 90-l911E 2.SOE+OO +/- l.6E-Ol 4.SlE+oO +/- 3.SE-01 5.24E+Ol +/- 2.8E-Ol

+S.5 90-l9llF 3.80E+OO +/- 3.0E-01 6.34E+OO +/- S.9E-01 S.S6E+ol +/- 3.8E-01

,_ +s.o 90-l9llG 4.S3E+o0 +/- 2.SE-01 7.93E+OO +/- 3.9E-Dl 6.llE+Ol +/- 3.0E-Dl

+4.5 90-l9llB 4.99E+OO +/- 4.4E-01 l.02E+ol +/- 7.SE-01 6.54E+Ol +/- S.OE-01

+4~0 90-l911I 5.76E+OO +/- 4.3E-01 l.OSE+Ol +/- 8.2E-Dl 6.93E+ol +/- S.lE-01

+3.5 90-l9llJ 6.89E+o0 +/- 4.SE-01 l.13E+Ol +/- 8.SE-01 7.38E+Ol +/- S.3E-01

+3.0 90-l911R 6.7SE+o0 +/- 4.SE-01 l.21E+ol +/- 8.3E-01 7.79E+Ol +/- S.SE-01 '

+2.5 90-l911L 7.02E+OO +/- 4.7E-01 l.19E+Ol +/- 7.4E-Ol 8.07E+ol +/- S.SE-01

+2.0 90-l9llM 7.32E+OO +/- 4.6E-01 l.29E+ol +/- 9.5E-01 8.37E+Ol +/- S.6E-Ol

+1.5 90-l911N 7.23E+o0 +/- 4.9E-Dl l.21E+ol +/- 8.3E-Ol 8.66E+ol +/- S.9E-Ol

+l.O 90-19110 7.61E+o0 +/- 5.4E-01 l.18E+ol +/- 8.5E-01 8.88E+Ol +/- S.8E-Ol

+o.s 90-l911P 7.3SE+o0 +/- S.OE-01 l.21E+Ol +/- 9.4E-Ol 9.06E+ol +/- S.8E-Ol o.o 90-19110 7.67E+o0 +/- 5.6E-01 l.32E+ol +/- l.lE+OO 9.16E+Ol +/- S.9E-Dl

-o.s 90-l9llR 7.26E+OO +/- 4.3E-01 l.27E+ol +/- l.OE+OO 9.l9E+Ol +/- S.9E-01

-1.0 90-19116 7.38E+OO +/- 5.0E-01 l.14E+ol +/- 9.lE-01 9.13E+Ol +/- S.9E-01

-1.s 90-l911T 7.20E+o0 +/- 4~9E-Ol l.lSE+ol +/- 8.6E-Ol 8.98E+ol +/- 5.SE-01

-2.0 90-19110 7.02E+OO +/- 4.7E-Ol l.22E+Ol +/- 8.9E-Ol 8.89E+ol +/- 5.8E-Ol

-2.5 90-1911V 6.94E+o0 +/- S.lE-01 l.lSE+Ol +/- 7.9E-Ol 8.59E+ol +/- 5.7E-Ol

-3.0 90-19llW 6.22E+OO +/- 3.7E-Ol l.19E+ol +/- 8.0E-01 8.43E+ol +/- 5.lE-01

-3.S 90-19llX 6.lSE+oO +/- S.OE-01 l.08E+ol +/- 8.lE-Dl* 7.89E+ol +/-.S.SE-01

-4.0 90-1911Y S.67E+o0 +/- 3.2E-Ol 9.58E+o0 +/- 5.SE-01 7.49E+ol +/- 4.8E-01

-4.5 90-l911Z 4.78E+00 +/- 4.0E-01 8.45E+o0 +/- 8.2E-Ol 6.80E+ol +/- S.lE-01 Renarks: *Results are in units of dps/(1113 of Dosimeter Material).

AL File: 14175

References:

Lab.Book~ 49 pages 3-20.

Procedures: A-524 *

~ "'"'['

Approved:~ ~

B-14

• Westinghluse Electric CorporatiR SA. CP{1 L 67 1 Advanced Energy Systems - Analytical Laboratory Waltz Mill Site Requestt 14175 TO: E.P.Lippinoott (W)Energy Center - East (4-17)

Received: 9/27/90 Reported: 5/13/91

[RFSlLTS CF ANl\LYSIS]

Palisades Cycle 8 Reactor Cavity Dosinetry Azinuth: 210 deg.

Bead Chain Tag ID: 210 deg.

Feet (< ~ of chain @12/12/90 >]

from Lab Mn-54 Co-58 c:crliC Midpl.ane Sanplei dps~ 2 sigma dps/mg 2 sigma dps/mg 2 sigma

+s.o 90-1912A S.16E-Ol +/- l.OE-01 9.26E-Ol +/- l.3E-01 4.20E+ol +/- 1.9£-01

+7.5 90-19128 7.64£-01 +/- l.OE-01 l.35E+o0 +/- l.SE-01 4.83E+ol +/- 2.lE-01

+7.0 90-1912C. l.04E+o0 +/- l.9E-Ol l.93E+o0 +/- 4.2E-Ol S.57E+ol +h.4.2E-Ol

+6.5 90-19120 l.47E+OO +/- 2.4E-Ol 3.00E+oo +/- 4.6E-Ol 6.22E+ol +/- 4.4£--01

+6.0 90-1912E 2.48E+o0 +/- 2.7E-Ol 4.68E+o0 +/- S.lE-01 6.70E+ol +/- 4.SE-01

+5.5 90-l912F 3.39E+o0 +/- 2.9E-ol S.92E+o0 +/- S.SE-01 7.2lE+ol +/- 4~3E-Ol.

+s.o 90-l912G 4.31E+o0 +/- 3.4£-ol 7.77E+o0 +/- 6.4E-Ol 7.82E+ol +/- 4.9E-01

+4.5 9D-1912B 5.22E+o0 +/- 3.7E-Ol 9.40E+o0 +/- 6.7E-Ol 8.56E+ol +/- S.lE-01

+4.0 90-l912I S.73E+o0 +/- 6.4E-Ol 9.7lE+oO +/- 7.SE-ol 9.06E+ol +/- S.6E-ol

+3.5 90-1912J 6.35E+o0 +/- 3.6E-Ol l.04E+ol +/- 6.SE-01 9.68E+ol +/- S.SE-ol

+3.0 90-l912X 6.SOE+oO +/- 4.9E-Ol l.llE+ol +/- 8.SE-ol 9.97E+ol +/- 6.lE-01

+2.5 90-1912L 7.28E+o0 +/- S.OE-01 l.07E+ol +/- 9.tt-ol l.04E+02 +/- 6.4E-Ol

+2.0 90-1912M 6.76E+o0 +/- 4.SE-01 l.l9E+ol +/- 9.lE-01 l.07E+o2 +/- 6.4£-ol *

+l.5 90-1912N 6.96E+o0 +/- S.2E-Ol l.17E+ol +/- 9.3E-ol l.l2E+o2 +/- 6.SE-01

+l.O . 90-19120 6.86E+o0 +/- S.OE-ol l.12E+Ol +/- 9.SE-01 l.14E+o2 +/- 6.7£-ol

+o.5 90-l912P 6.98E+o0 +/- S.SE-01 l.22E+ol +/- 9.tt-ol l.l6E+02 +/- 6.SE-01 o.o 90-19120 6.62E+OO +/- 6.4E-Ol l.17E+ol +/- l.4E+OO l.21E+o2 +/.;. l.lE+OO

-o.s 90-l912R 7.29E+OO +/- 7.2E-Ol l.l9E+ol +/- l.lE+oO l.20E+o2 +/- l.lE+oO

-1.0 90-l.S12S 6.89E+o0 +/- 6.9E-Ol l.18E+Ol +/- l.lE+oO l.20E+02 +/- l.lE+oO

-1.5 90-1912'1' *6.92E+o0 +/- 7.2E-Ol l.27E+ol +/- l.3E+o0 l.l8E+o2 +/- l.lE+oC

-2.0 90-19120 7.16E+oo +/- a. OE-ol l.l9E+ol +/- l.JE+oO l.l6E+o2 +/- l.lE+oO

-2.5 9D-l912V . 6.SlE+oO +/- 6.9E-Ol l.09E+ol +/- l.2E+o0 l.llE+o2 +/- l.OE+oO

-3.0 90-1912W 6.l9E+o0 +/- 4.lE-ol l.04E+Ol +/- 7.lE-ol l.OSE+o2 +/- S.7E-Ol

@ -3.5 90-l912X 6.26E+o0 +/- 4.4E-01 9.99E+o0 +/- 7.2E-Ol l.OOE+o2 +/- S.6E-Ol

-4.0 90-l9l2Y S.l4E+OO +/- 3.4E-Ol 9.45E+o0 +/- 7.0E-01 9.28E+ol +/- 5.4E-Ol

. -4.5 90-1Sl2Z 4.21E+o0 +/- 3.JE-01 8.06E+o0 +/- 6.lE-01 ~.42E+Ol +/- 5.lE-01 Rerarks: *Results are in WU.ts of dps/(mg of Dosimeter Material).

@ 5ar1i>le 190-1912-X: Corrected Co-60 value.

M. File: 14175

References:

Lab.Bookt 49 pages 3-20.

Procedures: A-524.

Analyst: WI'F, n\

• Approved: #L:JL .

• B-15

Westinghouse Electric Corporation .

.Advanced Energy Systems - Analytical Laboratory Waltz Mill Site Requesttl4175

'l'O: E.P.Lippincott (W) Energy Center - Fast (4-17)

Received: 9/27/90 Reported: 3/13/91

[RESJLTS CF NW.YSIS]

(1~ l>EG.)

Palisades Cycle 8 Reactor Cavity Dosimetry Azimlth: 260 deg.

Bead *Chain Tag m: (tmE)

Feet [< dps/mg of chain @12/12/90 >]

fran Lab Mn-54 Co-58 Co-60 Midplane Sanplet ~ 2 sic;ma dps/mg 2 sigma dps/mg 2 sic;ma

+e.o 9~1913A 6.58E-Ol +/- l.6E-ol l.36E+OO +/- 3.0E-ol 3.44E+Ol +/- 3.2£-ol

+7.5 9~1913B 8.56E-ol +/- l.OE-Ol l.SSE+-00 +/- 2.3E-Ol 3.94E+Ol +/- 2.lE-ol

+7.0 9~1913C l.33E+o0 +/- 2.lE-01 2.34E+OO +/- 4.3E-Ol 4.29E+ol +/- 3.6E-ol 9~1913D l.77E+OO +/- 2.0E-01 2.65E+OO +/- 3.9E-Ol 4.49E+Ol +/- 3.6E-Ol

-+(;.0 9~1913E 8.96£-Dl +/- 2.0E-Ol 2.25E+OO +/- 4.lE-ol 4.09E+Ol+/- 3.SE-01

+5.5 9~1913F 5.38E-Ol +/- l.SE-Dl l.35E+OO +/- 3.3E-Ol 3.68E+Ol +/- 3.3E-Ol

+5.o 5.42£-Dl +/- l.SE-01 9.06E-ol +/- 2.9E-Ol 3.lOE+Ol +/- 2.lE-01 e

9~191~

+4.5 9~19138 4.69E-Ol +/- l.4E-Ol S.9JE-Ol +/- 2.3E-Dl 2.77E+Ol +/- 2.0E-ol

+4.0 9~1913! . 2. 73E-Ol +/- 7.SE-D2 4.49E-Ol +/- 1.6£-Dl 2.63E+Ol +/- l.7E-Dl

+3.5 9~1913J 1.74£-Ql +/- 9.4E-02 5.0SE-Dl +/- 2.lE-ol 2.lSE+Ol +/- l.7E-Ql

_,"'~ +3.0

+2.5 9~1913!{

9~1913L l.68E-Dl +/- 7.8E-D2 1.93£-Ql +/- 8.7E-02 4.23E-Ol +/- l.SE-01 2.0SE-01 +/- l.7E-Ol 2.06E+ol +/- l.SE-01 l.69E+Ol +/- l.SE-01 '

~ +2.0 9~1913M l.27E-Ol +/- 8.9E-D2 2.SSE-Dl +/- l.SE-01 l.SlE+ol +/- l.SE-Dl

+1.5 9~1913N 1.32£-Dl +/- 6.0E-02 R> l.25E+Ol +/- l.3E-Ol

+l.O . 9~19130 R> R> l.06E+Ol +/- 1.2£-ol

+o.5 9~1913P R> R> s.m+oo +/- l.lE-01 o.o 9~19130 ID R> 8.03E+OO +/- l.JE-01

-0.5 9~1913R R> R> 7.07E+o0 +/- 9.9E-02

-1.0 9~1913S R> 8.13E-02 +/- S.2E-D2 7.00E+oO +/- 8.7E-D2

-1~5 9~1913T 5~9lE-02 +/- 3.6E-02 ID 6.3lE+oO +/- 8.4E-D2

-2.0 9~19130 R> R>

• 5.87E+OO +/- 8~1E-o2

-2.5 9~1913V R) R) S.42E+OO +/- 7.7E-02

-3.0 9~191~ R) R) 5.22E+OO +/- 7.6E-02

-3.5 9~1913X R) R) 4.83E+OO +/- 7.3E-02

-4.0 *9~1913Y R) R> 4.44E+o0 +/- 7.0E-02

-4.5 9~1913Z R) R) 4.28E+o0 +/- 6.9E-02 R>- Not Detected Remarks:

• Results are in units of dps/(mg of Dosineter Material).

AL File: 14175

References:

Lab.Book# 49 pages 3-20.

Procedures: A-524.

Analyst: WI'F, 'Ii{. Approved:~ c ~

B-16

RSACP,r~ l 671 Westinghouse Electric Corporation Advanced Energy Syst~ - Analytical Laboratory Waltz Mill Site Requesd 14175 TO: E.P.Lippinc:ott (W)Energy Center - F.ast (4-17)

Received: 9/27/90 Reported: 4/2/91

[RESJLTS OF ANALYSIS}

('iJ.o ~EG,) Palisades Cycle 8 Reactor cavity Dosimetiy Azimlth: 340 deg.

Bead Chain Tag m: (150).

Feet [< dps/mg of chain @ 12/12/90 >:

from Lab Mn-54 Co-58 Co-60 Midplane Sanplet dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma

+8.0 90-192~ 7.82E-Ol +/- l.SE-01 l.27E+OO +/- 3.0E-01 4.2SE+ol +/-.2.6E-Ol

+7.5 90-1920B 8.18E-01 +/- 2.3E-01 2.32E+o0 +/- S.2E-Ol 4.6SE+ol +/-*3.3E-01

+7.0 90-1920C *l.36E+OO +/~ l.6E-01 2.55E+o0 +/- 3.7E-Ol S.38E+ol +/-*2~9E-Ol

+6.5 90-19200 2.05E+o0 +/- l.6E-Ol 3.53E+o0 +/- 3.SE-01 S.62E+ol +/- 2.2E-Ol

+6.0 90-1920E 2.68E+OO +/- 2.0E-01 S.55E+o0 +/- 4.7E-Ol 6.53E+Ol +/- 3.2E-Dl .

+5.5 90-1920F 3.98E+o0 +/- 2.3E-01 7.26E+o0 +/- 4.7E--Ol 6.82E+ol +/- 3.3E-01

+5.0 90-19200 4.92E+o0 +/- 3.2E-Ol 8.42E+o0 +/- 7.6E-Ol 7.57E+ol +/- 4.2E-Ol

+4.5 90-19208 5.65E+o0 +/- 2.6E-01 l.02E+ol +/- S.SE--01 8.19E+ol +/- 3.6E-Ol

+4.0 90-19201 6.61E+o0 +/- 4.5E-01 l.07E+ol +/- 8.SE--01 8.78E+ol +/- 6.6E--Ol

+3.5 90-192Cll 7.20E+OO +/- 4.0E-01 l.lSE+ol +/- 7.4E--Ol 9.0lE+ol +/- 4.7E-Ol

+3.0 90-1920K 7.88E+o0 +/- 5.lE-01 l.24E+o1 +/- l.OE+oo 9.73E+Ol +/- 6.9E-Ol

+2.5 90-1920!. 8.0SE+oo +/- 4.7E-Ol l.32E+Ol +/- 8.6E-Ol 9.98E+ol +/- 4.9E--Ol

+2.0 90-1920M 8.21E+o0 +/- 4.9E-01 l.34E+ol +/- l.lE+oO l.07E+o2 +/- 7.2E-Dl

+l.5 . 90-1920N 8.lSE+oO +/- 5.2E-Ol l.34E+ol +/- l.lE+oO l.l1E+o2 +/- 7.4E-Ol

+l.O 90-19200 7.65E+o0 +/- 4.0E-01 l.35E+ol +/- 9.SE-01 l.12E+02 +/- S.2E-Dl

+o.5 90-1920P 7.89E+o0 +/- S.4E-Ol l.36E+ol +/- l.2E+OO l.18E+o2 +/- 7.6E-01 o.o 90-192D;l 8.29E+o0 +/- 4.6E-01 l.29E+ol +/- l.lE+oO l.20E+o2 +/- 7.6E-Ol

-0.5 90-1920R 8.00E+oO +/- 4.SE-01 l.39E+ol +/- 8.9E-Ol l.19E+o2 +/- S.3E-01

• -1.0 90-1920S 7.96E+OO +/- 5.4E-01 l.40E+ol +/- l.2E+o0 l.22E+02 +/- 7.7E-Ol

-1.5 90-19201' 7.96E+o0 +/- 4.lE-01 l.4SE+ol +/- 8.9E-01 l.18E+o2 +/- S.3E-Ol

-2.0 90-19200 8.22E+o0 +/- 6.0E-01 l.42E+ol +/- l.2E+OO l.19E+o2 +/- 7.7E-Ol

-2.5 90-1920\1 8.09E+o0 +/- 5.2E-Ol l.36E+ol +/- l.lE+oO l.16E+o2 +/- 7.SE-01

. -3.0 90-192tw 7.31E+o0 +/- 4.6E-Ol l.2SE+ol +/- 9.9E-Ol l.09E+o2 +/- S.3E-Ol

-3~5. 90-1920X 6.94E+o0 +/- 5.lE-01 l.09E+ol +/- 9.5E--01 .l.07E+o2 +/- 7.2E-Ol

-4.0 90-192oY 6.l4E+o0 +/- 2.9E-Ol l.OSE+Ol +/- 6.0E-01 9.83E+ol +/- 4.3E-Ol

-4~ 90-1920Z 2.02E+OO +/- l.5E-01 3.22E+o0 +/- 2.7E-Ol r 3.24E+Ol +/- 1. 7E-Ol Remarks:

• Results are in units of dps/(mg of Dosineter Material).

AL File: 14175

References:

Lab.Bookt 49 pages 3-20.

Procedures: A-524.

Analyst: Wl'F, TK

• 8-17

WESTINGHOUSE ELECTRIC CORPORA~~~.L NUCLEAR & ADVANCED TECHNOLOGY DIVISION AES ANALYTICAL LABORATORY WALTZ MILL SITE bf

• ARNOLD FERO AL REGUEST 14175

!! NATO, RADIATION ENGINEERING Receipt Date May 20, 1991 W284-4891 Report Date May 24, 1991

============================================mccccc============c

MATERIAL DESCRIPTION PALISADES CAVITY DOSIMETRY . CHAINS Al Serv. # Chain Identification ANALYSIS WEIGHT PERCENT ( x)

Azimuth ID Tag Fe Ni Co 91- 1909 s 300 30 64.4

---

B.74 0.126 1910 s 900 90 67.7 9.10 0.197 1911 1912 K

B

~'s"* none 210 66.7

-66.5 9.06 S.65 0.131 0.179 2100 1915 D 2800 280 70.6 B.Bl 0.176 1916 D 2900. 290 68.3 9.20 0.174 1918 c 3150 315 67.3 S.68 0.164 1919 M 3300 330 68.7 9.35 0.164 1920 z 3400 150 68.4 9.41 0.186 Average chain value 67.2 9.00 0.165 1 sigma  :;!:.1.7  :;!:.0.29 -!:,.0.022 NBS 1154 Certified 64.:S 12.26 0.101 Measured *61.6 11.77 0.099 NBS 160b Certified 65.1 10.25 0.12 Measured 66.1 10.25 0.1198 ICPS measurement at 2 sigma deviation  ::!:. 2.6  :.!:. 0.56  ::!:. 0.016 .

Method of Analysis Operator File Metals ICPS RMck AL14175 Approved by ~/~

Lawrence Kardos, Sr Scientist B-18

APPENDIX C SPECIFIC ACTIVITIES AND IRRADIATION IDSTORY OF REACTOR CAVITY SENSOR SETS - CYCLE 9 In this appendix, the irradiation history as extracted from NUREG-0020 and the measured specific activities of radiometric monitors irradiated in the reactor cavity during Cycle 9 are provided.

The irradiation history of Cycle 9 was as follows:

Cycle Startm> Shutdown Comment 9 03/10/91 02/06/92 Reference Core Power = 2530 MWt The monthly thermal generation applicable to the Palisades reactor is provided in addition to

• the specific activities of the sensors on the following pages .

• C-1

TABLE C-1 IRRADIATION HISTORY OF REACTOR CAVITY SENSOR SETS Cycle 9 Thermal Generation Date MW-hr Oct-90 0 Nov-90 0 Dec-90 0 Jan-91 0 Feb-91 0 Mar-91 480456 Apr-91 1809167 May-91 1885464 Jun-91 1818648 Jul-91 Aug-91 Sep-91 Oct-91 1143408 1837560 1818984 1882521 Nov-91 1712592 Dec-91 1513368 Jan-92 1867224 Feb-92 357888 C-2

TABLE C-2 CONTENTS OF MULTIPLE FOIL SENSOR SETS CYCLE 9 IRRADIATION Radiometric Monitor Foil ID Capsule ID/ Bare or Cd 23su 23su NBS Shielded Cnat) PUD 231Np Position Fe Ni Cu Ti Nb Co 1illall A-1 'B A - - - - A A-2 Cd K A BA AK BA K AA BG A-3 Cd - - - - - - - - llNl 8 C-1 B c - - - - c C-2 Cd M c BC AM BC M AC BI C-3 Cd - - - - - - - - 13N3 10 F-1 B F - - - - F F-2 Cd p F BF AP BF p AF AM F~3 Cd - - - - - - - - 16N6 13 J-1 B c - - - - 0 J-2 Cd M 0 AO BE AC AO c c J-3 Cd - - - - - - - - - 16 K-1 B D - - - - p K-2 Cd N p AP BF AD AP D D K-3 *. Cd - - - - - - - - - 17 L-1 ' B E - - - - R L-2 Cd 0 R AR BG AE AR E E L-3 Cd - - - - - - - - - 18 N-1 B G - - - - T.

N-2 Cd R T AT BI AG AT G u N-3 Cd - - - - - - - - - 20

• C-3

REPORT RSACPALtl' Westinghouse Electric Corporation Advanced Programs - Analytical Laboratory Waltz Mill Site

. // 1,N Requestl 14601

'ID: A.H.Fero (W)Energy Center - East (4-17)

~

Received: 3/3/92 Reported: 3/6/92

[RE&JL'IS <:R ANALYSIS]

Palisades Cycle 9 Reactor Cavity Dosilletry Lab Dosilleter (@ 3/4/92)

Foil m Sanplet Material Rlclide dps/1119 *' 2 sigira A 92-725 Co-Al Co-60 4.27Ei02 +/- 3.9E+o0 K 92-726 Co-Al Co-60 2.69Ei02 +/- 3.4E+o0 c 92-738 Co-Al* Co-60 2.lSEi-02 +/- 3.0E+oO M 92-739 Co-Al Co-60 l.49Ei02 +/- 2.SE+oO F 92-751 Co-Al Co-60 3.l4Ei02 +/- 3.7E+o0 p 92-752 Co-Al Co-60 2.16Ei02 +/- 3.lE+oO 0 92-764 Co-Al Co-60 l.77Ei02 +/- 2.7E+o0 AD 92-765 Co-Al Co-60 l.17Ei02 +/- l.6E+o0 p 92-m Co-Al Co-60 l.SSEi-02 +/- l.8E+o0 AP 92-n8 Co-Al Co-60 l.13Ei02 +/- l.6E+o0 R 92-790 Co-Al Co-60 9.SOEi-01 +/- l.4E+o0 AR 92-791 Co-Al Co-60 6.43Ei01 +/- l.2E+o0 T 92-803 Co-Al Co-60 l.76Ei02 +/- 2.0E+oO

,_ AT 92-804 Co-Al Co-60 l.llEi-02 +/- l.6E+o0

'IJ{.92-723 Ti Sc-46 6.73Ei00 +/- 7.2E-02 AH 92-736 Ti Sc-46 l.l4Ei00 +/- l.7E-02 AP 92-749 Ti Sc-46 2.07Ei00 +/- 2.6£-02 BE 92-762 Ti Sc-46 7.24Ei00 +/- 7.SE-02 BF 92-n5 Ti Sc-46 6.l2Ei00 +/- 7.0E-02 BG 92-788 Ti Sc-46 l.82Ei00 +/- 2.7E-02 BI 92-801 Ti Sc-46 4.l6Ei00 +;-* 4.lE-02 Remarks: *Results are in units of dps/(1119 of Dosinrter Material).

AL File: 14601

References:

Lab.Bookt 49 pages 300-301,IB 151 page 32.. )/J. . / / / / /

Procedures: A-524.

Analyst: W'IF,FRC,MRK. AWroved: f f/tiL'f(~

I C-4

e RSAC PAL t ' '

Westinghouse Electric Corporation Advanced Programs - Analytical Laborat-ory REPORT Waltz Mill Site Requestt 14601

'10: A.H.Fero (W) Energy Center - F.ast (4-17)

~

Received: 3/3/92 Reported: 3/6/92

[RESULTS OF ANALYSIS]

~alisades Cycle 9 Reactor Cavity Dosimetry *I,

. 1~

Ii Lab Dosimeter (@ 3/4/92) :ti Foil ID Sample I Material Nuclide dps/mg

• 2 sigma A

92-719 Fe Mn-54 l.91E+Ol +/- ----------

l.5E-Ol K 92-720 Fe Mn-54 l.84E+Ol +/- l.4E-Ol c 92-732 . Fe Mn-54 3.59E+OO +/- 8.0E-02 M 92-733 Fe Mn-54 3.22E+OO +/- 6.4£-02 F 92-745 Fe Mn-54 7.72E+OO +/- 7.0E-02 p 92-746 Fe Mn-54 7.51E+OO +/- 6.5E-02 c 92-758 Fe Mn-54 l.48E+Ol +/- 9.8£-02 M 92-759 Fe Mn-54 l.44E+Ol +/- l.3E-01 D 92-771 Fe Mn-54 l.22E+Ol +/- l.3E-Ol N 92-772 Fe Mn-54 l.21E+Ol +/- l.2E-Ol E 92-784 Fe Mn-54 3.95E+OO +/- 8.2£-02 0 92-785 Fe Mn-54 3.69E+OO +/- 7.lE-02 G 92-797 Fe Mn-54 8.58E+OO +/- l.2E-Ol R 92-798 Fe Mn-54 8.22E+OO +/- l.OE-01 A 92-721 Ni Co-58 3.16E+02 +/- 2.0E+OO c 92-734 Ni Co-58 6.02E+Ol +/- 8.7£-01 F 92-747 Ni Co-58 l.04E+02 +/- l.lE+OO 0 92-760 Ni Co-58 3.48E+02 +/- 2.lE+OO p 92-773 Ni Co-58 2.94E+02 +/- l.9E+00 R 92-786 Ni Co-58 9.52E+Ol +/- l.OE+OO T 92-799 Ni Co-58 2.01E+02 +/- l.6E+OO BA 92-722 Cu Co-60 9.62£-01 +/- 1.5£-02 BC 92-735 Cu Co-60 l.46E-Ol +/- 4.5E-03 BF 92-748 Cu Co-60 4.86£-01 +/- 7.6£-03 AO 92-761 Cu Co-60 4.65£-01 +/- 7.6£-03 AP 92-774 Cu Co-60 3.99£-01 +I-

+/-.

7.0E-03 AR 92-787 Cu Co-60 l.llE-01 2.9E-03 AT 92-800 Cu Co-60 2.76E-Ol +/- 6.0E-03 Remarks:

• Results are in units of dps/ (1113 of Dosimeter Material).

AL File: 14601

References:

Lab.Bookl 49 pages 300-301,IB 151 page 32 **

Procedures: A-524.

Analyst: W'IF, FR:, MRK. Approved:

C-5

e RSACPALtt:>

Westinghouse Electric Corporation Advanced Energy Systems - Analytical laboratory REPORT Waltz Mill Site Request# 14601 TO: A.H.Fero {W)Energy Center - East (4-17) ~

Radiation Engineering &Analysis Received: 3/3/92 Westinghouse Electric Corporation Reported: 3/9/92

[RESULTS OF ANALYSIS]

Palisades Cycle 9 Reactor Cavity Dosimetry Lab Dosimeter (~ 3/4/92)

.Foil ID Sample# Material Nuclide dps/mg

• 2 sigma 8

92-731 Np-237 Zr-95

---

+/- ----------

3.28E+02 2.6E+OO 10 92-744 Np-237 Zr-95 7.95E+Ol +/- 1. 6E+OO 13 92-757 Np-237 Zr-95 1. 24E+o2 +/- 1. 7E+OO 16 92-768 Np-237 Zr-95 3.65E+02 +/- 2.9E+OO 17 92-781 Np-237 Zr-95 2.82E+02 +/- 2.7E+OO 18 92-794 . Np-237 Zr-95 l. l 7E+o2 +/- 1.3E+OO 20 92-807 Np-237 Zr-95 2.13E+02 +/- 6.0E+OO 8 92-731 Np-237 Ru-103 3.08E+02 +/- 2.2E+OO 10 13 16 17 18 20 8

92-744 92-757 92-768 92-781 92-794 92-807 92-731 Np-237 Np-237 Np-237 Np-237 Np-237 Np-237 Np-237 Ru-103 Ru-103 Ru-103 Ru-103 Ru-103 Ru-103 Cs-137 8.40E+Ol l.16E+o2 3.42E+02 2.86E+02 l.13E+o2 2.22E+o2 2.81E+ol

+/- l. lE+OO

+/- l.5E+OO

+/- 2.4E+OO

+/- l .8E+OO

+/- l. lE+OO

+/- 4.2E+o0

+/- l.2E+OO 10 92-744 Np-237 Cs-137 7.03E+o0 +/- 6.lE-01 13 92-757 Np-237 Cs-137 l.57E+ol +/- 8.SE-01 16 92-768 Np-237 Cs-137 l.15E+ol +/- l.OE+oO 17 92-781 Np-237 Cs-137 l.09E+ol +/- 9.0E-01 18 92-794 Np-237 Cs-137 3.40E+o0 +/- 4.7E-01 20 92-807 Np-237 Cs-137 6.45E+o0 +/- l.5E+OO Remarks: **Results are in units of dps/(mg of Dosimeter Material).

Al File: 14601

References:

lab.Book# 49 pages 300-301,LB #51 page 32.. ',- / //

Procedures: A-524.

Analyst: WTF,FRC,MRK. Approved: , ( ~(_ ~ Ml*Ut4/

I C-6

• REPORT RSAC PAL It!)

Westinghouse Electric Corporation Advanced Energy Systems - Analytical Laboratory Waltz Mill Site Request# 14601

\ '

TO: A.H. Fero (W)Energy Center - East (4-17)

Radiation Engineering &Analysis Received: 3/3/92 Westinghouse Electric Corporation Reported: 3/16/92

[RESULTS OF ANALYSIS]

Palisades Cycle 9 Reactor Cavity Dosimetry Lab Dosimeter (@ 3/4/92)

Foi 1 ID Sample# Material Nuclide dps/mg

• 2 sigma AA 92-727 U (nat) Zr-95 6.22E+Ol +/-

1.5E-Ol BG 92-728 U (dep) Zr-95 2.0IE+Ol +/- 9.9E-02 11 92-729 U (dep) Zr-95 2.IOE+Ol +/- 1.9E-Ol NI 92-730 U (nat) Zr-95 6.18E+Ol +/- 2.SE-01 AC 92-740 U (nat) Zr-95 3.32E+Ol +/- 7.7E-02 BI 92-741 U (dep) Zr-95 5.66E+OO +/- 8.8E-02 13 92-742 U (dep) Zr-95 4.76E+OO +/- 9.lE-02 N3 92-743 U (nat) *Zr-95 3.ISE+Ol +/- 2.lE-01 AF 92-753 U (nat) Zr-95 3.67E+Ol +/- 9.2E-02 AM 92-754 U (dep) Zr-95 7.43E+OO +/- 6.8E*02 16 92-755 U (dep) Zr-95 6.63E+OO +/- 4.JE-02 N6 92-756 U (nat) Zr-95 3.48E+Ol +l- 8.8E*02 c 92-766 U (nat) Zr-95 6.66E+Ol +/- 1.SE-01 c 92-767 U (dep) Zr-95 2.32E+Ol +/- 1.3E*Ol D 92-779 U (nat) Zr-95 6. llE+Ol +/- 3.JE-01 D 92-780 U (dep) Zr-95 2.21E+Ol +/- 1.2E-Ol E 92-792 U (nat) Zr-95 3.SOE+Ol +/- 1.9E-Ol E 92-793 U (dep) Zr-95 7.33E+OO +/- 6.8E-02 G 92-805 U (nat) Zr-95 1.62E+02 +/- 9.0E-01 u 92-806 U (dep) Zr-95 l.42E+Ol +/- 9.7E-02 Remarks: *Results are in units of dps/(mg of Dosimeter Material).

AL File: 14601

References:

Lab.Book# 49 Procedures: A-524.

pages 261-265; 300-301' LB 151 pa71M}_ // / /

Analyst: WTF,FRC,MRK,TRK Approved: ~

West i n~~s~ Hec~rft ~oloZt ?on Advanced Energy Systems - Analytical Laboratory REPORT Waltz Hill Site Request# 14601 TO: A.H. Fero (W)Energy Center - East (4-17)

• Radiation Engineering &Analysis Received: 3/3/92 Westinghouse Electric Corporation Reported: 3/16/92

[RESULTS OF ANALYSIS]

Palisades Cycle 9 Reactor Cavity Dosimetry Lab Dosimeter (@ 3/4/92)

Foil ID Sample# Material Nuclide dps/mg

• 2 sigma AA 92-727 U (nat)

Ru-103 4.37E+Ol +/-

9.lE-02 BG 92-728 U (dep) Ru-103 2.08E+Ol +/- 9.9E-02 11 92-729 U (dep) Ru-103 2.20E+Ol +/- 1.6E-Ol NI 92-730 U (nat) Ru-103 4.26E+Ol +/- l.6E-01 AC 92-740 U (nat) Ru-103 1. 73E+Ol +/- 5.8E-02 BI 92-741 U (dep) Ru-103 5.12E+OO +/- 7.0E-02 13 92-742 U (dep) Ru-103 4.70E+OO +/- 7.8E-02 N3 92-743 U (nat) Ru-103 1.63E+Ol +/- 1.6E-01 AF 92-753 U (nat) Ru-103 2.12E+Ol +/- 7.0E-02 AM 92-754 U (dep) Ru-103 6.91E+OO +/- 4.7E-02 16 92-755 U (dep) Ru-103 7.19E+OO +/- 3.8E-02 N6 92-756 U (nat) Ru-103 1.96E+Ol +/- 6.7E-02 c 92-766 U (nat) Ru-103 4.70E+Ol +/- 9.5E-02 c 92-767 U (dep) Ru-103 2.31E+Ol +/- 1. lE-01 D 92-779 U (nat) Ru-103 4.17E+Ol +/- 2.lE-01 D 92-780 U (dep) Ru-103 1.94E+Ol +/- 8.0E-02 E 92-792 U (nat) Ru-103 1.96E+Ol +/- 1.4E-Ol E 92-793 . U (dep) Ru-103 6.87E+OO +/- 5.8E-02 G 92-805 U (nat) Ru-103 l.02E+02 +/- 5.7E-Ol u 92-806 U (dep) Ru-103 l.29E+Ol +/- 6.8E-02 Remarks: **Results are in units of dps/(mg of Dosimeter Material).

AL File: 14601

References:

Lab.Book# 49 pages 261-265; 300-301: LB 151 p~t,"' . .!' ~

Approved:_~

_______/(_llM~------

Procedures: A-524.

Analyst: WTF,FRC,MRK,TRK C-8

RSAC PAL//!>

Westiughouse Electric Corporation

\

Advanced Energy Systems - Analytical laboratory REPORT Waltz H111 Site Request# .14601 TO: A.H.Fero (W)Energy Center - East (4-17)

Radiation Engineering &Analysis Received: 3/3/92 Westinghouse Electric Corporation Reported: 3/16/92

[RESULTS OF ANALYSIS]

Palisades Cycle 9 Reactor Cavity Dosimetry Lab Dosimeter (@ 3/4/92)

Foil ID Sample# Material Nuclide dps/nig

• 2 sigma AA 92-727 U (nat)

Cs-137 5.42E+o0 +/- 5.3E-02 BG 92-728 U (dep) Cs-137 l.68E+OO +/- 5.0E-02 11 92-729 U (dep) Cs-137 l.84E+OO +/- 8.2E-02 Nl 92-730 U (nat) Cs-137 5.36E+OO +/- 8.8E-02 AC 92-740 U (nat) Cs-137 2.32E+o0 +/- 3.0E-02 BI 92-741 U (dep) Cs-137 4.17E-Ol +/- 3.0E-02 13 92-742 U (dep) Cs-137 3.46E-Ol +/- 4.2E-02 NJ 92-743 U (nat) Cs-137 2.19E+o0 +/- 8.SE-02 AF 92-753 U (nat) Cs-137 3.61E+OO +/- 3.9E-02 AM 92-754 U (dep) Cs-137 8.97E-Ol +/- 3.lE-02 16 92-755 U (dep) Cs-137 . 9.21£-01 +/- 2.0E-02 N6 92-756 U (nat) Cs-137 3.57E+o0 +/- 3.7E-02 c 92-766 U (nat) Cs-137 2.21E+o0 +/- 4.8E-02 c 92-767 U (dep) Cs-137 7.JSE-01 +/- 3.8E-02 D 92-779 U (nat) Cs-137 l.92E+o0 +/- 9.8£-02 D 92-780 U (dep) Cs-137 7.16£-01 +/- 3.8E-02 E 92-792 U (nat) Cs-137 9.48£-01 +/- 6.SE-02 E 92-793 U (dep) Cs-137 2.lOE-01 +/- 2.4E-02 G 92-805 U (nat) Cs-137 S.16E+OO +/-2.69£-01 u 92-806 U (dep) Cs-137 4.67£-01 +/-3.54£-02 Remarks:

• Results are in units of dps/(mg of Dosimeter Material).

Al File: 14601

References:

Lab.Book# 49 pages 261-265; 300-301: LB 151 p~ ,f' Procedures: A-524.

Analyst: WTF,FRC,MRK,TRK

• A~ro~d: *~

• C-9

REPORT REVISION

"" *1 I I,,,.._,,.

RequesU 14175 TO: A.H.Fero (W)Energy Center - East (4-17)

Radiation Engineering &Analysis Received: 9/27/90

. Westinghouse Electric Corporation Reported: 12/16/92

[RESULTS OF ANALYSIS]

Lab Dosimeter (f 12/12/90)

Foil ID Sample# Material Nuclide Bq/mg

• 2 s igiila BB 90-1862 Nb-93(Cd) Nb-93m 8.09E+ol +/- ----------

3.7E+o0 BD 90-1875 Nb-93(Cd) Nb-93m S.60E+ol +/- 3.lE+oO BE 90-1888 Nb-93(Cd) Nb-93m 1.67E+ol +/- l.7E+o0 BG 90-1901 Nb-93(Cd) Nb-93m 3.8SE+ol +/- 2.6E+o0 Remarks: *Results are in units of Bq/(mg of Dosimeter Material) *

.Nb-93m half-life: 5890 days AL File: Request# 14175

References:

Lab.Book# 51 page 61 Procedures: 01-Nb Analyst:.WTF,FRC,MRK C-10

r\uttlJ r HL. I I~

Westinghouse Electric Corporation Advanced Programs - Analytical Laboratory REPORT Waltz Mill Stte REVISION TO: A.H.Fero (W)Energy Center - East (4-17)

Radiation Engineering &Analysis Received: 3/4/92 Westinghouse Electric Corporation Reported: 12/17/92

[RESULTS OF ANALYSIS]

Lab Dosimeter (9 3/4/92)

Fof 1 ID Sample# Material Nuclide Bq/mg

• 2 sigma

------- 92-724 BA

---

Nb-93(Cd) ------- ----------

---

Nb-93m 1.28E+02 +/- ----------

4.6E+OO BC 92-737 Nb-93(Cd) Nb-93m 2.SOE+Ol +/- 2.lE+OO BF 92-750 Nb-93(Cd) Nb-93m 5.58E+Ol +/- 3.lE+oO AC 92-763 Nb-93(Cd) Nb-93m 4.37E+Ol +/- 2.8E+o0 AD 92-776 Nb-93(Cd) Nb-93m 4.SSE+Ol +/- 2.9E+OO AE 92-789 Nb-93(Cd) Nb-93m 1.59E+Ol +/- l.8E+o0 AG 92-802 Nb-93(Cd) Nb-93m 2.87E+Ol +/- 2.3E+o0 Remarks: *Results are in units of Bq/(mg of Dosimeter Material).

Nb-93m half-life: 5890 days AL File: Request# 14601

References:

Lab.Book# 51 page 61 Procedures: 01-Nb Analyst: WTF,FRC,MRK C-11

, RE~RT RSACPAL775 Westinghouse Advanced Energy Systems Analytical Laboratory - Waltz Mni Site RequesU 14601 Originator: A.H.Fero (W)Energy Center - East (4-17)

• Radiation Engineering &Analysis Received: 3/3/92 Westinghouse Electric Corporation Reported: 3/16/92

[RESULTS OF ANALYSIS]

PALISADES CYCLE 9 REACTOR CAVITY DOSIMETRY Bead Chain Tag ID: S-1 270 Degrees Feet [<-------------- dps/mg of chain @ 3/4/92 -------------->]

from Lab ------- Mn-54 ------- ------- Co-58 --~---- ------- Co-60 -------

Midplane Sample# dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma

+0.5 92-712A l.18E+ol +/- 7.lE-01 2.91E+Ol +/- 8.9E-Ol 1.54E+02 +/- 9.SE-01

-0.5 92-712C l.15E+Ol +/- 7.0E-01 2.84E+Ol +/- 8.6E-Ol l.54E+02 +/- 9.SE-01

-1.0 92-712D l.13E+Ol +/- 7.4E-Ol 2.76E+Ol +/- 8.6E-Ol 1.51E+02 +/- 9.7E-Ol

-1. 5 92-712E l.lOE+Ol +/- 7.6E-Ol 2.53E+Ol +/- 8.6E-01 1.47E+02 +/- 9.7E-Ol

-2.0 92-712F 9.39E+OO +/- 6.lE-01 2.50E+Ol +/* 8.4E-01 l.45E+02 +/- 9.5E-Ol

-2.5 92-712G 8.79E+OO +/- 7.lE-01 2.20E+Ol +/- 8.4E-Ol l.33E+02 +/* l.2E+OO

-3.0 92-712H 7.09E+OO +/- 6.lE-01 l.82E+Ol +/- 7.2E-01 l.35E+02 +/- 9.2E-Ol

-3.5 92-7121 6.70E+OO +/* 6.6E-Ol l.62E+Ol +/- 7.4E-01 l.26E+02 +/- 8.7E-Ol

-4.0 92-712J 6.42E+OO +/- 6.SE-01 l.48E+Ol +/- 7.2E-Ol l.15E+02 +/- 8.4E-Ol

-4.5 92-712K 3.93E+o0 +/- 4.7E-Ol

• l.02E+Ol +/- 5.6E-01 9.38E+Ol +/- 7.6E-Ol

-5.0 92-712l 2.76E+OO +/* 4.2E-Ol 7.33E+OO +/* 5.4E-01 8.33E+Ol +/- 7.2E-Ol Remarks: *Results are in units of dps /(mg of Dosimeter Material)

AL File: 14601

References:

lab Book# 49 pages 261-265; 300-301; Lab Book#jffe#51 Procedures: A-524.

p e 32. ;/µ~*

'1 Analyst: WTF, TRK, MRK, FRC Approved: f/

C-12

I_,

14601~

(\-.11'\U I ML' Westinghouse Advanced Energy Systems REPORT Analytical Laboratory - Waltz Hill Site Request*

Originator: Arnold Fero Radiation Engineering &Analysis Received: 3/3/92 Westinghouse Electric Corporation Reported: 3/9/92

[RESULTS OF ANALYSIS]

PALLISADES CYCLE 9 REACTOR CAVITY DOSIMETRY Bead Chain Tag ID: S-2 280 Degrees Feet [<----------,;, ___ dps/mg of chain @ 3/4/92 -------------->]

from Lab ------- Hn-54 ------- ------- Co-58 ------- ------- Co-60 -----~-

Hidplane Sample# dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma

+0.5 92-713A 9.33E+OO +/- 2.8E-Ol 3.25E+Ol +/- 3.7E-Ol 4.83E+Ol +/- 3.lE-01

-0.38 92-713X 9.75E+OO +/- 3.lE-01 3.26E+Ol +/- 4.lE-01 4.86E+Ol +/- 3.3E-Ol

-0.5 92-713C 9.43E+OO +/- 2.8E-01 3.20E+Ol +/- 3.6E-Ol 4.84E+Ol +/- 3.lE-01

-1.0 92-713D 9.41E+OO +/- 2.8E-Ol 3.llE+Ol +/- 3.6E-Ol 4.79E+Ol +/- 3.lE-01

-1.5 92-713E 8.97E+OO +/- 2.7E-Ol 3.03E+Ol +/- 3.5E-Ol 4.71E+Ol +/- 3.0E-01

-2.0 92-713F 8.29E+OO +/- 2.7E-Ol 2.87E+Ol +/- 3.4E-Ol 4.60E+Ol +/- 3.0E-01.

-2.5 92-713G 7.84E+OO +/- 2.6E-Ol 2.69E+Ol +/- 3.3E-Ol 4.49E+Ol +/- 2.9E-Ol

-3.0 92-713H 7.00E+OO +/- 2.4E-Ol 2.49E+Ol +/- 3.2E-Ol 4.32E+Ol +/- 2.9E-Ol

-3.5 92-7131 6.53E+OO +/- 2.5E-Ol 2.29E+Ol +/- 3.lE-01 4.lOE+Ol +/- 3.8E-Ol

-4.0 92-713J 5.38E+OO +/- 2.lE-01 l.90E+Ol +/- 2.9E-Ol 3.88E+Ol +/- 3.7E-Ol

-4.5 92-713K 3.93E+OO +/- l.SE-01 l.48E+Ol +/- 2.6E-Ol 3.37E+Ol +/- 2.5E-Ol

-5.0 92-713L 2.81E+OO +/- l.9E-Ol l.03E+Ol +/- 2.3E-Ql 2.84E+Ol +/- 2.3E-Ol

-5.5 92-713H 2.15E+OO +/- l.7E-Ol 7;5SE+OO +/- 2.lE-01 2.69E+Ol +/- 2.2E-Ol Remarks:

AL File: 14601

References:

Lab Book#49 pages 262-265; Lab Bookl51 page 32 Procedures: A-524.

Analyst: WTF, TRK, HRK, FRC

'f ii J t: ~ /"-

l "1,; zt°'Ut:.( _.

Approved:_~-~---....,......--..__ __

• C-13

, REPORT RSAC PAL 775 Westinghouse Advanced Energy Systems Analytical Laboratcry - Waltz Mill Site RequesU 14601 Originator: A.H.Fero (W)Energy Center - East (4-17)

• Radiation Engineering &Analysis Received: 3/3/92

-Westinghouse Electric Corporation Reported: 3/16/92

[RESULTS OF ANALYSIS]

PALISADES CYCLE 9 REACTOR CAVITY DOSIMETRY Bead Chain Tag ID: S-2 290 Degrees feet [<-------------- dps/mg of chain @3/4/92 -------------->]

from Lab ------- Hn-54 ------- . -------- Co-58 ------- ------- Co-60 -------

Midplane Sample# dps/mg 2 sigma dps/mg 2 sigma .dps/mg 2 sigma

+0.5 92-714A 7.93E+OO +/- 2.7E-Ol 2.67E+Ol +/- 3.4E-Ol ---------

4.38E+Ol +/- 2.9E-01

-0.5 92-714C 8.12E+OO +/- 3.3E-01 2.71E+Ol +/- 4.3E-Ol 4.43E+Ol +/- 3.6E-Ol

-1.0 92-714D 7.64E+OO +/- 3.3E-Ol 2.70E+Ol +/- 4.3E-Ol 4.44E+Ol +/- 3.7E-01

-1.5 92-714E 7.98E+OO +/- 3.4E-Ol 2.64E+Ol +/- 4.3E-Ol 4.37E+Ol +/- 3.6E-Ol

-2.0 92-714f 7.39E+OO +/- 3.0E-01 2.55E+Ol +/- 4.lE-01 4.31E+Ol +/- 3.6E-Ol

-2.5 92-7146 6.99E+OO +/- 3.2E-Ol 2.46E+Ol +/- 4.2E-Ol 4.lSE+Ol +/- 3.5E-01

-3.0 92-714H 6.42E+OO +/- 2.9E-Ol 2.32E+Ol +/- 4.0E-01 4.llE+Ol +/- 3.5E-Ol

-3.5 92-714I 6.0lE+OO +/- 2.4E-Ol 2.19E+Ol +/- 3.2E-Ol 3.94E+Ol +/- 2.SE-01 _

-4.0 92-714J 5.61E+OO +/- 2.4E-Ol l.93E+Ol +/- 3.0E-01 3.7SE+Ol +/- 2.7E-Ol

-4.5 92-714K 4.SSE+OO +/- 2.lE-01 l.65E+Ol +/- 2.SE-01 3.52E+Ol +/- 2.6E-Ol

-5.0 92-714L 3.20E+OO +/- l.SE-01 l.12E+Ol +/- 2.SE-01 2.85E+Ol +/- 2.3E-01 Remarks: *Results are in units of dps /(mg of Dosimeter Material)

AL FHe: 14601

References:

Lab Book# 49 pages 261-265; 300-301; Lab Book#Sl~~~

Procedures: A-524.

Analyst: WTF, TRK, MRK, FRC Approved:_~ __.....__ __._/!_-"'-lf,1~--~:;....;..;..~--

C-14

146~

Westinghouse Advanced Energy Systems REPC'RT Analytical Laboratory - Waltz Mill Site Requestf Criginator: A.H.Fero (W)Energy Center - East (4*17)

Radiation Engineering &Analysis Received: 3/3/92 Westinghouse Electric Corporation Reported: 3/16/92

[RESULTS OF ANALYSIS]

PALISADES CYCLE 9 REACTOR CAVITY DOSIMETRY Bead Chain Tag ID: S-1 300 Degrees Feet [<-------------- dps/mg of chain @3/4/92 -------------->]

from Lab ------- Mn-54 ------- ------- Co-58 ------- -~----- Co-60 -------

Midplane Sample# dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma

+0.5 92-715A S.38E+OO +/- 5.lE-01 l.33E+Ol +/- 6.5E-Ol ---------

l.18E+02 +/- 8.5E-Ol

-0.5 92-715C 4.68E+OO +/- 5.SE-01 7.21E+OO +/- 6.lE-01 l.07E+02 +/- 8.0E-01

-1.0 92-7150 3.73E+OO +/- S.SE-01 S.lOE+OO +/- 5.8E-Ol l.05E+02 +/- 7.9E-Ol

-1.5 ' 92-715E 3.49E+OO +/- 5.3E-Ol 3.lOE+OO +/- S.OE-01 l.03E+02 +/- 7.9E-Ol

-2.0 92-715F 2.52E+OO +/- 4.6E-Ol 2.SOE+OO +/- 5.0E-01 l.OOE+02 +/- 7.SE-01

-2.5 92-715G 2.46E+OO +/- 4.5E-Ol l.75E+OO +/- 3.SE-01 9.62E+Ol +/- 7.6E-Ol

-3.0 92-715H 2.13E+OO +/- 4.6E-Ol l.13E+OO +/- 3.0E-01 9.23E+Ol +/- 7.5E-Ol

-3.5 92-7151 2.lOE+OO +/- 4.7E-01 l.20E+OO +/- 4.lE-01 8.82E+Ol +/- 7.3E-Ol

-4.0 92-71SJ 2.13E+OO +/- 4.9E-Ol 7.81E-Ol +/- 3.0E-01 8.40E+Ol +/- 7.lE-01 .

-4.5 92-715K l.6SE+OO +/- 2.SE-01 8.00E-01 +/- 2.0E-01 7.84E+Ol +/- 3.9E-Ol

-5.0 92-715l l.32E+OO +/- 3.2E-Ol *6.05E-Ol +/- 2.9E-Ol 6.89E+Ol +/- 6.5E-01

-5.5 92-715H 8.33E-Ol +/- 3.3E-Ol 7.42E-Ol +/- 2.9E-Ol 5.97E+Ol +/- 6.0E-01

... Remarks: *Results are in units of dps /(mg of Dosimeter Material)

AL File: 14601

References:

Lab Book# 49 pages 261-265; 300-301; Lab BooktSI Procedures: A-524.

Analyst: WTF, TRK, MRK, FRC Approved:  ? pa~

C-15

RSACPAL775 Westinghouse Advanced Energy Systems REPORT AnalJtical Laboratory - Waltz Mill Site 'lequesU 14601 Originator: A.H.Fero (W)Energy Center - E,jst (4-17)

Radiation Engineering &Analysis Received: 3/3/92 Westinghouse Electric Corporation Reported: 3/16/92

[RESULTS OF ANALYSIS]

PALISADES CYCLE 9 REACTOR CAVITY DOSIMETRY Bead Chain Tag ID: S-2 315 Degrees Feet from Lab

[<-------------- dps/mg of chain @ 3/4/92 -------------->]

---

Mn-54 ------- ------- Co-58 ------- ------- Co-60 -------

Midplane Sample# dps/mg 2 s*igma dps/mg . 2 sigma dps/mg 2 sigma

+0.5 92-716A 5.52E+OO +/- 2.4E-Ol l.84E+Ol +/- 3.2E-01 4.85E+Ol +/- 3.0E-01

-0.5 92-716C 5.58E+OO +/- 2.5E-Ol l.83E+Ol +/- 3.2E-01 4.89E+Ol +/- 3.lE-01

-1.0 92-716D 5.46E+OO +/- 2.5E-Ol l.SOE+Ol +/- 3.lE-01 4.83E+Ol +/- 3.lE-01

-1.5 92-716E 4.82E+OO +/- 2.lE-01 l.71E+Ol +/- 3.lE-01 4.77E+Ol +/- 3.0E-01

-2.0 92-716F 4.68E+OO +/- 2.4E-Ol l.66E+Ol +/- 3.2E-Ol 4.65E+Ol +/- 3.0E-01

-2.5 92-7166 4.42E+OO +/- 2.3E-Ol l.53E+Ol +/- 3.0E-01 4.51E+Ol +/- 2.9E-Ol

-3.0 92-716H 4.37E+OO +/- 2.3E-Ol l.45E+Ol +/- 2.9E-Ol 4.33E+Ol +/- 2.9E-Ol

-3.5

-4.0

-4.5

-5.0

-5.S 92-7161 92-716J 92-716K 92-716L 92-716M 3.90E+OO 3.30E+OO 2.82E+OO 2.16E+OO l.66E+OO

+/-

+/-

+/-

+/-

+/-

2.4E-Ol 2.lE-01 2.0E-01 l.SE-01 l.SE-01 l.34E+Ol l.16E+Ol

. l.OlE+Ol 7.35E+OO 5.67E+OO

+/-

+/-

+/-

+/-

+/-

2.9E-01 2.7E-Ol 2.6E-Ol 2.2E-Ol 2.0E-01 4.12E+Ol 3.85E+Ol 3.54E+Ol 2.72E+Ol 2.38E+Ol

+/-

+/-

+/-

+/-

+/-

2.SE-01 2.7E-Ol 2.6E-Ol 2.3E-Ol 2.lE-01

• Remarks: *Results are in units dps /(mg of Dosimeter Material)

AL File: 14601

References:

Lab Bookl49 pages 261-272; 300-301; Lab BooklSI page~~ ~

Procedures: A-524.

Analyst: WTF, TRK, HRK, FRC Approved; ) ' /;IW.A C-16

RSAC PAL 775 ~,~

Westinghouse Advanced Energy Systems REPORT Analytical LaboratorJ - Waltz Mill Site Request# 14601 Originator: A.H.Fero (W)inergy Center - East (4-17)

Radiation Engineering &Analysis Received: 3/3/92 Westinghouse Electric.Corporation Reported: 3/16/92

[RESULTS OF ANALYSIS]

PALISADES CYCLE 9 REACTOR CAVITY DOSIMETRY Bead Chain Tag ID: S-2 330 degrees Feet [<-------------- dps/mg of chain @ 3/4/92 -------------->]

from Lab ------- Mn-54 ------- ------- Co-58 ------- ------- Co-60 -------

Midplane Sample# dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma

+0.5 92-717A 7.92E+OO +/- 5.4E-Ol 2.68E+Ol +/- 6.3E-Ol 5.43E+Ol +/- 5.6E-Ol 0.0 92-7178 8.07E+OO +/- 2.9E-Ol 2.68E+Ol +/- 3.7E-Ol 5.36E+Ol +/* 3.2E-01

-0.5 92-717C 7.76E+OO +/- 2.8E-Ol 2.64E+Ol +/- 3.6E-Ol 5.39E+Ol +/- 3.2E-01

-1.0 92-717D 7.93E+OO +/- 2.7E-Ol 2.62E+Ol +/- 3.6E-Ol 5.34E+Ol +/- 3.2E-01

-1.5 92-717E 7.69E+OO +/- 2.8E-Ol 2.61E+Ol +/- 3.7E-Ol S.26E+Ol +/- 3.2E-01

-2.0 92-717F 7.51E+OO +/- 2.8E-01 2.54E+Ol +/- 3.4E-Ol 5.13E+Ol +/- 3.lE.-01

-2.5 92- 717G 6.93[+00 +/- 2.6[-01 2.40E+Ol +/- 3.4E-Ol 4.97E+Ol +/- 3.lE-01

-3.0 . 92-717H 6.49E+OO +/- 2.5E-Ol 2.24E+Ol +/- 3.4E-Ol 4.75E+Ol +/- 3.0E-01

-3.5 92-717I 5.82E+OO +/- 2.4E-Ol 2.05E+Ol +/- 3.2E-Ol 4.49E+Ol +/- 2.9E-Ol

-4.0 92-717J 5.0lE+OO +/- 2.2E-Ol l.78E+Ol +/- 2.8E-Ol 4.18E+Ol +/- 2.6E-Ol

-4.5 92-717K 4.0lE+OO +/- 2.lE-01 l.47E+Ol +/- 2.8E-Ol 3.84E+Ol +/- 2.7E-Ol

-5.0 92-717L 2.68E+OO +/- I. 7E-01 9.75E+OO +/- 2.4E-01 2.96E+Ol +/- 2.4E-01

-5.5 92-717M 2.llE+OO +/- l.7E-Ol 7.llE+OO +/- 2.lE-01 2.64E+Ol +/- 2.2E-Ol Remarks:

• Results are in units dps /(mg of Dosimeter Material)

AL File: 14601

References:

Lab *Book#49 pages 261-272; 300-301; Lab Bookl~ate ~~ ~. ,!

Procedures: A-524.

Analyst: WTF, TRK, MRK, FRC Approved: __-++-~-------~-~~---

1

• C-17

, REPORT RSAt; t"AL t t :J Westinghouse Advanced Energy Systems .

Analytical Laboratory - Waltz Hill Site RequesU 14601 Originator: A.H.Fero (W)Energy Center - East (4-17)~

Radiation Engineering &Analysis Received: 3/3/92 Westinghouse Electric Corporation Reported: 3/17/92

[RESULTS OF.ANALYSIS]

PALISADES CYCLE 9 REACTOR CAVITY DOSIMETRY 3ead Chain Tag ID: S-2 30 Degrees Feet [<-------------- dps/mg of chain @ 3/4/92 -------------->]

from Lab ------- Hn-54 -----*-- ------- Co-58 ------- ------- Co-60 -------

~idplane Sample# dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma

+8.0 92-707A 4.54E-01 +/- 7.0E-02 1.72E+OO +/- 9.8E-02 1.96E+Ol +/- 1.2E-01

+7.5 92-7078 6.33E-01 +/- 9.6E-02 2.50E+OO +/- l.4E-01 2.24E+Ol +/- 1.6E-Ol

+7.0 92-707C 1.02E+OO +/- 1.lE-01 3.82E+OO +/- 1.5E-Ol 2.47E+Ol +/- 1.7E-Ol

+6.5 92-707D 1.56E+OO +/- 1.3E-01 5.63E+o0 +/- l.6E-Ol 2.68E+Ol +/- 1.BE-01

+6.0 92-707E 2.29E+OO +/- 1.9E-Ol 7.69E+OO +/- 2.3E-Ol 2.88E+Ol +/- 2.3E-Ol

+5.5 92-707F 2.98E+OO +/- 2.0E-01 l.09E+Ol +/- 2.5E-Ol 3.15E+Ol +/- 2.5E-Ol

+5.0 92-7076 3.84E+OO +/- 2.3E-01 1.31E+Ol +/- 2.8E-Ol 3.41E+Ol +/- 2.6E-Ol

+4.5 92-707H 4.38E+OO +/- 2.2E-Ol 1.56E+Ol +/- 2.8E-Ol 3.64E+Ol +/- 2.6E-Ol

+4.0 92-707I 5.09E+OO +/- 2.4E-Ol l.72E+Ol +/- 3.0E-01 3.84E+Ol +/- 2.7E-Ol

+3.5 92-707J 5.48E+OO +/- 2.4E-01 L88E+Ol +/- 3.0E-01 4.02E+Ol +/- 2.8E-Ol

+3.0 92-707K 5.77E+OO +/- 2.3E-Ol 1.93E+Ol +/- 3.lE-01 4.20E+Ol +/- 2.8E-Ol

+2.5 92-707L 6.00E+OO +/- 2.5E-Ol 2.03E+Ol +/- 3.3E-Ol 4.34E+Ol +/- 2.9E-Ol

+2.0 92-707H 5.89E+OO +/- 2.7E-Ol 2.08E+Ol +/- 3.4E-01 4.48E+Ol +/- 2.9E-Ol

+1.5 92-707N 6.lOE+OO +/- 2.7E-Ol 2.08E+ol +/- 3.4E-Ol 4.64E+Ol +/- 3.0E-01

+1.0 92-7070 6.14E+OO +/- 2.7E-Ol 2.llE+Ol +/- 3.4E-Ol 4.74E+Ol +/- 3.0E-01

+o.5 92-707P 6.28E+OO +/- 2.5E-Ol 2.13E+ol +/- 3.4E-Ol 4.88E+Ol +/- 3.0E-01 0.0 92-707Q 6.33E+OO +/- 2.7E-Ol 2.0SE+Ol +/- 3.4E-01 4.90E+Ol +/- 3.lE-01

-0.5 92-707R 6.37E+OO +/- 2.SE-01 2.12E+Ol +/- 3.5E-Ol 4.97E+Ol +/- 3.lE-01

-1.0 92-707S 6.09E+OO +/- 2.6E-Ol 2.14E+Ol +/- 3.4E-Ol 4.95E+Ol +/- 3.lE-01

-1.S 92-707T 6.30E+OO +/- 2.6E-Ol 2.14E+Ol +/- 3.5E-Ol 4.95E+Ol +/- 3.lE-01

-2.0 92-707U 6.16E+OO +/- 2.6E-01 2.14E+ol +/- 3.5E-01 4.91E+Ol +/- 3.lE-01

-2.S 92-707V 6.06E+OO +/- 2.7E-Ol 2.lOE+Ol +/- 3.5E-Ol 4.79E+Ol +/- 3.0E-01

-3.0 92-707W 6.00E+OO +/- 2.SE-01 l.98E+Ol +/- 3.4E-Ol 4.63E+Ol +/- 3.0E-01

-3.5 92-707X 5.21E+OO +/- 2.4E-Ol l.81E+Ol +/- 3.lE-01 4.40E+Ol +/- 2.9E-Ol

-4.0 92-707Y 4.62E+OO +/- 2.4E-Ol l.66E+Ol +/- 3.0E-01 4.13E+Ol +/- 2.SE-01

-4.5 92-707Z 4.0SE+OO +/- 2.3E-Ol l.40E+Ol +/- 3.0E-01 3.79E+Ol +/- 2.7E-Ol Remarks:

AL File: 14601

References:

Lab Bookl49 pages 259-271.

Procedures: A-524.

Analyst: WTF, TRK, HRK C-18

146~

RSACPALtt':J Westinghouse Advanced Energy Systems REPORT Analytical Laboratory - Waltz Mill Si~e Request*

Originator: A.H.Fero (W)Energy Cen~r (4-17)

Radiation Engineering &Analysis Received: 3/3/92 Westinghouse Electric Corporation Reported: 3/18/92

[RESULTS OF ANALYSIS]

PALISADES CYCLE 9 REACTOR CAVITY DOSIMETRY Bead Chain Tag IO: S-2 90 Degrees Feet [<-------------- dps/mg of chain @ 3/4/92 -------------->]

from Lab ------- Mn-54 ------- ------- Co-58 ------- ------- Co-60 -------

Midplane Sample# dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma 5.52E-Ol +/- 1.2E-Ol 2.17E+OO +/- 1.7E-Ol 2.14E+Ol +/- 2.0E-01

+8.0 92-708A

+7.5 92-7088 8.56E-Ol +/- 1.4E-Ol 3.56E+OO +/- I.SE-OJ 2.49E+Ol +/- 2.2E-Ol

+7.0 92-708C l.43E+o0 +/- l.6E-Ol 5.09E+OO +/- 2.lE-01 2.77E+Ol +/- 2.3E.-Ol

+6.5 92-7080 2.16E+OO +/- 1.9E-Ol 7.36E+OO +/- 2.2E-Ol 3.05E+Ol +/- 2.4E-Ol

+6.0 92-708E 3.UE+OO +/- 2.lE-01 I.04E+Ol +/- 2.5E-Ol 3.31E+Ol +/- 2.5E-Ol

+5.5 92-708F 4.06E+o0 +/- 2.2E-Ol J.38E+Ol +/- 2.7E-01 3.64E+Ol +/- 2.6E-Ol

+5.0 92-708G 4.93E+OO +/- 2.4E-Ol J.73E+Ol +/- 3.lE-01 3.93E+Ol +/- 2.7E-Ol

+4.5 92-708H 5.91E+o0 +/- 2.4E-Ol 2.05E+Ol +/- 3.3E-Ol 4.25E+Ol +/- 2.9E-Ol

+4.0 92-7081 6.33E+o0 +/- 2.6E-Ol 2.27E+Ol +/- 3.5E-Ol 4.55E+Ol +/- 3.0E-01

+3.5 92-708J 7.IIE+OO +/- 2.6E-Ol 2.45E+ol +/- 3.6E-Ol 4.SOE+Ol +/- 3.lE-01

+3.0 92-708K 7.43E+o0 +/- 2.9E-Ol 2.57E+ol +/- 3.7E-Ol 5.lOE+Ol +/- 3.lE-01

+2.5 92-708L 7.68E+o0 +/- 2.SE-01 2.59E+Ol +/- 3.7E-Ol *5.30E+Ol +/- 3.2E-Ol

+2.0 92-708M 7.70E+o0 +/- 2.SE-01 2.59E+ol +/- 3.SE-01 S.31E+Ol +/- 3.z'E-01

+1.5 92-708N 7.79E+o0 +/- 3.0E-01 2.65E+Ol +/- 3.9E-Ol S.76E+Ol +/- 3.3E-Ol

+1.0 92-7080 8.04E+OO +/- S.3E-Ol 2.69E+Ol +/- 6.9E-01 S.95E+Ol +/- 6.0E-01

+0.5 92-708P 7.96E+OO +/- S.3E-Ol 2.68E+Ol +/- 6.SE-01 6.04E+Ol +/- 6.0E-01 0 92-708Q 8.0SE+OO +/- S.SE-01 2.64E+Ol +/- 6.9E-Ol 6.13E+Ol +/- 6.0E-01

-0.S 92-708R 8.0lE+OO +/- S.3E-Ol 2.63E+Ol +/- 7.0E-01 6.13E+Ol +/- 6.2E-Ol

-1.0 92-7085 8.02E+OO +/- S.6E-Ol 2.61E+Ol +/- 6.9E-Ol 6.16E+Ol +/- 6.0E-01

-1.5 92-708T 7.13E+o0 +/- 4.SE-01 2.48E+Ol +/- 6.7E-Ol 6.0SE+Ol +/- 6.0E-01

-2.0 92-708U 6.96E+OO +/- S.OE-01 2.40E+Ol +/- 7.0E-01 S.9SE+Ol +/- 6.0E-01

-2.5 92-708V 6.40E+OO +/- 2.9E-Ol 2.30E+Ol +/- 3.7E-Ol S.69E+Ol +/- 3.3E-Ol

-3.0 92-708W 6.42E+OO +/- 3.SE-01 2.16E+Ol +/- 5.0E-01 5.46E+Ol +/- 4.SE-01

-3.5 92-708X 6.0lE+OO +/- 2.SE-01 2.02E+Ol +/- 3.SE-01 S.14E+Ol +/- 3;2E-Ol

-4.0 92-708Y S.33E+o0 +/- 2.SE-01 l.83E+Ol +/- 3.5E-Ol 4.84E+Ol +/- 3.lE-01

-4.5 92-708Z 4.57E+o0 +/- 2.5E-Ol l.56E+Ol +/- 3.lE-01 4.31E+Ol +/- 2.9E-Ol Remarks:

AL File: 14601 Approv~

References:

Lab Book#49 pages 259-271.

Procedures: A-524.

Analyst: WTF, TRK, MRK

• C-19

RSAC PAL 775 Westinghouse Advanced Energy Systems*

• REPORT Analytical L~boratory - Waltz Mill Site Request# 14601 Originator: A.H.Fero (W)Energy Center (E 4-17)

Radiation Engineering &Analysis Received: 3/3/92 Westinghouse Electric Corporation Reported: 3/20/92

[RESULTS OF ANALYSIS]

PALISADES CYCLE 9 REACTOR CAVITY DOSIMETRY Bead Chain Tag ID: S-2 150 Degrees Feet [<-------------- dps/mg of. chain @ 3/4/92 -------------->]

from Lab ------- Mn-54 ------- ------- Co-58 ------- ------- Co-60 -------

Midplane Sample# dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma

+8.0 92-709A 4.67E-Ol +/- l.3E-Ol l.75E+OO +/- l.6E-Ol 1.99E+Ol +/- ---------

l.9E-Ol

+7.5 92-7098 7.16E-Ol +/- l.3E-Ol 2.48E+OO +/- l.8E-Ol 2.32E+Ol +/- 2.IE-01

+7.0 92-709C l.16E+OO +/- l.6E-Ol 4.17E+OO +/- 2.2E-Ol 2.69E+Ol +/- 2.2E-Ol

+6.5 92-709D l.75E+OO +/- l.7E-Ol 5.98E+OO +/- 2.4E-Ol 2.95E+Ol +/- 2.4E-Ol

+6.0 92-709E 2.29E+OO +/- 2.lE-01 8.35E+OO +/- 2.SE-01 3.18E+Ol +/- 2.5E-Ol *

+5.5 92-709F 3.llE+OO +/- 2.2E-Ol l.lSE+Ol +/- 2.SE-01 3.45E+Ol +/- 2.6E-Ol

+5.0 92-709G 3.89E+OO +/- 2.lE-01 l.42E+Ol +/- 3.0E-01 3.70E+Ol +/- 2.7E-Ol

+4.5 92-709H 4.84E+OO +/- 2.5E-Ol l.69E+Ol +/- 3.lE-01 4.03E+Ol +/- 2.SE-01

+4.0 92-7091 5.21E+OO +/- 2.4E-Ol l.81E+Ol +/- 3.3E-Ol 4.30E+Ol +/- 2.9E~Ol

+3.5 92-709J 5.66E+OO +/- 2.3E-Ol 2.0SE+Ol +/- 3.SE-01 4.56E+Ol +/- 3.0E-01

+3.0 92-709K 5.95E+OO +/- 2.6E-Ol 2.19E+Ol +/- 3.6E-Ol 4.79E+Ol +/- 3.0E-01

+2.5 92-709L 6.50E+OO +/- 2.4E-Ol 2.27E+Ol +/- 3.6E-Ol 4.95E+Ol +/- 3.lE-01

+2.<i 92-709M 6.57E+OO +/- 2.6E-Ol 2.28E+Ol +/- 3.7E-Ol 5.14E+Ol +/~ 3.2E-Ol

+l.S 92-709N 6.94E+OO +/- 2.9E-Ol 2.30E+Ol +/- 3.SE-01 5.31E+Ol +/- 3.2E-Ol

+l.O 92-7090 6.48E+OO +/- 2.6E-Ol 2.30E+Ol +/- 3.SE-01 S.46E+Ol +/- 3.3E-Ol

+O.S 92-709P 6.66E+OO +/- 2.9E-Ol 2.36E+Ol +/- 3.9E-Ol S.54E+Ol +/- 3.3E-Ol 0.0 92-709Q 6.76E+OO +/- 2.8E-Ol 2.37E+Ol +/- 3.SE-01 S.63E+Ol +/- 3.3E-Ol

-0.5 92-709R 6.99E+OO +/- 2.8E-Ol 2.38E+Ol +/- 3.SE-01 S.68E+Ol +/- 3.3E-Ol

-1.0 92-709S 7.0lE+OO +/- 2.9E-Ol 2.34E+Ol +/- 3.7E-Ol S.60E+Ol +/- 3.3E-Ol

-1.5 92-709T 6.52E+OO +/- 2.7E-Ol 2.28E+Ol +/- 3.SE-01 S.SOE+Ol +/- 3.3E-Ol

-2.0 92-709U 6.63E+OO +/- 2.7E-01 2.25E+Ol +/- 3.SE-01 . S.SSE+Ol +/- 3.3E-Ol

-2.5 92-709V 6.31E+OO +/- 2.7E-Ol 2.22E+Ol +/- 3.7E-Ol S.36E+Ol +/- 3.2E-Ol

-3.0 92-709W 6.06E+OO +/- 2.8E-Ol 2.13E+Ol +/- 3.7E-Ol S.24E+Ol +/- 3.2E-Ol

-3.5 . 92-709X 5.83E+OO +/- 2.5E-Ol 2.03E+Ol +/- 3.SE-01 4.96E+Ol +/- 3.lE-01

-4.0 92-709Y 5.02E+OO +/- 2.4E-Ol l.82E+Ol +/- 3.4E-Ol 4.67E+Ol +/- 3.0E-01

-4.S 92-709Z 4.57E+OO +/- 2.5E-Ol l.53E+Ol +/- 3.2E-Ol 4.* 30E+Ol +/- 2.9E-Ol Remarks:

AL File: 14601

References:

Lab Book#49 pages 261-265; Lab BooklSJ page 32~

Procedures: A-524.

Analyst: WTF, TRK, MRK, FRC Approved: .

C-20

1460~

RSAC PAL 775 Westinghouse Advanced Energy Systems*

REPORT Analytical Laboratory - Waltz Hill Site Request#

O~iginator: *A.H.Fero (W)Energy Center (E 4-17)

Radiation Engineering &Analysis Received: 3/3/92 Westinghouse Electric Corporation Reported: 3/23/92

[RESULTS OF ANALYSIS]

PALISADES CYCLE 9 REACTOR CAVITY DOSIMETRY Bead Chain Tag ID: S-2 260 Degrees Feet from Lab

[<-------------- dps/mg of chain 9 3/4/92 -------------->]

---

Hn-54 ------- -----*-- Co-58 ------- ------- Co-60 -------

Midplane Sample# dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma

+8.0 92-711A S.OlE-01 +/- l.lE-01 l.6E-Ol ----------

l.89E+OO +/- --------- ---------

2.16E+Ol +/- 2.0E-01

+7.5 92-7118 7.43E-Ol +/- l.4E-Ol 2.89E+OO +/- l.9E-Ol 2.42E+Ol +/- 2.lE-01

+7.0 92-711C l.22E+OO +/- l.6E-Ol 4.40E+OO +/- 2.lE-01 2.68E+Ol +/- 2.2E-Ol.

+6.S 92-711D l.92E+OO +/- l.9E-Ol 6.63E+OO +/- 2.4E-Ol 2.95E+Ol +/- 2.4E-Ol

+6.0 92-711E 2.65E+OO +/- 2.2E-Ol 9.44E+OO +/- 2.7E-Ol 3.19E+Ol +/- 2.5E-Ol

+5.5 92-711F 3.70E+OO +/- 2.3E-Ol l.26E+Ol +/- 2.9E-Ol 3.44E+Ol +/- 2.6E-Ol

+5.0 92-7116 4.62E+OO +/- 2.4E-Ol l.66E+Ol +/- 3.lE-01 3.68E+Ol +/- 2.7E-Ol

+4.5 92-711H S.47E+OO +/- 2.3E-Ol l.94E+Ol +/- 3.2E-Ol 3.93E+Ol +/- 2.7E-Ol

+4.0 92-7111 6.20E+OO +/- 2.6E-Ol 2.17E+Ol +/- 3.4E-Ol 4.12E+Ol +/- 2.8E-Ol

+3.S 92*711J 6.73E+OO +/- 2.5E-Ol 2.37E+Ol +/- 3.5E-01 4.31E+Ol +/- 2.9E-Ol

.*3.0 92-711K 7.34E+OO +/- 2.7E-01 2.49E+Ol +/- 3.7E-Ol 4.51E+Ol +/- 2.9E-Ol

+2.5 92-711L 7 .6.2E+OO +/- 2.6E-Ol 2.56E+Ol +/- 3.8E-Ol 4.66E+Ol +/- 3.0E-01

+2.0 92-711M 7.46E+OO +/- 2.7E-Ol 2.58E+Ol +/- 3.7E-Ol 4.82E+Ol +/- 3.lE-01

+1.5 92-711N 7.53E+OO +/- 2.9E-Ol 2.62E+Ol +/- 3.SE-01 4.93E+Ol +/- 3.lE-01

+1.0 92-7110 7.90E+OO +/- 2.9E-Ol 2.64E+Ol +/- 3.SE-01 5.07E+Ol +/- 3.lE-01

+0.5 92-711P 8.08E+OO +/- 2.SE-01 2.66E+Ol +/- 3.9E-Ol 5.16E+Ol +/- 3.2E-Ol 0 92-7110 8.04E+OO +/- 3.0E-01 2.68E+Ol +/- 4.0E-01 5.lSE+Ol +/- 3.2E-Ol

-0.S 92-711R 8.04E+OO +/- 2.4E-Ol 2.70E+Ol +/- 3.6E-Ol 5.25E+Ol +/- 2.9E-Ol

-1.0 92-7115 8.07E+OO +/- 3.0E-01 2.73E+Ol +/- 3.8E-Ol 5.21E+Ol +/- 3.2E-Ol

-1.5 92-711T 8.20E+OO +/- 3.0E-01 2.73E+Ol +/- 3.9E-Ol 5.15E+Ol +/- 3.2E-Ol

-2.0 92-711U 7.96E+OO +/- 2.9E-Ol 2.63E+Ol +/- 3.9E-Ol 5.05E+Ol +/- 3.lE-01

-2.S 92-711V 7.47E+OO +/- 2.SE-01 2.SlE+Ol +/- 3.8E-Ol 4.94E+Ol +/- 3.lE-01

-3.0 92-711W *7.0SE+OO +/- 2.SE-01 2.42E+Ol +/- 3.8E-Ol 4~73E+Ol +/- 3.0E-01

-3.5 92-711X 6.68E+OO +/- 2.7E-Ol 2.24E+Ol +/- 3.SE-01 4.53E+Ol +/- 3.0E-01

-4.0 92-711Y S.72E+OO +/- 2.5E-Ol 2.03E+Ol +/- 3.SE-01 4.27E+Ol +/- 2.SE-01

-4.S 92-711Z 4.84E+OO +/- 2.4E-Ol l.72E+Ol +/- 3.3E-Ol 3.89E+Ol +/- 2.7E-Ol Remarks:

AL File: 14601

References:

Lab Book#49 pages 259-271.

Procedures: A-524.

Analyst: WTF, TRK, MRK Approved:~~

• C-21

, REPORT Analytic~l RSACPAL775 Westinghouse Advanced Energy Systems Laboratory - Waltz Hill Site Requ~.;U 14601 Origina.or: A.H.Fero (W)Energy Center (E 4-17)

Radiation Engineering &Analysis Received: 3/3/92 Westinghouse Electric Corporation Reported: 3/25/92

[RESULTS OF ANALYSIS]

PALISADES CYCLE 9 REACTOR CAVITY DOSIMETRY ead Chain Tag ID: S-2 340 Degrees Feet [<-------------- dps/mg of chain ll 3/4/92 -------------->]

from Lab ------- Hn-54 ------- ------~ Co-58 ------- ------- Co-60 -------

iiiplane Sample# dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma

+8.0 92-718A 5.44E-01 +/- 6.9E-02 2.08E+OO +/- l.lE-01 2.04E+Ol +/- l.2E-Ol

+7.5 92-7188 7.77E-01 +/- 7.4E-02 3.llE+OO +/- l.2E-Ol 2.29E+Ol +/- l.3E-Ol

+7.0 92-718C l.06E+OO +/- 1.2E-Ol 4.63E+OO +/- 2.lE-01 2.53E+Ol +/- 2.2E-Ol

+6.5 92-7180 l.87E+OO +/- l.6E-Ol 6.90E+OO +l- 2.5E-Ol 2.82E+Ol +/- 2.3E-Ol

+6.0 92-718E 2.72E+OO +/- 2.lE-01 9.86E+OO +/- 2.6E-Ol 3.09E+Ol +/- 2.4E-Ol

+5.5 92-718F 3.67E+OO +/- 2.lE-01 l.30E+Ol'+/- 3.0E-01 3.40E+Ol +/- 2.5E-Ol

+5.0 92-718G 4.83E+OO +/- 2.3E-Ol l.68E+Ol +/- 3.3£-01 3.64E+Ol +/- 2.7E-Ol

+4.5 92-718H 5.73E+OO +/- 2.4E-Ol l.99E+Ol +/- 3.4£-01 3.91E+Ol +/- 2.7£-01

+4.0 92-718I 6.50E+OO +/- 2.5E-Ol 2.21E+Ol +/- 3.6E-Ol 4.15E+Ol +/- 2.SE-01

+3.5 92-718J 6.94E+OO +/- 2.6E-Ol 2.40£+01 +/- 3.6E-01 4.38E+Ol +/- 2.9E-Ol

+3.0 92-718K 7.25E+OO +/- 2.7E-Ol 2.52E+Ol +/- 3.8E-Ol 4.58E+Ol +/- 2.9E-Ol

+2.5 92-718L 7.30£+00 +/- 2.7E-Ol 2.58E+Ol +/- 3.SE-01 4.81E+Ol +/- 3.lE-01

+2.0 92-718H 7.63E+OO +/- 2.6£-01 2.64£+01 +/- 3.6E-Ol 5.0lE+Ol +/- Z.9E-01

+1.5 92-718N 7.64E+OO +/- 2.8E-Ol 2.68E+Ol +/- 4.0E-01 5.16E+Ol +/- 3.2E-Ol

+1.0 92-7180 7.79E+OO +/- 3.0E-01 2.67E+Ol +/- 4.0E-01 5.31E+Ol +/- 3.2E-Ol.

+0.5 92-718P 7.68E+OO +/- 3.0E-01 Z.68E+Ol +/- 4.lE-01 5.45E+Ol +/- 3.3E-01 0 92-718Q 8.05£+00 +/- 3.lE-01 2.73E+Ol +/- 4.lE-01 5.49E+Ol +/- 3.3E-Ol

-0.5 92-718R 8.14E+OO +/- 2.9£-01 2.69E+Ol +/- 4.ZE-01 5.56E+Ol +/- 3.3E-Ol

-1.0 92-718S 8.Z6E+OO +/- 3.0E-01 2.73E+Ol +/- 4.lE-01 5.53E+Ol +/- 3.3£-01

-1.5 92-718T 8.18E+OO +/- Z.9£-01 Z.70E+Ol +/- 4.ZE-01 5.50E+Ol +/- 3.3E-Ol

-z.o 92-718U 7.93E+OO +/- 2.9E-Ol 2.70£+01 +/- 4.lE-01 5.34£+01 +/- 3.ZE-01

-2.5 92-718V 7.52E+OO +/- 2.9E-01 Z.61E+Ol +/- 4.lE-01 5.21E+Ol +/- 3.2E-01

-3.0 92-718W 7.06£+00 +/- 2.8£-01 2.46E+Ol +/- 3.9E-Ol 5.03E+Ol +/- 3.lE-01

-3.5 92-718X 6.59E+OO +/- 2.6E-OI . 2.28E+OI +/- 3.7£-01 4.78£+01 +/- 3.0E-01

-4.0 92-718Y 5.66E+OO +/- 2.5E-OI 2.0lE+OI +/- 3.6E-Ol 4.49E+Ol +/~ 2.9E-Ol

-4.5 92-7182 5.I3E+OO +/- 2.3E-OI l.71E+OI +/- 3.4E-Ol 4.09E+OI +/- 2.8E-01 Remarks:

Approv~:JJ ~~

AL File: 14601

References:

Lab Bookl49 pages 259-271.

Procedures: A-524.

Analyst: WTF, TRK, HRK C-22

wEST IeG§O~S~ ELE~Th ! C~R?aRAT I ON NUCLEAR o!c ADVANCED TECHNOLOGY DIVIS rcJN AES ANALYTICAL LABORATORY WALTZ MILL SITE .

ARNIE FERRO 1 M. Kawchek AL REQUEST 14601

~ NATO Energv Center Receict March 13. 199.2 W284-4891 m:E417 Report March 17 ~ 1992

===============================================================

MATERIAL DESCRIF'TION STAINLESS STEEL DOSIMETRY REACTOR CHAIN BEADS Al Serv. # ~ NATD Identification WEIGHT PERCENT C7.l CHAIN LOCATION F~ Ni Co 92- 707 s-:: 300 71.67 9.7'2.9 0.1345 92- 708 S-2 900 71.36 9.781 0.1387 92- 7(19 S-2 1500 72.07 9.873 0.1430 92- 711 S-2 2600 71.16 9.581 0.1384 92- 712 s-;t j 2700 70.88 9.768 0.1745 92- 713 S-.2 2800 72.13 9.887 0.1430

• 92- 714 S-2 29(10 69.13 10.184 0.1385 92- 715 s-~j 3000 71.97 10.234 0.1784 92- 716 S-2 3150 70.73 10.112 0.1404 92- 717 S-2 33(10 72.93 10.402 0.1417 92- 718 S-2 34(10 70.01 9.986 0.1363 aver std dev 71.28

~1.07 9.958 0.1394

!,.0.250 !,.0.0029 !,.0.0152 0.1461

• 1.RSD 1.57. 2.57. 2.0Y. 10.47. *

• two cobalt values significantly different than group, calculated average

• with and without values.

NIST 12ld SS Control certified 68.23 11.17 0.10 ft aver measured 69.67 11.463 0.0959

!. std dev !_O. 92  !,.0.270 t,O. 0011

7. RSD 1.3% 2.4% 1.27.

S-2 2100 received empty bag, no sample (92-710)

Method of Analysis Operator File Metals ICPS RMcK AL 14601 Approved by C-23

APPENDIX D SPECIFIC ACTIVITIES AND IRRADIATION HISTORY OF REACTOR CAVITY SENSOR SETS - CYCLES 10 & 11 In this appendix, the irradiation history as extracted from NUREG-0020 and the measured specific activities of radiometric monitors irradiated in the reactor cavity during Cycles 10 and 11 are provided.

The irradiation history of Cycles 10 and 11 were as follows:

Cycle Startup Shutdown Comment 10 04/18/92 06/05/93 11 11/06/93 05/22/95 Reference Core Power = 2530 MWt

• The monthly thermal generation applicable to the Palisades reactor is provided in addition to the specific activities of the sensors on the following pages.

• D-1

TABLE D-1 IRRADIATION HISTORY OF REACTOR CAVITY SENSOR SETS Cycle 10 Cycle 11 Thermal Thermal Generation Generation Date MW-hr Da,te MW-hr Mar-92 (j Jul-93 0 Apr-92 620112 Aug-93 0 May-92 1878432 Sep-93 0 Jun-92 1819464 Oct-93 0 Jul-92 1392552 Nov-93 1242336 Aug-92 1459272 Dec-93 1876608 Sep-92 1260672 Jan-94 1844112 Oct-92 1779079 Feb-94 1004688 Nov-92 1326168 Mar-94 0 Dec-92 1880496 Apr-94 0 Jan-93 1879536 May-94 0 Feb-93 1698408 Jun-94 666768 Mar-93 1880544 Jul-94 1874208 Apr-93 1688919 Aug-94 1874448 May-93 862632 Sep-94 1812408 Jun-93 237864 Oct-94 1869960 Nov-94 1813872 Dec-94 1867368 Jan-95 1874304 Feb-95 1655688

• Mar-95 1871832 Apr-95 1811689 May-95 1170744 D-2.

TABLE D-2 e CONTENTS OF MULTIPLE FOIL SENSOR SETS CYCLES 10/11 IRRADIATION Radiometric Monitor Foil ID Capsule ID/ Bare or Cd Position Shielded Fe Ni Cu Ti Nb Co 23su 231Np 0-1 B D - - - - A 0-2 Cd N AN AN BD AN N 0-3 Cd - - - - - - 1 1 P-1 B E - - - - B P-2 Cd 0 AO AO BE AO 0 P-3 Cd - - - - - - 2 2 Q-1 B F - - - - c Q-2 Cd p AP AP BF AP p e Q-3 R-1 Cd B G

- D

- 3 R-2 Cd R AR AR BG AR R R-3 Cd - - - - - - 4 3 S-1 B H - - - - E S-2 Cd s AS AS BH AS s S-3 Cd - - - - - - 5 4 T-1 B I - - - - F T-2 Cd T AT AT BI AT T T-3 Cd - - - - - - 6 5 U-1 B J - - - - G U-2 Cd u AU AU BJ AU u U-3 Cd - - - - - - 7 6

• D-3

e Westinghouse Electric Corporation RER:>RT Chemistry Operations Technology &Analysis Request# 15753 Waltz Mill Site Originator: J. Perock (W)NTD, Energy Center (4-36)

Received: 6/29/95 Reported: 8/24/95

[RESULTS OF ANALYSIS]

Palisades Reactor Cavity Dosimetry - Cycle 10 Lab Dosimeter (@ 8/15/95)

Foil ID Sample# Material Nuclide dpafmg

• 2 sigma D

95-1495 Fe Hn-54 1.29E+el +/- l.2E~l ..

N 95-1496 Fe Hn-54 l.24E+el +/- l.2E-01 E 95-1505 Fe Hn-54 l.34E+el +/- l.6E-01 0 95-1506 Fe Hn-54 l.30E+el +/- l~~E-01 F 95-1515 Fe Hn-54 3.62E+00 +/-:-. 8.5E-02 p 95-1516 Fe Hn-54 3.31E+00 +/- 8.2E-02 G 95-1524 Fe Hn-54 l.22E+el +/- l.5E-01 R 95-1525 Fe Hn-54 l.19E+el +/- l.4E-01 H 95-1534 Fe Hn-54 9.48E+00 +/- 1.4£-01 s 95-1535 Fe Hn-54 9.64E+00 +/- l.3E-01 I 95-1544 Fe Hn-54 9.43E+00 +/- l.4E-01 T 95-1545 Fe Hn-54 9.19B+00 +/- l.3B-01 J 95-1554 Fe Hn-54 l.llB+el +/- l.5B-01 u 95-1555 Fe Hn-54 l.08B+el +/- l.4E-01 A 95-1501 Al Co Co-60 3.16B+e2 +/- 4.6B+00 N 95-1502 Al Co Co-60 2.04E+e2 +/- 3.6B+00 B 95-1511 AlCo Co-60 3.04E+e2 +/- 4.5E+00 0 95-1512 Al Co Co-60 2.02E+e2 +/- 3. 7B+00.

c 95-1521 Al Co Co-60 1. 71E+e2 +/- 3.3B+00 p 95-1522 Al Co Co-60 l.15B+e2 +/- 2. 7B+00 D 95-1530 AlCo Co-60 2.85E+02 +/- 4.2E+00 R 95-1531 AlCo Co-60 l.97B+e2 +/- 3.5B+00 B 95-1540 Al Co Co-60 2.81E+e2 +/- 4.1E+00 s 95-1541 Al Co Co-60 2.00E+e2 +/- 3.5B+00 F 95-1550 AlCo Co-60 3.25B+e2 +/- 4.4E+00 T 95-1551 AlCo Co-60 2.06E+e2 +/- 3.5B+00 G 95-1560 AlCo Co-60 3.41E+e2 +/- 4.5B+00 u 95-1561 AlCo Co-60 2.13B+e2 +/- 3.6B+00 Remarks:

• Results are in units of dps/(mg of Dosimeter Material).

AL File: 15753 Procedures: A-524 Analyst: WTF D-4

e West~use Electric Corporation RBKlRT Chemistry rations Technology & Analysis Request# 15753 Waltz Mill Site Originator: J. Perock {W)NTD, Energy Center (4-36)

Received: 6/29/95 Reported: 8/24/95

[RESULTS OF ANALYSIS]

Palisades Reactor Cavity Dosimetry - Cycle 10 Lab Dosimeter (@ 8/15/95)

Foil ID Sample# Material Nuclide dps/mg

• 2 sisma

---BD 95-1499 Ti Sc-46 3.08B+ee +/- 8.8B-02 BR 95-1509 Ti Sc-46 3.23B+ee +/- 8.91-02 BF 95-1519 Ti Sc-46 8.49B-01 +/- 4.31-02 BG 95-1528 Ti Sc-46 3.0ll+ee +/- 8.51-02 BH 95-1538 Ti Sc-46 2.34E+ee +/- 7.51-02 BI 95-1548 Ti Sc-46 2.331+ee +/- 7.61-02 BJ 95-1558 Ti Sc-46 2.72B+ee +/- 8.2B-02 AN 95-1497 Ni C&-58 1.35B+02 +/- 1. 71+00 AO 95-1507 Ni C&-58 1.44B+02 +/- 1. 71+00 AP*

95-1517 Ni C&-58 3.94B+01 +/- 9.0B-01

. AR 95-1526 Ni C&-58 1.26B+02 +/- 1.6B+ee AS 95-1536 Ni C&-58 1.99B+02 +/- 2.0B+ee AT 95-1546 Ni C&-58 9.92B+01 +/- 1.4B+ee AU 95-1556 Ni C&-58 1.18B+02 +/- 1.61+00 AN 95-1498 ~ C&-60 7.671-01 +/- 2.0B-02 AO 95-1508 ~ C&-60 7.98B-01 +/- 2.11-02 AP 95-1518 ~ C&-60 1.931-01 +/- 1.0B-02 AR 95-1527 ~ C&-60 7.28B-01 +/- 1.9B-02 AS 95-1537 ~ C&-60 6.101-01 +/- 1.81-02 AT 95-1547 ~ C&-60 5.43B-01 +/- 1. 71-02 AU 95-1557 ~ C&-60 6.59B-01 +/- 1.9B-02 Remarks:.

• Results are in uni ts of dps/ (mg of Dosimeter Material) .

AL File: 15753 Procedures: A-524 Analyst: WI'F J.!l~L~lL Approved:_-r-+-__.L_ _ _ _ _ _ __

• D-5

e

• Westinghouse Electric Corporation RE FORT Chemistry Operations Technology & Analysis Request# 15753 Waltz Mill Site Originator: J. Perock (W)NTD, Energy Center (4-36)

Received: 6/29/95 Reported: 8/24/95

---

[RESULTS OF ANALYSIS]

Palisades Reactor Cavity Dosimetry - Cycle 10 Lab Dosimeter (@ 8/15/95)

Foil ID Sample# Material Nuclide dps/mg

• 2 sigma 1 95-1503

-002- - ---- Ce-137 1.55B+00 +/- 2.3B-01 2 95-1513 002 Ce-137 1.41B+00 +/- 2.2E-01 3 95-1523 U02 Ce-137 3.38B-01 +/- 1.3E-01 4 95-1532 002 Cs-137 1.32R+00 +/- 1.6B-01 5 95-1542 U02 Cs-137 1.09B+00 +/- 1.BR-01 6 95-1552 002 Cs--137 1.01B+00 +/- 1.6R-01 7 95-1562 .002 Cs-137 1.33B+00 +/- 2.9R-01 1 95-1503 U02 Ru-103 5.73B+00 +/- 3.6R-01 2 95-1513 002 Ru-103 5.78B+00 +/- 3.9R-01 3 95-1523 002 Ru-103 1. 74B+00 +/- 1.6R-01 4 95-1532 U02 Ru-103 5.37R+00 +/- 2.BB-01 5 95-1542 U02 Ru-103 4.14B+00 +/- 2.9R-01 6 95-1552 002 Ru-103 3.96B+00 +/- 2.3R-01 7 95-1562 U02 Ru-103 4.94B+00 +/- 3.9B-01 1 95-1503 002 Zr-95 7.60B+00 +/- 5.2R-01 2 95-1513 U02 Zr-95 8.16B+ee +/- 6.0R-01 3 95-1523 U02 Zr-95 2.32R+ee +/- 2.5R-01 4 95-1532 U02 Zr-95' 7.81E+00 +/- 4.5R-01 5 95-1542 U02 Zr-95 5.75B+00 +/- 4.2R-01 6 95-1552 002 Zr-95 5.87R+ee +/- 3.BR-01 7 95-1562 002 Zr-95 6.07R+00 +/- 5. 7E-01 Remarks:

• Results are in units of dps/(mg of Dosimeter Material).

AL File: 15753 Procedures: A-524 Analyst: WTF D-6

Westinghouse Electric Corporation RE FORT Chemistry Operations Technology & Analysis Request# 15753 Waltz Mill Site Originator: J. Perock (W)NTD, Energy Center (4-36)

Received: 6/29/95 Reported: 8/24/95

[RESULTS OF ANALYSIS]

Palisades Reactor Cavity Dosimetry - Cycle 10 Lab Dosimeter (@ 8/15/95)

Foil ID Sample# Material Nuclide dps/mg

• 2 sigma 1 95-1504 N}l02 Ce-137 1.92E+el +/- 2.1E+00 2 95-1514 N}l02 Ce-137 2.04B+el +/- 1.8E+00 3 95-1533 N}l02 Ce-137 1.95E+el +/- 1. 7E+00 4 95-1543 N}l02 Ce-137 1.56E+el +/- 1.4E+00 5 95-1553 N}l02 Ce-137 1.49E+el +/- 1.9K+00 6 95-1563 N}l02 Ce-137 1.52K+el +/- 1.6K+00 1 95-1504 N}l02 Ru-103 7.48E+el +/- 3.6E+00 2 95-1514 N}l02 Ru-103 7.54E+el +/- 3.2K+00 3 95-1533 N}l02 Ru-103 6.76E+el +/- 3.4E+00 4 95-1543 N}l02 Ru-103 5.62K+el +/- 2.7E+00 5 95-1553 N}l02 Ru-103 5.29K+el +/- 3.0&+00 6 95-1563 N}l02 Ru-103 6.26E+el +/- 3.0E+00 1 95-1504 N}l02 Zr-95 1.20K+e2 +/- 5.0&+00 2 95-1514 Np02 Zr-95 1.26E+e2 +/- 5.0&+00 3 95-1533 N}l02 Zr-95 1.11E+e2 +/- 4.7E+00 4 95-1543 N}l02 Zr-95 9.55K+el +/- 4.5&+00 5 95-1553 N}l02 Zr-95 9.13K+el +/- 4.7E+00 6 95-1563 N}l02 Zr-95 1.05K+e2 +/- 4.6E+00 Remarks:

• Results are in units of dpa/(mg of Dosimeter Material).

AL File: 15753 Procedures: A-524 Analyst: Wl'F D-7

REVISED Westinghouse Electric Corporation REPORT Chemistry Operations Technology & Analysis Request# 15753 Waltz Mill Site

• Originator: J~ Perock (W)NTD, Energy Center (4-36)

Received: 6/29/95 Reported: 11/27/95

[RESULTS OF ANALYSIS]

Palisades Reactor Cavity Dosimetry - Cycle 10 Bead Chain Tag ID: 270 deg.

Feet from Lab

[<------------

- - - - Mn - - -

dps/mg of chain @ 8/15/95

- * - Co - - -

---

>]

- - - - Co - -

Midplane Sample# dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma

+0.5 95-1488-A 8.40E+00 +/- 4.9E-01 l.21E+01

+/- 5.4E-01 l.10E+02 +/- 1.lE+00

-0.5 95-1488-B 8.27E+00 +/- 5.lE-01 l.18E+01 +/- 5.9E-01 l.07E+02 +/- 1.lE+00

-1.5 95-1488-C 7.15E+00 +/- 4.3E-01 l.05E+01 +/- 5.0E-01 l.05E+02 +/- 1.1E+00

-2.5 95-1488-D 6.17E+00 +/- 4.4E-01 9.82E+00 +/- 4.9E+00 9.79E+01 +/- 9.4E-01

-3.5 95-1488-E 4.70E+00 +/- 3.8E-01 7.39E+00 +/- 4.9B-01 8.76E+01 +/- 8.9B-01

-4.5 95-1488-F 2.96E+00 +/- 3.lE-01 4.54E+00 +/- 3.5E-01 6.64E+01 +/- 7.7E-01

-5.5 95-1488-G l.46E+00 +/- l.5E-01 2.38E+00 +/- l.6E-01 5.64E+01 +/- 4.lE-01 Bead Chain T8:g __ ID: 280 deg.

Feet from Midplane Lab Sample#

[<----------

- - - Mn-54 ------

dps/mg 2 sigma dps/mg of chain @ 8/15/95

- - - Co - -

dps/mg 2 sigma

---

>]

dps/mg Co - - -

2 sigma

+0.5 95-1489-A 8.81E+00.+/- 4.8E-01 1.32E+01 +/- 5.3E-01 1.02E+02 +/- 9.6E-01

-0.5 95-1489-B 8.42E+00 +/- 4.9E-01 l.31E+01 +/- 6.0E-01 1.01E+02 +/- 9.7E-01

-1.5 95-1489-C 8.52E+00 +/- 6.lE-01 l.24E+01 +/- 5.SE-01 9.99E+01 +/- 9.8E-01

-2.5 95-1489-D 7.30E+00 +/- 5.3E-01 l.08E+01 +/- 5.7E-01 9.40B+01 +/- 9.4E-01

-3.5 95-1489-E 6.44E+00 +/- 4.7E-01 9.89E+00 +/- 4.7E-01 8.74E+01 +/- 8.9E-01

-4.5 95-1489-F 4.17E+00 +/- 2.lE-01 6.79E+00 +/- 2.4E-01 7.27E+01 +/- 4.SE-01

-5.73 95-1489-G 1.66E+00 +/- l.7E-01 2.69E+00 +/- l.8E-01 5.57E+01 +/- 4.0E-01 Remarks:

• Results are in units of dps/(mg of Dosimeter Material).

• Correction of Co-58: 95-1489-F AL File: 15753 Procedures: A-524 Analyst: WI'F D-8

Westinghouse Electric Corporation REFORT Chemistry Operations Technology & Analysis Request# 15753 Waltz Mill Site Originator: J .. Perock (W)NTD, Energy Center (4-36)

Received: 6/29/95 Reported: 8/28/95

[RESULTS OF ANALYSIS]

Palisades Reactor Cavity Dosimetry - Cycle 10 Bead Chain Tag ID: 290 deg.

Peet [< dps/mg of chain @ 8/15/95 ------>]

from Lab Mn-54 Co-58 - - - Co - -

Hidplane Sample# dps/mg 2 sigma dPs/mg 2 sigma dps/mg 2 sisma

+0.5 95-1490-A 7.39B+00 +/- 4.7B-01 1.10B+e1 +/- 5.3B-01 9.48E+01 +/- 9.4E-01

-e.5 95-1490-8 7.65B+ee +/- 4.SB-01 1.11B+01 +/- 5.3B-01 9.58E+01 +/- 9.3!-01

-1.5 95-1490-C 7.73B+ee +/- 4.BB+ee 1.12B+01 +/- 5.lB-01 9.38B+01 +/- 9.3B-01

-2.5 95-1490-D 6.95B+ee +/- 4.SB-01 1.02B+01 +/- 5.4E-01 9.09B+01 +/- 9.lB-01

-3.5 95-1490-E 6.07E+ee +/- 4.SE-01 9.26E+ee +/- 5.0E-01 8.53E+01 +/- 8.BB-01

-4.5 95-1490-F 4.66B+ee +/- 2.5E-01 7.36E+ee +/- 2.9E-01 7.64E+01 +/- 4.8B-01

-5.5 95-1490-G 2.52B+00 +/- 2.0B-01 4.04E+ee +/- 2.0E-01 5.74E+01 +/- 4.lB-01

• Bead Chain Tag ID: 300 deg.

Feet from Midplane Lab Sample#

[<-------

- - - - Mn-54 dps/mg 2 sip.a dps/mg of chain @ 8/15/95 dps/mg Co-58 2 sigma

- - - - .---->]

dps/mg Co-60 2 sigma

+0.5 95-1491-A 6.46B+00 +/- 4.3E-01 9.48E+ee +/- 4.9B-01 9.54E+01 +/- 9.3B-01

-e.5 95-1491-8 6.30E+00 +/- 4.lE-01 9.42B+ee +/- 4.SE-01 9.57E+01 +/- 9.3E-01

-1.5 95-1491-C 6.11E+00 +/- 3.9E-01 9.13B+ee +/- 4.7B-01 9.28B+01 +/- 9.2B-01

-2.5 95-1491-D 6.23E+00 +/- 4.BE-01 8.35B+ee +/- 4.5E-01 8.90B+01 +/- 9.lB-01

-3.5 95-1491-E 5.09E+ee +/- 2.lB-01 7.78B+00 +/- 2.6E-01 8.29E+01 +/- 4.SE-01

-4.5 95-1491-F. 4.06E+00 +/- 2.lE-01 6.46E+ee +/- 2.4E-01 7.39B+01 +/- 4.SE-01

-5.5 95-1491-G 2.36E+ee +/- 1.BE-01 3.52B+ee +/- 1.BE-01 5.42B+01 +/- 4.0E-01 Remarks:

• Results are in uni ts of dps/(mg of Dosi.meter Material).

AL File: 15753 Procedures: A-524 Analyst: Wl'F

• D-9

Westinghouse Electric Corporation*

RE FORT Chemistry Operations Technology & Analysis Request# 15753 Waltz Hill Site .

Originator: J. Perock (W)NTD, Energy Cen1<er (4-36)

Received: 6/29/95 Reported: 8/28/95

[RESULTS OF ANALYSIS]

Palisades Reactor Cavity Dosimetry - Cycle 10 3ead Chain Tag ID: 315 deg.

Feet [< dpe/mg of chain @ 8/15/95 ------>]

from Lab - - - - Mn-54 Co-58 - - - Co - -

iidplane Sample# dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigina

+e.5 95-1492-A 6.05E+ee +/- 6.2E-01 8.90E+ee +/- 6.5E-01 1.07E+02 +/- 1.0K+ee

-0.5 95-1492-B 6.18E+e0 +/- 5.2E-01 8.41E+e0 +/- 5.5E-01 1.07E+02 +/- 1.0K+ee

-1.5 95-1492-C 5.59E+e0 +/- 5.4E-01 8.37E+ee +/- 6.4E-01 1.03K+02 +/- 1.0K+ee

-2.5 95-1492-D 4.68E+ee +/- 5.lE-01 7.75E+ee +/- 5.5E-01 9.83K+01 +/- 9.9E-01

-3.5 95-1492-E 4.17E+e0 +/- 3.0E-01 6.78K+ee +/- 3.9E-01 8.90E+01 +/- 6.0K-01

-4.5 95-1492-F 3.57E+e0 +/- 2.9E-01 5.56K+e0 +/- 3.5E-01 7.68K+01 +/- 5.5K~l

-5.5 95-1492-G 1.75E+ee +/- 1.BE-01 3.09K+ee +/- 2.4E-01 5.23E+01 +/- 4.lE-01 3ead Chain Tag ID: 330 deg.

Feet from 1idplane Lab Sample#

[< - - - - -

- - - Mn - -

dps/mg 2 sigma dps,lmg of chain dps/mg

@ 8/15/95 Co - -

2 sigma

- - - Co - -

dps/mg

>]

2 sigma

+e.5 95-1493-A 6.66E+ee +/- 6.3E-01 9.B0E+ee +/- 6.SE-01 1.13K+02 +/- 1.lE+ee

-0.5 95-1493-B 7.llE+ee +/- 5.BE-01 1.01K+01 +/- 7.3K-01 1.14E+e2 +/- 1.lE+ee

-1.5 95-1493-C 6.61.R+ee +/- 5.'TE-01 9.97E+ee +/- 6.'TE-01 1.12K+02 +/- 1.lE+ee

-2.5 95-1493-D 6.72E+ee +/- 4.4E-01 1.01E+01 +/- 5.6K-01 1.07E+02 +/- 9.9K-01

-3.5 95-1493-E 5.70E+ee +/- 4.0.R-01 B.94K+ee +/- 5.3K-01 9.56K+01 +/- 9.3K-01

-4.5 95-1493-F 4.20E+00 +/- 2.lE-01 6.63K+ee +/- 2.3E-01 B.13K+01 +/- 4.9K-01

-5.5 95-1493-G 2.01E+e0 +/- 1.6.R-01 3.31K+ee +/- 2.lE-01 5.79.R+el +/- 4.lE-01 Remarks:

• Results are in units of dps/(mg of Dosimeter Material).

AL File: 15753 Procedures: A-524 Analyst: WI'F D-10

Westinghouse Electric Corporation REroRT Chemistry Operations Technology & Analysis Request# 15753 Waltz Hill Site Originator: J. Perock (W)NTD, Energy Center (4-36)

Received: 6/29/95 Reported: 8/30/95

---

[RESULTS OF ANALYSIS]

Palisades Reactor Cavity Dosimetry - Cycle 10 I Cycle 11 3ead Chain Tag ID: 30 deg.

feet [< dps/mg of chain @ 8/15/95 >]

from Lab - - - Mn-54 Co-58 Co -

iidplane Sample# dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma

+8.0 95-1483-A 4.10K-01 +/- 1.2K-01 7.561-01 +/- 1.6K-01 4.25B+01 +/- 3.5B-01

+7.5 95-1483-B 7 .09E-01 +/- 1.51-01 1.24B+00 +/- 1. 7E-01 4.891+01 +/- 3.BE-01

+6.5 95-1483-C 1.81E+00 +/- 2.11-01 2.72K+00 +/- 1.8E-01 5.95B+01 +/- 4.2K..:e1

+5.5 95-1483-D 3.42K+00 +/- 2.31-01 5.361+00 +/- 2.6B-01 6.97E+01 +/- 4.6R-01 *.

.~ \

+4.5 95-1483-E 4.91E+00 +/- 3.61-01 7.661+00 +/- 4.0B-01 8.04R+01 +/- 8.6R-01

+3.5 95-1483-F 5.68E+00 +/- 3.91-01 9.11R+00 +/- 4.6R-01 9.01R+01 +/- 9.lE-01

+2.5 95-1483-G 6.42K+00 +/- 4.61-01 9.201+00 +/- 4.4B-01 9.701+01 +/- 9.4R-01

+1.5 95-1483-H 6.36E+00 +/- 4.31-01 9.57E+00 +/- 4.8R-01 1.02K+02 +/- 9. 7E-01

+0.5 95-1483-I 6.15R+00 +/- 4.2K-01 9.37E+00 +/- 4.7E-01 1.07E+02 +/- 9.9R-01 e.0 95-1483-J 6.18R+00 +/- 4.51-01 9.56B+00 +/- 5.4R-01 1.091+02 +/- 1.0R+00

-0.5 95-1483-K 6.57E+00 +/- 4.2K-01 . 9. 761+00 +/- 4. 7E-01 1.09B+02 +/- 1.0R+00

-1.5 95-1483-L 6.36E+00 +/- 4.4B-01 9.261+00 +/- 5.0R-01 1.101+02 +/- 1.0R+00

-2.5 95-1483-H 5.98R+00 +/- 4.0K-01 9.731+00 +/- 5.8R-01 1.051+02 +/- 9.9R-01

-3.5 95-1483-N 8.14R+00 +/- 5.01-01 8.951+00 +/- 5.81-01 9.76R+01 +/- 9.5R-01

-4.5 95-1483-0 4.35E+00 +/- 4.31-01 8.87E+00 +/- 4.5R-01 8.38R+01 +/- 8.8R-01

.Remarks:

• Results are in units of dps/(mg of Dosimeter Material) .

AL File: 15753 Procedures: A-524 Analyst: \r7l'F

• D-11

Westinghouse Electric Corporation RER>RT _Chemistry Operations Technology & Analysis Request# 15753 Waltz Hill Site Originator: J. Perock (W)NTD, Energy Center (4-36)

Received: 6/29/95 Reported: 8/30/95 (RESULTS OF ANALYSIS]

. Palisades Reactor Cavity Dosimetry - Cycle 10 / Cycle 11 Bead Chain Tag ID: 90 deg.

Feet (<--------- dps/mg.of cha.in @ 8/15/95 >]

from Lab - - - Hn-54 Co-58 Co-60 Midplane Sample# dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma

+a.0 95-1484-A 5.44E-01 +/- 1.2E-01 7 .98B-01 +/- 1.5B-01 4.46H+01 +/- 3.BB-01

+7.5 95-1484-B 6.07H-01 +/- 1.4E-01 1.24E+00 +/- 1. 7H-01 5.21B+01 +/- 3.9B-01

+a.5 95-1484-C 1.88B+00 +/- 1.9B-01 2.73B+00 +/- 2.lB-01 8.41B+01 +/- 4.4E-01

+5.5 95-1484-D 3.46B+00 +/- 2.lB-01 5.74E+00 +/- 3.0B-01 7.53B+01 +/- 4.8B-01

+4.5 95-1484-E 5.46E+00 +/- 4.4E-01 8.67H+00 +/- 6.lB-01 8.75B+01 +/- 9.lB-01

+3.5 95-1484-F 6.50B+00 +/- 5.2E-01 9.58B+00 +/- 6.4E-01 9.83H+01 +/- 9.7H-01

+2.5 95-1484.;_G 7.10B+00 +/- 5.8B-01 1.13B+01 +/- 7.0B-01 1.09E+02 +/- 1.0B+00

+l.5 95-1484-H 7.52E+00 +/- 6.3B-01 1.16B+01 +/- 6.9B-01 1.18E+02 +/- 1.1B+00

+0.5 95-1484-I 6.93E+00 +/- 5.7H-01 1.08B+01 +/- 6.4E-01 l.23B+02 +/- l.1B+00 0.0 95-1484-J 7.16B+00 +/- 6.0B-01 1.06B+01 +/- 7 .2E-01 l.23B+02 +/- l.1B+00

-0.5 95-1484-K 6.94E+00 +/- 6.3B-01 l.05B+01 +/- 6. 7H-01 l.25B+02 +/- l.lB+ee

-1.5 95-1484-L 6.86E+00 +/- 4.4B-01 l.06E+01 +/- 6.lB-01

• 1.26B+02 +/- 1.1B+00

-2.5 95-1484-M 6.40E+00 +/- 5.lB-01 9.84E+00 +/- 5.4E-01 l.20E+02 +/- l.1B+00

-3.5 95-1484-N 5.70E+00 +/- 4.9B-01 9.30E+00 +/- 5.lB-01 l.08E+02 +/- l.0B+00

-4.5 95-1484-0 4.45E+00 +/- 3.6B-01 6.68B+00 +/- 4.6B-01 8.43E+01 +/- 8.7H-01 Remarks:

• Results are in Wlits of dps/(mg of Dosimeter Material).

AL File: 15753 Procedures: A-524 Analyst: Wl'F D-12

Westinghouse Electric Corporation RBR>RT Chemistry Operations Technology & Analysis Request# 15753 Waltz Hill Site Originator: J. Perock {W)N'l'D, Energy Center {4-36)

Received: 6/29/95 Reported: 8/30/95

[RESULTS OF ANALYSIS]

Palisades Reactor Cavity Dosimetry - Cycle 10 I Cycle 11 3ead Chain Tag ID: 340 deg.

Feet [< dpajmg of chain @ 8/15/95 >]

from Lab ---Mn-54-- Co-58 Co-60 1id.plane Sample# dps/mg 2 sigma dpa/mg 2 sigma dpa/mg 2 sigma

+8.0 95-1494-A 6.leB-01 +/- l.lB-01 9.07K-01 +/- 1.4K-01 4.151+01 +/-

-3.4K-01

+7.5 95-1494-B B.66B-01 +/- l.3B-01 1.54K+ee +/- 1.BB-01 4.681+01 +/- 3.7i-01

+6.5 95-1494-C 1.70B+ee +/- 1.BB-01 3.01B+ee +/- 1. 7i-01 5.891+01 +/- 4.11-01 :c~~:;

+5.5 95-1494-D 3.13B+ee +/- l.9B-01 4.94B+ee +/- 2.lB-01 7.091+01 +/- 4.5B...;el

+4.5 95-1494-B 4.701+00 +/- 2.2B-01 7.67K+ee +/- 2.7i-01 8.291+01 +/- 4.91-01

+3.5 95-1494-F 5.91B+ee +/- 4.3B-01 9.13B+ee +/- 4.4K-01 9.31i+01 +/- 9.2B-01

+2.5 95-1494-G 6.42B+ee +/- 3.3B-01 l.00B+el +/- 4.eB-01 l.02B+e2 +/- 7 .31-01

+1.5 95-1494-H 6.41B+ee +/- 4.5B-01 1.03B+el +/- 5.51-01 1.111+02 +/- 1.0B+ee

+e.5 95-1494-I 6.791+00 +/- 4.5B-01 1.03B+el +/- 5.6B-01 l.17i+02 +/- l.0B+ee 0.0 95-1494-J 7.01B+ee +/- 5.0B-01 1.02B+el +/- 5.lB-01 1.19B+e2 +/- l.0B+ee

-e.5 95-1494-K 6.85B+ee +/- 4.BB-01 l.05B+el +/- 5.3B-01 l.20B+e2 +/- l.0B+ee

-1.5 95-1494-L 6.71i+ee +/- 4.BB-01 l.00B+el +/- 5.3B-01 1.19B+e2 +/- 1.lB+ee

-2.5 95-1494-H 6.46B+ee +/- 4.5B-01 9.82B+ee +/- 5.2B-01 1.14K+02 +/- l.0B+ee

-3.5 95-1494-N 5.77E+ee +/- 5.lB-01 B.81i+ee +/- 4.9B-01 1.05B+e2 +/- B.9B-01

-4.5 95-1494-0 4.85B+ee +/- 4.BB-01 7.73B+ee +/- 5.3B-01 9.05B+el +/~ 9.21!:-01 Remarks:

• Results are in units of dps/(mg of Dosimeter Material).

AL File: 15753 Procedures: A-524 Analyst: \ofl'F Approved:

• D-13

Westinghouse Electric Corporation Chemistry Operations Technology & Analysis Request# 15753 Waltz Mill Site Originator: J. Perock (W)NTD, Energy Center (4-36)

Received: 6/29/95 Reported: 8/31/95

[RESULTS OF ANALYSIS)

Palisades Reactor Cavity Dosimetry - Cycle 10 I Cycle 11 t

Bead Chain Tag ID: 150 deg.

feet from Lab

[<

- - - Mn-54 dps/mg of chain @ 8/15/95

- - Co-58--- --Co-60--

>] I Midplane Sample# dps/mg 2 sigma dps/mg 2 sigma dpe/mg 2 sigma

+8.0 95-1485-A 5.12B-01 +/- 1.3E-01 8.88B-01 +/- 1.5B-01 4.49B~l +/- 3.BB-01

+7.5 95-1485-B 7 .54B-01 +/- 1.2B-01 l.30B+00 +/- l.5B-01 5. llE~l +/- 3.9B-01

+6.5 95-1485-C 1.84B+00 +/- 1.5B-01 3.08B+00 +/- 2.2B-01 8.48B-te1 +/- 4.3B-01

+5.5 95-1485-D 3.61E+00 +/- 2.lE-01 5.71E+00 +/- 2.5B-01 7.55B-tel +/- 4.7B-01

+4.5 95-1485-B 5.34B+00 +/- 4.SB-01 7.99B+00 +/- 5.lE-01 8.80B-te1 +/- 9.0B-01

+3.5 95-1485-F 6.08B+00 +/- 4.lE-01 l.01B-tel +/- 5.4B-01 9.98B-tel +/- 9.5B-01

+2.5 95-1485-G 7.03B+00 +/- 4.7B-01 1.01E-tel +/- 6.lE-01 l.08B~2 +/- l.0B+00

+1.5 95-1485-H 6.62B+00 +/- 4.6B-01 l.03B-tel +/- 5.BB-01 l.15B-te2 +/- l.0B+00

-te.5 95-1485-I 7.03B+00 +/- 4.3B-01 1.02B-tel +/- 5.9B-01 1.20B-te2 +/- 1.1E+00 0.0 95-1485-J 6.88B+00 +/- 5.7B-01 l.03B-tel +/- 5.BB-01 l.21E-t02 +/- 1.1E+00

-0.5 95-1485-K

• 6.72B+00 +/- 5.2B-01 9.72B+00 +/- 5.SB-01 1.22B-te2 +/- 1.1E+00

-1.5 95-1485-L 6.59B+00 +/- 5.3B-01 9.69B+00 +/- 5.7B-01 1.21B-t02 +/- 1.1B+00

-2.5 95-1485-M. 6.41B+00 +/- 5.3B-01 l.01B-tel +/- 6.SB-01 1.17B-t02 +/- 1.1E+00

-3.5 95-1485-N 5.75B+00 +/- 5.3B-01 8.75B+00 +/- 5.7B-01 l.08B-te2 +/- 1.0B+00

-4.5 95-1485-0 4.29B+00 +/- 5.3B-01 8.77B+00 +/- 6.lB-01 9.24B-t01 +/- 9.4B-01 Remarks:

• Results are in Wlits of dps/(mg of Dosimeter Material).

AL File: 15753 Procedures: A-524 Analyst: Wl'F Approved:

D-14

Westinghouse Electric Corporation REPORT Chemistry Operations Technology &Analysis Request# 15753 Waltz Hill Site Originator: J. Perock (W)NTD, Energy Center (4-38)

Received: 6/29/95 Reported: 8/31/95

[RESULTS OF ANALYSIS]

Palisades Reactor Cavity Dosimetry - Cycle 10 I Cycle 11 ead Chain Tag ID: 210 deg.

• 1 Feet [<-------- dpe/mg of chain @ 8/15/95 >]

from Lab Mn - - ~58 ~60 idplane Sample# dpe/mg 2 sigma dpe/mg 2 sigma dps/mg 2 sigma

+8.0 95-1486-A 4.64£-01 +/- 1.5B-01 6.92B-01 +/-

L4B-01 4.29B+01 +/- 3.6B-01

+7.5 95-1486-B 7 .97B-01 +/- 1. 7B-01 1.33B+00 +/- 1.7B-01 4.95B+01 +/- 3.9B-01

+6.5 95-1486-C 1.61B+00 +/- 2.0B-01 2.78B+00 +/- 2.3B-01 6.30B+01 +/- 4.4B-01

+5.5 95-1486-D 3.56B+00 +/- 2.0B-01 5.64B+00 +/- 2.7B-01 7.36B+01 +/- 4.6B-01

+4.5 95-1486-B 5.31B+00 +/- 4.0B-01 8.78B+00 +/- 5.lB-01 8. 73B+01 +/- 8.9B-0_L

+3.5 95-1486-F 6.36B+00 +/- 4.2B-01 9.78B+00 +/- 4.9B-01 9.68B+01 +/- 9.5B-01'

+2.5 95-1486-G 6.33B+00 +/- 4.4B-01 1.01B+01 +/- 5.0B-01 1.06E+02 +/- 9.8B-01 *

+1.5 95-1486-H 6.92B+00 +/- 4.7B-01 1.02B+01 +/- 5.8B-01 1.13B+02 +/- 1.0B+00

+0.5 95-1486-I 6.95B+00 +/- 5.3B-01 1.06B+01 +/- 6.2B-01 1.17B+02 +/- 1.0E+00 0.0 95-1486-J 7.55B+00 +/- 5.8E-01 1.05B+01 +/- 6.6E-01 1.18E+02 +/- 1.0B+00

-0.5 95-1486-K 6.90B+00 +/- 5.8B-01 1.04B+01 +/- 6. 7B-01 1.19B+02 +/- 1.1B+00

-1.5 95-1486-L 6.96B+00 +/- 6.2B-01 1.02E+01 +/- 7.3E-01 1.17B+02 +/- 1.0B+00

-2.5 95-1486-H 6.54B+00 +/- 4.9B-01 1.00E+01 +/- 5.4B-01 1.12B+02 +/- 1.0E+00

-3.5 95-1486-N 5.49B+00 +/- 3.8B-01 8.42B+00 +/- 4.BB-01 1.02E+02 +/- 9.SE-01

-4.5 95-1486-0 4.32B+00 +/- 4.0B-01 6.91E+00 +/- 4.8E-01 8.51!+01 +/- 8.8E-01 Remarks:

• Results are in units of dps/(mg of Dosimeter Material).

AL File: 15753 Procedures: A-524 Analyst: WTF Approved:

• D-15

REVISED Westinghouse Electric Corporation RER:>RT Chemistry Operations Technology & Analysis Request# 15753 Waltz Mill Site Originator: J .. Perock (W)NTD, Energy Center (4-36)

Received: 6/29/95 Reported: 10/3/95

[RESULTS OF ANALYSIS]

Palisades Reactor Cavity Dosimetry - Cycle 10 I Cycle 11 Bead Chain Tag ID: 260 deg.

Feet [<--------- dps/mg of chain @ 8/15/95 >]

from Lab - - - Hn - - - Co-58 Co-60

!lidplane Sample# dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sisma

-+8.0 95-1487-A 4.70E-01 +/- l.4E-01 9.61E-01 +/- 1. 7E-01 4.62E+01 +/- 3.'TE-01

+7.5 95-1487-B 7 .68E-01 +/- 1.3E-01 1.29E+00 +/- 1. 7E-01 5.18E+01 +/- 3.9E-01

+6.5 95-1487-C 1.74E+00 +/- 1.9E-01 2.95E+00 +/- 2.31-01 6.30E+01 +/- 4.3E-01

+5.5 95-1487-D 3.36E+00 +/- 1.8E-01 5.30E+00 +/- 2.~-01

• T.321+01 +/- 4. 7E-01

+4.5 95-1487-E 5.08E+00 +/- 4.6E-01 8.21E+00 +/- 5.4E-01 8.26E+01 +/- 8.8E-01

+3.5 . 95-1487-F 6.28E+00 +/- 5.0E-01 9.99E+00 +/- 5.4E-01 9.181+01 +/- 9.3E-01

+2.5 95-1487-G 6.43E+00 +/- 4.7E-01 1.04E+01 +/- 5.51-01 9.931+01 +/- 9.7E-01

+1.5 95-1487-H 6.45E+00 +/- 4.7E-01 1.051+01 +/- 6. 7E-01 1.04E+02 +/- 9.91-01

+0.5 95-1487-I 6.78E+00 +/- 5.7E-01 9.96E+00 +/- 6.6E-01 1.07E+02 +/- 1.0B+00 0.0 95-1487-J 6.52E+00 +/- 5.8E-01  ; 1.03B+01 +/- 6.7E-01 1.081+02 +/- 1.01+00

-0.5 95-1487-K 6.77E+00 +/- 6.lE-01 1.04B+01 +/- 6.7E-01 1.101+02 +/- 1.01+00

-1.5 95-1487-L 6.90E+00 +/- 5.8E-01 1.07E+01 +/- 6.9B-01 1.07E+02 +/- 1.01+00

-2.5 95-1487-H 6.llE+00 +/- 6.3E-01 9.72E+00 +/- 6.9B-01 1.031+02 +/- 1.0B+00

-3.5 95-1487-N 5.40E+00 +/- 4.6E-01 8.54B+00 +/- 6.2E-01 9.44E+01 +/- 9.7E-01

-4.5 95-1487-0 3.82E+00 +/- 4.7E-01 6.64E+00 +/- 5.21-01 8.04E+01 +/- 9.01-01 Remarks:

• Results are in units of dps/(mg of Dosimeter Material).

• Corrections of Co-58: 95-1487-J and 95-1487-L.

AL File: 15753 Procedures: A-524 Analyst: WI'F Approved:

D-16

Customer John Perock AL*Requeat# 15753 W NTD, Energy Center win (8) 284-5788 . Receipt Date JUN.29, 1995 fax (8) 284-4697 Report Date DEC.11,1995.

Nuclear Plant Pallaadea - cycle #1 O Material Description Stalnleaa steel Doalmetry ReactDr Chain Beads ALServlce# W NTO ldentlftcatlon Weight Percent (%)

Chain Location Fe NI Co 9S- 1483 S-3 30 89.82 9.44 0.18 1484 S-3 90 89.83 9.43 0.18 1485 S-3 150 89.83 9.45 0.18 1486 S-3 210 71.49 9.45 0.18 1487 S-3 280 89.59 9.51 0.19 1488 S-3 270 89.97 9.28 0.18 1489 S-3 280 89.89 9.40 0.17 1490 S-3 290 89.91 9.37 0.18 1491 S-3 300 70.05 9.29 0.18 1492 S-3 315 89.83 9.38 0.18 1493 S-3 330 70.13 9.25 0.18 1494 S-3 340 70.03 9.39 0.18 Average 70.03 9.39 0.18 atd dev 0.48 0.08 0.00

%RSD 0.89 0.84 2.37 Method of Analysis

• Operator File Metals ICPS R.W.MCKINNY 15753 Approved by~~ ~c.cfJ....'

Lawrence Becker

• D-17

APPENDIX E Biological Shield Wall Heat Deposition The presence of heat generated in the biological shield wall, along with the various material temperatures, results in thermal gradients within and between components. These thermal gradients result in thermal stresses and thermal growth which must be accounted for in the design and analysis of the various components. This appendix presents the results of analysis of the heat generation rates in selected biological shield wall components and is based on the use of a core power distribution provided by Consumers Power for Palisades Cycle 11. Note that the DORT model used a reactor vessel thickness of 8.5 inches, which is conservative for this analysis.

Calculation of Heat Generation Rates The calculation of heat generation rates in the biological shield wall involves a number of different steps. First, the reactor core power distribution must be defined. Second, the energy and spatial distribution of the neutron flux must be calculated within the reactor and biological shield wall geometry. Third, the energy and spatial distribution of the gamma ray sources must be calculated. Fourth, the energy and spatial distribution of the gamma ray flux must be calculated. Finally, the neutron kinetic energy deposition and the gamma ray energy deposition in the wall must be calculated. The amount of energy deposited in a component per unit time is referred to as its heat generation rate.

Radiation Analysis Gamma rays (photons) are born in the reactor in several ways. Prompt fission gamma rays are emitted at the instant of fission by the fissioning nucleus. Delayed gamma rays are emitted by the decaying fission products 1* Secondary gamma rays .are also produced in the core, reactor coolant, and the reactor components by neutron interactions other than fission and are the primary contributor to the heat generation outside the reactor vessel. The two most important of these This heat deposition calculation does not include delayed gamma rays, however, this

• source is judged to be negligible. The heat deposition in the biological shield region is dominated by secondary gamma rays produced in the reactor vessel and biological shield.

E-1

neutron interaction processes are radiative capture and inelastic scattering.

In radiative capture of neutrons by nuclei, the kinetic energy of the incident neutron and its binding energy in the compound nucleus are emitted as one or more, generally high-energy gamma rays.

In neutron inelastic scattering, part of the energy of the incident neutron is carried off by the scattered neutron, and part is absorbed by the target nucleus, which subsequently emits one or more, generally low-energy gamma rays.

In the present biological shield wall heat generation rate calculation, the core power distribution defines the fission neutron source and the fission gamma-ray source. Neutron

• transport calculations are required to define the distribution of the other gamma ray sources both inside and outside the core. Once all the sources of gamma rays are defined, gamma-ray transport calculations are performed to determine where the gamma ray energy is ultimately deposited.

In practice, coupled neutron-gamma-ray transport calculations are run, wherein, gamma ray sources are represented as transfer matrices in the cross-section set from neutron groups to gamma ray groups.

There are three main .mechanisms through which gamma rays interact with and deposit energy in the biological shield wall. These mechanisms are the photoelectric effect, pair production, and Compton scattering.

In the photoelectric process, the entire energy of the gamma ray is transferred to an orbital electron which is ejected from its shell and emerges from the atom as a photoelectron. The photoelectric process is important only at low gamma ray energies.

In the process of pair production, a gamma ray interacts with the electric field of atomic electrons or the nucleus. The incident gamma ray is completely annihilated, and its energy is converted into the mass and kinetic energy of an electron-positron pair. Pair production is important only at high gamma ray energies.. For the purposes of understanding biological shield heating, both the photoelectric effect and pair production may be regarded as a total absorption of the gamma ray energy at the point of interaction, which then appears as heat in the material.

The Compton effect is the scattering of a gamma ray by a free electron. The gamma ray imparts energy to the electron and is altered in direction and energy. The energy given to the electron appears as heat in the material. A very important feature of the Compton effect is the fact that, E-2

except when the scattering angle is large, the gamma ray emerges from the interaction with a e significant fraction of the incident gamma ray energy. This fact accounts for much of the complexity associated with gamma ray transport analysis. Compton scattering is the most important gamma ray interaction process at intermediate gamma ray energies .

.l As with gamma rays, there are thr~ main mechanisms through which neutron interact with and deposit energy in the biological shield wall. These mechanisms are elastic scattering, inelastic scattering, and radiative capture.

In the elastic scattering process {n,n), the incident neutron strikes a nucleus in its ground state and when the neutron reappears, the nucleus is left in its ground state. Thus, the kinetic energy of the process is conserved and accounted for in the neutron kerma factors.

In the inelastic scattering process {n,n'), the incident neutron strikes a nucleus in its ground state, however, when the neutron is emitted by the compound nucleus the nucleus is left in an excited

.. l state. Thus, the kinetic energy of the scattered neutron is accounted for as well as the kinetic energy imparted on the excited nucleus as gamma ray emissions dllring decay. Some kinetic energy is left behind in the recoiling nucleus, however, it is negligible.

In radiative capture {n, y ), the neutron is captured by the nucleus and one or more gamma rays are emitted. Some kinetic energy is left behind in the recoiling nucleus, however, it is negligible.

The gamma rays emitted due to the decay of the excited nucleus in both the inelastic scattering and radiative capture processes are accounted for in the gamma ray transport.

Results of Analysis A complete two dimensional discrete ordinates transport theory calculation at core rnidplane was performed to determine the heat generation rate distribution in the Palisades plant biological shield wall. The model included the first 3~ ft of the biological shield wall. The biological shield geometry was modeled as eight separate zones.

This geometry is illustrated in Figures E-1 and E-2 which show the model and selected isometric heat deposition lines in the reactor cavity annulus and biological shield wall, respectively. These also show the azimuthal variation of the heat deposition {w/g of concrete) in the reactor cavity annulus and biological shield wall. Note that in Figure E-1 shows only the outer part of the

• . model, i.e. from the reactor vessel line to the biological shield wall. The model was started with E-3

a boundary source at the reactor vessel liner inner radius (218.44 cm) from the neutron transport calculations for Cycle 11. e Material Zone Inner radius 1 (cm)

Reactor Cavity Liner 335.28 Concrete . 335.92 Homogenized Concrete, Cooling Coils, & Water 341.57 Concrete 344.23 Homogenized Concrete and Rebar (vertical & horizontal) 349.19 Homogenized Concrete and Rebar (radial) 351.85 Homogenized Concrete and Rebar (vertical & horizontal) 372.05 Homogenized Concrete and Rebar (radial) 374.71 End of model 429.48 Table E-1 presents the radial heat deposition in W/g of concrete through the biological shield wall. Figure E-3 shows the radial heat deposition (Wig of concrete) through the model at azimuthal locations of 16° (1=20) and 45° (J=68). Note that the heat deposition shows an analytical oscillation that is near the convergence criteria for the DORT model, i.e. the heat e

deposition at the maximum flux location 16° does not show the maximum heat deposition at all radial locations.

Measured from the reactor vessel centerline E-4

TABLE E-1 e RADIAL VARIATION OF HEAT DEPOSITION IN.THE BIOLOGICAL SHIELD WALL I~

~

't.

. Heat Deposition (W/g of concrete)

Radius (cm) JE. ~ ~ 30° 45° 335.60 9.77e-<XJ 9.72e-<XJ 9.71e-<XJ 9.19e-<XJ 8.76e-<XJ 336.34 8.76e-<XJ 8.75e-<XJ 8.72e-<XJ 8.20e-<XJ 7.76e-<XJ 337.38 7.3 le-<XJ 7.33e-<XJ 7.30e-<XJ 6.83e-<XJ 6.42e-<XJ 339.00 5.55e-<XJ 5.59e-<XJ 5.56e-<XJ 5.18e-<XJ 4.84e-<XJ 340.79 4.04e-<XJ 4.05e-<XJ 4.05e-<XJ 3.75e-<XJ 3.47e-<XJ 342.90 2.76e-<XJ 2.80e-<XJ 2.78e-<XJ 2.57e-<XJ 2.38e-<XJ 344.62 2.06e-<XJ 2.03e-<XJ 2.08e-<XJ 1.90e-<XJ 1.71e-<XJ 347.10 1.47e-<XJ 1.50e-<XJ 1.48e-<XJ 1.36e-<XJ 1.25e-<XJ 350.52 9.29e-07 9.04e-07 9.17e-07 8.43e-07 7.54e-07 352.()C) 7.14e-07 6.99e-07 6.98e-07 6.39e-07 5.79e-07 356.16 4.04e-07 4.23e-07 4.15e-07 3.78e-07 3.47e-07 365.00 1.38e-07 1.28e-07 1.33e-07 1.21e-07 1.07e-07 371.03 7.55e-08 5.85e-08 6.89e-08 6.lle-08 5.04e-08 373.38 5.12e-08 4.36e-08 4.78e-08 4.17e-08 3.69e-08 377.86 2.80e-08 2.63e-08 2.76e-08 2.39e-08 2.18e-08 382.52 1.63e-08 1.47e-08 1.57e-08 1.37e-08 1.22e-08 385.54 1.lle-08 1.05e-08 1.09e-08 9.33e-W 8.64e-()C) 388.58 7.70e-()C) 7.52e..()C) 7.77e-W 6.68e-()C) 6.15e-W 391.61 5.49e-()C) 5.42e-()C) 5.59e-09 4.79e-W 4.41e-W 394.64 3.94e-()C) 3.91e-W 4.02e-09 3.46e-09 3.17e-09 397.67 2.83e-09 2.83e-09 2.91e-09 2.50e-W 2.30e-09 400.70 2.05e-W 2.06e-09 2.lle-09 1.82e-W 1.66e-09 403.73 1.48e-09 1.49e..()C) 1.53e-09 1.32e-()C) 1.21e-W 4<XJ.76 1.08e-()C) 1.09e..()C) 1.lle-09 9.60e-10 8.78e-10 4W.79 7.82e-10 7.92e-10 8.12e-10 6.98e-10 6.39e-10 412.82 5.68e-10 5.77e-10 5.90e-10 5.09e-10 4.65e-10 415.85 4.12e-10 4.19e-10 4.29e-10 3.69e-10 3.38e-10 418.88 2.97e-10 3.03e-10 3.09e-10 2.67e-10 2.44e-10 421.91 2.lle-10 2.16e-10 2.20e-10 1.90e-10 1.74e-10 424.94 1.46e-10 1.49e-10 1.53e-10 1.32e-10 1.20e-10 427.97 9.00e-11 9.21e-11 9.38e-11 8.12e-11 7.43e-11 E-5

FIGURE E-1 BIOLOGICAL SHIELD WALL MODEL

.. {"\

• )

/ *~

EGADS 1.1 X1995/12/18 2819452797599- '.1 Palisades RT Cycle 11 Bioshield Heat Deposition 0

0 0

0 0

0 N

E 0

0 N

0

~

0

~

0.,.,

0 50 100 150 200 250 300 350 400 r, cm E-6

-.* t.

FIGURE E-2 BIOLOGICAL SHIELD WALL ISOMETRIC HEAT DEPOSITION VARIATION PALISADES BIOSHIEL[}1{N ~G}"AEAT 0EPOSITION formw 1 17 priddh Heat Deposition (w/ g of concrete) 0 0

~-i;:-::::::>---

~ N f===--

E

Co 0

N 0

0 0

"' lsoconlou11 1 l.21- 5 .

7 I.lo- 5 l 1.01- 5 4 9.01- 6 5 *1.01- 6 0 so 100 150 200 250 JOO 350 400 6 1.01- 8 Radius - cm 7 1.01- 9 E-7

FIGURE E-3 BIOLOGICAL SHIELD WALL RADIAL HEAT DEPOSITION VARIATION  !~ .,

• ,-1 AT 16° (J=20) AND 45° (J=68) .,

4-l \

I X1995/12/18 16:50:34.60 27258 prideth aspectw 2.1 TP C1995/11/19 PALISADES BIOSHIELD CONCRETE (N,G) HEAT DEPOSITION

~ ~~

1.0-03 1.0-04 '

Q)

Qi.._ ""'\

u '\\._~

1.0-05 c:

0 u '

io 0

1.0-06 '\.

Cl

~

c: ~

- 0 "iii 0

1.0-07 0..

~

Q) 0 0

Q) 1.0-08 I

""~

1.0-09

~

1.0-10 1.0-11 210 260 310 360 410 Radius (cm)

E-8