Text

Reactor Cavity Neutron Measurement Program
	ML020920048
Person / Time
Site:	Mcguire, McGuire
Issue date:	03/19/2002
From:	Barron H; Duke Energy Corp
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	WCAP-15253
	Download: ML020920048 (176)
	v • d • e

Duke 0 Duke Energy Corporation McGuire Nuclear Station 1 Energy. 12700 Hagers Ferry Road Huntersville, NC 28078-9340 H. B. Barton (704) 875-4800 OFFICE Vice President (704) 875-4809 FAX March 19, 2002 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369, 50-370 Reactor Cavity Neutron Measurement Program During Cycle 12 of reactor operation, a reactor cavity measurement program was instituted at each McGuire unit to provided continuous monitoring of the beltline region of the reactor vessel.

The use of the cavity measurement program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel neutron exposure and the uncertainty associated with that exposure over the licensed life of the unit.

The results of the reactor cavity measurement program for McGuire Unit I are contained in WCAP- 15253, "Duke Power Company Reactor Cavity Neutron Measurement Program for William B. McGuire Unit 1 Cycle 12" (attached). Similarly, the results for McGuire Unit 2 are contained in WCAP- 15334, "Duke Power Company Reactor Cavity Neutron Measurement Program for William B. McGuire Unit 2 Cycle 12" (also attached).

By letter dated June 13, 2001, Duke Energy Corporation (DEC) submitted an Application to Renew the Facility Operating Licenses of McGuire Nuclear Station and Catawba Nuclear Station (Application). In a letter dated January 28, 2002, the staff provided requests for additional information (RAIs) based on its review of the reactor coolant system portion of the Application.

Both of the above WCAPs will be referenced in some of the Duke responses to these staff RAIs, which will be submitted on or about April 15, 2002. WCAP-15253 and WCAP-15334 may also be referenced in future licensing submittals pertaining the McGuire reactor vessels and neutron exposure.

If there are any questions concerning this submittal, please contact either Kay Crane at (704) 875-4306 or Bob Gill at (704) 382-3339.

-. B. Barron Attach nents

Westinghouse Non-Proprietary Class 3 Duke Power Company Reactor Cavity Neutron Measurement Program for William B. McGuire Unit I Cycle 12 Westinghouse Electric Company LLC

WESTINGHOUSE NGN-PROPRIETARY CLASS 3 WCAP-15253 Duke Power Company Reactor Cavity Neutron Measurement Program for William B. McGuire Unit 1 Cycle 12 John D. Perock, Stanwood L. Anderson and Thomas V. Congedo Radiation Engineering and Analysis July 1999 Approved: w. §ýQ M. C. Rood, Manager Radiation Engineering and Analysis Prepared by Westinghouse for the Duke Power Company Purchase Order No. MN19501 Work performed under Shop Order No. DIPP450 Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355

iii TABLE OF CONTENTS LIST O F TA BLES ........................................................................................................................................ v LIST O F FIG UR ES ................................................................................................................................... xiii EXECUTIVE

SUMMARY

.................................................................................................................. xv 1 OVERVIEW OF THE PROGRAM ........................................................................................... 1-1 2 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 NEUTRON TRANSPORT AND DOSIMETRY EVALUATION METHODOLOGIES ..... 3-1 3.1 NEUTRON TRANSPORT ANALYSIS METHODS .................................................. 3-1 3.2 NEUTRON DOSIMETRY EVALUATION METHODOLOGY ............................... 3-7 3.3 DETERMINATION OF BEST ESTIMATE PRESSURE VESSEL EXPOSURE ..... 3-12 4 RESULTS OF NEUTRON TRANSPORT CALCULATIONS ............................................... 4-1 4.1 REFERENCE FORWARD CALCULATION .............................................................. 4-1 4.2 FUEL CYCLE SPECIFIC ADJOINT CALCULATIONS ......................................... 4-15 5 EVALUATION 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 EVALUATIONS OF REACTOR CAVITY DOSIMETRY ....................................................... 6-1 6.1 CYCLE 12 RESULTS ..................................................................................................... 6-1 7 COMPARISON OF CALCULATIONS WITH MEASUREMENTS ..................................... 7-1 7.1 COMPARISON OF BEST ESTIMATE RESULTS WITH CALCULATION ........... 7-1 7.2 COMPARISONS OF MEASURED AND CALCULATED SENSOR REA CTIO N RA TES ...................................................................................................... 7-1 8 BEST ESTIMATE NEUTRON EXPOSURE OF PRESSURE VESSEL MATERIALS .......... 8-1 8.1 EXPOSURE DISTRIBUTIONS WITHIN THE BELTLINE REGION ..................... 8-1 8.2 EXPOSURE OF SPECIFIC BELTLINE MATERIALS ............................................. 8-15 8.3 UNCERTAINTIES IN EXPOSURE PROJECTIONS ............................................... 8-19 9 BUGLE-96 BEST ESTIMATE NEUTRON EXPOSURE RESULTS ....................................... 9-1 9.1 COMPARISON OF BUGLE-96 CALCULATIONS WITH MEASUREMENTS .... 9-1 9.2 BEST ESTIMATE NEUTRON EXPOSURE OF PRESSURE VESSEL M ATERIA LS .................................................................................................................. 9-7 9.3 EXPOSURE OF SPECIFIC BELTLINE MATERIALS ............................................. 9-21 10 R EFEREN C ES ........................................................................................................................... 10-1 WCAP-15253, Rev. 0 July 1999

iv TABLE OF CONTENTS (Continued)

A-1 APPENDIX A - SURVEILLANCE CAPSULE DATA ..................................................................

APPENDIX B - CAVITY DOSIMETRY DATA ............................................................................... B-1 July 1999 WCAP-15253, Rev. 00 July 1999

V LIST OF TABLES Table 4.1-1 Calculated Reference Neutron Energy Spectra at Cavity Sensor Set Locations 3411 Mw t; Fa = 1.2 ................................................................................................... 4-3 Table 4.1-2 Reference Neutron Sensor Reaction Rates and Exposure Parameters at the Cavity Sensor Set Locations - 3411 Mwt; Fa = 1.20 ............................................................... 4-4 Table 4.1-3 Calculated Reference Neutron Energy Spectra at Surveillance Capsule Locations - 3411 Mw t; Fa = 1.2 ..................................................................................... 4-5 Table 4.1-4 Reference Neutron Sensor Reaction Rates and Exposure Parameters at the Center of Surveillance Capsules - 3411 Mwt; F .= 1.20 ........................................................ 4-6 Table 4.1-5 Azimuthal Variation of Fast Neutron Flux (E > 1.0 MeV) at the Pressure Vessel Inner Radius .................................................................................................................. 4-7 Table 4.1-6 Summary of Exposure Rates at the Pressure Vessel Clad/Base Metal Interface ............................................................................................................ 4-11 Table 4.1-7 Relative Radial Distribution of Neutron Flux (E > 1.0 MeV) within the Pressure Vessel Wall ................................................................................................... 4-12 Table 4.1-8 Relative Radial Distribution of Neutron Flux (E > 0.1 MeV) within the Pressure Vessel Wall ................................................................................................... 4-13 Table 4.1-9 Relative Radial Distribution of Iron Displacement Rate (dpa) within the Pressure Vessel Wall ................................................................................................... 4-14 Table 4.2-1 Calculated Fast Neutron Flux (E > 1.0 MeV) at the Center of Reactor Vessel Surveillance Capsules ................................................................................................ 4-16 Table 4.2-2 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Center of Reactor Vessel Surveillance Capsules .................................................................................... 4-17 Table 4.2-3 Calculated Fast Neutron Flux (E > 1.0 MeV) at the Pressure Vessel Clad/Base M etal Interface ......................................................................................... 4-18 Table 4.2-4 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface ......................................................................................... 4-19 Table 4.2-5 Calculated Fast Neutron Flux (E > 1.0 MeV) at the Cavity Sensor Set Locations ................................................................................................................ 4-20 Table 4.2-6 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Cavity Sensor Set L ocations ...................................................................................................................... 4-21 WCAP-15253, Rev. 0 July 1999

vi LIST OF TABLES (Continued)

Table 4.2-7 Calculated Fast Neutron Flux (E > 0.1 MeV) at the Center of Reactor Vessel Surveillance C apsules ................................................................................................ 4-22 Table 4.2-8 Calculated Fast Neutron Fluence (E > 0.1 MeV) at the Center of Reactor Vessel Surveillance Capsules .................................................................................... 4-23 Table 4.2-9 Calculated Fast Neutron Flux (E > 0.1 MeV) at the Pressure Vessel Clad/Base M etal Interface ......................................................................................... 4-24 Table 4.2-10 Calculated Fast Neutron Fluence (E > 0.1 MeV) at the Pressure Vessel Clad/Base M etal Interface ......................................................................................... 4-25 Table 4.2-11 Calculated Fast Neutron Flux (E > 0.1 MeV) at the Cavity Sensor Set L ocation s ................................................................................................................ 4-26 Table 4.2-12 Calculated Fast Neutron Fluence (E > 0.1 MeV) at the Cavity Sensor Set L ocation s ...................................................................................................................... 4-27 Table 4.2-13 Calculated Iron Displacement Rate at the Center of Reactor Vessel Surveillance C apsules ................................................................................................ 4-28 Table 4.2-14 Calculated Iron Displacements at the Center of Reactor Vessel Surveillance C ap sules ....................................................................................................................... 4-29 Table 4.2-15 Calculated Iron Displacement Rate at the Pressure Vessel Clad/Base Metal In terface ........................................................................................................................ 4-30 Table 4.2-16 Calculated Iron Displacements at the Pressure Vessel Clad/Base M etal Interface ............................................................................................................ 4-31 Table 4.2-17 Calculated Iron Displacement Rate at the Cavity Sensor Set Locations ............. 4-32 Table 4.2-18 Calculated Iron Displacements at the Cavity Sensor Set Locations .................... 4-33 Table 5.1-1 Summary of Reaction Rates Derived from Multiple Foil Sensor Sets Withdrawn From Internal Surveillance Capsules .................................................... 5-3 Table 5.2.1 Best Estimate Exposure Rates from Surveillance Capsule U Dosimetry Withdrawn at the End of Fuel Cycle 1....................................................................... 5-4 Table 5.2-2 Best Estimate Exposure Rates from the Surveillance Capsule X Dosimetry Withdrawn at the End of Fuel Cycle 5 ....................................................................... 5-5 July 1999 WCAP-15253, Rev. 00 WCAP-15253, July 1999

vii LIST OF TABLES (Continued)

Table 5.2-3 Derived Exposure Rates from the Surveillance Capsule V Dosimetry Withdrawn at the End of Fuel Cycle 8 ....................................................................... 5-6 Table 5.2-4 Derived Exposure Rates from the Surveillance Capsule Z Dosimetry Withdrawn at the End of Fuel Cycle 8 ....................................................................... 5-7 Table 5.2-5 Derived Exposure Rates from the Surveillance Capsule Y Dosimetry Withdrawn at the End of Fuel Cycle 11 ..................................................................... 5-8 Table 6.1-1 Summary of Reaction Rates Derived from Multiple Foil Sensor Sets Cycle 12 Irradiation ..................................................................................................................... 6-3 Table 6.1-2 MFe(np), 58 Ni(np), and ' 8Co(ny) Reaction Rates Derived from the Stainless Steel Gradient Chain at 0.50 - Cycle 12 Irradiation ........................................................... 6-4 Table 6.1-3 MFe(np),

'Ni(np), and 58Co(ny) Reaction Rates Derived from the Stainless Steel Gradient Chain at 14.5' - Cycle 12 Irradiation ......................................................... 6-5 Table 6.1-4 -Fe(np), *Ni(np), and 58Co(ny) Reaction Rates Derived from the Stainless Steel Gradient Chain at 29.50 - Cycle 12 Irradiation ......................................................... 6-6 Table 6.1-5 4Fe(np), sNi(np), and -Co(ny) Reaction Rates Derived from the Stainless Steel Gradient Chain at 44.5' - Cycle 12 Irradiation ......................................................... 6-7 Table 6.1-6 Best Estimate Exposure Rates from the Capsule A Dosimetry Evaluation 0.5 Degree Azimuth - Core Midplane - Cycle 12 Irradiation .................................. 6-8 Table 6.1-7 Best Estimate Exposure Rates From the Capsule B Dosimetry Evaluation 14.5 Degree Azimuth - Core Midplane - Cycle 12 Irradiation ................................ 6-9 Table 6.1-8 Best Estimate Exposure Rates from the Capsule C Dosimetry Evaluation 29.5 Degree Azimuth - Core Midplane - Cycle 12 Irradiation .............................. 6-10 Table 6.1-9 Best Estimate Exposure Rates from the Capsule E Dosimetry Evaluation 44.5 Degree Azimuth - Core Midplane - Cycle 12 Irradiation .............................. 6-11 Table 6.1-10 Best Estimate Exposure Rates From The Capsule D Dosimetry Evaluation 44.5 Degree Azimuth - Top of Core - Cycle 12 Irradiation .................................... 6-12 Table 6.1-11 Best Estimate Exposure Rates from the Capsule F Dosimetry Evaluation 45.0 Degree Azimuth - Bottom of Core - Cycle 12 Irradiation ............................. 6-13 WCAP-15253, Rev. 0 July 1999

viii LIST OF TABLES (Continued) the Table 6.1-12 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position within 6-14 Reactor Cavity - Cycle 12 Irradiation .......................................................................

the Table 6.1-13 Fast Neutron Flux (E > 0.1 MeV) as a Function of Axial Position within 6-15 Reactor Cavity - Cycle 12 Irradiation .......................................................................

the Reactor Table 6.1-14 Iron Displacement Rate as a Function Of Axial Position within 6-16 Cavity - Cycle 12 Irradiation .....................................................................................

Surveillance Table 7.1-1 Comparison of Best Estimate and Calculated Exposure Rates from 7-3 Capsule and Cavity Dosimetry Irradiations .............................................................

Rates Table 7.2-1 Comparison of Measured and Calculated Neutron Sensor Reaction 7-6 from Surveillance Capsule and Cavity Dosimetry Irradiations ...........................

Table 8.1-1 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure 8-3 Vessel 0 D egree A zim uthal Angle ..............................................................................

Table 8.1-2 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure 8-4 Vessel 15 Degree Azim uthal Angle ............................................................................

Table 8.1-3 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure 8-5 Vessel 30 Degree Azimuthal Angle ............................................................................

Table 8.1-4 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure 8-6 Vessel 45 D egree Azimuthal Angle ............................................................................

Projections Table 8.1-5 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure for the Beitline Region of the McGuire Unit 1 Reactor Pressure 8-7 Vessel 0 Degree Azimuthal Angle ..............................................................................

Table 8.1-6 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor 8-8 Pressure Vessel 15 Degree Azimuthal Angle ............................................................

Table 8.1-7 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure 8-9 Vessel 30 Degree Azimuthal Angle ............................................................................

July 199¶ WCAP-15253, Rev. 0 Juy 1999

ix LIST OF TABLES (Continued)

Table 8.1-8 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beliline Region of the McGuire Unit 1 Reactor Pressure Vessel 45 Degree Azim uthal Angle .......................................................................... 8-10 Table 8.1-9 Summary of Best Estimate Iron Atom Displacement [dpa] Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 0 Degree Azim uthal Angle ............................................................................ 8-11 Table 8.1-10 Summary of Best Estimate Iron Atom Displacement [dpa] Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 15 Degree Azim uthal Angle .......................................................................... 8-12 Table 8.1-11 Summary of Best Estimate Iron Atom Displacement [dpa] Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 30 Degree Azim uthal Angle .......................................................................... 8-13 Table 8.1-12 Summary of Best Estimate Iron Atom Displacement [dpal projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 45 Degree Azim uthal Angle .......................................................................... 8-14 Table 8.2-1 Fast Neutron Fluence (E > 1.0 MeV) at Key Plate and Weld Locations of M cG uire U nit 1 ............................................................................................................ 8-16 Table 8.2-2 Fast Neutron Fluence (E > 0.1 MeV) at Key Plate and Weld Locations of M cG uire Unit 1 ............................................................................................................ 8-17 Table 8.2-3 Iron Atom Displacements (dpa) at Key Plate and Weld Locations of M cG uire U nit 1 ............................................................................................................ 8-18 Table 9.1-1 Comparison of Best Estimate and Calculated Exposure Rates from Surveillance Capsule and Cavity Dosimetry Irradiations - BUGLE-96 ....................................... 9-3 Table 9.1-2 Comparison of Measured and Calculated Neutron Sensor Reaction Rates from Surveillance Capsule and Cavity Dosimetry Irradiations - BUGLE-96 ..... 9-6 Table 9.2-1 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 0 Degree Azimuthal Angle - BUGLE-96 ........................................................ 9-9 Table 9.2-2 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 15 Degree Azimuthal Angle - BUGLE-96 .................................................... 9-10 July 1999 WCAP-15253, Rev. 0 WCA-P-15253, Rev. 0 July 1999

x LIST OF TABLES (Continued)

Table 9.2-3 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beitline Region of the McGuire Unit 1 Reactor Pressure Vessel 30 Degree Azimuthal Angle - BUGLE-96 .................................................... 9-11 Table 9.2-4 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 45 Degree Azimuthal Angle - BUGLE-96 .................................................... 9-12 Table 9.2-5 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 0 Degree Azimuthal Angle - BUGLE-96 ...................................................... 9-13 Table 9.2-6 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beitline Region of the McGuire Unit 1 Reactor Pressure Vessel 15 Degree Azimuthal Angle - BUGLE-96 .................................... 9-14 Table 9.2-7 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 30 Degree Azimuthal Angle - BUGLE-96 .................................................... 9-15 Table 9.2-8 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 45 Degree Azimuthal Angle - BUGLE-96 .................................................... 9-16 Table 9.2-9 Summary of Best Estimate Iron Atom Displacement [dpa] Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 0 Degree Azimuthal Angle - BUGLE-96 ...................................................... 9-17 Table 9.2-10 Summary of Best Estimate Iron Atom Displacement [dpa] Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 15 Degree Azimuthal Angle - BUGLE-96 .................................................... 9-18 Table 9.2-11 Summary of Best Estimate Iron Atom Displacement [dpa] Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 30 Degree Azimuthal Angle - BUGLE-96 .................................................... 9-19 Table 9.2-12 Summary of Best Estimate Iron Atom Displacement [dpa] projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 45 Degree Azimuthal Angle - BUGLE-96 .................................................... 9-20 Table 9.3-1 Fast Neutron Fluence (E > 1.0 MeV) at Key Plate and Weld Locations of McGuire U nit 1 - BUG LE-96 ...................................................................................................... 9-22 WCAP-15253, Rev. 0 July 1999

xi LIST OF TABLES (Continued)

Table 9.3-2 Fast Neutron Fluence (E > 0.1 MeV) at Key Plate and Weld Locations of McGuire U nit 1 - BU G LE-96 ...................................................................................................... 9-23 Table 9.3-3 Iron Atom Displacements (dpa) at Key Plate and Weld Locations of McGuire Unit 1 - BUG LE-96 ...................................................................................................... 9-24 Table A-1 McGuire Unit 1 Operating History - Cycles 1 Through 11 .................................... A-2 Table A-2 Radiometric Counting Results from Sensors Removed from Capsule U ............ A-5 Table A-3 Radiometric Counting Results from Sensors Removed from Capsule X ............ A-6 Table A-4 Radiometric Counting Results from Sensors Removed from Capsule V ............ A-7 Table A-5 Radiometric Counting Results from Sensors Removed from Capsule Z ..... A-8 Table A-6 Radiometric Counting Results from Sensors Removed from Capsule Y ............ A-9 Table B-1 McGuire Unit 1 Operating History - Cycle 12 .......................................................... B-2 Table B-2 McGuire Unit 1 Dosimeter Capsule Contents for Cycle 12 .................................... B-3 Table B-3 Radiometric Counting Results from Sensors Removed from Cycle 12 Cavity Dosimetry Set 1S-1 Capsules A, B, C, D, E, and F .................................................... B-4 VVLAF-1525, Rev. July 1999 WCZAI-15253, Rev. U 0 July 1999

xiii LIST OF FIGURES Figure 1-1 Description of Pressure Vessel Beitline Materials .................................................... 1-4 Figure 2.1-1 Azimuthal Location of Sensor Strings Cycle 1 Irradiation ..................................... 2-3 Figure 2.1-2 Axial Location of Multiple Foil Sensor Sets .............................................................. 2-4 Figure 2.1-3 Irradiation Capsule for Cavity Sensor Sets ............................................................... 2-5 Figure 2.2-1 Neutron Sensor Locations within Internal Surveillance Capsules ........................ 2-7 Figure 3.1-1 Reactor Geometry Showing A 450 RO Sector ...................................................... 3-5 Figure 3.1-2 Internal Surveillance Capsule Geometry .................................................................. 3-6 Figure 6.1-1 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position Along the 0.5 Degree Traverse in the Reactor Cavity Cycle 12 Irradiation .......................... 6-17 Figure 6.1-2 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position Along the 14.5 Degree Transverse in the Reactor Cavity Cycle 12 Irradiation .................... 6-18 Figure 6.1-3 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position Along the 29.5 Degree Traverse in the Reactor Cavity Cycle 12 Irradiation ........................ 6-19 Figure 6.1-4 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position Along the 44.5 Degree Traverse in the Reactor Cavity Cycle 12 Irradiation ........................ 6-20 WCAJ.'-15253, Rev. 00 July 1999 WCAP-15253, Rev. July 1999

XV EXECUTIVE

SUMMARY

All of the calculations and dosimetry evaluations presented in this report have been based on the BUGLE-93 nuclear cross-section data library derived from ENDF/B-VI. Additionally, evaluations have been performed with the BUGLE-96 nuclear cross-section data library, also derived from ENDF/B-VI, however interpretation of the results is continuing. The analysis presented in this report is consistent with the NRC approved methodology detailed in WCAP-14040-NP-A"'.

During Cycle 12 of reactor operation, a reactor cavity measurement program was instituted at McGuire Unit 1 to provide continuous monitoring of the beltline region of the reactor pressure vessel. The use of the cavity measurement program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit.

To date, reactor cavity dosimetry has been evaluated at the conclusion of Cycle 12, in addition to the internal surveillance capsules withdrawn following Cycle 1, 5, 8 and 11, resulting in the following projected fast neutron fluence levels at the inner radius of the reactor pressure vessel wall:

MIDPLANE/MAXIMUM O(E > 1.0 MeV) [n/cm 2]

EOC 11 EOC 12 00 3.47E+18 3.73E+18 150 5.11E+18 5.51E+18 300 4.86E+18 5.27E+18 450 5.50E+18 5.95E+18 The maximum exposure location occurs approximately at the axial core midplane.

Based on the continued use of an average (Cycles 9-12) low leakage fuel loading pattern, the projected maximum fast neutron exposure of the vessel beltline materials at 21, 34, and 54 effective full power years of operation is summarized as follows:

July 1999 WCAP-15253, Rev. 0 WCAP-15253, Rev. 0 July 1999

xvi 4D(E > 1.0 MeV) [n/cm 2]

21 EFPY 34 EFPY 51 EFPY Intermediate Shell Plate 1.07E+19 1.71E+19 2.54E+19 00 Longitudinal Weld 6.66E+18 1.06E+19 1.57E+19 1200/240' Longitudinal Weld 9.55E+18 1.52E+19 2.27E+19 Intermediate to Lower Shell Circumferential Weld 1.06E+19 1.69E+19 2.51E+19 Lower Shell Plate 1.06E+19 1.69E+19 2.51E+19 1800 Longitudinal Weld 6.58E+18 1.04E+19 1.55E+19 600/3000 Longitudinal Weld 9.43E+18 1.50E+19 2.24E+19 As further data are accumulated from subsequent irradiations, the neutron environment in the vicinity of the Unit 1 pressure vessel will become better characterized and the uncertainties in the vessel exposure projections will be reduced. Thus, the measurement program will permit the assessment of vessel condition to be based on realistic exposure levels with known uncertainties and will eliminate the need for any unnecessary conservatism in the determination of vessel operating parameters.

July 1999 Rev. 00 WCAP-15253, Rev. July 1999

1-1 1 OVERVIEW OF THE PROGRAM The Reactor Cavity Neutron Measurement Program initiated at McGuire Unit 1 at the onset of reactor operation was designed to provide a mechanism for the long term 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 (RTNDT) over the nuclear power plant lifetime. When used in conjunction with dosimetry from internal surveillance capsules"2 ', the reactor cavity neutron dosimetry provides an extensive plant specific measurement data base that can be used with the results of neutron transport calculations to provide best estimate neutron exposure projections for the pressure vessel with a minimum uncertainty. Minimizing the uncertainty in the neutron exposure projections will, in turn, help to assure that the reactor can be operated in the least restrictive mode possible with respect to

1. 10CFR50 Appendix G pressure/temperature limit curves for normal heatup and cooldown of the reactor coolant system.

2. Emergency Response Guideline (ERG) pressure/temperature limit curves.

3. Pressurized Thermal Shock (PTS) RT. screening criteria.

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

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

A schematic of the beltline region of the McGuire Unit I reactor pressure vessel is provided in Figure 1-1. In this case, the beliline region is constructed of six shell plates, six longitudinal welds, and one circumferential weld. Each of these thirteen materials must be considered in the overall embrittlement assessments of the pressure vessel.

In order to satisfy the requirements of 10CFR50 Appendix G for the calculation of pressure/temperature limit curves for normal heatup and cooldown of the reactor coolant system, fast neutron exposure levels must be defined at depths within the vessel wall equal to 25 and 75 percent of the wall thickness for each of the materials comprising the beltline region.

These locations are commonly referred to as the 1AT and 3/T positions in the vessel wall. The 1AT exposure levels are also used in the determination of upper shelf fracture toughness as specified in 10CFR50 Appendix G.

Overview of the Program July 1999 WCAP-15253, Rev. 0

1-2 In the determination of values of RTMs for comparison with applicable pressurized thermal shock screening criteria for plates, longitudinal welds, 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 probabilistic fracture mechanics evaluation of the pressure vessel is performed, or if an evaluation of thermal annealing and subsequent material re-embrittlement is undertaken, a complete embrittlement profile is required for the entire volume of the pressure vessel beltline. The determination of this embrittlement profile would, in turn, necessitate the evaluation of neutron exposure gradients throughout the entire beltline.

The methodology used to provide these required best estimate neutron exposure evaluations for the McGuire Unit 1 pressure 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 surveillance capsule and reactor cavity dosimetry data bases. That is, as a progression is made from the use of a purely analytical approach tied to experimental benchmarks to an approach that makes use of industry and plant specific power reactor measurements to remove potential biases in the analytical method, knowledge regarding the neutron environment applicable to 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

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

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

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

This report provides the results of neutron dosimetry evaluations performed subsequent to the completion of Cycle 12. 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 pressure vessel wall is described and used to project the exposure of the vessel itself.

Overview of the Program July 1999 WCAP-15253, Rev. 0

1-3 Results of exposure evaluations from surveillance capsule dosimetry withdrawn at the end of Cycles 1, 5, 8, and 11 as well as cavity dosimetry results from Cycle 12 are incorporated to provide the integrated exposure of the pressure vessel from plant startup through the end of Cycle 12. Also, uncertainties associated with the derived exposure parameters at the measurement locations and with the projected exposure of the pressure vessel are provided.

In addition to the evaluation of the current exposure of the reactor vessel beltline materials, projections of the future exposure of the vessel are also provided. Current evaluations and future projections are provided for each of the beltline weldments as well as for the plates comprising both the intermediate and lower shells.

All of the calculations and dosimetry evaluations presented in this report have been based on the BUGLE-93 nuclear cross-section data library derived from ENDF/B-VI.

Overview o0 the Program July 1999 WCAP-15253, Rev. 0

1-4 N

0 z

8-442 90° 2-442B 3or B5012-3 CD ~ 00 1800 0 /

9-442 2700 00 90° rB5013-1 9 -" 2 1802 3.442CA B5013-2 2700 Figure 1-1 Description of Pressure Vessel Beltline Materials Overview of the Program July 1999 WCAP-15253, Rev. 0

2-1 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 consisting of 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 environment over the regions of interest.

Placement of the multiple foil sensor sets was such that spectra evaluations could be made at four azimuthal locations at an axial elevation representative of the midplane of the reactor core.

The intent here was to determine changes in spectra caused by varying amounts of water located between the core and the pressure vessel. Due to the irregular shape of the reactor core, water thickness varies significantly as a function of azimuthal angle. In addition to the four midplane sensor sets, two multiple foil packages were positioned opposite the top and bottom of the active core at the azimuthal angle corresponding to the maximum neutron flux. Here the intent was to measure variations in neutron spectra over the core height; particularly near the top of the fuel where backscattering of neutrons from primary loop nozzles and vessel support structures could produce significant perturbations. At each of the four azimuthal locations selected for core midplane spectra measurements, gradient chains extended over a thirteen foot height centered on the core midplane.

2.1.1 Sensor Placement in the Reactor Cavity The placement of the individual multiple foil sensor sets and gradient chains within the reactor cavity is illustrated in Figures 2.1-1 and 2.1-2. In Figure 2.1-1 plan views of the azimuthal locations of the four strings of sensor sets are depicted. The strings were located at azimuthal positions of 0.5, 14.5, 29.5, and 44.5 degrees relative to the core cardinal axis. The sensor strings were hung in the annular gap between the pressure vessel insulation and the primary biological shield at a nominal radius of 101.75 inches relative to the core centerline.

In Figure 2.1-2, the axial extent of each of the sensor set strings is illustrated along with the locations of the multiple foil holders used during the Cycle 12 irradiation. At the 44.5' azimuth, multiple foil sets were positioned at the core midplane as well as opposite the top and bottom of the active fuel. At each of the remaining azimuthal locations, multiple foil sets were positioned only opposite the core midplane. In all cases, stainless steel gradient chains extended +/- 6.5 feet relative to the midplane of the active core.

The sensor sets and gradient chains were suspended from a support bar that hung underneath the Loop D hot leg nozzle. The support bar was suspended from support plates welded to the liner plate at El. 746'+101/2". The top edge of the support bar was positioned 10 inches above the top of the active fuel. The sensor sets and gradient chains were retained and supported at the bottom by chain damps attached to the mounting plates bolted to the bioshield wall in the Description of the Measurement Program July 1999 WCAP-15253, Rev. 0

2-2 sump area below the reactor vessel. The design of the dosimetry support bar along with the gradient chains and stops ensured correct axial and azimuthal positioning of the dosimetry relative to well known reactor features.

2.1.2 Description of Irradiation Capsules The sensor sets used to characterize the neutron spectra within the reactor cavity were retained in 3.87 inch x 1.00 inch x 0.50 inch rectangular aluminum 6061 capsules such as that shown in Figure 2.1-3. Each capsule included three compartments to hold the neutron sensors. The top compartment (position 1) was intended to accommodate bare radiometric monitors, whereas, the two remaining compartments (positions 2 and 3) were meant to house cadmium shielded packages. The separation between positions I 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 the Cycle 12 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 four traverses. Subsequent to irradiation these gradient chains were removed from the cavity and segmented to provide neutron reaction rate measurements at one foot intervals over the height of the axial traverses. When coupled with a chemical analysis, the stainless steel gradient chains yielded activation results for the 'Fe(n,p), 'Ni(n,p), and 59Co(n,y) reactions. The high purity iron, nickel, and cobalt-aluminum foils contained in the multiple foil sensor sets irradiated during Cycle 12 established a direct correlation with the measured reaction rates from the stainless steel chain and provided an confirmation of the type 304 SS chemical analysis.

Description of the Measurement Program July 1999 WCAP-15253, Rev. 0

2-3 90O 00 1800 0 0.50

14.50

29.50

44.50 270D Locations of Axial Traverses in the Reactor Cavity 900 N

0° 1800

0.50

14.50

29.50

44.50 270D Figure 2.1-1 Azimuthal Location of Sensor Strings Description of the Measurement Program July 1999 WCAP-15253, Rev. 0

2-4 0.50 14.50 29.50 44.50 Core Midplane N Multiple Foil Set

Gradient Chain Figure 2.1-2 Axial Location of Multiple Foil Sensor Sets Description of the Measurement Program July 1999 WCAP-15253, Rev. 0

2-5 II I

Figure 2.1-3 Irradiation Capsule for Cavity Sensor Sets T-1 44 czuLLI.UUIon of ule Tweasurement .rogram July 1999 WCAP-15253, Rev. 0

2-6

2.2 DESCRIPTION

OF SURVEILLANCE CAPSULE DOSIMETRY capsule At the conclusion of the first fuel cycle at McGuire Unit 1, the first material surveillance second was withdrawn from its position between the neutron pad and the reactor vessel. The 5, two more following internal surveillance capsule was withdrawn at the conclusion of Cycle capsules Cycle 8, and a fifth after Cycle 11. The neutron dosimetry contained within these their respective provided a measure of the integral exposure received by the capsules during I through irradiation periods; i.e., Cycle 1, Cycles 1 through 5, Cycles 1 through 8, and Cycles 11.

program are The type and location of the neutron sensors included in the materials surveillance Figure 2.2-1 of this described in some detail in Reference 2; and, are illustrated schematically in report.

in wire form, were Relative to Figure 2.2-1, copper, iron, nickel, and cobalt-aluminum monitors, placed in holes drilled in spacers at several axial levels within each capsule. The within a cadmium-shielded uranium and neptunium fission monitors were accommodated to the dosimeter block located near the center of the capsule. Specific information pertinent the appendices to individual sensor sets included in Capsules U, X, V, Z, and Y is provided in this report.

July iSi3 Description of the Measurement Program JulyRe.9 WCAP-15253, Rev. 0

2-7 Gk Figure 2.2-1 Neutron Sensor Locations within Internal Surveillance Capsules iesrippuon of me Mveasurement 1-rogram July 1999 WCAP-15253, Rev. 0

3-1 3 NEUTRON TRANSPORT AND DOSIMETRY EVALUATION METHODOLOGIES As noted in Section 1.0 of this report, the best estimate exposure of the reactor pressure vessel was developed using a combination of absolute plant specific neutron transport calculations and plant specific measurements from the reactor cavity and internal surveillance capsules. In this section, the neutron transport and dosimetry evaluation methodologies are discussed in some detail and the approach used to combine the calculations and measurements to produce the best estimate vessel exposure is presented.

3.1 NEUTRON TRANSPORT ANALYSIS METHODS Fast neutron exposure calculations for the reactor and cavity geometry were carried out using both forward and adjoint discrete ordinates transport techniques. A single forward calculation provided the relative energy distribution of neutrons for use as input to neutron dosimetry evaluations as well as for use in relating measurement results to the actual exposure at key locations in the pressure vessel wall. A series of adjoint calculations, on the other hand, established the means to compute absolute exposure rate values using fuel cycle specific core power distributions; thus, providing a direct comparison with all dosimetry results obtained over the operating history of the reactor.

In combination, the absolute cycle specific data from the adjoint evaluations together with relative neutron energy spectra distributions 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 pressure vessel.

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

3.1.1 Reference Forward Calculation A plan view of the reactor geometry at the core midplane elevation is shown in Figure 3.1-1.

Since the reactor exhibits 1/8" core symmetry only a 0-45 degree sector is depicted. In addition to the core, reactor internals, pressure vessel, and the primary biological shield, the model also included explicit representations of the surveillance capsules, the pressure vessel cladding, and the mirror insulation located external to the vessel.

A description of a dual surveillance capsule holder attached to the neutron pad is shown in Figure 3.1-2. From a neutronic standpoint, the inclusion of the surveillance capsules and Neutron Transport and Dosimetry Evaluation Methodologies July 1999 WCAP-15253, Rev. 0

3-2 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 model was not required.

The forward transport calculation for the reactor model depicted in Figures 3.1-1 and 3.1-2 was 31 carried out in re geometry using the DORT two-dimensional discrete ordinates code" and the BUGLE-93 cross-section library1 41. The BUGLE-93 library is a 47 neutron group, ENDF/B-VI based, data set produced specifically for light water reactor applications. In these analyses, anisotropic scattering was treated with a P3 expansion of the scattering cross-sections and the angular discretization was modeled with an S, order of angular quadrature. The reference forward calculation was normalized to a core mnidplane power density characteristic of operation at a thermal power level of 3411 MWt. The 3411 MWt power level represents the rated operating power for the McGuire Unit I reactor.

The spatial core power distribution utilized in the reference forward calculation was derived from statistical studies of long-term operation of Westinghouse 4-loop plants. Inherent in the development of this reference core power distribution was the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2o uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used.

Due to the use of this bounding spatial power distribution, the results from the reference forward calculation establish conservative exposure projections for reactors of this design operating at the rated power of 3411 MWt. Since it is unlikely that actual reactor operation would result in the implementation of a power distribution at the nominal +20 level for a large number of fuel cycles and, further, because of the widespread implementation of low leakage fuel management strategies, the fuel cycle specific calculations for this reactor result in exposure rates well below these conservative predictions. This difference between the conservative forward calculation and the fuel cycle specific best estimate computations is illustrated by a comparison of the analytical results given in Section 4.0 of this report.

3.1.2 Cycle Specific Adjoint Calculations All adjoint analyses were also carried out using an S, order of angular quadrature and the P3 cross-section approximation from the BUGLE-93 library. Adjoint source locations were chosen at each of the azimuthal locations containing cavity dosimetry as well as at several key azimuths on the pressure vessel inner radius. In addition, adjoint calculations were carried out Neutron Transport and Dosimetry Evaluation Methodologies July 1999 WCAP-15253, Rev. 0

3-3 for sources positioned at the geometric center of capsules located at 31.5 and 34.0 degrees relative to the core cardinal axes.

Again, these calculations were run in rO geometry to provide neutron source distribution importance functions for the exposure parameter of interest, in this case, O(E > 1.0 MeV).

The importance functions generated from these individual adjoint analyses provided the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for each of the fuel cycles to date; and, established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles.

Having the importance functions and appropriate core source distributions, the response of interest can be calculated as:

R(r,0) = f f f I(r,0, E) S(r,0,E) r dr dt dE r 0 E where:

R(r,G) = (E > 1.0 MeV) at radius r and azimuthal angle e.

I(r,O,E)= Adjoint source importance function at radius r, azimuthal angle 0, and neutron source energy E.

S(r,O,E)= Neutron source strength at core location rE and energy E.

It is important to note that the cycle specific neutron source distributions, S(rG,E), utilized with the adjoint importance functions, I(r,e,E), permitted the use not only of fuel cycle specific spatial variations of fission rates within the reactor core, but, also allowed for the inclusion of the effects of the differing neutron yield per fission and the variation in fission spectrum introduced by the build-in of plutonium isotopes as the burnup of individual fuel assemblies increased.

Although the adjoint importance functions used in these analyses were based on a response function defined by the threshold neutron flux (E > 1.0 MeV), prior calculations"5 ' have shown that, while the implementation of low leakage loading patterns significantly impact the magnitude and the spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the exposure parameter ratios such as [dpa/sec]/[ O(E > 1.0 MeV)] are insensitive to changing core source distributions. In the application of these adjoint importance functions to the current evaluations, therefore, calculation of the iron displacement rates (dpa/sec) and the neutron flux (E > 0.1 MeV) were computed on a cycle specific basis by using the appropriate [dpa/sec]/[ O(E > 1.0 MeV)] and

[O(E > 0.1 MeV)]/[ O(E > 1.0 MeV)] ratios from the reference forward analysis in conjunction with the cycle specific O(E > 1.0 MeV) solutions from the individual adjoint evaluations.

Neutron Transport and Dosimetry Evaluation Methodologies July 1999 WCAP-15253, Rev. 0

3-4 In particular, after defining the following exposure rate ratios,

[dpa / sec ]

O(E > 1.0 MeV)

R R (E > 0.1 MeV)

O(E > 1.0 MeV) the corresponding fuel cycle specific exposure rates at the adjoint source locations were computed from the following relations:

dpa/sec = [O(E > 1.OMeV)] R1 O(E > 0.1 MeV) = [ O(E > 1.0 MeV)] R2 All absolute calculations were also normalized to the current rated power level for McGuire Unit 1, 3411 MWt.

Neutron Transport and Dosimetry Evaluation Methodologies July 1999 WCAP-15253, Rev. 0

3-5

mO*

'1.

I Figure 3.1-1 Reactor Geometry Showing A 450 RE Sector Neutron Transport and Dosimetry Evaluation Methodologies July 1999 WCAP-15253, Rev. 0

3-6 OETC PAD Figure 3.1-2 Internal Surveillance Capsule Geometry JuLy 1999 Neutron Transport and Dosimetry Evaluation Methodologies July 1999 WCAP-15253, Rev. 0

3-7 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.

These procedures apply to all of the evaluations provided in this report. For McGuire Unit 1, the measurement of specific activities were performed by Westinghouse or a qualified supplier and Westinghouse carried out the evaluation of all measured data. Thus, the measurement and evaluation procedures were consistent for all surveillance capsule and cavity dosimetry evaluations.

3.2.1 Determination of Sensor Reaction Rates Following irradiation, the multiple foil sensor sets along with reactor cavity gradient wires or chains were recovered and transported to Pittsburgh for evaluation. Analysis of all radiometric foils and gradient chains was performed at the Westinghouse Waltz Mill Facility by Antech Ltd.

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" '", 6 1 . Following sample preparation and weighing, the specific activity of each sensor was determined by means of a lithium drifted germanium, Ge(Li), gamma spectrometer. In the case of the surveillance capsule and cavity multiple foil sensor sets, these analyses were performed by direct counting of each of the individual foils or wires; or, as in the case of mU and "7Np 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 or wires used in the cavity Neutron Transport and Dosimetry Evaluation Methodologies July 1999 WCAP-15253, Rev. 0

3-8 irradiations, individual sensors were obtained by cutting the chains or wires into a series of segments to provide data points at one foot intervals over an axial span encompassing +/- 6 feet relative to the reactor core midplane.

The irradiation history of the reactor over its operating lifetime was obtained from NUREG-0020, :Licensed Operating Reactors Status Summary Report" and from data supplied 1 . For the sensor sets utilized in surveillance capsule and reactor cavity by the utilityý 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 3411 MWt were determined from the following equation:

A NOFY P C1 [1-e-tf ]e-td j Pref 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 Pf (rps/nucleus).

N. = number of target element atoms per gram of sensor.

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

Y = number of product atoms produced per reaction.

Pj = average core power level during irradiation period j (MW).

P = maximum or reference core power level of the reactor (MW).

C = calculated ratio of cp(E > 1.0 MeV) during irradiation period j to the time weighted average O(E > 1.0 MeV) over the entire irradiation period.

= decay constant of the product isotope (sec').

Tf = 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 Pj/P,,f accounts for month by month variation of power level within a given fuel cycle. The ratio C, is calculated for each fuel cycle using the adjoint Neutron Transport and Dosimetry Evaluation Methodologies July 1999 WCAP-15253, Rev- 0

3-9 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 C. = 1.0. However, for multiple cycle irradiations, particularly those employing low leakage fuel management the additional C, 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 'U foil measurements to account for the presence of '3U 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 calculations.

In addition to the corrections made for the presence of 2U in the 23U fission sensors, corrections were also made to both the .. U and "7Np sensor reaction rates to account for gamma ray induced fission reactions occurring over the course of the irradiation. These photo-fission corrections were, likewise, location dependent and were based on the reference transport calculations described in Section 3.1.1.

For the cavity fission monitors, additional corrections were made to the 238U and 2 7Np reactions for the vanadium encapsulated oxide detectors. Since these sensors are counted directly with the Ge(Li) detector, a correction has to be made to account for self-absorption of the fission fragment gamma rays by the U0 2 or NpO 2 oxide material and the vanadium tubing.

These correction factors were determined by J. M. Adams, et al and are reported in a recent paper (Reference 18). For the three fission product isotopes measured the correction factors are:

Daughter Correction 95Zr 1.048 103 Ru 1.073 137 Cs 1.055 3.2.2 Least Squares Adjustment Procedure Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code*' 91. 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 best estimate exposure parameters along with the associated uncertainties were then obtained from the best estimate spectrum. This methodology is fully consistent with that described in Reference 1.

Neutron Transport and Dosimetry Evaluation Methodologies July 1999 WCAP-15253, Rev. 0

3-10 In the FERRET evaluations, a log-normal least squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertainties and correlations.

matrix, A:

In general, the measured values, f, are linearly related to the flux, 0, by some response f (sl) = , A( Oc) i 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, R 1 = Z ig C relates a set of measured reaction rates, R, to a single spectrum, 0g, by the multi-group reaction cross-section, Gag. The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with large assigned uncertainties.

cross In the least squares adjustment, the continuous quantities (i.e., neutron spectra and of 53 energy groups. The trial sections) were approximated in a multi-group format consisting 2 °.

input spectrum was converted to the FERRET 53 group structure using the SAND-il code This procedure was carried out by first expanding the 47 group calculated spectrum into the group SAND-II 620 group structure using a SPLINE interpolation procedure in regions where boundaries do not coincide. The 620 point spectrum was then re-collapsed into the group structure used in FERRET.

The sensor set reaction cross-sections, obtained from the ENDF/B-VI dosimetry file [211, were also collapsed into the 53 energy group structure using the SAND-il code. In this instance, the trial spectrum, as expanded to 620 groups, was employed as a weighting function in the cross 53 section collapsing procedure. Reaction cross-section uncertainties in the form of a 53 x 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 taken from the center of the surveillance capsule modeled in the reference forward transport calculation. While the 53 x 53 group covariance matrices applicable to the sensor reaction cross-sections were developed from the ENDF/B-VI data files, the covariance matrix for the input trial spectrum was constructed from the following relation:

Mgg R + Rg RgPgg Neutron Transport and Dosimetry Evaluation Methodologies July 1999 WCAP-15253, Rev. 0

3-11 where R. specifies an overall fractional normalization uncertainty (i.e., complete correlation) for the set of values. The fractional uncertainties, Rg, specify additional random uncertainties for group g that are correlated with a correlation matrix given by:

P = [I.-18g + o e-H where:

2 H (g-g')

2 r2 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes short range correlations over a group range y (0 specifies the strength of the latter term). The value of 8 is 1 when g = g' and 0 otherwise. For the trial spectrum used in the current evaluations, a short range correlation of y = 6 groups was used. This choice implies that neighboring groups are strongly correlated when 0 is close to 1. Strong long-range correlations (or anti-correlations) were justified based on information presented by R. E. Maerker ~'. The uncertainties associated with the measured reaction rates included both statistical (counting) and systematic components. The systematic component of the overall uncertainty accounts for counter efficiency, counter calibrations, irradiation history corrections, and corrections for competing reactions in the individual sensors.

In performing the least squares adjustment with the FERRET code, the input spectra from the reference forward transport calculations were normalized to the absolute calculations from the cycle specific adjoint analyses. The specific normalization factors for individual evaluations depended on the location of the sensor set as well as on the neutron flux level at that location.

The specific assignment of uncertainties in the measured reaction rates and the input (trial) spectra used in the FERRET evaluations was as follows:

REACTION RATE UNCERTAINTY 5%

FLUX NORMALIZATION UNCERTAINTY 15%

FLUX GROUP UNCERTAINTIES (E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 29%

(E < 0.68 eV) 52%

SHORT RANGE CORRELATION (E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Neutron Transport and Dosimetry Evaluation Methodologies July 1999 WCAP-15253, Rev. 0

3-12 FLUX GROUP CORRELATION RANGE (E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 It should be noted that the uncertainties listed for the upper energy ranges extend down to the lower range. Thus, the 29% group uncertainty in the second range is made up of a 15%

uncertainty with a 0.9 short range correlation and a range of 6, and a second part of magnitude 25% with a 0.5 correlation and a range of 3.

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

3.3 DETERMINATION OF BEST ESTIMATE PRESSURE VESSEL EXPOSURE As noted earlier in this report, the best estimate exposure of the reactor pressure vessel was developed using a combination of absolute plant specific transport calculations based on the methodology discussed in Section 3.1 and plant specific measurement data determined using the measurement evaluation techniques described in Section 3.2. In particular, the best estimate vessel exposure is obtained from the following relationship:

(DBest Est. = K (DCalc.

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

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

(DCalc = 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 pressure vessel wall, additional uncertainty is incurred due to the analytically determined relative ratios among the various measurement points and locations within the pressure vessel wall.

Neutron Transport and Dosimetry Evaluation Methodologies July 1999 WCAP-15253, Rev. 0

3-13 The implementation of this approach acts to remove plant specific biases associated with the definition of the core source, actual vs. assumed reactor dimensions, and operational variations in water density within the reactor. As a result, the overall uncertainty in the best estimate exposure projections within the vessel wall depend on the individual uncertainties in the measurement process, the uncertainty in the dosimetry location, and in the uncertainty in the calculated ratio of the neutron exposure at the point of interest to that at the measurement location.

The uncertainties in the measured flux were derived directly from the results of the least squares evaluation of dosimetry data. The positioning uncertainties were taken from parametric studies of sensor position performed as part of an analytical sensitivity evaluation of the McGuire Unit 1 reactor. The uncertainties in the exposure ratios relating dosimetry results to positions within the vessel wall were based on analytical sensitivity studies of the vessel thickness tolerance for the cavity data and on downcomer water density variations and vessel inner radius tolerance for the surveillance capsule measurements.

Neutron Transport and Dosimetry Evaluation Methodologies July 1999 WCAP-15253, Rev. 0

4-1 4 RESULTS OF NEUTRON TRANSPORT CALCULATIONS 4.1 REFERENCE FORWARD CALCULATION As noted in Section 3.0 of this report, data from the reference forward transport calculation were used in evaluating dosimetry from both reactor cavity and surveillance capsule irradiations as well as in relating the results of these evaluations to the neutron exposure of the pressure vessel wall. In this section, the key data extracted from the reference forward calculation is presented and its relevance to the dosimetry evaluations and vessel exposure projections is discussed. The reader should recall that the results of the reference forward transport calculation were intended for use on a relative basis and, therefore, should not be used for absolute comparison with measurement. All absolute comparisons were based on the results of the fuel cycle specific adjoint calculations discussed in Section 4.2.

4.1.1 Cavity Sensor Set Locations Data from the reference forward calculation pertinent to cavity sensor evaluations are provided in Tables 4.1-1 and 4.1-2.

In Table 4.1-1, the calculated neutron energy spectra applicable to the permanent sensor locations at 0.5, 14.5, 29.5 and 44.5 degrees relative to the core cardinal axes are listed. These data represent the trial spectra used as the starting guess in the FERRET least squares adjustment evaluations of the cavity sensor sets. On a relative basis these calculated energy distributions establish a baseline against which adjusted spectra may be compared; and, when coupled with the adjoint results of Section 4.2, provide an analytical prediction of absolute neutron spectra at the sensor set locations for each irradiation period.

In Table 4.1-2, the calculated neutron sensor reaction rates associated with the spectra from Table 4.1-1 are provided along with the reference exposure rates in terms of (D(E > 1.0 MeV),

D(E < 0.1 MeV) and dpa/sec. Also listed are the associated exposure rate ratios calculated for each of the cavity sensor set locations.

The reference reaction rates, exposure rates, and exposure rate ratios were used in conjunction with fuel cycle specific adjoint transport calculations from Section 4.2 to provide calculated sensor set reaction rates and to project sensor set exposures in terms of 1I(E > 0.1 MeV) and dpa/sec for each irradiation period.

4.1.2 Surveillance Capsule Locations Data from the reference forward calculation pertinent to surveillance capsule evaluations are provided in Tables 4.1-3 and 4.1-4.

In Table 4.1-3, the calculated neutron energy spectra at the geometric center of surveillance capsules located at 34 and 31.5 degrees relative to the core cardinal axes are listed. In Table 4.1-4, the calculated neutron sensor reaction rates and exposure rate ratios associated with Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-2 the spectra from Table 4.1-3 are provided along with the calculated exposure rates in terms of of (D(E > 1.0 MeV), c1(E < 0.1 MeV) and dpa/sec. Again, these data are applicable to the geometric center of each surveillance capsule. These tabulated data were used in the surveillance capsule dosimetry evaluations and exposure calculations in the same fashion as was the case for the cavity sensor sets.

4.1.3 Pressure Vessel Wall Data from the reference forward calculation pertinent to the pressure vessel wall are provided in Tables 4.1-5 through 4.1-9.

In Table 4.1-5, the calculated azimuthal distribution of fast neutron flux (E > 1.0 MeV) is listed for the center of the vessel cladding, at the pressure vessel clad/base metal interface, and at the center of the first DORT mesh interval in the base metal. The interface information (base metal inner radius) was obtained from a linear interpolation of the two sets of data obtained directly from the reference forward calculation. In this detailed tabulation, calculated flux levels are given for each of the 98 azimuthal mesh intervals included in the analytical model.

In Table 4.1-6, the calculated azimuthal distribution of exposure rates in terms of c1(E > 1.0 MeV), (D(E < 0.1 MeV), and dpa/sec are listed at approximately 5 degree intervals over the reactor geometry. These data are applicable to the clad/base metal interface. Also given in Table 4.1-6 are the exposure rate ratios [(D(E > 0.1 MeV)]/[ (D(E > 1.0 MeV)] and

[dpa/sec]/[ (D(E > 1.0 MeV)] that provide an indication of the variation in neutron spectrum as a function of azimuthal angle at the pressure vessel inner radius.

Radial gradient information for 1(E > 1.0 MeV), 1(E > 0.1 MeV), and dpa/sec is given in Tables 4.1-7, 4.1-8, and 4.1-9, respectively. These data are presented on a relative basis for each exposure parameter at the 0, 15, 30, and 45 degree azimuthal locations. Exposure rate distributions within the vessel wall were obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data given in Tables 4.1-7 through 4.1-9.

Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-3 Table 4.1-1 Calculated Reference Neutron Energy Spectra at Cavity Sensor Set Locations 3411 MWt, F,=1.20 Lower Energy

[MeV] 0.50 14.50 29.50 44.50 1.42E+01 3.32E+05 3.83E+05 3.37E+05 2.58E+05 1.22E+01 8.98E+05 1.06E+06 9.39E+05 7.26E+05 1.OOE+01 3.27E+06 3.95E+06 3.58E+06 2.84E+06 8.61E+00 5.74E+06 7.06E+06 6.43E+06 5.12E+06 7.41E+00 8.28E+06 1.05E+07 9.62E+06 7.72E+06 6.07E+00 1.70E+07 2.18E+07 2.OOE+07 1.57E+07 4.97E+00 2.23E+07 2.98E+07 2.78E+07 2.22E+07 3.68E+00 3.80E+07 5.30E+07 5.14E+07 4.25E+07 3.01E+00 2.95E+07 4.17E+07 4.14E+07 3.50E+07 2.73E+00 2.29E+07 3.27E+07 3.30E+07 2.86E+07 2.47E+00 2.84E+07 4.10E+07 4.18E+07 3.63E+07 2.37E+00 1.46E+07 2.10E+07 2.16E+07 1.91E+07 2.35E+00 4.32E+06 6.19E+06 6.48E+06 6.02E+06 2.23E+00 2.20E+07 3.14E+07 3.29E+07 3.05E+07 1.92E+00 6.06E+07 8.58E+07 9.03E+07 8.41E+07 1.65E+00 8.37E+07 1.19E+08 1.28E+08 1.21E+08 1.35E+00 1.43E+08 2.09E+08 2.26E+08 2.11E+08 1.OOE+00 3.41E+08 4.96E+08 5.47E+08 5.20E+08 8.21E-01 3.35E+08 4.89E+08 5.47E+08 5.27E+08 7.43E-01 1.63E+08 2.63E+08 2.93E+08 2.49E+08 6.08E-01 7.30E+08 1.05E+09 1.20E+09 1.18E+09 4.98E-01 7.15E+08 1.07E+09 1.23E+09 1.16E+09 3.69E-01 7.22E+08 1.11E+09 1.28E+09 1.16E+09 2.97E-01 1.08E+09 1.58E+09 1.84E+09 1.77E+09 1.83E-01 1.39E+09 2.23E+09 2.59E+09 2.22E+09 1.IlE-01 1.53E+09 2.38E+09 2.79E+09 2.47E+09 6.74E-02 1.03E+09 1.63E+09 1.91E+09 1.65E+09 4.09E-02 7.84E+08 1.25E+09 1.47E+09 1.24E+09 3.18E-02 2.67E+08 4.45E+08 5.20E+08 4.13E+08 2.61E-02 1.74E+08 3.03E+08 3.55E+08 2.68E+08 2.42E-02 4.31E+08 5.76E+08 6.92E+08 7.24E+08 2.19E-02 2.73E+08 4.01E+08 4.83E+08 4.63E+08 1.50E-02 5.04E+08 8.40E+08 1.OOE+09 8.23E+08 7.10E-03 7.15E+08 1.19E+09 1.40E+09 1.11E+09 3.36E-03 7.26E+08 1.18E+09 1.40E+09 1.13E+09 1.59E-03 6.34E+08 1.04E+09 1.22E+09 9.78E+08 4.54E-04 9.83E+08 1.58E+09 1.86E+09 1.50E+09 2.14E-04 5.23E+08 8.41E+08 9.91E+08 7.91E+08 1.01E-04 5.24E+08 8.25E+08 9.72E+08 7.89E+08 3.73E-05 6.64E+08 1.04E+09 1.22E+09 9.97E+08 1.07E-05 7.71E+08 1.20E+09 1.41E+09 1.16E+09 5.04E-06 4.29E+08 6.60E+08 7.76E+08 6.39E+08 1.86E-06 5.45E+08 8.33E+08 9.78E+08 8.09E+08 8.76E-07 3.97E+08 6.02E+08 7.05E+08 5.86E+08 4.14E-07 3.55E+08 5.34E+08 6.24E+08 5.20E+08 1.OOE-07 7.91E+08 1.14E+09 1.33E+09 1.15E+09 O.OOE+00 2.76E+09 3.17E+09 3.69E+09 4.05E+09 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-4 Table 4.1-2 Reference Neutron Sensor Reaction Rates and Exposure Parameters at the Cavity Sensor Set Locations - 3411 Mwt; F = 1.20 0.50 14.50 29.50 44.50

'Cu (n,co) (Cd) 7.28E-19 9.14E-19 8.36E-19 6.65E-19

-r'i (n,p) (Cd) 9.92E-18 1.28E-17 1.19E-17 9.48E-18 "4Fe (n,p) (Cd) 5.34E-17 7.23E-17 6.95E-17 5.77E-17 SSNi (n~p) (Cd) 7.42E-17 1.01E-16 9.85E-17 8.30E-17 2U (n,f) (Cd) 2.61E-16 3.68E-16 3.79E-16 3.39E-16

"=7Np (n,f) (Cd) 3.54E-15 5.20E-15 5.79E-15 5.38E-15 "95Co (n,y) 1.53E-13 1.99E-13 2.32E-13 2.26E-13 "95Co (n,y) (Cd) 4.78E-14 7.47E-14 8.79E-14 7.18E-14 2-Np (n,y) 1.35E-17 1.60E-17 1.58E-17 1.40E-17

'37Np (n,y) 3.79E-17 4.50E-17 4.46E-17 3.94E-17 Neutron Flux [n/cm2 -s]

Op(E > 1.0 MeV) 8.65E+08 1.24E+09 1.32E+09 1.22E+09 O(E > 0.1 MeV) 7.72E+09 1.17E+10 1.35E+10 1.23E+10 dpa/sec Displacement Rate 2.76E-12 4.13E-12 4.65E-12 4.21E-12 O(E > 0.1 MeV) / O(E > 1.0 MeV) 8.92 9.44 10.16 10.03

[dpa/sec] / O(E > 1.0 MeV) 3.19E-21 3.33E-21 3.51E-21 3.44E-21 2U (n,y) / 2U (n,f) 0.052 0.043 0.042 0.041 27Np (n,y) / 237Np (n,f) 0.011 0.009 0.008 0.007 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-5 Table 4.1-3 Calculated Reference Neutron Energy Spectra at Surveillance Capsule Locations - 3411 Mwt; F = 1.2 Lower Energy O(E) [n/cm 2-sl Lower Energy O(E) [n/cm 2-sl

[MeVI 31.50 34.00 [MeVI 31.5' 34.00 1.42E+01 1.52E+07 1.58E+07 1.83E-01 7.43E+10 9.OOE+10 1.22E+01 4.92E+07 5.17E+07 1.11E-01 7.62E+10 9.27E+10 1.0OE+01 2.20E+08 2.34E+08 6.74E-02 5.62E+10 6.82E+10 8.61E+00 4.36E+08 4.65E+08 4.09E-02 4.43E+10 5.37E+10 7.41E+00 7.78E+08 8.37E+08 3.18E-02 1.44E+10 1.73E+10 6.07E+00 1.91E+09 2.07E+09 2.61E-02 5.54E+09 6.66E+09 4.97E+00 3.01E+09 3.30E+09 2.42E-02 1.68E+10 2.05E+10 3.68E+00 6.45E+09 7.23E+09 2.19E-02 1.11E+10 1.36E+10 3.01E+00 5.50E+09 6.29E+09 1.50E-02 1.84E+10 2.22E+10 2.73E+00 4.44E+09 5.09E+09 7.10E-03 3.65E+10 4.37E+10 2.47E+00 5.39E+09 6.21E+09 3.35E-03 5.24E+10 6.28E+10 2.37E+00 2.72E+09 3.14E+09 1.58E-03 4.87E+10 5.86E+10 2.35E+00 7.63E+08 8.79E+08 4.54E-04 7.49E+10 9.03E+10 2.23E+00 3.94E+09 4.54E+09 2.14E-04 3.92E+10 4.72E+10 1.92E+00 1.15E+10 1.33E+10 1.01E-04 4.48E+10 5.40E+10 1.65E+00 1.44E+10 1.69E+10 3.73E-05 5.85E+10 7.05E+10 1.35E+00 2.32E+10 2.73E+10 1.07E-05 6.89E+10 8.31E+10 1.OOE+00 5.17E+10 6.16E+10 5.04E-06 3.78E+10 4.55E+10 8.21E-01 3.75E+10 4.50E+10 1.86E-06 4.62E+10 5.55E+10 7.43E-01 1.87E+10 2.24E+10 8.76E-07 3.17E+10 3.79E+10 6.08E-01 6.47E+10 7.85E+10 4.14E-07 2.22E+10 2.65E+10 4.98E-01 5.35E+10 6.51E+10 1.OOE-07 3.72E+10 4.42E+10 3.69E-01 6.13E+10 7.45E+10 0.OOE+00 4.47E+10 5.36E+10 2.97E-01 6.22E+10 7.62E+10 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-6 Table 4.1-4 Reference Neutron Sensor Reaction Rates and Exposure Parameters at the Center of Surveillance Capsules - 3411 Mwt; F = 1.20 31.50 34.00 6Cu (n,a) 6.89E-17 7.43E-17 "UFe (np) 7.94E-15 8.88E-15 58 Ni (n,p) 1.12E-14 1.26E-14

"*U (n,f) (Cd) 4.22E-14 4.86E-14 7

2 Np (n,f) (Cd) 4.08E-13 4.84E-13 59 Co (n,'y) 5.91E-12 7.10E-12

' 9Co (ny) (Cd) 4.05E-12 4.87E-12 2U (n,,y) 1.48E-15 1.74E-15 7

3 Np (ny) 4.12E-15 4.84E-15 Neutron Flux [n/cm 2-s]

O(E > 1.0 MeV) 1.37E+11 1.60E+11 O(E > 0.1 MeV) 5.96E+11 7.18E+11 dpa/sec Displacement Rate 2.61E-10 3.10E-10 O(E > 0.1 MeV) / O(E > 1.0 MeV) 4.36 4.49

[dpa/sec] / O(E > 1.0 MeV) 1.91E-21 1.94E-21 2U (n,y) / 2U (nf) 0.035 0.036 7

23Np (n,7) /

7Np (nf) 0.010 0.010 July 1999 Results of Transport Calculations Neutron Transport of Neutron Calculations July 1999 WCAP-15253, Rev. 0

4-7 Table 4.1-5 Azimuthal Variation of Fast Neutron Flux (E > 1.0 MeV) at the Pressure Vessel Inner Radius Radius (cm) 0 (deg) 220.27 220.35 221.00 0.25 1.79E+10 1.76E+10 1.69E+10 0.75 1.79E+10 1.76E+10 1.69E+10 1.25 1.80E+10 1.77E+10 1.69E+10 1.85 1.81E+10 1.77E+10 1.70E+10 2.35 1.82E+10 1.79E+10 1.71E+10 2.75 1.82E+10 1.79E+10 1.72E+10 3.25 1.84E+10 1.80E+10 1.73E+10 3.75 1.85E+10 1.82E+10 1.74E+10 4.25 1.87E+10 1.84E+10 1.76E+10 4.75 1.89E+10 1.86E+10 1.78E+10 5.25 1.91E+10 1.88E+10 1.80E+10 5.75 1.94E+10 1.90E+10 1.82E+10 6.25 1.96E+10 1.93E+10 1.85E+10 6.75 1.99E+10 1.96E+10 1.88E+10 7.25 2.02E+10 1.99E+10 1.91E+10 7.75 2.06E+10 2.02E+10 1.94E+10 8.25 2.09E+10 2.06E+10 1.97E+10 8.81 2.13E+10 2.10E+10 2.01E+10 9.44 2.18E+10 2.14E+10 2.06E+10 10.00 2.23E+10 2.19E+10 2.10E+10 10.44 2.27E+10 2.23E+10 2.14E+10 10.82 2.31E+10 2.27E+10 2.17E+10 11.25 2.35E+10 2.31E+10 2.21E+10 11.75 2.40E+10 2.35E+10 2.26E+10 12.31 2.45E+10 2.41E+10 2.31E+10 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-8 Table 4.1-5 Azimuthal Variation of Fast Neutron Flux (E > 1.0 MeV) at the Pressure (cont.) Vessel Inner Radius Radius (cm) e (deg) 220.27 220.35 221.00 12.94 2.52E+10 2.47E+10 2.37E+10 13.50 2.58E+10 2.53E+10 2.43E+10 13.94 2.63E+10 2.58E+10 2.47E+10 14.32 2.67E+10 2.62E+10 2.51E+10 14.75 2.71E+10 2.67E+10 2.56E+10 15.25 2.77E+10 2.72E+10 2.61E+10 15.75 2.82E+10 2.77E+10 2.65E+10 16.25 2.87E+10 2.82E+10 2.70E+10 16.75 2.92E+10 2.87E+10 2.75E+10 17.25 2.97E+10 2.92E+10 2.80E+10 17.75 3.02E+10 2.97E+10 2.84E+10 18.25 3.06E+10 3.01E+10 2.88E+10 18.75 3.11E+10 3.05E+10 2.92E+10 19.25 3.14E+10 3.09E+10 2.96E+10 19.75 3.17E+10 3.12E+10 2.98E+10 20.25 3.20E+10 3.14E+10 3.01E+10 20.75 3.21E+10 3.15E+10 3.02E+10 21.25 3.21E+10 3.16E+10 3.02E+10 21.75 3.21E+10 3.15E+10 3.01E+10 22.25 3.19E+10 3.13E+10 3.OOE+10 22.75 3.17E+10 3.12E+10 2.98E+10 23.25 3.16E+10 3.10E+10 2.97E+10 23.75 3.14E+10 3.08E+10 2.95E+10 24.25 3.13E+10 3.07E+10 2.94E+10 24.75 3.12E+10 3.06E+10 2.93E+10 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-9 Table 4.1-5 Azimuthal Variation of Fast Neutron Flux (E > 1.0 MeV) at the Pressure (cont.) Vessel Inner Radius Radius (cm) e (deg) 220.27 220.35 221.00 25.25 3.11E+10 3.05E+10 2.92E+10 25.75 3.09E+10 3.03E+10 2.90E+10 26.25 3.06E+10 3.01E+10 2.88E+10 26.75 3.03E+10 S2.98E+10 2.85E+10 27.25 2.99E+10 2.94E+10 2.82E+10 27.75 2.94E+10 2.89E+10 2.77E+10 28.25 2.88E+10 2.83E+10 2.71E+10 28.75 2.81E+10 2.76E+10 2.64E+10 29.25 2.73E+10 2.69E+10 2.57E+10 29.75 2.68E+10 2.63E+10 2.52E+10 30.25 2.63E+10 2.59E+10 2.48E+10 30.56 2.64E+10 2.59E+10 2.46E+10 30.69 2.63E+10 2.58E+10 2.45E+10 30.84 2.61E+10 2.56E+10 2.44E+10 31.11 2.55E+10 2.51E+10 2.41E+10 31.50 2.51E+10 2.47E+10 2.38E+10 31.89 2.48E+10 2.44E+10 2.34E+10 32.23 2.47E+10 2.43E+10 2.33E+10 32.56 2.47E+10 2.42E+10 2.33E+10 32.94 2.47E+10 2.43E+10 2.34E+10 33.27 2.49E+10 2.45E+10 2.35E+10 33.62 2.51E+10 2.46E+10 2.36E+10 34.00 2.53E+10 2.49E+10 2.39E+10 34.39 2.56E+10 2.52E+10 2.41E+10 34.67 2.60E+10 2.55E+10 2.43E+10 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-10 Table 4.1-5 Azimuthal Variation of Fast Neutron Flux (E > 1.0 MeV) at the Pressure (cont.) Vessel Inner Radius Radius (cm) 220.35 221.00 e0(deg) 220.27 2.62E+10 2.56E+10 2.44E+10 34.82 2.63E+10 2.57E+10 2.45E+10 34.94 2.63E+10 2.58E+10 2.46E+10 35.10 2.64E+10 2.59E+10 2.48E+10 35.35 2.67E+10 2.62E+10 2.51E+10 35.75 2.71E+10 2.66E+10 2.55E+10 36.25 2.76E+10 2.71E+10 2.59E+10 36.75 2.80E+10 2.75E+10 2.63E+10 37.25 2.84E+10 2.78E+10 2.66E+10 37.75 2.87E+10 2.82E+10 2.70E+10 38.25 2.90E+10 2.85E+10 2.73E+10 38.75 2.94E+10 2.88E+10 2.76E+10 39.25 2.97E+10 2.92E+10 2.79E+10 39.75 3.OOE+10 2.95E+10 2.82E+10 40.25 3.03E+10 2.98E+10 2.85E+10 40.75 3.06E+10 3.OOE+10 2.87E+10 41.25 3.09E+10 3.03E+10 2.90E+10 41.75 3.11E+10 3.06E+10 2.92E+10 42.25 3.13E+10 3.08E+10 2.94E+10 42.65 3.15E+10 3.09E+10 2.96E+10 43.10 3.16E+10 3.10E+10 2.97E+10 43.70 3.17E+10 3.12E+10 2.98E+10 44.25 3.18E+10 3.12E+10 2.99E+10 44.75 July i99i Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-11 Table 4.1-6 Summary of Exposure Rates at the Pressure Vessel Clad/Base Metal Interface Neutron Flux [nrcm 2-sec]

E > 0.1 MeV dpa/sec 0 (deg) (E > 1.0 MeV) (E > 0.1 MeV) dpa/sec E > 1.0 MeV E > 1.0 MeV 0.25 1.76E+10 3.71E+10 2.73E-11 2.10 1.55E-21 5.25 1.88E+10 3.97E+10 2.91E-11 2.11 1.55E-21 10.00 2.19E+10 4.65E+10 3.38E-11 2.12 1.54E-21 15.25 2.72E+10 5.79E+10 4.17E-11 2.13 1.53E-21 20.25 3.14E+10 6.70E+10 4.80E-11 2.13 1.53E-21 25.25 3.05E+10 6.59E+10 4.67E-11 2.16 1.53E-21 30.25 2.59E+10 5.96E+10 4.03E-11 2.30 1.56E-21 35.10 2.58E+10 6.43E+10 4.08E-11 2.49 1.58E-21 40.25 2.95E+10 7.50E+10 4.68E-11 2.54 1.59E-21 44.75 3.12E+10 7.88E+10 4.94E-11 2.53 1.58E-21 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-12 Table 4.1-7 Relative Radial Distribution of Neutron Flux (E > 1.0 MeV) within the Pressure Vessel Wall Radius Azimuthal Angle

[cml 0.0° 15.00 30.00 45.00 1.000 1.000 1.000 220.35 1.000 0.958 0.960 0.957 221.00 0.959 0.850 0.851 0.846 222.30 0.852 0.736 0.737 0.729 223.60 0.739 0.630 0.632 0.622 224.89 0.634 0.557 0.559 0.547 225.87 0.561 0.481 0.484 0.471 227.01 0.486 0.390 0.392 0.380 228.63 0.395 0.320 0.323 0.311 230.09 0.325 0.269 0.271 0.259 231.39 0.273 0.225 0.227 0.216 232.68 0.229 0.183 0.186 0.175 234.14 0.187 0.146 0.148 0.139 235.76 0.149 0.124 0.126 0.117 236.90 0.127 0.107 0.109 0.101 237.88 0.110 0.088 0.090 0.082 239.18 0.091 0.074 0.067 240.47 0.074 0.072 0.058 0.060 0.053 241.77 0.061 0.055 0.057 0.050 242.42 0.058 Note:

Base Metal Inner Radius = 220.35 cm.

Base Metal 11/44T = 225.87 cm.

Base Metal 1/2,T = 231.39 cm.

Base Metal 33/44T = 236.90 cm.

Base Metal Outer Radius = 242.42 cm.

juiy 1' Results of Neutron Transport Calculations JulyRe. 0 WCAP-15253, Rev. 0

4-13 Table 4.1-8 Relative Radial Distribution of Neutron Flux (E > 0.1 MeV) within the Pressure Vessel Wall Radius Azimuthal Angle

[cml 0.00 15.0° 30.00 45.00 220.35 1.000 1.000 1.000 1.000 221.00 1.014 1.012 1.015 1.009 222.30 1.002 0.996 1.002 0.988 223.60 0.966 0.957 0.965 0.943 224.89 0.920 0.908 0.918 0.890 225.87 0.882 0.868 0.880 0.848 227.01 0.835 0.820 0.833 0.797 228.63 0.768 0.752 0.767 0.726 230.09 0.708 0.691 0.707 0.663 231.39 0.654 0.637 0.654 0.608 232.68 0.602 0.585 0.602 0.554 234.14 0.544 0.528 0.545 0.496 235.76 0.481 0.467 0.484 0.434 236.90 0.438 0.425 0.442 0.392 237.88 0.401 0.389 0.405 0.356 239.18 0.353 0.343 0.359 0.309 240.47 0.307 0.298 0.313 0.263 241.77 0.262 0.250 0.264 0.216 242.42 0.253 0.240 0.254 0.206 Note:

Base Metal Inner Radius = 220.35 cm.

Base Metal 1AT = 225.87 cm.

Base Metal 1/2T = 231.39 cm.

Base Metal 3/4T = 236.90 cm.

Base Metal Outer Radius = 242.42 cm.

Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-14 Table 4.1-9 Relative Radial Distribution of Iron Displacement Rate (dpa) within the Pressure Vessel Wall Radius Azimuthal Angle

[cm] 0.00 15.0° 30.00 45.0' 220.35 1.000 1.000 1.000 1.000 221.00 0.965 0.965 0.968 0.965 222.30 0.877 0.876 0.882 0.878 223.60 0.785 0.783 0.793 0.787 224.89 0.699 0.696 0.708 0.702 225.87 0.638 0.635 0.649 0.642 227.01 0.575 0.571 0.587 0.579 228.63 0.495 0.491 0.508 0.499 230.09 0.432 0.428 0.446 0.436 231.39 0.383 0.378 0.397 0.386 232.68 0.339 0.334 0.352 0.341 234.14 0.295 0.291 0.308 0.296 235.76 0.252 0.248 0.264 0.251 236.90 0.224 0.221 0.237 0.223 237.88 0.202 0.199 0.214 0.199 239.18 0.175 0.172 0.186 0.171 240.47 0.150 0.147 0.160 0.144 241.77 0.127 0.123 0.135 0.118 242.42 0.123 0.118 0.130 0.113 Note:

Base Metal Inner Radius = 220.35 cm.

Base Metal 'AT = 225.87 cm.

Base Metal 1/2T = 231.39 cm.

Base Metal 33/4T = 236.90 cm.

Base Metal Outer Radius = 242.42 cm.

Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-15 4.2 FUEL CYCLE SPECIFIC ADJOINT CALCULATIONS Results of the fuel cycle specific adjoint transport calculations for the first 12 cycles of operation at McGuire Unit 1 are summarized in Tables 4.2-1 through 4.2-18. The data listed in these tables establish the means for absolute comparison of analysis and measurement for the Cycle 12 cavity dosimetry irradiation as well as for the five sets of surveillance capsule dosimetry withdrawn to date. These results also provide the fuel cycle specific relationship among the surveillance capsule and reactor cavity measurement locations and key positions at the inner radius of the pressure vessel wall.

The core power distributions used in the cycle specific fast neutron exposure calculations for Cycles 1 through 12 were taken from the fuel cycle design reports applicable to McGuire Unit 1.13 311 The data extracted from the fuel cycle design reports represented cycle averaged relative fuel assembly powers and bumups as well as cycle averaged relative axial distributions. Therefore, the results of the adjoint evaluation provided data in terms of fuel cycle averaged neutron flux which, when multiplied by the appropriate fuel cycle length, produced the incremental fast neutron exposure for the fuel cycle.

The calculated fast neutron flux (E > 1.0 MeV) and cumulative fast neutron fluence at the center of surveillance capsules located at 31.5 and 34.0 degrees are provided for each of the twelve operating fuel cycles in Tables 4.2-1 and 4.2-2, respectively. The data as tabulated are applicable to the axial core midplane. Similar data applicable to the pressure vessel inner radius are given in Tables 4.2-3 and 4.2-4 and data pertinent to the cavity dosimetry sensor locations are listed in Tables 4.2-5 and 4.2-6.

Exposure parameter ratios necessary to convert the cycle specific data listed in Tables 4.2-1 through 4.2-6 to other key fast neutron exposure units are given in Section 4.1 of this report.

Application of these ratios to the data from Tables 4.2-1 through 4.2-6 yielded corresponding exposure data in terms of flux/fluence (E > 0.1 MeV) (Tables 4.2-7 through 4.2-12) and iron atom displacements (Tables 4.2-13 through 4.2-18).

Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-16 Table 4.2-1 Calculated Fast Neutron Flux (E > 1.0 MeV) at the Center of Reactor Vessel Surveillance Capsules 2

Neutron Flux [n/cm -secl 340 Cycle No. 31.50 1 1.01E+11 1.18E+11 2 1.26E+11 1.45E+11 3 9.16E+10 1.05E+11 4 8.43E+10 9.69E+10 5 8.01E+10 9.04E+10 6 8.52E+10 9.66E+10 7 8.34E+10 9.53E+10 8 8.30E+10 9.39E+10 9 8.60E+10 9.86E+10 10 8.35E+10 9.61E+10 11 7.94E+10 8.93E+10 12 7.69E+10 8.71E+10 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-17 Table 4.2-2 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Center of Reactor Vessel Surveillance Capsules Cycle Length Cumulative Fluence [n/cm 2]

Cycle EFPS 31.50 340 1 3.437E+07 3.46E+18 4.05E+18 2 2.321E+07 6.37E+18 7.41E+18 3 2.492E+07 8.66E+18 1.OOE+19 4 2.592E+07 1.08E+19 1.25E+19 5 2.733E+07 1.30E+19 1.50E+19 6 2.575E+07 1.52E+19 1.75E+19 7 3.526E+07 1.82E+19 2.08E+19 8 3.176E+07 2.08E+19 2.38E+19 9 2.857E+07 2.33E+19 2.66E+19 10 3.396E+07 2.61E+19 2.99E+19 11 3.119E+07 2.86E+19 3.27E+19 12 3.108E+07 3.10E+19 3.54E+19 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-18 Table 4.2-3 Calculated Fast Neutron Flux (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface Neutron Flux [nlcm 2-sec]

Cycle 00 150 300 450 1 1.28E+10 1.92E+10 1.92E+10 2.34E+10 2 1.67E+10 2.54E+10 2.43E+10 2.79E+10 3 1.28E+10 1.86E+10 1.79E+10 1.99E+10 4 1.21E+10 1.83E+10 1.66E+10 1.92E+10 5 1.30E+10 1.81E+10 1.58E+10 1.71E+10 6 1.20E+10 1.82E+10 1.68E+10 1.84E+10 7 1.13E+10 1.65E+10 1.62E+10 1.84E+10 8 1.20E+10 1.81E+10 1.65E+10 1.79E+10 9 1.11E+10 1.71E+10 1.68E+10 1.91E+10 10 1.16E+10 1.61E+10 1.61E+10 1.86E+10 11 1.16E+10 1.64E+10 1.57E+10 1.67E+10 12 9.37E+09 1.47E+10 1.50E+10 1.66E+10 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-19 Table 4.2-4 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface CYCLE Cumulative Fluence [n/cm 2]

LENGTH Cycle EFPS 00 150 300 450 1 3.437E+07 4.39E+17 6.61E+17 6.60E+17 8.03E+17 2 2.321E+07 8.27E+17 1.25E+18 1.22E+18 1.45E+18 3 2.492E+07 1.15E+18 1.72E+18 1.67E+18 1.95E+18 4 2.592E+07 1.46E+18 2.15E+18 2.10E+18 2.45E+18 5 2.733E+07 1.82E+18 2.68E+18 2.53E+18 2.91E+18 6 2.575E+07 2.13E+18 3.15E+18 2.96E+18 3.38E+18 7 3.526E+07 2.52E+18 3.73E+18 3.54E+18 4.03E+18 8 3.176E+07 2.91E+18 4.31E+18 4.06E+18 4.60E+18 9 2.857E+07 3.22E+18 4.80E+18 4.54E+18 5.15E+18 10 3.396E+07 3.62E+18 5.34E+18 5.08E+18 5.78E+18 11 3.119E+07 3.98E+18 5.85E+18 5.57E+18 6.30E+18 12 3.108E+07 4.27E+18 6.31E+18 6.04E+18 6.82E+18 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-20 Table 4.2-5 Calculated Fast Neutron Flux (E > 1.0 MeV) at the Cavity Sensor Set Locations Neutron Flux [n/cm 2-sec]

Cycle 0.50 14.50 29.50 44.50 1 6.17E+08 8.81E+08 9.47E+08 8.93E+08 2 8.07E+08 1.15E+09 1.20E+09 1.08E+09 3 6.12E+08 8.53E+08 8.78E+08 7.77E+08 4 5.85E+08 8.27E+08 8.27E+08 7.43E+08 5 6.15E+08 8.25E+08 7.86E+08 6.73E+08 6 5.82E+08 8.26E+08 8.30E+08 7.21E+08 7 5.39E+08 7.57E+08 7-96E+08 7.14E+08 8 5.81E+08 8.19E+08 8.16E+08 7.03E+08 9 5.41E+08 7.81E+08 8.23E+08 7.40E+08 10 5.44E+08 7.45E+08 7.88E+08 7.21E+08 11 5.47E+08 7.53E+08 7.68E+08 6.59E+08 12 4.61E+08 6.74E+08 7.28E+08 6.50E+08 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-21 Table 4.2-6 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Cavity Sensor Set Locations CYCLE Cumulative Fluence [n/cm 2]

LENGTH Cycle EFPS 0.50 14.50 29.50 44.50 1 3.437E+07 2.12E+16 3.03E+16 3.25E+16 3.07E+16 2 2.321E+07 3.99E+16 5.71E+16 6.04E+16 5.58E+16 3 2.492E+07 5.52E+16 7.83E+16 8.23E+16 7.51E+16 4 2.592E+07 7.04E+16 9.98E+16 1.04E+17 9.44E+16 5 2.733E+07 8.72E+16 1.22E+17 1.25E+17 1.13E+17 6 2.575E+07 1.02E+17 1.44E+17 1.47E+17 1.31E+17 7 3.526E+07 1.21E+17 1.70E+17 1.75E+17 1.56E+17 8 3.176E+07 1.40E+17 1.96E+17 2.01E+17 1.79E+17 9 2.857E+07 1.55E+17 2.19E+17 2.24E+17 2.OOE+17 10 3.396E+07 1.74E+17 2.44E+17 2.51E+17 2.24E+17 11 3.119E+07 1.91E+17 2.67E+17 2.75E+17 2.45E+17 12 3.108E+07 2.05E+17 2.88E+17 2.97E+17 2.65E+17 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-22 Table 4.2-7 Calculated Fast Neutron Flux (E > 0.1 MeV) at the Center of Reactor Vessel Surveillance Capsules Neutron Flux [n/cm 2-sec]

31.50 340 Cycle No.

1 4.39E+11 5.29E+11 2 5.47E+11 6.51E+11 3 3.99E+11 4.69E+11 4 3.68E+11 4.35E+11 5 3.49E+11 4.06E+11 6 3.72E+11 4.34E+11 7 3.64E+11 4.28E+11 8 3.62E+11 4.22E+11 9 3.75E+11 4.43E+11 10 3.64E+11 4.31E+11 11 3.46E+11 4.01E+11 12 3.35E+11 3.91E+11 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-23 Table 4.2-8 Calculated Fast Neutron Fluence (E > 0.1 MeV) at the Center of Reactor Vessel Surveillance Capsules Cycle Length Cumulative Fluence [n/cm2 ]

Cycle EFPS 31.50 340 1 3.437E+07 1.51E+19 1.82E+19 2 2.321E+07 2.78E+19 3.33E+19 3 2.492E+07 3.77E+19 4.50E+19 4 2.592E+07 4.73E+19 5.63E+19 5 2.733E+07 5.68E+19 6.73E+19 6 2.575E+07 6.64E+19 7.85E+19 7 3.526E+07 7.92E+19 9.36E+19 8 3.176E+07 9.07E+19 1.07E+20 9 2.857E+07 1.01E+20 1.20E+20 10 3.396E+07 1.14E+20 1.34E+20 11 3.119E+07 1.25E+20 1.47E+20 12 3.108E+07 1.35E+20 1.59E+20 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-24 Table 4.2-9 Calculated Fast Neutron Flux (E > 0.1 MeV) at the Pressure Vessel Clad/Base Metal Interface Neutron Flux [n/cm 2-secl Cycle 00 150 300 450 2.69E+10 4.10E+10 4.42E+10 5.90E+10 1

3.52E+10 5.41E+10 5.59E+10 7.05E+10 2

2.70E+10 3.97E+10 4.11E+10 5.02E+10 3

2.55E+10 3.89E+10 3.82E+10 4.86E+10 4

2.75E+10 3.85E+10 3.64E+10 4.31E+10 5

2.53E+10 3.88E+10 3.87E+10 4.64E+10 6

2.37E+10 3.51E+10 3.74E+10 4.64E+10 7

2.53E+10 3.85E+10 3.79E+10 4.53E+10 8

2.33E+10 3.65E+10 3.86E+10 4.82E+10 9

2.43E+10 3.42E+10 3.71E+10 4.71E+10 10 2.43E+10 3.49E+10 3.61E+10 4.22E+10 11 1.97E+10 3.14E+10 3.45E+10 4.20E+10 12 juiy Results of Neutron Transport Calculations WCAP-15253, Rev. 0

4-25 Table 4.2-10 Calculated Fast Neutron Fluence (E > 0.1 MeV) at the Pressure Vessel Clad/Base Metal Interface CYCLE Cumulative Fluence In/cm']

LENGTH Cycle EFPS 00 150 300 450 1 3.437E+07 9.25E+17 1.41E+18 1.52E+18 2.03E+18 2 2.321E+07 1.74E+18 2.66E+18 2.82E+18 3.66E+18 3 2.492E+07 2.41E+18 3.65E+18 3.84E+18 4.92E+18 4 2.592E+07 3.07E+18 4.66E+18 4.83E+18 6.18E+18 5 2.733E+07 3.83E+18 5.71E+18 5.83E+18 7.35E+18 6 2.575E+07 4.48E+18 6.71E+18 6.82E+18 8.55E+18 7 3.526E+07 5.31E+18 7.95E+18 8.14E+18 1.02E+19 8 3.176E+07 6.12E+18 9.17E+18 9.34E+18 1.16E+19 9 2.857E+07 6.78E+18 1.02E+19 1.04E+19 1.30E+19 10 3.396E+07 7.61E+18 1.14E+19 1.17E+19 1.46E+19 11 3.119E+07 8.37E+18 1.25E+19 1.28E+19 1.59E+19 12 3.108E+07 8.98E+18 1.34E+19 1.39E+19 1.72E+19 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-26 Table 4.2-11 Calculated Fast Neutron Flux (E > 0.1 MeV) at the Cavity Sensor Set Locations Neutron Flux [n/cm 2-sec]

Cycle 0.50 14.50 29.50 44.50 1 5.50E+09 8.31E+09 9.62E+09 8.95E+09 2 7.21E+09 1.09E+10 1.22E+10 1.08E+10 3 5.46E+09 8.05E+09 8.92E+09 7.79E+09 4 5.22E+09 7.80E+09 8.40E+09 7.45E+09 5 5.49E+09 7.78E+09 7.99E+09 6.74E+09 6 5.19E+09 7.80E+09 8.44E+09 7.22E+09 7 4.81E+09 7.15E+09 8.08E+09 7.16E+09 8 5.18E+09 7.73E+09 8.29E+09 7.05E+09 9 4.83E+09 7.37E+09 8.36E+09 7.42E+09 10 4.86E+09 7.03E+09 8.01E+09 7.23E+09 11 4.88E+09 7.11E+09 7.80E+09 6.61E+09 12 4.11E+09 6.36E+09 7.39E+09 6.51E+09 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-27 Table 4.2-12 Calculated Fast Neutron Fluence (E > 0.1 MeV) at the Cavity Sensor Set Locations CYCLE Cumulative Fluence In/cm2 ]

LENGTH Cycle EFPS 0.50 14.50 29.50 44.50 1 3.437E+07 1.89E+17 2.86E+17 3.31E+17 3.08E+17 2 2.321E+07 3.56E+17 5.39E+17 6.14E+17 5.59E+17 3 2.492E+07 4.93E+17 7.39E+17 8.36E+17 7.53E+17 4 2.592E+07 6.28E+17 9.41E+17 1.05E+18 9.46E+17 5 2.733E+07 7.78E+17 1.15E+18 1.27E+18 1.13E+18 6 2.575E+07 9.12E+17 1.36E+18 1.49E+18 1.32E+18 7 3.526E+07 1.08E+18 1.61E+18 1.77E+18 1.57E+18 8 3.176E+07 1.25E+18 1.85E+18 2.04E+18 1.79E+18 9 2.857E+07 1.38E+18 2.06E+18 2.28E+18 2.OOE+18 10 3.396E+07 1.55E+18 2.30E+18 2.55E+18 2.25E+18 11 3.119E+07 1.70E+18 2.52E+18 2.79E+18 2.46E+18 12 3.108E+07 1.83E+18 2.72E+18 3.02E+18 2.66E+18 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-28 Table 4.2-13 Calculated Iron Displacement Rate at the Center of Reactor Vessel Surveillance Capsules Displacement Rate [dpa/secl 340 Cycle No. 31.50 1 1.92E-10 2.28E-10 2 2.39E-10 2.81E-10 3 1.75E-10 2.02E-10 4 1.61E-10 1.88E-10 5 1.53E-10 1.75E-10 6 1.63E-10 1.87E-10 7 1.59E-10 1.85E-10 8 1.58E-10 1.82E-10 9 1.64E-10 1.91E-10 10 1.59E-10 1.86E-10 11 1.52E-10 1.73E-10 12 1.47E-10 1.69E-10 July 1999 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-29 Table 4.2-14 Calculated Iron Displacements at the Center of Reactor Vessel Surveillance Capsules Cycle Length Cumulative Displacements [dpal CQycle EFPS 31.50 340 1 3.437E+07 6.60E-03 7.83E-03 2 2.321E+07 1.22E-02 1.43E-02 3 2.492E+07 1.65E-02 1.94E-02 4 2.592E+07 2.07E-02 2.42E-02 5 2.733E+07 2.49E-02 2.90E-02 6 2.575E+07 2.90E-02 3.38E-02 7 3.526E+07 3.46E-02 4.03E-02 8 3.176E+07 3.97E-02 4.61E-02 9 2.857E+07 4.44E-02 5.16E-02 10 3.396E+07 4.98E-02 5.79E-02 11 3.119E+07 5.45E-02 6.33E-02 12 3.108E+07 5.91E-02 6.85E-02 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-30 Table 4.2-15 Calculated Iron Displacement Rate at the Pressure Vessel Clad/Base Metal Interface Displacement Rate [dpalsecl Cycle 00 150 300 450 1 1.98E-11 2.95E-11 2.99E-11 3.70E-1I 2 2.59E-11 3.90E-11 3.78E-11 4.42E-11 3 1.99E-11 2.86E-11 2.78E-11 3.15E-11 4 1.88E-11 2.80E-11 2.58E-11 3.04E-11 5 2.02E-11 2.78E-11 2.46E-11 2.70E-11 6 1.86E-11 2.80E-11 2.61E-11 2.91E-11 7 1.75E-11 2.53E-11 2.53E-11 2.91E-11 8 1.86E-11 2.77E-11 2.56E-11 2.84E-11 9 1.72E-11 2.63E-11 2.61E-11 3.02E-11 10 1.79E-11 2.46E-11 2.51E-11 2.95E-11 11 1.79E-11 2.51E-11 2.44E-11 2.64E-11 12 1.45E-11 2.26E-11 2.33E-11 2.63E-11 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-31 Table 4.2-16 Calculated Iron Displacements at the Pressure Vessel Clad/Base Metal Interface CYCLE Cumulative Displacements [dpal LENGTH Cycle EFPS 00 150 300 450 1 3.437E+07 6.80E-04 1.01E-03 1.03E-03 1.27E-03 2 2.321E+07 1.28E-03 1.92E-03 1.91E-03 2.30E-03 3 2.492E+07 1.78E-03 2.63E-03 2.60E-03 3.08E-03 4 2.592E+07 2.26E-03 3.36E-03 3.27E-03 3.87E-03 5 2.733E+07 2.81E-03 4.12E-03 3.94E-03 4.61E-03 6 2.575E+07 3.29E-03 4.84E-03 4.61E-03 5.36E-03 7 3.526E+07 3.91E-03 5.73E-03 5.50E-03 6.38E-03 8 3.176E+07 4.50E-03 6.61E-03 6.32E-03 7.28E-03 9 2.857E+07 4.99E-03 7.36E-03 7.06E-03 8.14E-03 10 3.396E+07 5.60E-03 8.20E-03 7.91E-03 9.15E-03 11 3.119E+07 6.16E-03 8.98E-03 8.68E-03 9.97E-03 12 3.108E+07 6.61E-03 9.68E-03 9.40E-03 1.08E-02 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-32 Table 4.2-17 Calculated Iron Displacement Rate at the Cavity Sensor Set Locations Displacement Rate Idpalsec]

Cycle 0.50 14.50 29.50 44.50 1 1.97E-12 2.93E-12 3.32E-12 3.07E-12 2 2.58E-12 3.84E-12 4.22E-12 3.72E-12 3 1.95E-12 2.84E-12 3.08E-12 2.68E-12 4 1.87E-12 2.75E-12 2.90E-12 2.56E-12 5 1.96E-12 2.74E-12 2.76E-12 2.32E-12 6 1.86E-12 2.75E-12 2.92E-12 2.48E-12 7 1.72E-12 2.52E-12 2.79E-12 2.46E-12 8 1.85E-12 2.73E-12 2.86E-12 2.42E-12 9 1.73E-12 2.60E-12 2.89E-12 2.55E-12 10 1.74E-12 2.48E-12 2.77E-12 2.48E-12 11 1.74E-12 2.51E-12 2.70E-12 2.27E-12 12 1.47E-12 2.24E-12 2.55E-12 2.24E-12 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

4-33 Table 4.2-18 Calculated Iron Displacements at the Cavity Sensor Set Locations CYCLE Cumulative Displacements Idpal LENGTH C~ycle EFPS 0.50 14.50 29.50 44.50 1 3.437E+07 6.76E-05 1.01E-04 1.14E-04 1.06E-04 2 2.321E+07 1.27E-04 1.90E-04 2.12E-04 1.92E-04 3 2.492E+07 1.76E-04 2.61E-04 2.89E-04 2.59E-04 4 2.592E+07 2.25E-04 3.32E-04 3.64E-04 3.25E-04 5 2.733E+07 2.78E-04 4.07E-04 4.40E-04 3.88E-04 6 2.575E+07 3.26E-04 4.78E-04 5.15E-04 4.52E-04 7 3.526E+07 3.87E-04 5.66E-04 6.13E-04 5.39E-04 8 3.176E+07 4.45E-04 6.53E-04 7.04E-04 6.16E-04 9 2.857E+07 4.95E-04 7.27E-04 7.87E-04 6.89E-04 10 3.396E+07 5.54E-04 8.11E-04 8.81E-04 7.73E-04 11 3.119E+07 6.08E-04 8.89E-04 9.65E-04 8.44E-04 12 3.108E+07 6.54E-04 9.59E-04 1.04E-03 9.14E-04 Results of Neutron Transport Calculations July 1999 WCAP-15253, Rev. 0

5-1 5 EVALUATION OF SURVEILLANCE CAPSULE DOSIMETRY In this section, the results of the evaluations of the three neutron sensor sets withdrawn as a part of the McGuire Unit 1 Reactor Vessel Materials Surveillance Program are presented. The capsule designation, location within the reactor, and time of withdrawal of each of these dosimetry sets were as follows:

Azimuthal Withdrawal Irradiation Capsule ID Location Time Time (EFPS)

U 34.00 End Of Cycle 1 3.437E+07 X 34.00 End Of Cycle 5 1.358E+08 V 31.50 End Of Cycle 8 2.285E+08 Z 34.0° End Of Cycle 8 2.285E+08 Y 31.50 End Of Cycle 11 3.222E+08 5.1 MEASURED REACTION RATES The radiometric counting of each of these capsule dosimetry data sets was accomplished by Westinghouse for Capsules U, X, V, and Z, and by Antech, Ltd. for Capsule Y using the procedures discussed in Section 3.0 of this report. The measured specific activities are included in Appendix A to this report.

The irradiation history of the McGuire Unit 1 reactor during the first eleven fuel cycles of operation is also listed in Appendix A. The irradiation history was obtained from NUREG-0020, "Licensed Operating Reactors Status Summary Report" and from plant personnel"' for the applicable operating periods Based on the irradiation history, the individual sensor characteristics, and the measured specific activities, reaction rates averaged over the appropriate irradiation periods and referenced to a core power level of 3411 MWt were computed for the sensor set removed from Capsules U, X, V, Z, and Y The computed reaction rates for the multiple foil sensor sets from Capsules U, X, V, Z, and Y are provided in Table 5.1-1.

In regard to the data listed in Table 5.1-1, the fission rate measurements for the 'U sensors include corrections for 2U impurities, the build-in of plutonium isotopes during the long irradiations, and for the effects of yf reactions. Likewise, the fission rate measurements for the

`7Np sensors include adjustments for yf reactions occurring over the course of the respective irradiation periods.

Evaluation of Surveillance Capsule Dosimetry July 1999 WCAP-15253, Rev. 0

5-2 5.2 RESULTS OF THE LEAST SQUARES ADJUSTMENT PROCEDURE The results of the application of the least squares adjustment procedure to the five sets of surveillance capsule dosimetry are provided in Tables 5.2-1 through 5.2-5. In these tables, the best estimate exposure experienced by the capsule along with data illustrating the fit of both the trial and best estimate spectra to the measurements are given. Also included in the tabulations are the lo uncertainties associated with each of the derived exposure rates.

In regard to the comparisons listed in Tables 5.2-1 through 5.2-5, it should be noted that the columns labeled "Calculated" were obtained by normalizing the neutron spectral data from Table 4.1-3 to the absolute calculated neutron flux (E > 1.0 MeV) averaged over the applicable irradiation periods (Cycle I for Capsule U, Cycles I through 5 for Capsule X, Cycles 1 through 8 for Capsules V and Z, and Cycles 1 though 12 for Capsule Y) as discussed in Section 3.0. Thus, the comparisons illustrated in Tables 5.2-1 through 5.2-5 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 7.0 of this report.

JUj Evaluation of Surveillance Capsule Dosimetry WCAP-15253, Rev. 0

5-3 Table 5.1-1 Sum mary of Reaction Rates Derived from Multiple Foil Sensor Sets Witt Ldrawn From Internal Surveillance Capsules Reaction Rate [rps/nucleus]

Reaction Capsule U Capsule X Capsule V Capsule Z Capsule Y

"*Cu (nc) 6°Co 6.02E-17 5.32E-17 4.96E-17 5.17E-17 4.31E-17 "Fe (np)SdMn 6.39E-15 5.24E-15 4.55E-15 4.83E-15 4.41E-15 "SNi (n~p) mCo 9.15E-15 7.45E-15 6.43E-15 7.03E-15 6.14E-15 23U (n,f) '37Cs (Cd) 4.10E-14 3.59E-14 2.95E-14 3.33E-14 3.24E-14

'37Np (n,f) '37Cs (Cd) 4.31E-13 2.70E-13 1.35E-13 2.86E-13 1.88E-13

-9Co (n,Y)6Co 6.39E-12 5.30E-12 3.95E-12 4.50E-12 3.99E-12 59 "Co (n,y) 6Co (Cd) 3.60E-12 2.94E-12 2.31E-12 2.58E-12 2.26E-12 Note: The ' 7Np (nf)137Cs (Cd) measurement for Capsule V was not used in the least squares evaluation because it fell outside the +/-3o range of the 4-loop neutron pad reactor normalized reaction rate database for 31.50 capsules documented in WCAP-14044, "Westinghouse Surveillance Capsule Neutron Fluence Reevaluation," April, 1994.

Dosimetry July 1999 Evaluation of Evaluation Surveillance Capsule of Surveillance Capsule Dosimetry July 1999 WCAP-15253, Rev. 0

5-4 Table 5.2.1 Best Estimate Exposure Rates from Surveillance Capsule U Dosimetry Withdrawn at the End of Fuel Cycle 1 Reaction Rate (rps/nucleus)

Best Reaction Measured Calculated Estimate BE / Meas BE I Calc Meas / Calc 5.77E-17 0.96 1.05 1.10

'Cu (na) 6.02E-17 5.47E-17 6.53E-15 6.68E-15 1.05 1.02 0.98

'Fe (n,p) 6.39E-15 9.24E-15 9.46E-15 1.03 1.02 0.99

'Ni (n,p) 9.15E-15 3.77E-14 0.92 1.03 1.12

'U (n,f) (Cd) 4.10E-14 3.67E-14 0.94 1.10 1.17 "27Np (n,f) (Cd) 4.31E-13 3.67E-13 4.05E-13 59 6.30E-12 0.99 1.21 1.22

" Co (n,y) 6.39E-12 5.22E-12 5 3.65E-12 1.01 0.99 0.98 "Co (n,y) (Cd) 3.60E-12 3.67E-12 Best 1(y Calculated Estimate BE / Calc Uncertainty Exposure Rate 1.22E+11 1.04 6%

O(E > 1.0 MeV) [n/cm 2-sec] 1.18E+11 5.29E+11 5.72E+11 1.08 10%

O(E > 0.1 MeV) [n/cm2-sec]

1.10E+11 1.55 22%

O(E < 0.414 eV) [n/cm 2-sec] 7.1OE+10 2.28E-10 2.43E-10 1.07 7%

dpa/sec Evaluation of Surveillance Capsule Dosimetry July 1999 WCAP-15253, Rev. 0

5-5 Table 5.2-2 Best Estimate Exposure Rates from the Surveillance Capsule X Dosimetry Withdrawn at the End of Fuel Cycle 5 Reaction Rate (rps/nucleus)

Best Reaction Meas ured Calculated Estimate BE / Meas BE ! Calc Meas / Calc

"'Cu (na) 5.321E-17 5.13E-17 5.02E-17 0.94 0.98 1.04 5Fe (np) 5.241E-15 6.13E-15 5.55E-15 1.06 0.91 0.85 "8Ni (np) 7.451E-15 8.67E-15 7.80E-15 1.05 0.90 0.86 2U (n,f) (Cd) 3.591--14 3.44E-14 3.02E-14 0.84 0.88 1.04 2

7 Np (nf) (Cd) 2.701E-13 3.45E-13 2.87E-13 1.06 0.83 0.78 59Co (n,,y) 5.301E-12 4.90E-12 5.22E-12 0.98 1.07 1.08 59Co (n,y) (Cd) 2.941--12 3.45E-12 2.98E-12 1.01 0.86 0.85 Best la Exposure Rate Calculated Estimate BE / Calc Uncertainty p(E > 1.0 MeV) [n/cm2-sec] 1.11E+11 9.55E+10 0.86 6%

O(E > 0.1 MeV) [n/cmn2-sec] 4.96E+11 4.37E+11 0.88 10%

O(E < 0.414 eV) [n/cm2 -sec] 6.67E+10 9.32E+10 1.40 22%

dpa/sec 2.14E-10 1.88E-10 0.88 7%

Evaluation of Surveillance Capsule Dosimetry July 1999 WCAP-15253, Rev. 0

5-6 Table 5.2-3 Best Estimate Exposure Rates from the Surveillance Capsule V Dosimetry Withdrawn at the End of Fuel Cycle 8 Reaction Rate (rps/nucleus)

Best Reaction Measured Calculated Estimate BE / Meas BE / Calc Meas / Calc

'Cu (n,ox) 4.96E-17 4.58E-17 4.60E-17 0.93 1.00 1.08

'Fe (np) 4.55E-15 5.28E-15 4.85E-15 1.07 0.92 0.86

'Ni (n,p) 6.43E-15 7.42E-15 6.78E-15 1.05 0.91 0.87 2U (n,f) (Cd) 2.95E-14 2.88E-14 2.56E-14 0.87 0.89 1.02 7

1 Np (n,f) (Cd) 59 Co (n,7) 3.95E-12 3.93E-12 3.91E-12 0.99 0.99 1.01

'9Co (n,y) (Cd) 2.31E-12 2.76E-12 2.34E-12 1.01 0.85 0.84 Best 1(y Exposure Rate Calculated Estimate BE / Calc Uncertainty p(E > 1.0 MeV) [n/cm 2-sec] 9.10E+10 8.01E+10 0.88 7%

O(E > 0.1 MeV) [n/cm'-sec] 3.97E+11 3.62E+11 0.91 11%

O(E < 0.414 eV) [n/cm -sec] 5.38E+10 6.83E+10 1.27 22%

dpa/sec 1.74E-10 1.57E-10 0.91 9%

Evaluation of Surveillance Capsule Dosimetry July 1999 WCAP-15253, Rev. 0

5-7 Table 5.2-4 Best Estimate Exposure Rates from the Surveillance Capsule Z Dosimetry Withdrawn at the End of Fuel Cycle 8 Reaction Rate (rps/nucleus)

Best Reaction Measured Calculated Estimate BE / Meas BE / Calc Meas / Calc

"'Cu (nax) 5.17E-17 4.84E-17 4.79E-17 0.93 0.99 1.07 "TFe (n,p) 4.83E-15 5.79E-15 5.23E-15 1.08 0.90 0.83 SNi (np) 7.03E-15 8.18E-15 7.35E-15 1.05 0.90 0.86 2U (nf) (Cd) 3.33E-14 3.25E-14 2.85E-14 0.86 0.88 1.02 7

23 Np (nf) (Cd) 2.86E-13 3.25E-13 2.86E-13 1.00 0.88 0.88 59 Co (n,'y) 4.50E-12 4.63E-12 4.45E-12 0.99 0.96 0.97 5

-Co (n,,y) (Cd) 2.58E-12 3.25E-12 2.62E-12 1.02 0.81 0.79 Best 1y Exposure Rate Calculated Estimate BE I Calc Uncertainty O(E > 1.0 MeV) [n/cm 2-sec] 1.04E+11 9.07E+10 0.87 6%

p(E > 0.1 MeV) [n/cm 2-sec] 4.68E+11 4.24E+11 0.91 10%

4(E < 0.414 eV) [n/ cm -sec] 6.29E+10 7.99E+10 1.27 22%

dpa/sec 2.02E-10 1.81E-10 0.90 7%

Evaluation of Surveillance Capsule Dosimetry July 1999 WCAP-15253, Rev. 0

5-8 Table 5.2-5 Best Estimate Exposure Rates from the Surveillance Capsule Y Dosimetry Withdrawn at the End of Fuel Cycle 11 Reaction Rate (rps/nucleus)

Best Reaction Measured Calculated Estimate BE / Meas BE / Calc Meas / Calc

'Cu (n,cL) 4.31E-17 4.47E-17 4.16E-17 0.97 0.93 0.96

Fe (n,p) 4.41E-15 5.14E-15 4.60E-15 1.04 0.89 0.86 "SgNi (np) 6.14E-15 7.23E-15 6.45E-15 1.05 0.89 0.85 8U (n,f) (Cd) 3.24E-14 2.80E-14 2.48E-14 0.77 0.89 1.16 1 Np (n,f) (Cd) 1.88E-13 2.72E-13 2.16E-13 1.15 0.79 0.69

-9Co (n,y) 3.99E-12 3.83E-12 3.93E-12 0.98 1.03 1.04 59 "Co (n,y) (Cd) 2.26E-12 2.69E-12 2.29E-12 1.01 0.85 0.84 Best 1l7 Calculated Estimate BE / Calc Uncertainty Exposure Rate O(E > 1.0 MeV) [n/cm 2-sec] 8.86E+10 7.74E+10 0.87 6%

3.35E+11 0.87 10%

O(E > 0.1 MeV) [n/cm 2-sec] 3.87E+11 1.32 22%

O(E < 0.414 eV) [n/cm2 -sec] 5.24E+10 6.91E+10 dpa/sec 1.69E-10 1.47E-10 0.87 7%

Evaluation of Surveillance Capsule Dosimetry July 1999 WCAP-15253, Rev. 0

6-1 6 EVALUATION OF REACTOR CAVITY DOSIMETRY In this section, the results of the evaluations of the neutron sensor sets irradiated since the inception of the Reactor Cavity Measurement Program are presented. At McGuire Unit 1, the program was initiated prior to Cycle 12 operation and includes one set of measurement evaluations at the conclusion of Cycle 12. The evaluation of this set of measured data was accomplished using a consistent approach based on the methodology discussed in Section 3.0, resulting in an accurate data base defining the exposure of the reactor vessel wall.

6.1 CYCLE 12 RESULTS 6.1.1 Measured Reaction rates During the Cycle 12 irradiation, six multiple foil sensor sets and four stainless steel gradient chains were deployed in the reactor cavity as depicted in Figures 2.1-1 through 2.1-3. The capsule identifications associated with each of the multiple foil sensor sets were as follows:

CAPSULE IDENTIFICATION Azimuth Core Core Core (Degrees) Top Midplane Bottom 0.5 A 14.5 B 29.5 C 44.5 D E F The contents of each of these irradiation capsules is specified in Appendix B to this report.

The irradiation history of the McGuire Unit 1 reactor during Cycle 12 is listed in Appendix B.

The irradiation history was obtained from plant personneltf7 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 3411 MWt were computed for each multiple foil sensor and gradient wire segment.

The computed reaction rates for the multiple foil sensor sets irradiated during Cycle 12 are provided in Table 6.1-1. Corresponding "Fe (np), 5Ni (np), and ' 9Co (ny) reaction rate data from segments of the four stainless steel gradient chains are recorded in Tables 6.1-2 through 6.1-5 for the 0.5-, 14.50, 29.50, and 44.50, respectively.

In regard to the data listed in Table 6.1-1, the 'Fe (np) reaction rates represent an average of the bare and cadmium covered measurements for each capsule. The "U (n,f) reaction rates include corrections for 2.U impurities in the 'Ut sensors as well as corrections for photo-fission reactions in both the .. U and "7Np sensors. The 'U and "7Np sensors also include corrections Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-2 for the y-ray self-absorption in the oxide matrix and vanadium capsule as described in Section 3.2.1.2.

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 12 irradiation are provided in Tables 6.1-6 through 6.1-11. In these tables, the best estimate exposure experienced at each sensor set location along with data illustrating the fit of both the trial and best estimate spectra to the measurements are given. Also included in the tabulations are the 1y uncertainties associated with each of the derived exposure rates.

In regard to the comparisons listed in Tables 6.1-6 through 6.1-11, it should be noted that the columns labeled "Calculated" were obtained by normalizing the neutron spectral data from Table 4.1-2 to the absolute calculated neutron flux (E > 1.0 MeV) averaged over the Cycle 12 irradiation period as discussed in Section 3.0. Thus, the comparisons illustrated in Tables 6.1-6 through 6.1-11 indicate the degree to which the calculated neutron energy spectra matched the measured data before and after adjustment. Absolute comparisons of calculation and measurement are discussed further in Section 7.0 of this report.

Complete traverses of fast neutron exposure rates in the reactor cavity were developed by combining the results of the least squares adjustment of the multiple foil data with the 'Fe (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 O(E > 1.0 MeV), O(E > 0.1 MeV), and dpa/sec in the reactor cavity.

The resultant axial distributions of O(E > 1.0 MeV), O(E > 0.1 MeV), and dpa/sec are given in Tables 6.1-12, 6.1-13, and 6.1-14, respectively. The distributions of O(E > 1.0 MeV) are depicted graphically in Figures 6.1-1 through 6.1-4. In these graphical presentations, results for axial locations of -6.0, 0.0, and +6.0 feet relative to the core midplane represent the explicit results of the FERRET evaluations, while results at the remaining axial locations depict the normalized data from the gradient chains.

Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-3 Table 6.1-1 Summary of Reaction Rates Derived from Multiple Foil Sensor Sets Cycle 12 Irradiation Reaction Rate [rps/nucleus]

Reaction Capsule A Capsule B Capsule C Capsule E Capsule D Capsule F

'Cu (na)6Co (Cd) 3.36E-19 4.62E-19 3.79E-19 2.91E-19 1.08E-19 1.19E-19 "Ti (n,p) 4Sc (Cd) 4.30E-18 6.55E-18 5.77E-18 4.23E-18 1.68E-18 1.71E-18 "Fe (n,p) -Mn (Cd) 2.07E-17 3.17E-17 3.09E-17 2.44E-17 9.12E-18 1.03E-17

'8Ni (n~p) -ýCo 2.91E-17 4.50E-17 4.19E-17 3.44E-17 1.43E-17 1.55E-17 21U (nf) '37Cs (Cd) 1.23E-16 1.84E-16 1.77E-16 1.72E-16 7.62E-17 7.32E-17 7

23 Np (n,f) '37Cs (Cd) 1.85E-15 2.64E-15 3.16E-15 3.30E-15 1.33E-15 1.39E-15

' 9Co (n,Y) 60Co 4.24E-14 6.43E-14 7.42E-14 4.65E-14 1.68E-14 2.45E-14 "9Co (ny) "Co (Cd) 1.75E-14 2.77E-14 3.45E-14 2.77E-14 9.72E-15 1.63E-14 Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-4 Table 6.1-2 p Ni (n,p) and -9Co (n,y) Reaction Rates Derived from the RCND

'Fe (np),

Stainless Steel Gradient Chain at 0.50 - Cycle 12 Irradiation Distance From REACTION RATE Core Midplane [rps/nucleus]

(ft) 'Fe (n,p) 'Ni (n.p) "5Co (ny) 5.5 1.10E-17 1.58E-17 1.38E-14 4.5 1.65E-17 2.30E-17 1.84E-14 3.5 1.92E-17 2.40E-17 2.12E-14 2.5 1.93E-17 2.74t-17 2.31E-14 1.5 1.98E-17 2.85E-17 2.52E-14 0.5 2.OOE-17 2.79E-17 3.53E-14

-0.5 2.18E-17 2.84E-17 3.69E-14

-1.5 1.95E-17 2.64E-17 3.64E-14

-2.5 2.02E-17 2.82E-17 3.40E-14

-3.5 1.87E-17 2.71E-17 3.01E-14

-4.5 1.65E-17 2.33E-17 2.12E-14

-5.5 1.11E-17 1.62E-17 1.28E-14

-6.5 4.50E-18 6.53E-18 9.79E-15 Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-5 Table 6.1-3 'Fe (n,p), 'Ni (np) and 5 9Co (ny) Reaction Rates Derived from the RCND Stainless Steel Gradient Chain at 14.50 - Cycle 12 Irradiation Distance From REACTION RATE Core Midplane [rps/nucleus]

(ft) ýFe (np) gNi (np) "SCo (nv) 5.5 1.55E-17 2.29E-17 2.75E-14 4.5 2.42E-17 3.47E-17 3.92E-14 3.5 3.OOE-17 4.16E-17 4.75E-14 2.5 3.06E-17 4.13E-17 5.10E-14 1.5 3.07E-17 4.18E-17 5.21E-14 0.5 3.03E-17 4.23E-17 5.48E-14

-0.5 3.19E-17 3.99E-17 5.58E-14

-1.5 3.06E-17 3.95E-17 5.28E-14

-2.5 2.76E-17 3.96E-17 4.82E-14

-3.5 2.72E-17 3.67E-17 4.39E-14

-4.5 2.42E-17 3.50E-17 3.64E-14

-5.5 1.61E-17 2.24E-17 2.63E-14

-6.5 6.45E-18 8.99E-18 1.51E-14 Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-6 Table 6.1-4 'Fe (n,p), 5SNi (np) and 59Co (n,y) Reaction Rates Derived from the RCND Stainless Steel Gradient Chain at 29.50 - Cycle 12 Irradiation Distance From REACTION RATE Core Midplane [rps/nucleus]

5 (ft) "Fe (np) *Ni (n~p) "'Co (ny) 5.5 1.55E-17 2.31E-17 3.13E-14 4.5 2.25E-17 3.30E-17 4.34E-14 3.5 2.80E-17 4.20E-17 5.52E-14 2.5 2.98E-17 4.22E-17 6.15E-14 1.5 2.86E-17 4.20E-17 6.30E-14 0.5 2.98E-17 3.96E-17 6.28E-14

-0.5 3.08E-17 4.15E-17 6.61E-14

-1.5 2.81E-17 3.99E-17 6.36E-14

-2.5 2.83E-17 3.99E-17 6.17E-14

-3.5 2.61E-17 3.72E-17 5.44E-14

-4.5 2.45E-17 3.66E-17 4.35E-14

-5.5 1.53E-17 2.25E-17 3.17E-14

-6.5 6.49E-18 9.19E-18 1.85E-14 Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-7 Table 6.1-5 "Fe (np), 'Ni (np) and 'Co (ny) Reaction Rates Derived from the RCND Stainless Steel Gradient Chain at 44.5' - Cycle 12 Irradiation Distance From REACTION RATE Core Midplane [rps/nucleus]

(ft) "*Fe (n~p) *Ni (nop) ' 9Co (ny) 5.5 1.28E-17 1.87E-17 1.86E-14 4.5 1.84E-17 2.85E-17 2.66E-14 3.5 2.27E-17 3.16E-17 3.27E-14 2.5 2.28E-17 3.22E-17 3.71E-14 1.5 2.28E-17 3.41E-17 3.98E-14 0.5 2.47E-17 3.38E-17 4.12E-14

-0.5 2.24E-17 3.39E-17 4.20E-14

-1.5 2.37E-17 3.22E-17 4.19E-14

-2.5 2.49E-17 3.44E-17 4.04E-14

-3.5 2.35E-17 3.35E-17 3.65E-14

-4.5 2.02E-17 2.93E-17 3.13E-14

-5.5 1.38E-17 2.12E-17 2.27E-14

-6.5 6.32E-18 9.24E-18 1.86E-14 Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-8 Table 6.1-6 Best Estimate Exposure Rates from the Capsule A Dosimetry Evaluation 0.50 Azimuth - Core Midplane - Cycle 12 Irradiation Reaction Rate (rps/nucleus)

Best Reaction Measured Calculated Estimate BE / Meas BE / Calc Meas I Calc

'Cu (n,a) (Cd) 3.96E-19 3.18E-19 0.95 0.80 0.85 3.36E-19 5.39E-18 4.23E-18 0.98 0.78 0.80

'Ti (n,p) (Cd) 4.30E-18 2.22E-17 1.07 0.76 0.71

"*Fe (n,p) 2.07E-17 2.91E-17 4.04E-17 3.09E-17 1.06 0.76 0.72

'Ni (n,p) (Cd) 2.91E-17 1.43E-16 1.10E-16 0.89 0.77 0.86 "U (n,f) (Cd) 1.23E-16 1.95E-15 1.71E-15 0.92 0.88 0.95

-'Np (n,f) (Cd) 1.85E-15 8.18E-14 4.27E-14 1.01 0.52 0.52

-"Co (n,y) 4.24E-14 59 2.61E-14 1.76E-14 1.01 0.67 0.67

" Co (n,y) (Cd) 1.75E-14 Best 1G Calculated Estimate BE / Calc Uncertainty Exposure Rate 4.61E+08 3.63E+08 0.79 6%

O(E > 1.0 MeV) [n/cm2-sec]

4.11E+09 3.62E+09 0.88 11%

p(E > 0.1 MeV) [n/cm 2-sec]

1.88E+09 1.04E+09 0.55 22%

O(E < 0.414 eV) [n/cm-sec]

1.47E-12 1.25E-12 0.85 9%

dpa/sec July 1999 Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

Table 6.1-7 Best Estimate Exposure Rates From the Capsule B Dosimetry Evaluation 14.50 Azimuth - Core Midplane - Cycle 12 Irradiation Reaction Rate (rps/nucleus)

Best Reaction Measured Calculated Estimate BE / Meas BE / Calc Meas / Calc

'Cu (n,a) (Cd) 4.62E-19 5.05E-19 4.52E-19 0.98 0.90 0.91

Ti (np) (Cd) 6.55E-18 7.08E-18 6.20E-18 0.95 0.88 0.93

'Fe (n,p) 3.17E-17 4.00E-17 3.38E-17 1.07 0.85 0.79 3 4.50E-17 5.60E-17 4.72E-17 1.05 0.84 0.80 Ni (np) (Cd)

U (nf) (Cd) 1.84E-16 2.04E-16 1.70E-16 0.92 0.83 0.90 7

Np (n,f) (Cd) 2.64E-15 2.91E-15 2.54E-15 0.96 0.87 0.91 59

" Co (n,y) 6.43E-14 1.08E-13 6.48E-14 1.01 0.60 0.60 59 0.67 "Co (n,'y) (Cd) 2.77E-14 4.15E-14 2.79E-14 1.01 0.67 Best Exposure Rate Calculated Estimate BE / Calc Uncertainty 2

(E > 1.0 MeV) [n/cm -sec] 6.74E+08 5.60E+08 0.83 6%

ý(E > 0.1 MeV) [n/cm -sec]

6.36E+09 5.59E+09 0.88 11%

4(E < 0.414 eV) [n/cM2-sec] 2.32E+09 1.52E+09 0.65 21%

dpa/sec 2.24E-12 1.94E-12 0.87 9%

Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-10 Table 6.1-8 Best Estimate Exposure Rates From the Capsule C Dosimetry Evaluation 29.50 Azimuth - Core Midplane - Cycle 12 Irradiation Reaction Rate (rps/nucleus)

Best Reaction Measured Calculated Estimate BE I Meas BE / Calc Meas / Calc

'Cu (n,a) (Cd) 3.79E-19 4.69E-19 3.86E-19 1.02 0.82 0.81

'Ti (n,p) (Cd) 5.77E-18 6.65E-18 5.42E-18 0.94 0.82 0.87

Fe (n,p) 3.09E-17 3.90E-17 3.14E-17 1.02 0.81 0.79

Ni (n,p) (Cd) 4.19E-17 5.54E-17 4.46E-17 1.06 0.81 0.76

'-'U (n,f) (Cd) 1.77E-16 2.13E-16 1.73E-16 0.98 0.81 0.83

=7Np (nf) (Cd) 3.16E-15 3.28E-15 2.93E-15 0.93 0.89 0.96 "59Co (n,y) 7.42E-14 1.28E-13 7.48E-14 1.01 0.58 0.58 59 Co (n,y) (Cd) 3.45E-14 4.95E-14 3.46E-14 1.00 0.70 0.70 Best 1(y Exposure Rate Calculated Estimate BE / Calc Uncertainty O(E > 1.0 MeV) [n/cm 2 -sec] 7.28E+08 5.99E+08 0.82 6%

?(E > 0.1 MeV) [n/cm 2 -sec] 7.40E+09 6.54E+09 0.88 11%

O(E < 0.414 eV) [n/cm 2-sec] 2.74E+09 1.67E+09 0.61 21%

dpa/sec 2.56E-12 2.22E-12 0.87 9%

Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-11 Table 6.1-9 Best Estimate Exposure Rates From the Capsule E Dosimetry Evaluation 44.50 Azimuth - Core Midplane - Cycle 12 Irradiation Reaction Rate (rps/nucleus)

Best Reaction Measured Calculated Estimate BE ! Meas BE / Calc Meas / Calc

'Cu (na) (Cd) 2.91E-19 3.61E-19 2.89E-19 0.99 0.80 0.81 4Ti (np) (Cd) 4.23E-18 5.14E-18 4.09E-18 0.97 0.80 0.82

'Fe (n,p) 2.44E-17 3.13E-17 2.51E-17 1.03 0.80 0.78

'Ni (np) (Cd) 3.44E-17 4.51E-17 3.66E-17 1.06 0.81 0.76

'U (nf) (Cd) 1.72E-16 1.85E-16 1.56E-16 0.91 0.84 0.93

' 7Np (n,f) (Cd) 3.30E-15 2.95E-15 2.94E-15 0.89 1.00 1.12 59 4.65E-14 1.20E-13 4.78E-14 Co (ny) 1.03 0.40 0.39 "95Co (n,y) (Cd) 2.77E-14 3.91E-14 2.74E-14 0.99 0.70 0.71 Best la Exposure Rate Calculated Estimate BE / Calc Uncertainty

ý(E > 1.0 MeV) [n/cm 2-sec] 6.50E+08 5.68E+08 0.87 6%

O(E > 0.1 MeV) [n/cm 2-sec] 6.52E+09 6.26E+09 0.96 11%

O(E < 0.414 eV) [n/crn2-secj 2.75E+09 9.69E+08 0.35 22%

dpa/sec 2.24E-12 2.09E-12 0.93 9%

Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-12 Table 6.1-10 Best Estimate Exposure Rates From the Capsule D Dosimetry Evaluation 44.5' Azimuth - Top of Core - Cycle 12 Irradiation Reaction Rate (rps/nucleus)

Best Reaction Measured Calculated Estimate BE I Meas BE I Calc Meas / Calc

'Cu. (n,a) (Cd) 1.08E-19 1.50E-19 1.11E-19 1.03 0.74 0.72 63 Ci (np) (Cd) 1.68E-18 2.15E-18 1.59E-18 0.95 0.74 0.78

'Fe (np) 9.12E-18 1.31E-17 9.98E-18 1.09 0.76 0.70 5

Ni (n,p) (Cd) 1.43E-17 1.88E-17 1.47E-17 1.03 0.78 0.76

'U (nf) (Cd) 7.62E-17 7.72E-17 6.41E-17 0.84 0.83 0.99

' 7 Np (n,f) (Cd) 1.33E-15 1.23E-15 1.21E-15 0.91 0.98 1.08 5

'Co (ny) 1.68E-14 5.01E-14 1.73E-14 1.03 0.35 0.34 59 "Co (n,y) (Cd) 9.72E-15 1.63E-14 9.67E-15 0.99 0.59 0.60 Best la Exposure Rate Calculated Estimate BE I Calc Uncertainty O(E > 1.0 MeV) [n/cmn2-sec] 2.71E+08 2.36E+08 0.87 6%

p(E > 0.1 MeV) [n/cm2-sec] 2.72E+09 2.59E+09 0.95 11%

O(E < 0.414 eV) [n/cm2-sec] 1.15E+09 3.73E+08 0.32 22%

dpa/sec 9.34E-13 8.61E-13 0.92 9%

Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-13 Table 6.1-11 Best Estimate Exposure Rates From the Capsule F Dosimetry Evaluation 44.50 Azimuth - Bottom of Core - Cycle 12 Irradiation Reaction Rate (rps/nucleus)

Best Reaction Measured Calculated Estimate BE / Meas BE / Calc Meas / Calc 1.19E-19 1.43E-19 1.19E-19 1.00 0.83 0.83

'Cu (na) (Cd) 1.70E-18 0.99 0.83 0.84

' 6Ti (np) (Cd) 1.71E-18 2.04E-18 1.07E-17 1.04 0.86 0.83 "UFe(np) 1.03E-17 1.24E-17 1.55E-17 1.79E-17 1.57E-17 1.01 0.88 0.87

'Ni (np) (Cd) 7.32E-17 7.34E-17 6.79E-17 0.93 0.93 1.00

'U (nf) (Cd) 7 1.39E-15 1.17E-15 1.27E-15 0.91 1.09 1.19 2 Np (nf) (Cd)

"S9Co (ny) 2.45E-14 4.77E-14 2.53E-14 1.03 0.53 0.51 1.59E-14 0.98 1.03 1.05

' 9Co (n,y) (Cd) 1.63E-14 1.55E-14 Best la Calculated Estimate BE / Calc Uncertainty Exposure Rate 0.96 6%

O(E > 1.0 MeV) [n/cm2-sec] 2.58E+08 2.49E+08 1.05 11%

O(E > 0.1 MeV) [n/cm2-sec] 2.59E+09 2.70E+09 2

-sec] 1.09E+09 4.21E+08 0.39 23%

O(E < 0.414 eV) [n/crm dpa/sec 8.88E-13 9.03E-13 1.02 9%

Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-14 Table 6.1-12 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position within the Reactor Cavity - Cycle 12 Irradiation Distance From Core Midplane Neutron Flux (n/cm2-sec)

(ft) 0.50 14.50 29.50 44.50 6.0 2.36E+08 5.5 1.91E+08 2.80E+08 3.06E+08 3.10E+08 4.5 2.86E+08 4.36E+08 4.45E+08 4.43E+08 3.5 3.35E+08 5.40E+08 5.55E+08 5.47E+08 2.5 3.36E+08 5.50E+08 5.89E+08 5.49E+08 1.5 3.45E+08 5.53E+08 5.66E+08 5.50E+08 0.5 3.47E+08 5.46E+08 5.89E+08 5.95E+08 0.0 3.63E+08 5.60E+08 5.99E+08 5.68E+08

-0.5 3.78E+08 5.75E+08 6.09E+08 5.41E+08

-1.5 3.39E+08 5.50E+08 5.57E+08 5.73E+08

-2.5 3.51E+08 4.98E+08 5.60E+08 6.02E+08

-3.5 3.24E+08 4.91E+08 5.16E+08 5.68E+08

-4.5 2.86E+08 4.36E+08 4.85E+08 4.88E+08

-5.5 1.93E+08 2.89E+08 3.03E+08 3.32E+08

-6.0 2.49E+08

-6.5 7.82E+07 1.16E+08 1.28E+08 1.52E+08 Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-15 Table 6.1-13 Fast Neutron Flux (E > 0.1 MeV) as a Function of Axial Position within the Reactor Cavity - Cycle 12 Irradiation Distance From Core Midplane Neutron Flux (nlcm 2-sec)

(ft) 0.50 14.50 29.50 44.50 6.0 2.37E+09 5.5 1.7E+09 2.64E+09 3.11E+09 3.10E+09 4.5 2.56E+09 4.11E+09 4.52E+09 4.45E+09 3.5 2.99E+09 5.09E+09 5.64E+09 5.49E+09 2.5 3.OOE+09 5.19E+09 5.99E+09 5.50E+09 1.5 3.08E+09 5.21E+09 5.75E+09 5.52E+09 0.5 3.10E+09 5.15E+09 5.99E+09 5.97E+09 0.0 3.24E+09 5.29E+09 6.09E+09 5.69E+09

-0.5 3.38E+09 5.43E+09 6.19E+09 5.42E+09

-1.5 3.03E+09 5.19E+09 5.65E+09 5.74E+09

-2.5 3.13E+09 4.70E+09 5.69E+09 6.03E+09

-3.5 2.89E+09 4.63E+09 5.24E+09 5.69E+09

-4.5 2.56E+09 4.11E+09 4.92E+09 4.90E+09

-5.5 1.72E+09 2.73E+09 3.07E+09 3.33E+09

-6.0 2.49E+09

-6.5 6.98E+08 1.10E+09 1.31E+09 1.53E+09 Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-16 Table 6.1-14 Iron Displacement Rate as a Function Of Axial Position within the Reactor Cavity - Cycle 12 Irradiation Distance From Core Midplane Displacement Rate (dpalsec)

(ft) 0.50 14.50 29.50 44.50 6.0 8.14E-13 5.5 6.09E-13 9.31E-13 1.08E-12 1.07E-12 4.5 9.14E-13 1.45E-12 1.56E-12 1.53E-12 3.5 1.07E-12 1.79E-12 1.95E-12 1.88E-12 2.5 1.07E-12 1.83E-12 2.07E-12 1.89E-12 1.5 1.10E-12 1.84E-12 1.99E-12 1.90E-12 0.5 1.11E-12 1.81E-12 2.07E-12 2.05E-12 0.0 1.16E-12 1.86E-12 2.10E-12 1.96E-12

-0.5 1.21E-12 1.91E-12 2.14E-12 1.86E-12

-1.5 1.08E-12 1.83E-12 1.95E-12 1.97E-12

-2.5 1.12E-12 1.66E-12 1.97E-12 2.07E-12

-3.5 1.03E-12 1.63E-12 1.81E-12 1.96E-12

-4.5 9.14E-13 1.45E-12 1.70E-12 1.68E-12

-5.5 6.17E-13 9.63E-13 1.06E-12 1.14E-12

-6.0 8.57E-13

-6.5 2.50E-13 3.86E-13 4.51E-13 5.25E-13 Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-17 Figure 6.1-1 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position Along the 0.50 Traverse in the Reactor Cavity Cycle 12 Irradiation MoGuire 1 Cycle 12 - 0.5 deg Cavity 1.E+09 0

"1.E+08 z

z 1 .E+07

-8 -6 -4 -2 0 2 4 6 8 Distance from Core Midplane (ft)

July i33 Evaluation of Reactor Cavity Dosimetry July 199R WCAP-15253, Rev. 0

6-18 Figure 6.1-2 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position Along the 14.5' Traverse in the Reactor Cavity Cycle 12 Irradiation McGuire 1 Cycle 12 - 14.5 deg Cavity 1.E+09 E 9 U,I 0

X 1.E+08 o

L-.

Z0) 1.E+07 I I

-8 -6 -4 -2 0 2 4 6 8 Distance from Core Midplane (ft)

Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-19 Figure 6.1-3 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position along the 29.50 Traverse in the Reactor Cavity Cycle 12 Irradiation ivlGuire 1 Cycle 12 - 29.5 deg Cavity 1.E+09 U,t , ..

E 0

a.

1.E+08 z

4.L a,

1.E+07 I I

-8 -6 -4 -2 0 2 4 6 8 Distance from Core Midplane (ft)

Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

6-20 Figure 6.1-4 Fast Neutron Flux (E > 1.0 MeV) as a Function of Axial Position Along the 44.5' Traverse in the Reactor Cavity Cycle 12 Irradiation McGuire 1 Cycle 12 - 44.5 deg Cavity 1.E+09 E

1.E+08 x

0 z

1. E+07

-8 -6 -4 -2 0 2 4 6 8 Distance from Core Midplane (ft)

Evaluation of Reactor Cavity Dosimetry July 1999 WCAP-15253, Rev. 0

7-1 7 COMPARISON OF CALCULATIONS WITH MEASUREMENTS As described in Section 3.3, the best estimate neutron exposure projections for the McGuire Unit 1 pressure vessel were based on a combination of plant specific neutron transport calculations and plant specific measurements. Direct comparisons of the transport calculations with the McGuire Unit 1 measurement data base were used to quantify the biases that may exist due to the transport methodology, reactor modeling, and/or reactor operating characteristics over the respective irradiation periods.

In this section, comparisons of the measurement results from surveillance capsule and reactor cavity dosimetry with corresponding analytical predictions at the measurement locations are presented. These comparisons are provided on two levels. In the first instance, predictions of fast neutron exposure rates in terms of ý(E > 1.0 MeV), .O(E > 0.1 MeV), and dpa/sec are compared with the Best Estimate 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 best estimate/calculation (BE/C) comparisons yield an accurate plant specific bias factor that can be applied to neutron transport calculations performed for the McGuire Unit 1 reactor to produce "best estimate" exposure projections for the pressure vessel wall.

7.1 COMPARISON OF BEST ESTIMATE RESULTS WITH CALCULATION In Table 7.1-1, comparisons of best estimate and calculated exposure rates for the five surveillance capsule dosimetry sets withdrawn to date as well as for the reactor cavity midplane dosimetry sets irradiated during Cycle 12 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 midplane data, the best estimate integrated exposures were less than calculated values by factors of 0.87, 0.92, and 0.90 for ct(E > 1.0 MeV), 4(E > 0.1 MeV), and dpa/sec, respectively.

The standard deviations associated with each of the 9 sample data sets were 8.0%, 7.4%, and 7.2%, 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 standard deviations observed for the six fast neutron reactions range from 6.5% to Comparison of Calculations with Measurements July 1999 WCAP-15253, Rev. 0

7-2 17.1% on an individual reaction basis; whereas, the overall average M/C ratio for the entire data set has an associated la standard deviation of 13.6%. Furthermore, the average M/C bias of 0.90 observed in the reaction rate comparisons is in excellent agreement with the values observed in the integrated exposure comparisons shown in Table 7.1-1.

with Measurements July 1999 Comparison of Calculations Comparison of Calculations with Measurements July 1999 WCA-P-15253, Rev. 0

7-3 Table 7.1-1 Comparison of Best Estimate and Calculated Exposure Rates from Surveillance Capsule and Cavity Dosimetry Irradiations Neutron Fluence (E > 1.0 MeV) [n/cm2 ]

Calculated Best Estimate BE/C Surveillance Capsules Capsule U 4.05E+18 4.20E+18 1.04 Capsule X 1.50E+19 1.30E+19 0.86 Capsule V 2.08E+19 1.83E+19 0.88 Capsule Z 2.38E+19 2.07E+19 0.87 Capsule Y 2.86E+19 2.49E+19 0.87 0.50 Cavift Cycle 12 1.43E+16 1.13E+16 0.79 14.50 Cavity Cycle 12 2.09E+16 1.74E+16 0.83 29.5' Cavity Cycle 12 2.26E+16 1.86E+16 0.82 44.50 Cavity Cycle 12 2.02E+16 1.77E+16 0.87 Average BE/C Bias Factor 0.87

% Standard Deviation (1a) 8.0%

Note: The lc standard deviation represents the uncertainty associated with the derivation of the bias factor from all available measurements. The bias factor uncertainty is just one part of the total uncertainty associated with the pressure vessel exposure projections as described in Section 8.3.

Comparison of Calculations with Measurements July 1999 WCAP-15253, Rev. 0

7-4 Table 7.1-1 Comparison of Best Estimate and Calculated Exposure Rates from (cont.) Surveillance Capsule and Cavity Dosimetry Irradiations Neutron Fluence (E > 0.1 MeV) [n/cm2 ]

Calculated Best Estimate BE/C Surveillance Capsules Capsule U 1.82E+19 1.97E+19 1.08 Capsule X 6.73E+19 5.93E+19 0.88 Capsule V 9.07E+19 8.28E+19 0.91 Capsule Z 1.07E+20 9.68E+19 0.90 Capsule Y 1.25E+20 1.08E+20 0.87 0.50 Cavity Cycle 12 1.28E+17 1.12E+17 0.88 14.5" Cavity Cycle 12 1.98E+17 1.74E+17 0.88 29.5' Cavity Cycle 12 2.30E+17 2.03E+17 0.88 44.5 Cavity Cycle 12 2.02E+17 1.95E+17 0.96 Average BE/C Bias Factor 0.92

% Standard Deviation (1a) 7.4%

Note: The l( standard deviation represents the uncertainty associated with the derivation of the bias factor from all available measurements. The bias factor uncertainty is just one part of the total uncertainty associated with the pressure vessel exposure projections as described in Section 8.3.

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7-5 Table 7.1-1 Comparison of Best Estimate and Calculated Exposure Rates from (cont.) Surveillance Capsule and Cavity Dosimetry Irradiations Iron Displacements [dpa]

Calculated Best Estimate BE/C Surveillance Capsules Capsule U 7.83E-03 8.35E-03 1.07 Capsule X 2.90E-02 2.55E-02 0.88 Capsule V 3.97E-02 3.59E-02 0.90 Capsule Z 4.61E-02 4.13E-02 0.90 Capsule Y 5.45E-02 4.75E-02 0.87 0.50 Cavity Cycle 12 4.57E-05 3.89E-05 0.85 14.5' Cavity Cycle 12 6.96E-05 6.04E-05 0.87 29.5' Cavity Cycle 12 7.94E-05 6.89E-05 0.87 44.50 Cavity Cycle 12 6.95E-05 6.48E-05 0.93 Average BE/C Bias Factor 0.90

% Standard Deviation (1a) 7.2%

Note: The la standard deviation represents the uncertainty associated with the derivation of the bias factor from all available measurements. The bias factor uncertainty is just one part of the total uncertainty associated with the pressure vessel exposure projections as described in Section 8.3.

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7-6 Table 7.2-1 Comparison of Measured and Calculated Neutron Sensor Reaction Rates from Surveillance Capsule and Cavity Dosimetry Irradiations "3Cu (n,cc) 4

'Ti (n ) 'Fe (np) 'Ni (np) 2-8 (n,f) 2 7

Np (n,f)

Surveillance Capsules Capsule U 1.10 0.98 0.99 1.12 1.17 Capsule X 1.04 0.85 0.86 1.04 0.78 Capsule V 1.08 0.86 0.87 1.02 Capsule Z 1.07 0.83 0.86 1.02 0.88 Capsule Y 0.96 0.86 0.85 1.16 0.69 0.5° Cavity Cycle 12 0.85 0.80 0.71 0.72 0.86 0.95 14.50 Cavity Cycle 12 0.91 0.93 0.79 0.80 0.90 0.91 29.5° Cavity Cycle 12 0.81 0.87 0.79 0.76 0.83 0.96 44.50 Cavity Cycle 12 0.81 0.82 0.78 0.76 0.93 1.12 Average 0.96 0.85 0.83 0.83 0.99 0.93

% Std. Dev. (la) 12.4% 6.5% 8.9% 9.7% 11.5% 17.1%

Overall M/C Average 0.90

%Std. Dev. (1(y) 13.2%

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8-1 8 BEST ESTIMATE NEUTRON EXPOSURE OF PRESSURE VESSEL MATERIALS In this section the measurement results provided in Sections 5.0 and 6.0 are combined with the results of the neutron transport calculations described in Section 4.0 to establish a mapping of the best estimate neutron exposure of the beltline region of the McGuire Unit 1 reactor pressure vessel through the completion of Cycle 12. Based on the continued use of the core power distributions producing the Cycles 9-12 measured results, projections of future vessel exposure to 21, 34, and 51 effective full power years of operation are also provided. In addition to the spatial mapping over the beltline region, data pertinent to the maximum exposure experienced by the intermediate and lower shell plates as well as the beltline circumferential and longitudinal welds are highlighted.

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:

(DBest Est.= K(4 Caic.

where: ODt,. = The best estimate fast neutron exposure at the location of interest.

K = The plant specific best estimate/calculation (BE/C) bias factor derived from all available surveillance capsule and reactor cavity dosimetry data.

Dc.1c. = 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 4.2 were as follows:

(D(E > 1.0 MeV) 0.87 + 8.0%

'I(E > 0.1 MeV) 0.92 + 7.4%

dpa 0.90 + 7.2%

These bias factors were based on the results of the continuous monitoring program at McGuire Unit 1 that has provided measured data from five internal surveillance capsules and one reactor cavity sensor set through the first 12 cycles of operation.

The uncertainties listed with the individual bias factors are at the 1(y level. Additional uncertainties associated with the evaluation of the best estimate vessel exposure are discussed in Section 8.3.

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8-2 8.1.1 Exposure Accrued During Cycles 1 through 12 To assess the incremental exposure resulting from the Cycles 12 irradiation, the bias factors listed in Section 8.1 were applied directly to the calculated values from Section 4.2 for the vessel clad/base metal interface to produce best estimate fluence levels characteristic of the midplane of the reactor core. The axial gradient chain measurements 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 8.1-12 to establish the exposure accrued by the reactor vessel through the end of Cycles 11 and 12.

Exposure distributions through the vessel wall can be developed using these surface exposures and radial distribution functions from Section 4.0. This exposure information, applicable through the end of Cycle 12, 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 McGuire Unit 1 will expand resulting in reduced uncertainties and an improved accuracy in the assessment of vessel condition.

8.1.2 Projection of Future Vessel Exposure At the end of Cycle 12, the McGuire Unit 1 reactor had accrued 11.20 effective full power years (EFPY) of operation. In order to establish a framework for the assessment of future vessel condition, exposure projections to 21, 34, and 51 EFPY are also included in Tables 8.1-1 through 8.1-12 in addition to the plant specific exposure assessments through the end of Cycle 12.

These extrapolations into the future were based on the assumption that the data from the Cycles 9-12 irradiation were representative of all future fuel cycles. That is, that future fuel designs would incorporate the low leakage fuel management concept employed during Cycles 9 through 12. Examination of these projected exposure levels establishes the long term effectiveness of the low leakage fuel management incorporated to date and can be used as a guide in assessing strategies for future vessel exposure management. The validity of these projections for future operation will be confirmed via the continued cavity monitoring program.

Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-3 Table 8.1-1 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 0' Azimuthal Angle Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 2.14E+18 2.29E+18 5.23E+18 9.13E+18 1.42E+19

-4.5 3.03E+18 3.26E+18 6.19E+18 1.01E+19 1.52E+19

-3.5 3.24E+18 3.48E+18 6.42E+18 1.03E+19 1.54E+19

-2.5 3.34E+18 3.59E+18 6.52E+18 1.04E+19 1.55E+19

-1.5 3.42E+18 3.67E+18 6.60E+18 1.05E+19 1.56E+19

-0.5 3.45E+18 3.71E+18 6.64E+18 1.05E+19 1.56E+19 0.0 3.47E+18 3.72E+18 6.66E+18 1.06E+19 1.56E+19 0.5 3.47E+18 3.73E+18 6.66E+18 1.06E+19 1.57E+19 1.5 3.45E+18 3.71E+18 6.64E+18 1.05E+19 1.56E+19 2.5 3.41E+18 3.66E+18 6.60E+18 1.05E+19 1.56E+19 3.5 3.31E+18 3.55E+18 6.49E+18 1.04E+19 1.55E+19 4.5 3.01E+18 3.24E+18 6.17E+18 1.01E+19 1.52E+19 5.5 2.09E+18 2.24E+18 5.18E+18 9.07E+18 1.42E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-4 Table 8.1-2 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 150 Azimuthal Angle Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 3.15E+18 3.39E+18 7.73E+18 1.35E+19 2.10E+19

-4.5 4.47E+18 4.81E+18 9.15E+18 1.49E+19 2.24E+19

-3.5 4.78E+18 5.15E+18 9.48E+18 1.52E+19 2.27E+19

-2.5 4.92E+18 5.30E+18 9.64E+18 1.54E+19 2.29E+19

-1.5 5.03E+18 5.42E+18 9.75E+18 1.55E+19 2.30E+19

-0.5 5.08E+18 5.48E+18 9.81E+18 1.56E+19 2.31E+19 0.0 5.10E+18 5.50E+18 9.83E+18 1.56E+19 2.31E+19 0.5 5.11E+18 5.51E+18 9.84E+18 1.56E+19 2.31E+19 1.5 5.08E+18 5.48E+18 9.82E+18 1.56E+19 2.31E+19 2.5 5.02E+18 5.41E+18 9.75E+18 1.55E+19 2.30E+19 3.5 4.87E+18 5.25E+18 9.58E+18 1.53E+19 2.28E+19 4.5 4.44E+18 4.78E+18 9.12E+18 1.49E+19 2.24E+19 5.5 3.07E+18 3.31E+18 7.65E+18 1.34E+19 2.09E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-5 Table 8.1-3 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 300 Azimuthal Angle Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY 3.OOE+18 3.25E+18 7.53E+18 1.32E+19 2.06E+19

-5.5

-4.5 4.25E+18 4.61E+18 8.89E+18 1.46E+19 2.20E+19

-3.5 4.55E+18 4.93E+18 9.21E+18 1.49E+19 2.23E+19

-2.5 4.68E+18 5.07E+18 9.36E+18 1.50E+19 2.25E+19

-1.5 4.79E+18 5.18E+18 9.47E+18 1.51E+19 2.26E+19 4.84E+18 5.24E+18 9.52E+18 1.52E+19 2.26E+19

-0.5 0.0 4.86E+18 5.26E+18 9.54E+18 1.52E+19 2.26E+19 4.86E+18 5.27E+18 9.55E+18 1.52E+19 2.27E+19 0.5 1.5 4.84E+18 5.24E+18 9.53E+18 1.52E+19 2.26E+19 2.5 4.78E+18 5.18E+18 9.46E+18 1.51E+19 2.26E+19 3.5 4.63E+18 5.02E+18 9.30E+18 1.50E+19 2.24E+19 4.5 4.22E+18 4.58E+18 8.86E+18 1.45E+19 2.20E+19 5.5 2.93E+18 3.17E+18 7.45E+18 1.31E+19 2.06E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-6 Table 8.1-4 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 450 Azimuthal Angle Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 3.39E+18 3.66E+18 8.45E+18 1.48E+19 2.31E+19

-4.5 4.80E+18 5.20E+18 9.99E+18 1.63E+19 2.46E+19

-3.5 5.14E+18 5.56E+18 1.03E+19 1.67E+19 2.50E+19

-2.5 5.29E+18 5.73E+18 1.05E+19 1.69E+19 2.52E+19

-1.5 5.41E+18 5.85E+18 1.06E+19 1.70E+19 2.53E+19

-0.5 5.47E+18 5.92E+18 1.07E+19 1.71E+19 2.54E+19 0.0 5.49E+18 5.94E+18 1.07E+19 1.71E+19 2.54E+19 0.5 5.50E+18 5.95E+18 1.07E+19 1.71E+19 2.54E+19 1.5 5.47E+18 5.92E+18 1.07E+19 1.71E+19 2.54E+19 2.5 5.40E+18 5.84E+18 1.06E+19 1.70E+19 2.53E+19 3.5 5.24E+18 5.67E+18 1.05E+19 1.68E+19 2.51E+19 4.5 4.77E+18 5.17E+18 9.95E+18 1.63E+19 2.46E+19 5.5 3.31E+18 3.58E+18 8.37E+18 1.47E+19 2.30E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-7 Table 8.1-5 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 00 Azimuthal Angle Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 4.73E+18 5.08E+18 1.16E+19 2.02E+19 3.15E+19

-4.5 6.72E+18 7.21E+18 1.37E+19 2.23E+19 3.36E+19

-3.5 7.18E+18 7.71E+18 1.42E+19 2.28E+19 3.41E+19

-2.5 7.40E+18 7.94E+18 1.44E+19 2.31E+19 3.43E+19

-1.5 7.56E+18 8.11E+18 1.46E+19 2.32E+19 3.45E+19

-0.5 7.64E+18 8.20E+18 1.47E+19 2.33E+19 3.46E+19 0.0 7.67E+18 8.23E+18 1.47E+19 2.34E+19 3.46E+19 0.5 7.69E+18 8.25E+18 1.48E+19 2.34E+19 3.47E+19 1.5 7.65E+18 8.21E+18 1.47E+19 2.33E+19 3.46E+19 2.5 7.55E+18 8.10E+18 1.46E+19 2.32E+19 3.45E+19 3.5 7.32E+18 7.86E+18 1.44E+19 2.30E+19 3.43E+19 4.5 6.67E+18 7.16E+18 1.37E+19 2.23E+19 3.36E+19 5.5 4.62E+18 4.96E+18 1.15E+19 2.01E+19 3.14E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-8 Table 8.1-6 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 150 Azimuthal Angle Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 7.05E+18 7.60E+18 1.73E+19 3.02E+19 4.70E+19

-4.5 1.OOE+19 1.08E+19 2.05E+19 3.34E+19 5.02E+19

-3.5 1.07E+19 1.15E+19 2.12E+19 3.41E+19 5.10E+19

-2.5 1.10E+19 1.19E+19 2.16E+19 3.45E+19 5.13E+19

-1.5 1.13E+19 1.21E+19 2.18E+19 3.47E+19 5.16E+19

-0.5 1.14E+19 1.23E+19 2.20E+19 3.49E+19 5.17E+19 0.0 1.14E+19 1.23E+19 2.20E+19 3.49E+19 5.17E+19 0.5 1.14E+19 1.23E+19 2.21E+19 3.49E+19 5.18E+19 1.5 1.14E+19 1.23E+19 2.20E+19 3.49E+19 5.17E+19 2.5 1.12E+19 1.21E+19 2.18E+19 3.47E+19 5.15E+19 3.5 1.09E+19 1.18E+19 2.15E+19 3.43E+19 5.12E+19 4.5 9.94E+18 1.07E+19 2.04E+19 3.33E+19 5.01E+19 5.5 6.89E+18 7.42E+18 1.71E+19 3.OOE+19 4.68E+19 Vessel Materials July 1999 Best Estimate Best Neutron Exposure Estimate Neutron of Pressure Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-9 Table 8.1-7 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 300 Azimuthal Angle Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 7.26E+18 7.86E+18 1.82E+19 3.20E+19 5.OOE+19

-4.5 1.03E+19 1.12E+19 2.15E+19 3.53E+19 5.33E+19

-3.5 1.10E+19 1.19E+19 2.23E+19 3.61E+19 5.40E+19

-2.5 1.13E+19 1.23E+19 2.27E+19 3.64E+19 5.44E+19

-1.5 1.16E+19 1.26E+19 2.29E+19 3.67E+19 5.47E+19

-0.5 1.17E+19 1.27E+19 2.31E+19 3.68E+19 5.48E+19 0.0 1.18E+19 1.27E+19 2.31E+19 3.69E+19 5.49E+19 0.5 1.18E+19 1.28E+19 2.31E+19 3.69E+19 5.49E+19 1.5 1.17E+19 1.27E+19 2.31E+19 3.68E+19 5.48E+19 2.5 1.16E+19 1.25E+19 2.29E+19 3.67E+19 5.46E+19 3.5 1.12E+19 1.22E+19 2.25E+19 3.63E+19 5.43E+19 4.5 1.02E+19 1.11E+19 2.15E+19 3.52E+19 5.32E+19 5.5 7.09E+18 7.68E+18 1.81E+19 3.18E+19 4.98E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-10 Table 8.1-8 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 45' Azimuthal Angle Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 9.OOE+18 9.74E+18 2.25E+19 3.93E+19 6.14E+19

-4.5 1.28E+19 1.38E+19 2.65E+19 4.34E+19 6.55E+19

-3.5 1.37E+19 1.48E+19 2.75E+19 4.44E+19 6.64E+19

-2.5 1.41E+19 1.52E+19 2.79E+19 4.48E+19 6.69E+19

-1.5 1.44E+19 1.56E+19 2.83E+19 4.51E+19 6.72E+19

-0.5 1.45E+19 1.57E+19 2.84E+19 4.53E+19 6.74E+19 0.0 1.46E+19 1.58E+19 2.85E+19 4.54E+19 6.74E+19 0.5 1.46E+19 1.58E+19 2.85E+19 4.54E+19 6.75E+19 1.5 1.45E+19 1.57E+19 2.85E+19 4.53E+19 6.74E+19 2.5 1.43E+19 1.55E+19 2.82E+19 4.51E+19 6.72E+19 3.5 1.39E+19 1.51E+19 2.78E+19 4.47E+19 6.67E+19 4.5 1.27E+19 1.37E+19 2.65E+19 4.33E+19 6.54E+19 5.5 8.79E+18 9.51E+18 2.22E+19 3.91E+19 6.12E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-11 Table 8.1-9 Summary of Best Estimate Iron Atom Displacement [dpal Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 0' Azimuthal Angle Distance from Cycle 11 Cycle 12 Projected Exposures 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY Core Midplane 3.69E-03 8.40E-03 1.47E-02 2.28E-02

-5.5 3.43E-03 5.23E-03 9.95E-03 1.62E-02 2.44E-02 4.5 4.87E-03 5.59E-03 1.03E-02 1.66E-02 2.47E-02

-3.5 5.21E-03 5.76E-03 1.05E-02 1.67E-02 2.49E-02

-2.5 5.37E-03 5.89E-03 1.06E-02 1.69E-02 2.50E-02

-1.5 5.49E-03 5.95E-03 1.07E-02 1.69E-02 2.51E-02

-0.5 5.55E-03 5.97E-03 1.07E-02 1.69E-02 2.51E-02 0.0 5.57E-03 5.99E-03 1.07E-02 1.70E-02 2.51E-02 0.5 5.58E-03 5.96E-03 1.07E-02 1.69E-02 2.51E-02 1.5 5.55E-03 5.88E-03 1.06E-02 1.69E-02 2.50E-02 2.5 5.48E-03 5.70E-03 1.04E-02 1.67E-02 2.49E-02 3.5 5.31E-03 5.20E-03 9.92E-03 1.62E-02 2.44E-02 4.5 4.84E-03 3.60E-03 8.32E-03 1.46E-02 2.28E-02 5.5 3.35E-03 July 1999 Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-12 Table 8.1-10 Summary of Best Estimate Iron Atom Displacement [dpal Projections for the BeItline Region of the McGuire Unit I Reactor Pressure Vessel 150 Azimuthal Angle Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 5.01E-03 5.40E-03 1.23E-02 2.14E-02 3.34E-02

-4.5 7.11E-03 7.66E-03 1.46E-02 2.37E-02 3.57E-02

-3.5 7.60E-03 8.19E-03 1.51E-02 2.42E-02 3.62E-02

-2.5 7.83E-03 8.44E-03 1.53E-02 2.45E-02 3.65E-02

-1.5 8.OOE-03 8.63E-03 1.55E-02 2.47E-02 3.66E-02

-0.5 8.09E-03 8.72E-03 1.56E-02 2.48E-02 3.67E-02 0.0 8.12E-03 8.75E-03 1.57E-02 2.48E-02 3.68E-02 0.5 8.13E-03 8.77E-03 1.57E-02 2.48E-02 3.68E-02 1.5 8.09E-03 8.73E-03 1.56E-02 2.48E-02 3.67E-02 2.5 7.99E-03 8.61E-03 1.55E-02 2.47E-02 3.66E-02 3.5 7.75E-03 8.35E-03 1.53E-02 2.44E-02 3.64E-02 4.5 7.06E-03 7.61E-03 1.45E-02 2.37E-02 3.56E-02 5.5 4.89E-03 5.28E-03 1.22E-02 2.13E-02 3.33E-02 Vessel Materials July 1999 Best Estimate Best Neutron Exposure Estimate Neutron of Pressure Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-13 Table 8.1-11 Summary of Best Estimate Iron Atom Displacement [dpal Projections for the Beltline Region of the McGuire Unit I Reactor Pressure Vessel 300 Azimuthal Angle Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 4.84E-03 5.24E-03 1.22E-02 2.13E-02 3.33E-02

-4.5 6.86E-03 7.44E-03 1.44E-02 2.35E-02 3.55E-02

-3.5 7.34E-03 7.96E-03 1.49E-02 2.40E-02 3.60E-02

-2.5 7.56E-03 8.20E-03 1.51E-02 2.43E-02 3.63E-02

-1.5 7.73E-03 8.37E-03 1.53E-02 2.45E-02 3.64E-02

-0.5 7.81E-03 8.47E-03 1.54E-02 2.45E-02 3.65E-02 0.0 7.84E-03 8.50E-03 1.54E-02 2.46E-02 3.66E-02 0.5 7.86E-03 8.51E-03 1.54E-02 2.46E-02 3.66E-02 1.5 7.82E-03 8.47E-03 1.54E-02 2.46E-02 3.65E-02 2.5 7.72E-03 8.36E-03 1.53E-02 2.44E-02 3.64E-02 3.5 7.48E-03 8.11E-03 1.50E-02 2.42E-02 3.62E-02 4.5 6.82E-03 7.39E-03 1.43E-02 2.35E-02 3.55E-02 5.5 4.73E-03 5.12E-03 1.20E-02 2.12E-02 3.32E-02 Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-14 Table 8.1-12 Summary of Best Estimate Iron Atom Displacement [dpal Projections for the BeItline Region of the McGuire Unit 1 Reactor Pressure Vessel 450 Azimuthal Angle Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 5.56E-03 6.02E-03 1.39E-02 2.43E-02 3.79E-02

-4.5 7.89E-03 8.54E-03 1.64E-02 2.68E-02 4.05E-02

-3.5 8.44E-03 9.13E-03 1.70E-02 2.74E-02 4.10E-02

-2.5 8.69E-03 9.41E-03 1.73E-02 2.77E-02 4.13E-02

-1.5 8.88E-03 9.61E-03 1.75E-02 2.79E-02 4.15E-02

-0.5 8.98E-03 9.72E-03 1.76E-02 2.80E-02 4.16E-02 0.0 9.01E-03 9.75E-03 1.76E-02 2.80E-02 4.17E-02 0.5 9.03E-03 9.77E-03 1.76E-02 2.81E-02 4.17E-02 1.5 8.98E-03 9.72E-03 1.76E-02 2.80E-02 4.16E-02 2.5 8.87E-03 9.59E-03 1.75E-02 2.79E-02 4.15E-02 3.5 8.60E-03 9.31E-03 1.72E-02 2.76E-02 4.12E-02 4.5 7.84E-03 8.48E-03 1.63E-02 2.68E-02 4.04E-02 5.5 5.43E-03 5.88E-03 1.37E-02 2.42E-02 3.78E-02 Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-15 8.2 EXPOSURE OF SPECIFIC BELTLINE MATERIALS As shown in Figure 1.1-1, the beltline region of the McGuire Unit 1 reactor pressure vessel is comprised of a series of six shell plates (3 intermediate shell and three lower shell), six longitudinal welds (3 intermediate and 3 lower), and a circumferential weld joining the two shells. The circumferential weld is centered below the axial midplane of the active core slightly below the axial location of the maximum vessel exposure; while the intermediate shell extends upward to an elevation above the active fuel and the lower shell extends downward to an elevation below the bottom of the active fuel. The maximum neutron exposure experienced by each of these beltline materials can be extracted from the data provided in Tables 8.1-1 through 8.1-12.

The current (End of Cycle 12) and projected maximum.exposures of the beltline circumferential weld, the intermediate and lower shell plates, and the lower and intermediate shell longitudinal welds are listed in Table 8.2-1 through 8.2-3. In these tables, the weld and plate exposure is expressed in terms of 1(E > 1.0 MeV), 4)(E > 0.1 MeV), and dpa.

The peak axial fluence occurs at the 450 azimuth behind the neutron pad throughout the service life of the unit on the intermediate shell plate. Longitudinal weldments are placed at discrete azimuthal angles of 0 and 30 degrees; and each is exposed to the peak in the axial exposure distribution.

In regard to the exposure of the longitudinal welds, it should be noted that due to the non symmetry of the neutron pads variability in the exposure of the 300 longitudinal welds occurs from octant to octant. With no capsule holder present, the pad span ranges from 30' to 450 in the respective octant. Likewise, pad spans of 27.50 to 450 and 250 to 450 exist in octants containing single and double surveillance capsule holders, respectively. The presence of these extended pads acts to reduce the overall neutron exposure at the 30' locations behind the edge of the pad. The data presented in Tables 8.2-1 through 8.2-3 are characteristic of an octant with a 150 pad span and, thus represent the maximum vessel exposure at all azimuthal locations.

For specific longitudinal welds located at 1200 and 3000 the tabulated data for the 300 location should be reduced by a factor of 0.80; and for welds located at 600 and 2400 the 300 data should be reduced by a factor of 0.69. Longitudinal welds positioned at the 0' azimuth are not impacted by the span of the neutron pad.

Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-16 Table 8.2-1 Fast Neutron Fluence (E > 1.0 MeV) at Key Plate and Weld Locations of McGuire Unit I Best Estimate 1)(E>1.0 MeV) [n/cm 2]

Cycle 11 Cycle 12 Projected Exposures Location 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY Intermediate Shell Plate 00 3.47E+18 3.73E+18 6.66E+18 1.06E+19 1.57E+19 150 5.11E+18 5.51E+18 9.84E+18 1.56E+19 2.31E+19 300 4.86E+18 5.27E+18 9.55E+18 1.52E+19 2.27E+19 450 5.50E+18 5.95E+18 1.07E+19 1.71E+19 2.54E+19 Intermediate Shell Longitudinal Welds (00, 1200, 2400) 00 3.47E+18 3.73E+18 6.66E+18 1.06E+19 1.57E+19 300 4.86E+18 5.27E+18 9.55E+18 1.52E+19 2.27E+19 Intermediate to Lower Shell Circumferential Weld 00 3.42E+18 3.67E+18 6.58E+18 1.04E+19 1.55E+19 150 5.04E+18 5.43E+18 9.72E+18 1.54E+19 2.28E+19 300 4.80E+18 5.20E+18 9.43E+18 1.50E+19 2.24E+19 450 5.42E+18 5.87E+18 1.06E+19 1.69E+19 2.51E+19 Lower Shell Plate 00 3.42E+18 3.67E+18 6.58E+18 1.04E+19 1.55E+19 150 5.04E+18 5.43E+18 9.72E+18 1.54E+19 2.28E+19 300 4.80E+18 5.20E+18 9.43E+18 1.50E+19 2.24E+19 450 5.42E+18 5.87E+18 1.06E+19 1.69E+19 2.51E+19 Lower Shell Longitudinal Welds (600, 1800, 3000) 00 3.42E+18 3.67E+18 6.58E+18 1.04E+19 1.55E+19 300 4.80E+18 5.20E+18 9.43E+18 1.50E+19 2.24E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-17 Table 8.2-2 Fast Neutron Fluence (E > 0.1 MeV) at Key Plate and Weld Locations of McGuire Unit 1 Best Estimate 4(E>0.1 MeV) [n/cm 2]

Cycle 11 Cycle 12 Projected Exposures Location 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY Intermediate Shell Plate 00 7.69E+18 8.25E+18 1.48E+19 2.34E+19 3.47E+19 150 1.14E+19 1.23E+19 2.21E+19 3.49E+19 5.18E+19 300 1.18E+19 1.28E+19 2.31E+19 3.69E+19 5.49E+19 450 1.46E+19 1.58E+19 2.85E+19 4.54E+19 6.75E+19 Intermediate Shell Longitudinal Welds (00, 1200, 2400) 00 7.69E+18 8.25E+18 1.48E+19 2.34E+19 3.47E+19 300 1.18E+19 1.28E+19 2.31E+19 3.69E+19 5.49E+19 Intermediate to Lower Shell Circumferential Weld 00 7.58E+18 8.14E+18 1.46E+19 2.31E+19 3.42E+19 150 1.13E+19 1.22E+19 2.18E+19 3.45E+19 5.11E+19 300 1.16E+19 1.26E+19 2.28E+19 3.64E+19 5.42E+19 450 1.44E+19 1.56E+19 2.82E+19 4.48E+19 6.66E+19 Lower Shell Plate 00 7.58E+18 8.14E+18 1.46E+19 2.31E+19 3.42E+19 150 1.13E+19 1.22E+19 2.18E+19 3.45E+19 5.11E+19 300 1.16E+19 1.26E+19 2.28E+19 3.64E+19 5.42E+19 450 1.44E+19 1.56E+19 2.82E+19 4.48E+19 6.66E+19 Lower Shell Longitudinal Welds (600, 1800, 300*)

00 7.58E+18 8.14E+18 1.46E+19 2.31E+19 3.42E+19 300 1.16E+19 1.26E+19 2.28E+19 3.64E+19 5.42E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-18 Table 8.2-3 Iron Atom Displacements [dpa] at Key Plate and Weld Locations of McGuire Unit 1 Best Estimate Iron Atom Displacements [dpa]

Cycle 11 Cycle 12 Projected Exposures Location 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY Intermediate Shell Plate 00 5.58E-03 5.99E-03 1.07E-02 1.70E-02 2.51E-02 150 8.13E-03 8.77E-03 1.57E-02 2.48E-02 3.68E-02 300 7.86E-03 8.51E-03 1.54E-02 2.46E-02 3.66E-02 450 9.03E-03 9.77E-03 1.76E-02 2.81E-02 4.17E-02 Intermediate Shell Longitudinal Welds (00, 1200, 240')

00 5.58E-03 5.99E-03 1.07E-02 1.70E-02 2.51E-02 300 7.86E-03 8.51E-03 1.54E-02 2.46E-02 3.66E-02 Intermediate to Lower Shell Circumferential Weld 00 5.50E-03 5.90E-03 1.06E-02 1.67E-02 2.48E-02 150 8.02E-03 8.65E-03 1.55E-02 2.45E-02 3.63E-02 300 7.75E-03 8.40E-03 1.52E-02 2.43E-02 3.61E-02 450 8.91E-03 9.64E-03 1.74E-02 2.77E-02 4.12E-02 Lower Shell Plate 00 5.50E-03 5.90E-03 1.06E-02 1.67E-02 2.48E-02 150 8.02E-03 8.65E-03 1.55E-02 2.45E-02 3.63E-02 300 7.75E-03 8.40E-03 1.52E-02 2.43E-02 3.61E-02 450 8.91E-03 9.64E-03 1.74E-02 2.77E-02 4.12E-02 Lower Shell Longitudinal Welds (600, 1800, 3000) 00 5.50E-03 5.90E-03 1.06E-02 1.67E-02 2.48E-02 300 7.75E-03 8.40E-03 1.52E-02 2.43E-02 3.61E-02 Best Estimate Neutron Exposure of Pressure Vessel Materials July 1999 WCAP-15253, Rev. 0

8-19 8.3 UNCERTAINTIES IN EXPOSURE PROJECTIONS The overall uncertainty in the best estimate exposure projections within the pressure vessel wall stem primarily from two sources; 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, 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 relative to the cavity data and on downcomer water density variations and vessel inner radius tolerance relative to the surveillance capsule data.

The la uncertainties associated with the bias factors applicable to O(E > 1.0 MeV),

4)(E > 0.1 MeV), and dpa are given in Section 8.1 of this report. The additional information pertinent to the required analytical uncertainty for vessel locations has been obtained from 321 benchmarking studies using the Westinghouse neutron transport methodology' and from several comparisons of power reactor internal surveillance capsule dosimetry and reactor cavity dosimetry for which the irradiation history of all sensors was the same.

Based on these benchmarking evaluations the additional uncertainty associated with the tolerances in dosimetry positioning, vessel thickness, vessel inner radius and downcomer temperature was estimated to be approximately 6% for all exposure parameters. These uncertainty components were then combined as follows:

la UNCERTAINTY D(E> 1.0 MeV) 4(E > 0.1 MeV)

Bias Factor 8.0% 7.4% 7.2%

Analytical 6.0% 6.0% 6.0%

Combined 10.0% 9.6% 9.4%

These uncertainty values are well within the 20% la uncertainty in vessel fluence projections required by the PTS rule.

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9-1 9 BUGLE-96 BEST ESTIMATE NEUTRON EXPOSURE RESULTS The reactor vessel neutron fluence exposure projections provided in Section 8.0 were evaluated with BUGLE-93 neutron transport cross sections. In conjunction with the BUGLE-93 analysis, a similar evaluation was performed utilizing the BUGLE-96 cross section library'331, which is also derived from the ENDF/B-VI data set. Both of these libraries were produced for light water reactor shielding and reactor pressure vessel dosimetry applications.

This section is provided to present the results of the BUGLE-96 evaluation as a comparison to the BUGLE-93 results. Interpretation of the BUGLE-96 results is ongoing to determine the cause of the large overprediction of the calculations compared to the best estimate in the cavity region. The data provided in this section is for information only.

9.1 COMPARISON OF BUGLE-96 CALCULATIONS WITH MEASUREMENTS As described in Section 3.3 and illustrated in Section 7, the best estimate neutron exposure projections for the McGuire Unit 1 pressure vessel were based on a combination of plant specific neutron transport calculations and plant specific measurements. Direct comparisons of the transport calculations with the McGuire Unit 1 measurement data base were used to quantify the biases that may exist due to the transport methodology, reactor modeling, and/or reactor operating characteristics over the respective irradiation periods.

In this section, comparisons of the measurement results from surveillance capsule and reactor cavity dosimetry with corresponding analytical predictions at the measurement locations are presented. These comparisons are provided on two levels. In the first instance, predictions of fast neutron exposure rates in terms of O(E > 1.0 MeV), O(E > 0.1 MeV), and dpa/sec are compared with the Best Estimate 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 best estimate/calculation (BE/C) comparisons yield an accurate plant specific bias factor that can be applied to neutron transport calculations performed for the McGuire Unit 1 reactor to produce "best estimate" exposure projections for the pressure vessel wall.

9.1.1 Comparison of Best Estimate Results with Calculation In Table 9.1-1, comparisons of best estimate and calculated exposure rates for the five surveillance capsule dosimetry sets withdrawn to date as well as for the reactor cavity midplane dosimetry sets irradiated during Cycle 12 are given. In all cases, the calculated values were based on the fuel cycle specific exposure calculations, using the BUGLE-96 library, averaged over the appropriate irradiation period. An examination of Table 9.1-1 indicates that, considering all of the available core mnidplane data, the best estimate integrated exposures were less than calculated values by factors of 0.85, 0.88, and 0.88 for 4)(E > 1.0 MeV), CD(E > 0.1 MeV),

BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

9-2 and dpa/sec, respectively. The standard deviations associated with each of the 9 sample data sets were 10.1%, 8.9%, and 8.9%, respectively.

9.1.2 Comparisons of Measured and Calculated Sensor Reaction Rates In Table 9.1-2, 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 9.1-2 shows consistent behavior for all reactions and all measurement points. The standard deviations observed for the six fast neutron reactions range from 6.2% to 16.7% on an individual reaction basis; whereas, the overall average M/C ratio for the entire data set has an associated 1y standard deviation of 14.3%. Furthermore, the average M/C bias of6 0.88 observed in the reaction rate comparisons is in excellent agreement with the values observed in the integrated exposure comparisons shown in Table 9.1-1.

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9-3 Table 9.1-1 Comparison of Best Estimate and Calculated Exposure Rates from Surveillance Capsule and Cavity Dosimetry Irradiations - BUGLE-96 Neutron Fluence (E > 1.0 MeV) [n/crn2 ]

Calculated Best Estimate BE/C Surveillance Capsules Capsule U 4.07E+18 4.20E+18 1.03 Capsule X 1.51E+19 1.30E+19 0.86 Capsule V 2.09E+19 1.84E+19 0.88 Capsule Z 2.43E+19 2.10E+19 0.86 Capsule Y 2.87E+19 2.50E+19 0.87 0.5' Cavity Cycle 12 1.52E+16 1.13E+16 0.74 14.50 Cavity Cycle 12 2.23E+16 1.75E+16 0.78 29.5° Cavity Cycle 12 2.42E+16 1.87E+16 0.77 44.50 Cavity Cycle 12 2.17E+16 1.78E+16 0.82 Average BE/C Bias Factor 0.85

% Standard Deviation (1a) 10.1%

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9-4 Table 9.1-1 Comparison of Best Estimate and Calculated Exposure Rates from (cont.) Surveillance Capsule and Cavity Dosimetry Irradiations - BUGLE-96 Neutron Fluence (E > 0.1 MeV) [n/cm2 ]

Calculated Best Estimate BE/C Surveillance Capsules Capsule U 1.85E+19 1.99E+19 1.07 Capsule X 6.86E+19 5.99E+19 0.87 Capsule V 9.22E+19 8.38E+19 0.91 Capsule Z 1.15E+20 1.03E+20 0.89 Capsule Y 1.27E+20 1.09E+20 0.86 0.50 Cavity Cycle 12 1.43E+17 1.18E+17 0.82 14.50 Cavity Cycle 12 2.22E+17 1.82E+17 0.82 29.50 Cavity Cycle 12 2.61E+17 2.14E+17 0.82 44.50 Cavity Cycle 12 2.30E+17 2.05E+17 0.89 Average BE/C Bias Factor 0.88

% Standard Deviation (1a) 8.9%

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9-5 Table 9.1-1 Comparison of Best Estimate and Calculated Exposure Rates from (cont.) Surveillance Capsule and Cavity Dosimetry Irradiations - BUGLE-96 Iron Displacements [dpa]

Calculated Best Estimate BE/C Surveillance Capsules Capsule U 7.92E-03 8.40E-03 1.06 Capsule X 2.94E-02 2.57E-02 0.87 Capsule V 4.01E-02 3.62E-02 0.90 Capsule Z 4.83E-02 4.29E-02 0.89 Capsule Y 5.51E-02 4.77E-02 0.87 0.50 Cavity Cycle 12 5.01E-05 4.01E-05 0.80 14.50 Cavity Cycle 12 7.66E-05 6.23E-05 0.81 29.50 Cavity Cycle 12 8.82E-05 7.14E-05 0.81 44.50 Cavity Cycle 12 7.73E-05 6.71E-05 0.87 Average BE/C Bias Factor 0.88

% Standard Deviation (1a) 8.9%

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9-6 Table 9.1-2 Comparison of Measured and Calculated Neutron Sensor Reaction Rates from Surveillance Capsule and Cavity Dosimetry Irradiations - BUGLE-96

"*Cu (nce) "Ti (np2) *Fe (nop) -Ni (n~p) 'U (n,f) 37Np (nf)

Surveillance Capsules Capsule U 1.10 0.98 0.99 1.11 1.16 Capsule X 1.04 0.85 0.86 1.04 0.78 Capsule V 1.08 0.86 0.87 1.02 Capsule Z 1.08 0.84 0.86 1.02 0.84 Capsule Y 0.96 0.86 0.85 1.15 0.68 0.5° Cavity Cycle 12 0.83 0.78 0.70 0.70 0.80 0.86 14.50 Cavity Cycle 12 0.89 0.90 0.77 0.78 0.84 0.82 29.50 Cavity Cycle 12 0.79 0.85 0.77 0.73 0.76 0.86 44.5' Cavity Cycle 12 0.79 0.81 0.76 0.74 0.85 1.00 Average 0.95 0.84 0.82 0.82 0.95 0.88

% Std. Dev. (la) 13.6% 6.2% 9.9% 11.1% 15.0% 16.7%

Overall M/C Average 0.88

% Std. Dev. (lo) 14.3%

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9-7 9.2 BEST ESTIMATE NEUTRON EXPOSURE OF PRESSURE VESSEL MATERIALS Similar to the BUGLE-93 best estimate evaluation described in Section 8.0 this section provides the BUGLE -96 based best estimate neutron exposure of the beltline region of the McGuire Unit 1 reactor pressure vessel through the completion of Cycle 12. Based on the continued use of the core power distributions producing the Cycles 9-12 measured results, projections of future vessel exposure to 21, 34, and 51 effective full power years of operation are also provided. In addition to the spatial mapping over the beltline region, data pertinent to the maximum exposure experienced by the intermediate and lower shell plates as well as the beltline circumferential and longitudinal welds are highlighted.

9.2.1 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:

(I)Best Est. = K DCaic.

where: *,*,. = The best estimate fast neutron exposure at the location of interest.

K The plant specific best estimate/calculation (BE/C) bias factor derived from all available surveillance capsule and reactor cavity dosimetry data.

Dcalc. = The absolute calculated fast neutron exposure at the location of interest.

From the data provided in Table 9.1-1, the plant specific bias factors (K) to be applied to the BUGLE-96 calculated exposure values were as follows:

(D(E > 1.0 MeV) 0.85 + 10.1%

(D(E > 0.1 MeV) 0.88 + 8.9%

dpa 0.88 + 8.9%

These bias factors were based on the results of the continuous monitoring program at McGuire Unit 1 that has provided measured data from five internal surveillance capsules and one reactor cavity sensor set through the first 12 cycles of operation.

9.2.2 Exposure Accrued During Cycles 1 through 12 To assess the incremental exposure resulting from the Cycles 12 irradiation, the bias factors listed in Section 9.1 were applied directly to the BUGLE-96 calculated values for the vessel clad/base metal interface to produce best estimate fluence levels characteristic of the midplane BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

9-8 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 9.2-1 through 9.2-12 to establish the exposure accrued by the reactor vessel through the end of Cycles 11 and 12.

9.2.3 Projection of Future Vessel Exposure At the end of Cycle 12, the McGuire Unit I reactor had accrued 11.20 effective full power years (EFPY) of operation. In order to establish a framework for the assessment of future vessel condition, exposure projections to 21, 34, and 51 EFPY are also included in Tables 9.2-1 through 9.2-12 in addition to the plant specific exposure assessments through the end of Cycle 12.

These extrapolations into the future were based on the.assumption that the data from the Cycles 9-12 irradiation were representative of all future fuel cycles. That is, that future fuel designs would incorporate the low leakage fuel management concept employed during Cycles 9 through 12. Examination of these projected exposure levels establishes the long term effectiveness of the low leakage fuel management incorporated to date and can be used as a guide in assessing strategies for future vessel exposure management. The validity of these projections for future operation will be confirmed via the continued cavity monitoring program.

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9-9 Table 9.2-1 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 00 Azimuthal Angle - BUGLE-96 Distance from Cycle 11 Cycle 12 Projected Exposures 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY Core Midplane 2.23E+18 5.08E+18 8.86E+18 1.38E+19

-5.5 2.07E+18 2.94E+18 3.16E+18 6.01E+18 9.79E+18 1.47E+19

-4.5 3.15E+18 3.38E+18 6.23E+18 1.OOE+19 1.50E+19

-3.5 3.24E+18 3.48E+18 6.33E+18 1.01E+19 1.51E+19

-2.5 3.31E+18 3.56E+18 6.41E+18 1.02E+19 1.51E+19

-1.5 3.35E+18 3.60E+18 6.45E+18 1.02E+19 1.52E+19

-0.5 3.36E+18 3.61E+18 6.46E+18 1.02E+19 1.52E+19 0.0 3.37E+18 3.62E+18 6.47E+18 1.02E+19 1.52E+19 0.5 3.35E+18 3.60E+18 6.45E+18 1.02E+19 1.52E+19 1.5 3.31E+18 3.55E+18 6.40E+18 1.02E+19 1.51E+19 2.5 3.21E+18 3.44E+18 6.29E+18 1.01E+19 1.50E+19 3.5 2.92E+18 3.14E+18 5.99E+18 9.77E+18 1.47E+19 4.5 2.03E+18 2.18E+18 5.03E+18 8.81E+18 1.37E+19 5.5 BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

9-10 Table 9.2-2 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit I Reactor Pressure Vessel 150 Azimuthal Angle - BUGLE-96 Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 3.07E+18 3.31E+18 7.54E+18 1.31E+19 2.05E+19

-4.5 4.35E+18 4.69E+18 8.92E+18 1.45E+19 2.19E+19

-3.5 4.66E+18 5.02E+18 9.25E+18 1.49E+19 2.22E+19

-2.5 4.80E+18 5.17E+18 9.40E+18 1.50E+19 2.23E+19

-1.5 4.90E+18 5.29E+18 9.51E+18 1.51E+19 2.25E+19

-0.5 4.96E+18 5.34E+18 9.57E+18 1.52E+19 2.25E+19 0.0 4.97E+18 5.36E+18 9.59E+18 1.52E+19 2.25E+19 0.5 4.98E+18 5.37E+18 9.60E+18 1.52E+19 2.25E+19 1.5 4.96E+18 5.35E+18 9.57E+18 1.52E+19 2.25E+19 2.5 4.89E+18 5.28E+18 9.50E+18 1.51E+19 2.24E+19 3.5 4.75E+18 5.12E+18 9.35E+18 1.50E+19 2.23E+19 4.5 4.33E+18 4.67E+18 8.89E+18 1.45E+19 2.18E+19 5.5 3.OOE+18 3.23E+18 7.46E+18 1.31E+19 2.04E+19 July 1999 BUGLE-96 Best BUGLE-96 Exposure Results Estimate Exposure Best Estimate Results July 1999 WCAP-15253, Rev. 0

9-11 Table 9.2-3 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit I Reactor Pressure Vessel 300 Azimuthal Angle - BUGLE-96 Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY 2.89E+18 3.13E+18 7.26E+18 1.27E+19 1.99E+19

-5.5 4.10E+18 4.44E+18 8.57E+18 1.41E+19 2.12E+19

-4.5 4.38E+18 4.75E+18 8.88E+18 1.44E+19 2.15E+19

-3.5 4.52E+18 4.90E+18 9.03E+18 1.45E+19 2.17E+19

-2.5 4.62E+18 5.OOE+18 9.13E+18 1.46E+19 2.18E+19

-1.5 4.67E+18 5.06E+18 9.19E+18 1.47E+19 2.18E+19

-0.5 4.68E+18 5.08E+18 9.21E+18 1.47E+19 2.18E+19 0.0 4.69E+18 5.08E+18 9.22E+18 1.47E+19 2.19E+19 0.5 4.67E+18 5.06E+18 9.19E+18 1.47E+19 2.18E+19 1.5 4.61E+18 4.99E+18 9.12E+18 1.46E+19 2.18E+19 2.5 4.47E+18 4.84E+18 8.97E+18 1.45E+19 2.16E+19 3.5 4.07E+18 4.42E+18 8.55E+18 1.40E+19 2.12E+19 4.5 2.82E+18 3.06E+18 7.19E+18 1.27E+19 1.98E+19 5.5 BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

9-12 Table 9.2-4 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 450 Azimuthal Angle - BUGLE-96 Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 3.28E+18 3.55E+18 8.20E+18 1.44E+19 2.24E+19

-4.5 4.66E+18 5.04E+18 9.68E+18 1.58E+19 2.39E+19

-3.5 4.98E+18 5.39E+18 1.OOE+19 1.62E+19 2.42E+19

-2.5 5.13E+18 5.55E+18 1.02E+19 1.64E+19 2.44E+19

-1.5 5.24E+18 5.68E+18 1.03E+19 1.65E+19 2.45E+19

-0.5 5.30E+18 5.74E+18 1.04E+19 1.65E+19 2.46E+19 0.0 5.32E+18 5.76E+18 1.04E+19 1.66E+19 2.46E+19 0.5 5.33E+18 5.77E+18 1.04E+19 1.66E+19 2.46E+19 1.5 5.31E+18 5.74E+18 1.04E+19 1.65E+19 2.46E+19 2.5 5.24E+18 5.67E+18 1.03E+19 1.65E+19 2.45E+19 3.5 5.08E+18 5.50E+18 1.01E+19 1.63E+19 2.43E+19 4.5 4.63E+18 5.01E+18 9.65E+18 1.58E+19 2.39E+19 5.5 3.21E+18 3.47E+18 8.11E+18 1.43E+19 2.23E+19 BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

9-13 Table 9.2-5 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 00 Azimuthal Angle - BUGLE-96 Cycle 11 Cycle 12 Projected Exposures Distance from 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY Core Midplane 10.21 EFPY 1.10E+19 1.92E+19 2.99E+19

-5.5 4.49E+18 4.82E+18 1.30E+19 2.12E+19 3.19E+19

-4.5 6.37E+18 6.84E+18 1.35E+19 2.17E+19 3.24E+19

-3.5 6.81E+18 7.31E+18 1.37E+19 2.19E+19 3.26E+19

-2.5 7.02E+18 7.53E+18 1.39E+19 2.20E+19 3.27E+19

-1.5 7.17E+18 7.70E+18 1.40E+19 2.21E+19 3.28E+19

-0.5 7.25E+18 7.78E+18 1.40E+19 2.22E+19 3.29E+19 0.0 7.28E+18 7.81E+18 1.40E+19 2.22E+19 3.29E+19 0.5 7.29E+18 7.82E+18 1.40E+19 2.21E+19 3.28E+19 1.5 7.25E+18 7.79E+18 1.39E+19 2.20E+19 3.27E+19 2.5 7.16E+18 7.68E+18 1.36E+19 2.18E+19 3.25E+19 3.5 6.94E+18 7.45E+18 1.30E+19 2.11E+19 3.18E+19 4.5 6.33E+18 6.79E+18 1.09E+19 1.91E+19 2.98E+19 5.5 4.39E+18 4.71E+18 July 1999 BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

9-14 Table 9.2-6 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 150 Azimuthal Angle - BUGLE-96 Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 6.71E+18 7.23E+18 1.65E+19 2.87E+19 4.48E+19

-4.5 9.52E+18 1.03E+19 1.95E+19 3.18E+19 4.78E+19

-3.5 1.02E+19 1.10E+19 2.02E+19 3.25E+19 4.85E+19

-2.5 1.05E+19 1.13E+19 2.06E+19 3.28E+19 4.88E+19

-1.5 1.07E+19 1.16E+19 2.08E+19 3.31E+19 4.91E+19

-0.5 1.08E+19 1.17E+19 2.09E+19 3.32E+19 4.92E+19 0.0 1.09E+19 1.17E+19 2.10E+19 3.32E+19 4.93E+19 0.5 1.09E+19 1.17E+19 2.10E+19 3.33E+19 4.93E+19 1.5 1.08E+19 1.17E+19 2.09E+19 3.32E+19 4.92E+19 2.5 1.07E+19 1.15E+19 2.08E+19 3.30E+19 4.91E+19 3.5 1.04E+19 1.12E+19 2.04E+19 3.27E+19 4.87E+19 4.5 9.46E+18 1.02E+19 1.94E+19 3.17E+19 4.77E+19 5.5 6.55E+18 7.07E+18 1.63E+19 2.86E+19 4.46E+19 BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

9-15 Table 9.2-7 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 300 Azimuthal Angle - BUGLE-96 Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 6.85E+18 7.42E+18 1.72E+19 3.02E+19 4.72E+19

-4.5 9.72E+18 1.05E+19 2.03E+19 3.33E+19 5.03E+19

-3.5 1.04E+19 1.13E+19 2.11E+19 3.40E+19 5.10E+19

-2.5 1.07E+19 1.16E+19 2.14E+19 3.44E+19 5.14E+19

-1.5 1.09E+19 1.19E+19 2.16E+19 3.46E+19 5.16E+19

-0.5 1.11E+19 1.20E+19 2.18E+19 3.48E+19 5.17E+19 0.0 1.11E+19 1.20E+19 2.18E+19 3.48E+19 5.18E+19 0.5 1.11E+19 1.21E+19 2.18E+19 3.48E+19 5.18E+19 1.5 1.11E+19 1.20E+19 2.18E+19 3.48E+19 5.17E+19 2.5 1.09E+19 1.18E+19 2.16E+19 3.46E+19 5.16E+19 3.5 1.06E+19 1.15E+19 2.13E+19 3.43E+19 5.12E+19 4.5 9.66E+18 1.05E+19 2.03E+19 3.32E+19 5.02E+19 5.5 6.69E+18 7.25E+18 1.70E+19 3.OOE+19 4.70E+19 BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

9-16 Table 9.2-8 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 450 Azimuthal Angle - BUGLE-96 Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 8.52E+18 9.22E+18 2.13E+19 3.72E+19 5.81E+19

-4.5 1.21E+19 1.31E+19 2.51E+19 4.11E+19 6.20E+19

-3.5 1.29E+19 1.40E+19 2.60E+19 4.20E+19 6.29E+19

-2.5 1.33E+19 1.44E+19 2.65E+19 4.24E+19 6.33E+19

-1.5 1.36E+19 1.47E+19 2.68E+19 4.27E+19 6.36E+19

-0.5 1.38E+19 1.49E+19 2.69E+19 4.29E+19 6.38E+19 0.0 1.38E+19 1.49E+19 2.70E+19 4.30E+19 6.39E+19 0.5 1.38E+19 1.50E+19 2.70E+19 4.30E+19 6.39E+19 1.5 1.38E+19 1.49E+19 2.69E+19 4.29E+19 6.38E+19 2.5 1.36E+19 1.47E+19 2.67E+19 4.27E+19 6.36E+19 3.5 1.32E+19 1.43E+19 2.63E+19 4.23E+19 6.32E+19 4.5 1.20E+19 1.30E+19 2.50E+19 4.10E+19 6.19E+19 5.5 8.32E+18 9.01E+18 2.11E+19 3.70E+19 5.79E+19 July 1999 BUGLE-96 Best BUGLE-96 Exposure Results Estimate Exposure Best Estimate Results July 1999 WCAP-15253, Rev. 0

9-17 Table 9.2-9 Summary of Best Estimate Iron Atom Displacement [dpal Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 00 Azimuthal Angle - BUGLE-96 Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY 3.29E-03 3.53E-03 8.04E-03 1.40E-02 2.19E-02

-5.5 4.66E-03 5.01E-03 9.52E-03 1.55E-02 2.33E-02

-4.5 4.99E-03 5.35E-03 9.87E-03 1.59E-02 2.37E-02

-3.5 5.14E-03 5.52E-03 1.OOE-02 1.60E-02 2.39E-02

-2.5 5.25E-03 5.64E-03 1.02E-02 1.61E-02 2.40E-02

-1.5 5.31E-03 5.70E-03 1.02E-02 1.62E-02 2.40E-02

-0.5 5.33E-03 5.72E-03 1.02E-02 1.62E-02 2.41E-02 0.0 5.34E-03 5.73E-03 1.02E-02 1.62E-02 2.41E-02 0.5 5.31E-03 5.70E-03 1.02E-02 1.62E-02 2.40E-02 1.5 5.24E-03 5.63E-03 1.01E-02 1.61E-02 2.40E-02 2.5 5.08E-03 5.46E-03 9.97E-03 1.60E-02 2.38E-02 3.5 4.63E-03 4.97E-03 9.49E-03 1.55E-02 2.33E-02 4.5 5.5 3.21E-03 3.45E-03 7.96E-03 1.40E-02 2.18E-02 BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

9-18 Table 9.2-10 Summary of Best Estimate Iron Atom Displacement [dpal Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 150 Azimuthal Angle - BUGLE-96 Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 4.81E-03 5.19E-03 1.18E-02 2.06E-02 3.21E-02

-4.5 6.83E-03 7.36E-03 1.40E-02 2.28E-02 3.43E-02

-3.5 7.30E-03 7.88E-03 1.45E-02 2.33E-02 3.48E-02

-2.5 7.52E-03 8.11E-03 1.47E-02 2.35E-02 3.50E-02

-1.5 7.69E-03 8.29E-03 1.49E-02 2.37E-02 3.52E-02

-0.5 7.77E-03 8.38E-03 1.50E-02 2.38E-02 3.53E-02 0.0 7.80E-03 8.41E-03 1.50E-02 2.38E-02 3.53E-02 0.5 7.82E-03 8.43E-03 1.51E-02 2.39E-02 3.53E-02 1.5 7.78E-03 8.38E-03 1.50E-02 2.38E-02 3.53E-02 2.5 7.67E-03 8.28E-03 1.49E-02 2.37E-02 3.52E-02 3.5 7.44E-03 8.03E-03 1.47E-02 2.35E-02 3.49E-02 4.5 6.79E-03 7.32E-03 1.39E-02 2.27E-02 3.42E-02 5.5 4.70E-03 5.07E-03 1.17E-02 2.05E-02 3.20E-02 BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

9-19 Table 9.2-11 Summary of Best Estimate Iron Atom Displacement [dpa] Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 300 Azimuthal Angle - BUGLE-96 Distance from Cycle 11 Cycle 12 Projected Exposures 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY Core Midplane 10.21 EFPY 1.16E-02 2.03E-02 3.17E-02

-5.5 4.60E-03 4.98E-03 1.36E-02 2.24E-02 3.38E-02

-4.5 6.52E-03 7.07E-03 1.41E-02 2.29E-02 3.42E-02

-3.5 6.98E-03 7.56E-03 1.44E-02, 2.31E-02 3.45E-02

-2.5 7.19E-03 7.79E-03 1.45E-02 2.32E-02 3.46E-02

-1.5 7.34E-03 7.96E-03 1.46E-02 2.33E-02 3.47E-02

-0.5 7.43E-03 8.05E-03 1.46E-02 2.34E-02 3.48E-02 0.0 7.45E-03 8.08E-03 1.47E-02 2.34E-02 3.48E-02 0.5 7.47E-03 8.09E-03 1.46E-02 2.33E-02 3.47E-02 1.5 7.43E-03 8.05E-03 1.45E-02 2.32E-02 3.46E-02 2.5 7.33E-03 7.95E-03 1.43E-02 2.30E-02 3.44E-02 3.5 7.11E-03 7.71E-03 1.36E-02 2.23E-02 3.37E-02 4.5 6.48E-03 7.03E-03 1.14E-02 2.02E-02 3.16E-02 5.5 4.49E-03 4.87E-03 July 1' BUGLE-96 Best Estimate Exposure Results July e.0 WCAP-15253, Rev. 0

9-20 Table 9.2-12 Summary of Best Estimate Iron Atom Displacement [dpal Projections for the Beltline Region of the McGuire Unit 1 Reactor Pressure Vessel 45' Azimuthal Angle - BUGLE-96 Distance from Cycle 11 Cycle 12 Projected Exposures Core Midplane 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY

-5.5 5.30E-03 5.74E-03 1.32E-02 2.32E-02 3.62E-02

-4.5 7.53E-03 8.14E-03 1.56E-02 2.56E-02 3.86E-02

-3.5 8.05E-03 8.71E-03 1.62E-02 2.62E-02 3.92E-02

-2.5 8.29E-03 8.97E-03 1.65E-02 2.64E-02 3.94E-02

-1.5 8.47E-03 9.17E-03 1.67E-02 2.66E-02 3.96E-02

-0.5 8.57E-03 9.27E-03 1.68E-02 2.67E-02 3.97E-02 0.0 8.60E-03 9.30E-03 1.68E-02 2.68E-02 3.98E-02 0.5 8.61E-03 9.32E-03 1.68E-02 2.68E-02 3.98E-02 1.5 8.57E-03 9.27E-03 1.68E-02 2.67E-02 3.97E-02 2.5 8.46E-03 9.15E-03 1.67E-02 2.66E-02 3.96E-02 3.5 8.20E-03 8.88E-03 1.64E-02 2.63E-02 3.93E-02 4.5 7.48E-03 8.09E-03 1.56E-02 2.55E-02 3.85E-02 5.5 5.18E-03 5.61E-03 1.31E-02 2.31E-02 3.61E-02 BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

9-21 9.3 EXPOSURE OF SPECIFIC BELTLINE MATERIALS As shown in Figure 1.1-1, the beltline region of the McGuire Unit 1 reactor pressure vessel is comprised of a series of six shell plates (3 intermediate shell and three lower shell), six longitudinal welds (3 intermediate and 3 lower), and a circumferential weld joining the two shells. The circumferential weld is centered below the axial midplane of the active core slightly below the axial location of the maximum vessel exposure; while the intermediate shell extends upward to an elevation above the active fuel and the lower shell extends downward to an elevation below the bottom of the active fuel. The maximum neutron exposure experienced by each of these beltline materials can be extracted from the data provided in Tables 9.2-1 through 9.2-12.

The current (End of Cycle 12) and projected maximum exposures of the beltline circumferential weld, the intermediate and lower shell plates, and the lower and intermediate shell longitudinal welds are listed in Table 9.3-1 through 9.3-3. In these tables, the weld and plate exposure is expressed in terms of 4(E > 1.0 MeV), 1(E > 0.1 MeV), and dpa.

The peak axial fluence occurs at the 450 azimuth behind the neutron pad throughout the service life of the unit on the intermediate shell plate. Longitudinal weldments are placed at discrete azimuthal angles of 0 and 30 degrees; and each is exposed to the peak in the axial exposure distribution.

In regard to the exposure of the longitudinal welds, it should be noted that due to the non symmetry of the neutron pads variability in the exposure of the 300 longitudinal welds occurs from octant to octant. With no capsule holder present, the pad span ranges from 30' to 45' in the respective octant. Likewise, pad spans of 27.50 to 45' and 250 to 450 exist in octants containing single and double surveillance capsule holders, respectively. The presence of these extended pads acts to reduce the overall neutron exposure at the 30' locations behind the edge of the pad. The data presented in Tables 9.3-1 through 9.3-3 are characteristic of an octant with a 150 pad span and, thus represent the maximum vessel exposure at all azimuthal locations.

For specific longitudinal welds located at 1200 and 300' the tabulated data for the 30' location should be reduced by a factor of 0.80; and for welds located at 600 and 2400 the 30' data should be reduced by a factor of 0.69. Longitudinal welds positioned at the 00 azimuth are not impacted by the span of the neutron pad.

BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

9-22 Table 9.3-1 Fast Neutron Fluence (E > 1.0 MeV) at Key Plate and Weld Locations of McGuire Unit 1 - BUGLE-96 Best Estimate 'D(E>1.0 MeV) In/cm 2I Cycle 11 Cycle 12 Projected Exposures Location 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY Intermediate Shell Plate 00 3.37E+18 3.62E+18 6.47E+18 1.02E+19 1.52E+19 150 4.98E+18 5.37E+18 9.60E+18 1.52E+19 2.25E+19 300 4.69E+18 5.08E+18 9.22E+18 1.47E+19 2.19E+19 450 5.33E+18 5.77E+18 1.04E+19 1.66E+19 2.46E+19 Intermediate Shell Longitudinal Welds (00, 1200, 2400) 00 3.37E+18 3.62E+18 6.47E+18 1.02E+19 1.52E+19 300 4.69E+18 5.08E+18 9.22E+18 1.47E+19 2.19E+19 Intermediate to Lower Shell Circumferential Weld 00 3.32E+18 3.57E+18 6.38E+18 1.01E+19 1.50E+19 150 4.91E+18 5.30E+18 9.48E+18 1.50E+19 2.23E+19 300 4.63E+18 5.01E+18 9.10E+18 1.45E+19 2.16E+19 450 5.26E+18 5.69E+18 1.03E+19 1.64E+19 2.43E+19 Lower Shell Plate 00 3.32E+18 3.57E+18 6.38E+18 1.01E+19 1.50E+19 150 4.91E+18 5.30E+18 9.48E+18 1.50E+19 2.23E+19 300 4.63E+18 5.01E+18 9.10E+18 1.45E+19 2.16E+19 450 5.26E+18 5.69E+18 1.03E+19 1.64E+19 2.43E+19 Lower Shell Longitudinal Welds (600, 180', 300')

00 3.32E+18 3.57E+18 6.38E+18 1.01E+19 1.50E+19 300 4.63E+18 5.01E+18 9.10E+18 1.45E+19 2.16E+19 BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

9-23 Table 9.3-2 Fast Neutron Fluence (E > 0.1 MeV) at Key Plate and Weld Locations of McGuire Unit 1 - BUGLE-96 Best Estimate 4D(E>0.1 MeV) [n/cm2]

Cycle 11 Cycle 12 Projected Exposures Location 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY Intermediate Shell Plate 00 7.29E+18 7.82E+18 1.40E+19 2.22E+19 3.29E+19 150 1.09E+19 1.17E+19 2.10E+19 3.33E+19 4.93E+19 300 1.11E+19 1.21E+19 2.18E+19 3.48E+19 5.18E+19 450 1.38E+19 1.50E+19 2.70E+19 4.30E+19 6.39E+19 Intermediate Shell Longitudinal Welds (00, 1200, 2400) 00 7.29E+18 7.82E+18 1.40E+19 2.22E+19 3.29E+19 300 1.11E+19 1.21E+19 2.18E+19 3.48E+19 5.18E+19 Intermediate to Lower Shell Circumferential Weld 00 7.19E+18 7.72E+18 1.38E+19 2.19E+19 3.25E+19 150 1.07E+19 1.16E+19 2.07E+19 3.28E+19 4.87E+19 300 1.10E+19 1.19E+19 2.16E+19 3.44E+19 5.12E+19 450 1.36E+19 1.48E+19 2.67E+19 4.25E+19 6.31E+19 Lower Shell Plate 00 7.19E+18 7.72E+18 1.38E+19 2.19E+19 3.25E+19 150 1.07E+19 1.16E+19 2.07E+19 3.28E+19 4.87E+19 300 1.10E+19 1.19E+19 2.16E+19 3.44E+19 5.12E+19 450 1.36E+19 1.48E+19 2.67E+19 4.25E+19 6.31E+19 Lower Shell Longitudinal Welds (600, 1800, 3000) 00 7.19E+18 7.72E+18 1.38E+19 2.19E+19 3.25E+19 300 1.10E+19 1.19E+19 2.16E+19 3.44E+19 5.12E+19 BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

9-24 Table 9.3-3 Iron Atom Displacements [dpal at Key Plate and Weld Locations of McGuire Unit 1 - BUGLE-96 Best Estimate Iron Atom Displacements Idpal Cycle 11 Cycle 12 Projected Exposures Location 10.21 EFPY 11.20 EFPY 21 EFPY 34 EFPY 51 EFPY Intermediate Shell Plate 00 6.09E-03 6.54E-03 1.17E-02 1.85E-02 2.75E-02 150 8.92E-03 9.62E-03 1.72E-02 2.72E-02 4.04E-02 300 8.53E-03 9.24E-03 1.67E-02 2.67E-02 3.97E-02 450 9.83E-03 1.06E-02 1.92E-02 3.06E-02 4.54E-02 Intermediate Shell Longitudinal Welds (00, 1200, 2400) 00 6.09E-03 6.54E-03 1.17E-02 1.85E-02 2.75E-02 300 8.53E-03 9.24E-03 1.67E-02 2.67E-02 3.97E-02 Intermediate to Lower Shell Circumferential Weld 00 6.01E-03 6.45E-03 1.15E-02 1.83E-02 2.71E-02 150 8.80E-03 9.49E-03 1.70E-02 2.69E-02 3.99E-02 300 8.41E-03 9.11E-03 1.65E-02 2.64E-02 3.92E-02 450 9.70E-03 1.05E-02 1.90E-02 3.02E-02 4.49E-02 Lower Shell Plate 00 6.01E-03 6.45E-03 1.15E-02 1.83E-02 2.71E-02 150 8.80E-03 9.49E-03 1.70E-02 2.69E-02 3.99E-02 300 8.41E-03 9.11E-03 1.65E-02 2.64E-02 3.92E-02 450 9.70E-03 1.05E-02 1.90E-02 3.02E-02 4.49E-02 Lower Shell Longitudinal Welds (600, 1800, 3000) 00 6.01E-03 6.45E-03 1.15E-02 1.83E-02 2.71E-02 300 8.41E-03 9.11E-03 1.65E-02 2.64E-02 3.92E-02 BUGLE-96 Best Estimate Exposure Results July 1999 WCAP-15253, Rev. 0

10-1 10 REFERENCES

1. "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Curves," WCAP-14040-NP-A, January 1996.

2. "Duke Power Company William B. McGuire Unit 1 Reactor Vessel Radiation Surveillance Program", WCAP-9195, November 1977.

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

4. RSIC Data Library Collection DLC-175, "BUGLE-93, Production and Testing of the VITAMIN-B6 Fine Group and the BUGLE-93 Broad Group Neutron/Photon Cross Section Libraries Derived from ENDF/B-VI Nuclear Data", April 1994.

5. Maerker, R. E., et. al., "Accounting for Changing Source Distributions in Light Water Reactor Surveillance Dosimetry Analysis", Nuclear Science and Engineering, Volume 94, pp 291-308, 1986.

6. ASTM Designation E706-87 (Re-approved 1994), "Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1998.

7. ASTM Designation E853-87 (Re-approved 1995), "Practice for Analysis and Interpretation of Light -Water Reactor Surveillance Results," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1998.

8. ASTM Designation E261-98, "Practice for Determining Neutron Fluence Rate, Fluence, and Spectra by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1998.

9. ASTM Designation E262-97, "Test Method for Determining Thermal Neutron Reaction and Fluence Rates by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1998.

10. ASTM Designation E263-93, "Test Method for Determining Fast Neutron Reaction Rates by Radioactivation of Iron," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1998.

11. ASTM Designation E264-92 (Re-approved 1996), "Test Method for Determining Fast Neutron Reaction Rates by Radioactivation of Nickel," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1998.

12. ASTM Designation E481-97, "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, 1998.

13. ASTM Designation E523-92 (Re-approved 1996), "Test Method for Determining Fast Neutron Reaction Rates by Radioactivation of Copper," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1998.

References July 1999 WCAP-15253, Rev. 0

10-2

14. ASTM Designation E704-96, "Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1998.

15. ASTM Designation E705-96, "Test Method for Measuring Reaction Rates by Radioactivation of Neptunium-237," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1998.

16. ASTM Designation E1005-97, "Test 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, 1998.

17. Electronic mail transmissions of McGuire Unit 1 reactor power history dated September 24,1998 and January 12 1999, from R. A. Williams of Duke Power Company to A. H. Fero of Westinghouse.

18. J.M. Adams et al, "The Materials Dosimetry Reference Facility Round Robin Tests of 37Np and `*U Fissionable Dosimeters," pp. 124-129, Proceedingsof the 9"h International Symposium on Reactor Dosimetry, Prague, Czech Republic, 2-6 September 1996.

19. Schn-ittroth, E. A., "FERRET Data Analysis Code", HEDL-TME-79-40, Hanford Engineering Development Laboratory, Richland, Washington, September 1979.

20. 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 Laborator3y Kirkland AFB, NM, July 1967.

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

22. Maerker, R. E. as reported by Stallman, F. W., "Workshop on Adjustment Codes and Uncertainties - Proc. of the 4th ASTM/EURATOM Symposium on Reactor Dosimetry,"

NUREG/CP-0029, NRC, Washington, D.C., July 1982.

23. "The Nuclear Design of the W. B. McGuire Unit 1 Nuclear Power Plant - Cycle 1",

WCAP-9323-R1, August 1978. [Westinghouse Proprietary Class 2]

24. "The Nuclear Design of the W. B. McGuire Unit 1 Nuclear Power Plant - Cycle 2",

WCAP-10463, January 1984. [Westinghouse Proprietary Class 2]

25. "The Nuclear Design of the W. B. McGuire Unit 1 Nuclear Power Plant - Cycle 3",

WCAP-10782, February 1985. [Westinghouse Proprietary Class 2]

26. "The Nuclear Design of the W. B. McGuire Unit 1 Nuclear Power Plant - Cycle 4",

WCAP-11141, May 1986. [Westinghouse Proprietary Class 2]

27. "The Nuclear Design of the W. B. McGuire Unit 1 Nuclear Power Plant - Cycle 5",

WCAP-11589, October 1987. [Westinghouse Proprietary Class 2]

28. "The Nuclear Design of the W. B. McGuire Unit 1 Nuclear Power Plant - Cycle 6",

WCAP-12044, November 1988. [Westinghouse Proprietary Class 2]

References July 1999 WCAP-15253, Rev. 0

10-3

29. "The Nuclear Design of the W. B. McGuire Unit 1 Nuclear Power Plant - Cycle 7",

WCAP-12544, April 1990. [Westinghouse Proprietary Class 2]

30. D. E. Bortz, memorandum from Duke Power Company (DPC) to S. Zorichak (Westinghouse) transmitting the McGuire Unit 1 Cycles 8 through 11 core inventory, average assembly burnups, and average axial power distributions, March 31, 1997. [DPC Proprietary Information]

31. D. E. Bortz, memorandum from Duke Power Company (DPC) to A. H. Fero (Westinghouse) transmitting the McGuire Unit 1 Cycle 12 core inventory, average assembly burnups, and average axial power distributions, October, 2, 1998. [DPC Proprietary Information]

32. "Westinghouse Fast Neutron Exposure Methodology for Pressure Vessel Fluence Determination and Dosimetry Evaluation," WCAP-13362, May 1992. [Westinghouse Proprietary Class 2]

33. RSIC Data Library Collection DLC-185, "BUGLE-96, Coupled 47 Neutron, 20 Gamma Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications", March 1996.

References July 1999 WCAP-15253, Rev. 0

A-1 APPENDIX A SPECIFIC ACTIVrITES AND IRRADIATION HISTORY OF SENSORS FROM SURVEILLANCE CAPSULES U, X, V, Z, AND Y In this appendix, the irradiation history, as extracted from NUREG-0020 and reference 17, and the measured specific activities of radiometric sensors irradiated in Surveillance Capsules U, X, V, Z, and Y are provided.

The startup and shutdown dates for each fuel cycle comprising the irradiation history of the surveillance capsules were as follows:

Cycle Startup Shutdown 1 10/01/81 02/24/84 Capsule U Withdrawn 2 05/05/84 04/19/85 3 06/27/85 05/16/86 4 09/14/86 09/04/87 5 11/14/87 10/12/88 Capsule X Withdrawn 6 12/31/88 01/15/90 7 05/20/90 09/20/91 8 12/10/91 03/12/93 Capsules V and Z Withdrawn 9 06/14/93 08/19/94 10 10/27/94 12/14/95 11 01/25/96 02/14/97 Capsule Y Withdrawn The detailed operating history of the reactor over the course of these eleven fuel cycles is provided in Table A-1. Note that the reference full power for McGuire Unit 1 is 3411 MWt.

The measured specific activities of the monitors removed from Capsules U, X, V, Z, and Y are provided in Tables A-2 through A-6, respectively. The locations of the various monitors within the capsule may be obtained from reference 2.

July 1999 Appendix A Appendix A July 1999 WCAP-15253, Rev. 0

A-2 Table A-1 McGuire Unit I Operating History - Cycles 1 Through 11 Cycle 1 Cycle 2 Cycle 3 Cycle 4 Date MWt-hr Date MWt-hr Date MWt-hr Date MWt-hr Oct-81 222129 Mar-84 0 May-85 0 Jun-86 0 Nov-81 791796 Apr-84 0 Jun-85 117845 Jul-86 0 Dec-81 84955 May-84 1994665 Jul-85 2344677 Aug-86 0 Jan-82 1273713 Jun-84 2266395 Aug-85 2533610 Sep-86 944223 Feb-82 1054504 Jul-84 2319518 Sep-85 2329794 Oct-86 2293046 Mar-82 530357 Aug-84 2370549 Oct-85 2533965 Nov-86 44556 Apr-82 1192627 Sep-84 2480184 Nov-85 1540353 Dec-86 2516554 May-82 1484279 Oct-84 2042345 Dec-85 2425449 Jan-87 2534776 Jun-82 1266005 Nov-84 1553094 Jan-86 2471699 Feb-87 1861001 Jul-82 561572 Dec-84 226629 Feb-86 2022499 Mar-87 2537575 Aug-82 1467549 Jan-85 2356943 Mar-86 2089263 Apr-87 2287291 Sep-82 1454192 Feb-85 2088791 Apr-86 2204598 May-87 2495567 Oct-82 1338949 Mar-85 1694129 May-86 1001197 Jun-87 2441831 Nov-82 527736 Apr-85 596719 Jul-87 2534622 Dec-82 1265980 Aug-87 1847703 Jan-83 857172 Sep-87 221936 Feb-83 0 Mar-83 0 Apr-83 0 May-83 187441 Jun-83 1953836 Jul-83 1975352 Aug-83 1129545 Sep-83 2260850 Oct-83 1935163 Nov-83 1623905 Dec-83 2011387 Jan-84 2222932 Feb-84 1889758 Th. Gen. 3.256E+07 2.199E+07 2.361E+07 2.456E+07 MWt-hr Cy. Length 3.437E+07 2.321E+07 2.492E+07 2.592E+07 EFPS Appendix A July 1999 WCAP-15253, Rev. 0

A-3 Table A-1 McGuire Unit I Operating History - Cycles 1 Through 11 (cont.)

Cycle 5 Cycle 6 Cycle 7 Cycle 8 Date MWt-hr Date MWt-hr Date MWt-hr Date MWt-hr Oct-87 0 Nov-88 0 Feb-90 0 Oct-91 0 Nov-87 1101430 Dec-88 9881 Mar-90 0 Nov-91 0 Dec-87 2251637 Jan-89 2245259 Apr-90 0 Dec-91 1566100 Jan-88 2379899 Feb-89 2286701 May-90 521935 Jan-92 1277016 Feb-88 2314778 Mar-89 569136 Jun-90 2156153 Feb-92 752634 Mar-88 2380942 Apr-89 0 Jul-90 2494895 Mar-92 2438606 Apr-88 2314604 May-89 1664673 Aug-90 2124989 Apr-92 2390345 May-88 2528964 Jun-89 2412733 Sep-90 2200794 May-92 305612 Jun-88 2281133 Jul-89 2368074 Oct-90 1063172 Jun-92 880274 Jul-88 2485619 Aug-89 2300333 Nov-90 1146029 Jul-92 2285861 Aug-88 2522007 Sep-89 2425699 Dec-90 2518560 Aug-92 2514974 Sep-88 2437962 Oct-89 2518039 Jan-91 2519888 Sep-92 2450774 Oct-88 898152 Nov-89 2456058 Feb-91 1818751 Oct-92 2514483 Dec-89 2530503 Mar-91 2521731 Nov-92 2451002 Jan-90 608302 Apr-91 1993480 Dec-92 2526799 May-91 1274443 Jan-93 2543424 Jun-91 2446944 Feb-93 2288198 Jul-91 2516832 Mar-93 904384 Aug-91 2534664 Sep-91 1557266 Th. Gen. 2.590E+07 2.440E+07 3.341E+07 3.009E+07 MWt-hr Cy. Length 2.733E+07 2.575E+07 3.526E+07 3.176E+07 EFPS Appendix A July 1999 WCAP-15253, Rev. 0

A-4 Table A-1 McGuire Unit 1 Operating History - Cycles 1 Through 11 (cont.)

Cycle 9 Cycle 10 Cycle 11 Date MWt-hr Date MWt-hr Date MWt-hr Apr-93 0 Sep-94 0 Jan-96 408518 May-93 0 Oct-94 262464 Feb-96 2060351 Jun-93 1020090 Nov-94 2451277 Mar-96 2533375 Jul-93 2531661 Dec-94 2532823 Apr-96 2449220 Aug-93 1726608 Jan-95 2373265 May-96 2529255 Sep-93 0 Feb-95 2275482 Jun-96 2382373 Oct-93 1065271 Mar-95 2533681 Jul-96 2532247 Nov-93 2338426 Apr-95 2448229 Aug-96 2533802 Dec-93 2532733 May-95 2533705 Sep-96 2452128 Jan-94 1795522 Jun-95 2255438 Oct-96 2451562 Feb-94 298968 Jul-95 2124395 Nov-96 1363900 Mar-94 2426383 Aug-95 2487089 Dec-96 2369229 Apr-94 2437007 Sep-95 2124775 Jan-97 2428964 May-94 2454102 Oct-95 2343625 Feb-97 1058748 Jun-94 2447588 Nov-95 2443794 Jul-94 2533830 Dec-95 988198 Aug-94 1457626 Th. Gen. 2.707E+07 3.218E+07 2.955E+07 MWt-hr Cy. Length 2.857E+07 3.396E+07 3.119E+07 EFPS Appendix A July 1999 WCAP-15253, Rev. 0

A-5 Table A-2 Radiometric Counting Results For Sensors Removed From Capsule U EANERGYSYSEMS INLS~ GNTN CHMCLNLSI sAr ETNGHuADVANCED CHMIALANLSI WRTWEANALYOUUE LAMVANCEDEESYAAL SYSWFaq N IWALTZ MXLL SXTE 1117 1 4 Si-rAJ Ar~.bEJ-51SOYNe fFE~fmaMET"M ANALYST I ANALYST IE0C GAMMA~ SPECTO-c-T4 'AiT FILE AL ?tP Igr .EDC f-Si 1____ -..

O~NT~eS SAMPO fMCGUIgE "U: u~pfULE UNIT"I I______

ALOCK SOL.. Qcj21aq085 DLOS I . 01 __ pT.2.I1g j

LAS.NO. MATERIAL ISOTOPE QDPS/m _____.__I e6i-a72Ae NP CAPSULE _________I'S U rBP-b3L r,__13

.,30 F. 'reP MW S I 0q tI

- 273-01016.1 1

-. 73 -Co~ Ci,-L5 a_4_

- 3L 1 E27J -S +/-La a.2

- 2 a-7_ jI i- / rt1 Al 2 763-

~ROT 1fhN-Sq.LM ____

-~2I$.0_0 -(_0 V 6.3 .

~~~~ Ca Al ~-. -~

-- A IU 58

-a7j4 r L.aoi K-TI-li

_I_ _ ...

Appendix A July 1999 WCAP-15253, Rev. 0

A-6 Table A-3 Radiometric Counting Results For Sensors Removed From Capsule X westirnhouse Electric Corporation Advanced Energy Systems - Analytical Laboratory Waltz Mill Site Request# 13642 TO: S.L.Anderson (W)NAM, Energy Center (478)

S.E.Yanichko (W)GTI, R&D Center 701 Bldg. (208) Received: 3/22/89 Reported: 4/10/89 fRESLTS OF RALYSIS]

Dosimetry: MGuire Unit #I Capsule X (Feb.22, 1989)

Originator Lab. Dosimeter ID Sanple # Material Nuclide s/mg_

2 sigma FISSON MNTORS /

U-238 89-331 U-238 Cs-137 6.27E+02 +/ 6.9E+00 Np-237 89-332 Np-237 Cs-137 3.91E+03 +/- 2.8E+01 TOP WIRES AlCo 89-333 AICo CO-60 2.76E+04 +1 2.66E+02 AlCo 89-334 AlCo Co-60 3.14E+04 4./ 3.01E+02 AlCo (Cd)89-335 AICo (Cd) Co-60 1.68E+04 2. 06E+02 Cu 89-336 Cu Co-60 1.26E+02 4./ 1. 39E+00 Fe 89-337 Fe ?k-54 1.65E+03 4./ 2.56E+01 Ni 89-338 Ni Co-58 1.10E+04 4.,, 6.70E+01 MID WIRES AlCo 89-339 APCo Co-60 2.95E+04 4./ 2.91E+02 AlCo 89-340 AlCo Co-60 3.35E+04 4./ 3.16E+02 AlCo (Cd)89-341 AICo (Cd) Co-60 1. 68E+04 4./ 2.04E+02 Cu 89-342 Cu Co-60 1.25E+02 4./ 1.47E+O0 Fe 89-343 Fe Mn-54 1. 64E+03 +1 2.43E+01 Ni 89-344 Ni Co-58 1.07E+04 +1 6.26E+01 BOTTOM WnIS AICo 89-345 AlCo Co-60 2.49E+04 +1 2.75E+02 AlCo 89-346 AlCo Co-60 3.00E+04 4./ 3.33E+02 AICo (Cd)89-347 AlCo (Cd) Co-60 1.55E+04 41 2.09E+02 Cu 89-348 Cu CO-60 1.29E+02 4./ 1.37E+00 Fe 89-349 Fe Ior-54 1.73E+03 4./ 2.74E+01 Ni 89-350 Ni Co-58 1.12E+04 +1- 7.25E+01 Remarks:

Results are in units of dps/(mg of Dosimeter material).

References:

Request# 13642 Lab.Book#41 page 8.

Procedures: A-512,A-513.

Analyst: WTF,WRM 12 Approved:Ce.'4 -1 July 1999 A

Appendix A July 1999 WCAP-15253, Rev. 0

/

A-7 Table A-4 Radiometric Counting Results For Sensors Removed From Capsule V Westinghouse Electric Corporation Chemistry & Materials Technology - Analytical Laboratory REPORT Waltz Mill Site Request# 15212A Originator: Ed Terek (W)NATD Structural Reliability & Plant Life Optimization Received: 10/11/93 Westinghouse Electric Corporation Reported: 12/16/93

[RESULTS OF ANALYSIS]

Dosimetry: McGuire Unit I "V" Capsule Originator Lab. Dosimeter (October 12,1993)

ID Sample # Material Nucl ide dps/mg

2 sigma FISSON MONITORS U-238 #285 93-4153 U-238 Cs-137 8.38E+02 +/- 9.51E+00 Np-237 #155 93-4154 Np-237 Cs-137 3.11E+03 +/- 5.01E+01 TOP WIRES (579)

Al Co 93-4155 Al Co Co-60 2.74E+04 +/

4/

2.OOE+02 Al Co 93-4156 Al Co Co-60 3.22E+04 2.19E+02

+/ 1.50E+02 AlCo(Cd) 93-4157 Al Co Co-60 1.69E+04 Cu 93-4158 Cu Co-60 1.47E+02 4/ 9.69E-01 Ni 93-4159 Ni Co-58 4.73E+03 +/ 2.811E+O1 Fe Fe Nn-54 1.28E+03 4/ 1.40E+01 93-4160 MID WIRES (590)

Al Co 93-4161 Al Co Co-60 2.94E+04 +/ 2.12E+02 Al Co 93-4162 Al Co Co-60 2.43E+04 4/ 1.91E+02

+/ 1.48E+02 A]Co(Cd) 93-4162B Al Co Co-60 1.62E+04 +/

Cu 93-4163 Cu Co-60 1.61E+02 1.69E+00 Ni 93-4164 Ni Co-58 4.81E+03 4/ 2.80E+03

+/ 1.86E+01 Fe 93-4165 Fe Mn-54 1.28E+03 BOTTOM WIRES (579) +/

AlCo 93-4166 Al Co Co-60 2.60E+04 1.93E+02 AlCo 93-4167 Al Co Co-60 2.97E+04 4/ 2.16E+02 AlCo(Cd) 93-4168 Al Co Co-60 1.63E+04 4/ 1.56E+02 1.45E+02 4/ 1.04E+00 Cu 93-4169 Cu Co-60 +/

Ni 93-4170 Ni Co-58 4.74E+03 2.76E+O1 Fe 93-4171 Fe Mn-54 1.20E+03 4/ - 1.75E+01 Remarks:

Results are in units of dps/(mg of Dosimeter Material).

/ ,1 Procedures:A-512,A-513,A-524 Analyst:WTF,FRC,N4RK Approved Appendix A July 1999 WCAP-15253, Rev. 0

A-8 Table A-5 Radiometric Counting Results For Sensors Removed From Capsule Z Westinghouse Electric Corporation Chemistry & Materials Technology - Analytical Laboratory REPORT Waltz Mill Site Request# 1521:

Originator: Ed Terel (W)NATO Structural Rel iability & Plant Life Optimization Received: 10/11/93 Westinghouse Electric Corporation Reported: 10/25/93

[RESULTS OF ANALYSIS)

Dos imetry: McGuire Unit I "Z"Capsule Originator Lab. Dosimeter (October 12,1993)

ID Sample M Material Nuclide dps/mg

2 sigma FISSON MONITORS U-238 #295 93-4172 U-238 Cs-137 9.57E+02 4/- 2.21E+01 Np-237 #29-1S9 93-4173 Np-237 Cs-137 6.S8E+03 */- 7.6$E+01 TOP WIRES (079)

AlCo 93-4174 AlCo Co-60 3.43E+04 ÷/ 2.34E+02 AlCo(Cd) 93-4175 Al Co Co-60 1.82E÷04 4/ 1.58E+02 Cu 93-4176 Cu Co-60 1.53E÷02 4/ 1.53E+00 Ni 93-4177 *Ni Co.58 5.09E+03 4/ 3.59E+01 Fe 93-4178 Fe Mn-54 1.29E+03 1.33E+01 4/

MID WIRES (590)

Al Co 93-4179 Al Co Co-60 2.80E÷04 4/ 1.96E+02 Al Co 93-4180 Al Co Co-60 3.01E04 2.12E÷02 Al Co Co-60 3.35E+04 ÷/. 2.40E+02 Al Co 93-4181 4/ 1.60E+02 AlCo(Cd) 93-4182 Al Co Co-60 1.78E÷04 Cu 93-4183 Cu Co-60 1.61E+02 1.69E+00 Ni Co-SB 6.22E+03 4/ 3.78E+01 Ni 93-4184 1.38E+01 Fe 93-4185 Fe Mn-54 1.35E÷03

÷/

BOTTOM WIRES "(o1-4186 AlCo AlCo Co-60 3.03E÷04 ÷/ 2.01E+02 AlCo 93-4187 AlCo Co-60 3.54E+04 2.36E+02 AlCo(Cd) 93-4188 Al Co Co-60 1.89E+04 4/. 1.59E+02 Cy 93-4189 Cu Co-60 1.S7E+02 ÷/ 1 .S7E÷O0 Ni 93-4190 Ni Co-SB 5.11E+03 4/- 3.76E+01 Fe 93-4191 Fe Mn-54 1.31E+03 1.38E+01 Rm.. t.......kt.::R:u:t::arm:::units:of::of:Dosimeter ::aterial.....

Procedures:A-512,Am513,A-524 Analyst:WTFFRC,MRK Approved Appendix A July 1999 WCAP-15253, Rev. 0

A-9 Table A-6 Radiometric Counting Results For Sensors Removed From Capsule Y Antech Ltd.

One Trianl Drive . E~pon, Pemnsyjvam 15632

Phone. (412) 733-1161 F Iax: (412) 327-7793 September 16, 1997 Mr. Ed Torek Westinghouse Electric Corporation Energy Center Pittsburgh, PA 15230 Dosimetry Characterixation; Purchase Order No. DA7DC Mcouire Dosinetry T capsule.s; S!c; nergy Center Antech Ltd. Project NO. 97-0227W

Dear Mr. Torok:

Enclosed are analytical results for samples submitted by Westinghouse Electric Corporation. Samples were received and logged in for analysis on August 20, 1997.

Appropriate AST?4/ZPA methods were used and are indicated accordingly on the data tables. Appropriate quality assurance/quality control analyses were performed in accordance with Antech Ltd. I. Statement of Qualifications. If you hawe any questions, please call me at 412-722-5219.

Encore" Enclosures An Aenurca Whaft Sa~ic Campy Appendix A July 1999 WCAP-15253, Rev. 0

A-10 Table A-6 Radiometric Counting Results For Sensors Removed From Capsule Y (cont.)

mwzCA L2D.

CAMS NAB3AMVE I. PROJECT LMIX INFORKAWION A: PP.OJMCE NUNUZUtS ANJTECH LTD.: 97-0;27W purechASO Ord*er Number: D 7DEC C%,IEIT %

5: SMEPLZ IDENTIFXCATIONSt Antach ID Client IV A~taCh ID Client ID 9708-0054W s/S NP CAPSULE 20-158 9708.-0053W BRhSS U CAPSULE 9708-0056W TOP ALCO 9708-0055W TOP CDALCO 9708-00S8W TOP CU 9708-0057W TOP ALCO 9708-0060W TOP 7Z 9708-0059W TOP NI 9708-0062W MID ALCO 9708-0061W KID CDALCO 9708-0064W MID CU 9708-0063W MID ILCO 9708-0066W MID 91 9708-0065W KID 12 9708-0068W BOTTOM A140 9708-0067W BOTTOM CDALCO 9708-0070W BOTTOM CU 9708-0069W BOTTOM ALCO 9708-0072W BOTTOX FE 9708-0071W BOTTOM NI c: SNIPPIXG/R3CEXVXNG COMM6S3 None IZ. PREWAfl2IONIANALSIS CONMUEMS III. azu3RA CoIAUTsI cgs uja augearing on the data ahould not TrgiliflO zeroom and degimal urocedure, but rather by interyreted a. preciyi~fl of the onalXtiaal As a result of reworttingl forMat.

Appendix A July 1999 WCAP-15253, Rev. 0

A-11 Table A-6 Radiometric Counting Results For Sensors Removed From Capsule Y (cont.)

Table I ameeraL Data 2able Westinbhouse Electric Corporation Antech Ltd. Project No. 97-0227N Dosizmetry Cbaractet=ati*oln; Mcouire DOsiustry y Capsules Charge order No. DA7DCZACA1 STC ftrameter Identification Cs-137 Antech Client Date A-524 Sam=le ID s$mle ID Collected dva /ma FISSION MONITORS 9708-00S3W RR?.SS U CAPSI.3-U-238 (8/20/97) 1.29Z3 t 4.6E2 9708-0054W 9/$ UP CAPS*DE 20-158-NP-237 (8/20/97) 5.8723 t 7.1Z2 Table 2 oeneral Date Table westinghouse zleal*Le Corporation Antechi Ltd. Project Xo. 97-0227W Dosimetry CharactetrizattLon McGuire Dosizetry T Capsules Charge Order No. DA7DCZACt*. TC Parapeter Ident ification Co-60

Antech Client Date A-524

§aMple ID sample ID Collected dmm TOP WIRES 9708-0055W TOP CDALOO (8/20/97) 2.0234

5.032 9708-0056W TOP ALM (8/20/97) 3.8634 & 9.332 9708-0057W TOP &LcO (8/20/97) 3.4124

8.2E2 9708-0058W TOP CU (8/20/97) 1.$932 +/- 5.320 KID WIRES 9708-0061W MID CDALCO (8/20/97) 1.87Z4 +/- 4.692 9708-0062W MID ALOO (8/20/97) 2.91E4

7.032 9708-0063W MID ALCO (8/20/97) 3.5334 t 8.SE2 9708-0064W MID CU (8/20/97) 1.7022 i S.790 BOTTOM WIRES 9708-0067W BOTTOM CDAILCO (8/20/97) 1.9334 1 4.732 9708-0068W BOTTOM ALCO (8/20/97) 3.1624

7.632 9708-0069W BOTTOM ALCO (8/20/97) 3.70E4 t 8.932 9708-0070W BOTTOM CU (8/20/97) 1.4622

4.9go Appendix A July 1999 WCAP-15253, Rev. 0

A-12 Table A-6 Radiometric Counting Results For Sensors Removed From Capsule Y (cont.)

Table 3 General Data Table Westinghouse le*Ctr1.c Corporation Antech Ltd. Project Wo. 97-0227W Dosinetry Characterizatioul mcuirs DosiuItry Y Capsules Charge Order No. DA7D0C3A0; STC Parameter Identif i-at £ion Co-5B Date J-S24 Antech client Sarnmle IV Sarplme TV Collected domi TOP WIRE (8/20/97) 5.49E3 t 5.2E2 9708-0059W TOP NI MID WIRE MID NI "(8/20/97) 5.59Z3

5.3E2 9708-0066W BOTTOM WIVZ 1T1 MI (8/20/97) 5.58Z3 + 5.232 9708-0071W -- _-B Table 4 general Data Table Westingho.use -lectric Corporation Antech Ltd. Project No. 97-0227W Vosimetry Characterisation; McGuire boxiamt7? Capsules charge Order No. DA7DCBAC.; STC Parameter Identi fiAtion Mn-54 Antech Client Date A-524 Sample ID Sample ID C..11ecte dos*/=

TOP W!Rz 9708-0060W TOP FE (8/20/97) 1.4023 t 1.2E2 MID WIRE 9708-0063W MID FE (8/20/97) 1.48E3 +/- 1.3*2 BOTTOM WIRE 9708-0072W BOTTOM ]E (8/20/97) 1.4133

1.232 July 1999 Appendix A A July 1999 W/CAP-15253, Rev. 0

B-1 APPENDIX B SPECIFIC ACTIVITIES AND IRRADIATION HISTORY OF REACTOR CAVITY SENSOR SETS IRRADIATED DURING CYCLE 12 In this appendix, the irradiation history, as extracted from reference 17, and the measured specific activities of radiometric sensors irradiated in reactor cavity during Cycle 12 are provided.

The startup and shutdown dates for Cycle 12 were as follows:

Cycle Startup Shutdown 12 05/19/97 05/29/98 RCND set 1S-1 irradiated The detailed operating history of the reactor over the course of Cycle 12 is provided in Table B-1. Note that the reference full power for McGuire Unit 1 is 3411 MWt.

The capsule loading table for the RCND set 1S-1 irradiated during Cycle 12 is provided in Table B-2. The measured specific activities of the monitors removed from RCND set 1S-1 are provided in Table B-3. For the multiple foil sensor sets, the individual foil ID can be correlated with the capsule position description in Section 6.1.1 in order to determine the location of the foil within the reactor cavity July 1999 Appendix B B July 1999 WCAP-15253, Rev. 0

B-2 Table B-1 McGuire Unit 1 Operating History - Cycle 12 Cycle 12 Date MWt-hr Mar-97 0 Apr-97 0 May-97 543543 Jun-97 2338395 Jul-97 2496031 Aug-97 2529024 Sep-97 2224656 Oct-97 2545651 Nov-97 2468702 Dec-97 2545265 Jan-98 2530756 Feb-98 2059033 Mar-98 2428309 Apr-98 2449705 May-98 2291166 Th. Gen. 2.945E+07 MWt-hr Cy. Length 3.108E+07 EFPS Appendix B July 1999 WCAP-15253, Rev. 0

B-3 Table B-2 McGuire Unit 1 Dosimeter Capsule Contents for Cycle 12 Radiometric Monitor ID Capsule ID Bare or and Cadmium Position Shielded Fe Ni Cu Ti Co U-238 Np-237 A-1 B A - - - A - -

A-2 Cd CA A A A CA - -

A-3 Cd . . . . . 40 35 B-1 B B .. . . B - -

B-2 Cd CB B B B CB - -

B-3 Cd . . .. . . 41 36 C-1 B C .. . . C ....

C-2 Cd CC C C C CC - -

C-3 Cd . . .. . . 42 37 D-1 B D .. . . D -

D-2 Cd CD D D D CD - -

D-3 Cd . . .. . . 43 38 E-1 B E - - - E -- -

E-2 Cd CE E E E CE - -

E-3 Cd . . . . . 44 40 F-1 B F - - - F --

F-2 Cd CF F F F CF - -

F-3 Cd . . . .. . 45 41 Note: Each capsule contains two iron foils, one nickel foil, one copper foil, one titanium foil, two cobalt-aluminum foils, one vanadium encapsulated "U oxide detector, one vanadium encapsulated ' 7Np oxide detector, and two cadmium covers.

Appendix B July 1999 WCAP-15253, Rev. 0

B4 Table B-3 Radiometric Counting Results For Sensors Removed From Cycle 12 Cavity Dosimetry Set 1S-1 Capsules A, B, C, D, E, and F

- Antech Ltd.

alt/ '%MillSiie -I O. 11,\ 158 - MNadism. KA 15663-0158 -Phone: (724)722-5214 -Fax: (724)722"5UX December 23, 1998 Certificate of Conformance Mr. Larry Becker Westinghouse Electric Company SGO/Chem & Materials F Building, MS 60 Madison, PA 15663 Dosimetry Characterization; Purchase Order No. ACZP7500A McGuire Unit 1; WALTZ MILL; NUCLEAR SERVICES DIVISION Antech Ltd. Project No. 98-0418W

Dear Mr. Becker:

Electric Enclosed are analytical results for samples submitted by Westinghouse Company. Samples were received on September 29, 1998 and logged in for 'Inalysis on September 30, 1998.

accordingly on the data Appropriate ASTM/EPA methods were used and are indicated Appropriate quality assurance/quality control measures were performed tables.

10 CFR, part 50 in accordance with Antech Ltd. 's Quality Assurance Plan and Appendix B. If you have any questions, please call me at 724-722-5219.

Sincerely,._

Brian M. Carson Emery i. rohregin QA/QC Coordinator Supervisor EJG:mlm Enclosures Appendix B July 1999 WCAP-15253, Rev. 0

B-5 Table B-3 Radiometric Counting Results For Sensors Removed From Cycle 12 (cont.) Cavity Dosimetry Set IS-1 Capsules A, B, C, D, E, and F ANTCACE LTD CASE NARRATIVE I. PROJECT LOGIN INFORMATION:

A: PROJECT NUMBERS:

ANTECH LTD.: 98-0418W ACZP7500A CLIENT: Purchase Order Number:

B: SAMPLE IDENTIFICATIONS:

Antech ID Client ID Antech ID Client ID 9809-0464W CA FE (CAP;ULE A) 9809-0463W A FE (CAPSULE A)

A NI (CAPSULE A) 9809-0466W "A CU (CAPS';LE A) 9809-0465W "A CO (CAPS';LE A) 9809-0468W 9809-0467W A TI (CAPSULE A) 9809-0470W 40 (U-238 ,:APSULE A) 9809-0469W CA (CO-CAPSULE A) 9809-0472W B (FE CAPS;LE B) 9809-0471W 35 (NP-237 CAPSULE A 9809-0474W B (NI CAPS ;LE B) 9809-0473W CB (FE CAPSULE B) 9809-0476W B (TI CAPSULE B) 9809-0475W B (CU CAPSULE B) 9809-0478W CB (CO CAP.;ULE B) 9809-0477W B (CO CAPSULE B) 7 9809-0480W 36 (NP-23 CAPSULE B 9809-0479W 41 (U-238 CAPSULE B) 9809-0482W CC (FE CAP::ULE C) 9809-0481W "C (FE CAPSULE C) 9809-0484W "C (CU CAPS"LE C) 9809-0483W "C (NI CAPSULE C) 9809-0486W "C (CO CAPS"LE C) 9809-0485W "C (TI CAPSULE C) 9809-0488W 42 (U-238 ":APSULE C) 9809-0487W CC (CO CAPSULE C) 9809-0490W D (FE CAPS::LE D) 9809-0489W 37 (NP-237 CAPSULE C 9809-0492W D (NI CAPS':LE D) 9809-0491W CD (FE CAPSULE D) 9809-0494W D TI (CAPS'LE D) 9809-0493W D (CU CAPSULE D) 9809-0496W CD CO (CAP*ULE D) 9809-0495W D CO (CAPSULE D) 9809-0498W 38 NP-237 CAPSULE D 9809-0497W 43 U-238 (CAPSULE D) 9809-0500W CE FE (CAP:'ULE E) 9809-0499W E FE (CAPSULE E) 9809-0502W E CU (CAPS::LE E) 9809-0501W E NI (CAPSULE E) 9809-0504W E CO (CAPS':LE E) 9809-0503W E TI (CAPSULE E) 9809-0506W 44 U-238 (,:APSULE E) 9809-0505W CE CO (CAPSULE E) 40 NP-237 (CAPSULE E 9809-0508W "F FE (CAPS:LE F) 9809-0507W "F NI (CAPS':LE F) 9809-0510W 9809-0509W CF FE (CAPSULE F)

F CU (CAPSULE F) 9809-0512W "F TI (CAPS"LE F) 9809-0511W CF CO (CAP::ULE F) 9809-0514W 9809-0513W F CO (CAPSULE F) 9809-0516W 41 NP-237 -CAPSULE F 9809-0515W 45 U-238 (CAPSULE F) 9809-OS18W BLOCK 8 1S-1-15 9809-0517W BLOCK A 1S-1-00 +4.5 (0)

+5.5 (0) 9809-0520W 9809-0519W +2.5 (0)

+3.5 (0) 9809-0522W 9809-0521W +0.5 (0)

+1.5 (0) 9809-0524W 9809-0523W -1.5 (0)

(0) 9809-0526W 9809-0525W -0.5 9809-0528W -3.5 (0) 9809-0527W -2.5 (0) 9809-0530W -5.5 (0) 9809-0529W -4.5 (0) 9809-O532W BLOCK C 1S-1-30 9809-0531W -6.5 (0) 9809-O534W +6.5 (15) 9809-0533W BLOCK D,EF !S-1-45 9809-0536W +4.5 (15) 9809-0535W +5.5 (15)

Appendix B July1999 July 1999 WCA.P-15253, Rev. 0

B-6 Table B-3 Radiometric Counting Results For Sensors Removed From Cycle 12 (cont.) Cavity Dosimetry Set lS-1 Capsules A, B, C, D, E, and F ANTECE LTID CASE NARRATIVE (Continued) 9809-0537W +3.5 (15) 9809-0538W +2.5 (15) 9809-0539W +1.5 (15) 9809-0540W +0.5 (15) 9809-0541W -0.5 (15) 9809-0542W -1.5 (15) 9809-0543W -2.5 (15) 9809-0544W -3.5 (15) 9809-0545W -4.5 (15) 9809-0546W -5.5 (15) 9809-0547W -6.5 (15) 9809-0551W +5.5 (30) 9809-0552W +4.5 (30) 9809-0553W +3.5 (30) 9809-0554W +2.5 (30) 9809-0555W +1.5 (30) 9809-0556W +0.5 (30) 9809-0557W -0.5 (30) 9809-0559W -2.5 (30) 9809-0558W -1.5 (30) 9809-0560W -3.5 (30) 9809-0561W -4.5 (30) 9809-0562W -5.5 (30) 9809-0563W -6.5 (30) 9809-0564W NIST 121D 9809-0565W True Value of NIST 1 9809-0567W +5.5 (45) 9809-0568W +4.5 (45) 9809-0569W +3.5 (45) 9809-0570W +2.5 (45) 9809-0571W +1.5 (45) 9809-0572W +0.5 (45) 9809-0573W -0.5 (45) 9809-0574W -1.5 (45) 9809-0575W -2.5 (45) 9809-0576W -3.5 (45) 9809-0577W -4.5 (45) 9809-0578W -5.5 (45) 9809-0579W -6.5 (45)

C: SHIPPING/RECEIVING COMMENTS:

FINAL REPORT: 12/17/98 II. PREPARATION/ANALYSIS COMMENTS:

A: METALS:

NONE B: RADIOLOGICAL:

Samples have been decay corrected to December 1,1998. All ra..-Lochemistry analysis results are reported with a +/- 2 sioma error value. The following samples (client ID) do not exist: 9809-0548W (-7.5.1511, 9809-0549W (+7.5(30)), 9809-0550W (+6.5(30)), 9809-0566W (+6.5(451).

III. GENERAL COMMOENTS:

Trailing zeroes and decimal Places appearing on the data s-tould not be interpreted as precision of the analytical procedure. bit rather as a result of reporting format.

Appendix B July 1999 WCAP-15253, Rev. 0

0 m

It Table 1 General Data Table Uestinghouse Electric Conpany Antech Ltd. Project No. 98-0418W Dosimetry Characterization; McGuire Unit 1 Charge Order No. ACZP7SOOA; UALTZ HILL 81 81 Page 1 of 3 0 Parameter Identification Antech Client Sc-46 A-524 Mn-54 A-524 Co-58 A-524 Co-60 A-524 Zr-95 A-524 Ru-lO3 A-524 Cs-137 A-524 0

0 Sample ID Sample ID dMps/nq dps/mq dps/nig dps/mg dps/me dps/mg dps/rag U) rA 9809-0463W 9809-0464u 9809-0465W A FE (CAPSULE CA FE (CAPSULE A NI (CAPSULE A)

A)

NA NA NA 4.71EO t 5.OE-1 4.67E0 a 5.OE-1 NA 3.1t1EI NA NA 3.3E0 "A

NA NA (1) NA NA NA

'I 18 U.

m If.

9809-0466w A CU (CAPSULE A) NA NA NA 2.50E-1 1.2E-2 NA I-a 9809-0461W A I1 (CAPSULE A) 8.87E-1 i 5.8E-2 NA NA NA WA

'-a 9809-0468w A CO (CAPSULE A) NA NA NA 2.37E2 a 5.8E0 NA 9809-0469W CA (CO-CAPSULE A) NA NA NA 9.79EI a 2.71O NA 81 9809-0470w 40 (U-238 CAPSULE A) NA NA NA NA 1.62E0 a 1.9E-1 6.72E-1 1.3E-1 3.63E-1 a 7.4E-2 U) 9809-0471W 35 (HP-237 CAPSULE A) NA NA NA NA 2.69E1 t 1.1EO 7.90E0 +/- 8.7TE.I 5.64E0 S.2E-1 18 9809-0472W B (FE CAPSULEB) NA 7.34E0 a 7.7E-1 NA NA NA U)

U) 9809-0473W CB (FE CAPSULE 6) NA 7.07EO i 7.4E-1 NA NA NA 9809-0474u B (NI CAPSULE B) NA NA 4.81E1 i 4.4E0. NA NA 9809-0475W B (CU CAPSULE 8) NA NA NA 3.43E-1 t 1.8E-2 NA 9809-0476W B (TI CAPSULE B) 1.35E0 i 7.2E-2 NA NA NA NA 9809-0477W B (CO CAPSULE B) NA NA NA 3.60E2 a 9.OEO NA TI 9809-0478w CB (CO CAPSULE B) NA NA NA 1.55E2 a 4.2EO NA 9809-0479w 41 (U-238 CAPSULE B) NA NA NA NA 2.57E0 t 1.7E-1 9.49E-1 t 1.3E-1 5.38E-1 +/- 8.8E-2

'ii See footnotes at end of table. CA n

0D 0

A H

0 Table I *9 (Continued)

(JJ Page 2 of3 Parameter Identification Sc-46 Mn-54 Co-58 Co-60 Antech Client Zr-95 Ru-103 Cs-137 A-524 A-52'4 A-524 A-524 Sample ID SAmple I0 A-524 A-524 A-524 dps/mg dps/mg dps/mg dps/nm dps/mg dps/mg dps/mg 9809-0480w 36 (NP-237 CAPSULE 8) NA NA NA NA 4.13EI I i.JEO I.10EI i 1.5E0 7.68E0 i 8.3E-1 9809-0481W C (FE CAPSULE C) NA 7.01EO s 7.4E-I NA NA NA 9809-0482W CC (FE CAPSULE C) NA 7.02EO a 7.4E-1 NA NA 9809-0483w C (NI CAPSULE C) RA NA 4.48EI i 3.3EO NA NA *m 9809-0484W C (CU CAPSULE C) NA NA NA 2.82E-1 3 1.3E'2 NA 9809-0485W C (11 CAPSULE C) 1.19E0 i 7.IE-2 NA NA NA NA nE..

9809-0486w C (CO CAPSULE C) NA NA NA 4,15E2 i I.IEI NA 9809-0487W CC (CO CAPSULE C) NA NA NA NA 1.93E2 t 5.3E0 l-Ao NA 9809-0488w 42 (U-238 CAPSULE C) NA NA NA NA 2.55EO t 1.9E-1 8.56E-1 t 1.3E-1 5.32E-1 i 8.8E-2 9809-0489W 37 (NP-237 CAPSULE C) NA NA NA NA 4.68El t 3.lEO 1.37EI t 1.SEO 9.27EO t 1.0EO 9809-0490W 0 (FE CAPSULE 0) NA 2.OIEO a 2.3E-1 NA NA RA 9809-0491W CO (FE CAPSULE D) NA 2.13EO a 2.4E-1 NA NA MA " 4t 9809-0492W 0 (NI CAPSULE 0) NA NA 1.53EI i 1.6E0 NA NA 9809-0493w D (CU CAPSULE 0) NA NA NA 8.04E-2 i 5.3E.3 NA 9809-0494W D TI (CAPSULE 0) 3.47E-1 i 2.2E-2 NA NA NA NlA 9809-0495W D CO (CAPSULE 0) NA NA NA NA 9.42EI t 2.6E0 9809-0496w CD CO (CAPSULE D) NA NA NA NA 5.44EI A 1.7E0 9809-0497W 43 U-238 (CAPSULE 0) NA NA NA NA 1.04EO t 1.3E-1 3.49E-1 i 8.5E-2 2.51E-1 +/- 6.2E-2 Ai See footnotes at end of table. 0

,p_ .

01 -<

Table 1 (Continued)

Page 3 of 3 Parameter Identification Sc-46 Mn-54 Co-58 Co-60 Zr-95 Ru-103 Cs-137 Antech Client A-524 A-524 A'524 A-524 A.524 A-524 A-524 Sanple lO Sample I dps/mg dps/n dps mg dps/mg dps/mg dps/mg dps/mg 38 NP-237 (CAPSULE D) NA NA NA NA 2.03E1 i 8.8E-1 6.01EO , 7.SE-1 3.62E0 i 3.6E-1 9809-0498W NA NA NA 9809-0499w E FE (CAPSULE E) NA 5.66E0 , 2.5E-1 NA NA NA 9809-0500W CE FE (CAPSULE E) NA 5.41EO 5.7E-1 NA NA NA NA 3.68E1 i 1.7E0 9809-0501W E NI (CAPSULE E)

WA 2.16E-I i 8.4E.3 NA E CU (CAPSULE E) NA NA 9809-0502W NA NA NA 9809.0SOSW EII (CAPSULE E) 8.72E'l i 4.6E-2 NA NA NA NA 2.60E2 t 6.6E0 NA 9809.0S04W E CO (CAPSULE E)

NA NA NA 1.55E2 1 3.3E0 NA 9809-0505w CE CO (CAPSULE E)

NA NA 2.21E0 x 1.8E-1 9.60E-1 i 1.3E-1 4.B7E-1 +/- 1.01-1 44 U-238 (CAPSULE E) NA NA 9809.0506w NA NA 4.95E1 t 1.5E0 1.47E1 i 1.3E0 9.27EO t 5.9E-1 40 NP-237 (CAPSULE E) NA NA 9809.0507w NA NA NA F FE (CAPSULE F) NA 2.50E0 i 2.8E-1 9809.0508W NA NA NA CF FE (CAPSULE F) NA 2.18E0 +/- ?.4E-l 9809-0509w NA !too 1.66E1 t 6.4E'1 NA F NI (CAPSULE F) NA NA 9809-051OW NA NA NA 8.84E.2 i 8.8E-3 f CU (CAPSULE F) NA 9809-0511W NA 0A NA NA F TI (CAPSULE F) 3.52E-1 i 2.6E-2 NA 9809-0512w NA NA NA 1.37E2 i 4.OEO F CO (CAPSULE F) NA 9809-0513W NA NA 9.14E1 +/- 2.4E0 NA NA 9809-0514W CF CO (CAPSULE F)

NA NA 1.OOEO t 1.3E-1 4.05E-1

9.1E-2 1.99E-1 e 5.OE-2 NA NA I-I" 9809-0515W 45 U-238 (CAPSULE F)

NA NA 2.18E1 i 7.1E-1 6.08E0 t 5.5E-1 3.76E0 a 2.8E-1 NA NA 9809-0516W 41 NP-237 (CAPSULE F) 1

( )Dash denotes not analyzed.

(1

0 A,)

Table 2 General Data Table Westinghouse Electric Company Antech Ltd. Project No. 98-0418w Dosimetry Characterization; McGuire Unit 1 Charge Order No. ACZP750OA; WALTZ MILL Parameter Identification Antech Cobalt (Total Iron (Total) Ni ckel (Total)

Client Date Sample ID 6010(1) 6010(1) 6010(1)

Sample ID Collected ma /ka no I ko 9809-0517W BLOCK A 1S-1-00 (9/28/98) 1400 66.60 9.800 9809-0518W BLOCK B 1S-1-15 (9/28/98) 1500 67.60 9.900 9809-0532W BLOCK C IS-1-30 (9/28/98) 1400 67.80 9.900 9809-0533W BLOCK D,E,F 1S-1-45 (9/28/98) 1500 67.50 9.900 9809-0564W NIST 121D (9/28/98) rCM~

1100 63.40 11.30 9809-0565W True Value of NIST 121d (9/28/98) 1000 68.23 11.17 9809-0580W METHOD BLANK (9/28/98) <2.0 <0.001000 <0.001000 (1)U.S. Environmental Protection Agency, 1987, Test Methods for Evaluating Solid Waste, SW-846, 3rd ed., Office of Solid Waste and Emergency Response, Washington, DC.

r.'

(Ji Q1 .

C:)

I-.

0

0 tH Table 3 General Data Table Westinghouse Electric Company Antech Ltd. Project No. 98-041814 Dosimetry Characterization; McGuire Unit 1 Charge Order No. ACZP7SOOA; WALTZ MILL Page I of-2 Parameter Identification Co-58 Co-60 A-524 A-524

t,*l','* A'524

.4n./Inn Client Date dns/na Antech ops i n*j Sample ID Sample In Collected Dolr~r Caaceizton Antech~.

Charge~~~~

AZ70A cupLre~

WL UntK HL o Poj.t1o29-011 s/e~~~ 72 atd 1.66E0 I 1.AE-1 1.57E0 t 1.4E-1 2.17E I t 5.1E'!

9809-051914 45.5 (0) (9/28/98) 2.4920 t 2.0E-1 2.28E0 a 2.0E-1 2.90EI a 6.8E.1 ri)

4.5 (0) (9/28/98) 2.91E0 i 2.0E-I 2.38E0 a 2.2E-1 3.332i a7.7f-1 9809-0520W (9/28198)

.3.5 (0) 2.72E0 a 1.9E-1 3.63EI a 7.9E- 1 9809-0521I (9/28/98) 2.92E0 a 1.9E-1

.2.5 (0) 3.0OE0 i 2.0E-I 2.83E0 a 1.9E-1 3.96EI t8.6E- I 9809-05221 (9/28/98)

-1.5 (0) 3.02E0 t 2.3E-1 2.7720 a 2.4E-1 9809-0523U 5.56E1

.0.5 (0) (9/28/98) 3.29E0 t 2.12E-1 2.82E0 a 2.2E-1 5.8121 a 1.3E0 1,3E0 9809-0524W (9/28/98)

.0.5 (0) 2.62E0 +/- 2.1E-1 5.7301 a 1.2E0 9809-05251 2.95E0 i 2.2E-1 (9/28/98) 5.35E1 a 1.220

-1.5 (0) 3.05E0 t 2.2E-1 2.80E0 a 2.2E-1 9809-0526W (9/28/98)

-2.5 (0) 2.69E0 a 1.9E-1 4.73Et a 1.020 9809-0527W 2.82E0 i 2.2E-1 (9/28/98) 3.34E1 t 7.6E.1

-3.5 (0) 2.49E0 a 1.5E-1 2.31E0 I 1.5E-1 9809-0528W (9/28/98) . n

-4.5 (0) 1.61E0 t 1.5E-1 2.02Ei t 4.8E-1 9809-052914 1.68E0 t 1.4E-1

-5.5 (0) (9/28/98) 1.54E1 a 3.7E-1 6.80E-1 t 7.7E-2 6.48E-1 t 8.2E-2 9809-05301 (9/28/98)

-6.5 (0) 9809-05311 4.32E1 a 1.02O 2.35E0 t 1.9E-1 2.2720 t 2.1E-1 rO,,ftgBa 9809-0535W *:I.;%,a ......... 6.1i70 a *.*Lu U~T 3.66E0 i 2.8E-1 3.44E0 t 2.9E-1 9809-0536U +4.5 (15) (9/28/98) 4.13E0 a 2.7E-1 7.47E1 a 1.6E0 (9/28/98) 4.53E0 t 2.8E-1 ri) 9809-0537+ .3.5 (15) (9/28/98) 4.62E0 t 3.DE-1 4.10E0 t 2.7E-I 8.03E1 a 1.7EO Ot 9809-0538U .2.5 (15) 9809-053918 .1.5 (15) (9/28/98) 4.64E0 t 3.0E-1 4.1520 a 2.82-1 8.20E1 a 1.8E0

+0.5 (15) (9/28/98) 4.58E0 a 3.2E-1 4.20EO a 3.0E-1 8.63EI a 1.9E0 9809-05403 4.83E0 a 2.92E-1 3.96E0 a 2.8E-1 8.78E1 t 2.OEO rA0 9809-0541U -0.5 (15) (9/28/98)

-1.5 (15) (9/28/98) 4.62E0 a 3.0E-1 3.92E0 +/- 2.8E-1 8.31E1 a 1.9E0 9809-05421

-2.5 (15) (9/28/98) 4.1820 t 2.5E-1 3.93E0 t 2.6E-1 7.59E1 a 1.6E0 9809-05432

-3.5 (15) (9/28/98) 4.12E0 +/- 2.6E-1 3.64E0 t 2.6E-I 6.91E1 a 1.520 9809-05443

-4.5 (15) (9/28/98) 3.66E0 t 2.8E'1 3.472Et 2.9E-1 5.73E1 a 1.3EO 9809-05454 9809.0546s -5.5 (15) (9128198) 2.4320 a 2.0E-1 2.22E0 t 2.01-1 4.14E1 a 9.5E-1 9809-05476 .6.5 (15) (9/28/98) 9.521E a 9.2.2- 8.92E-1 a 9.0E-2 2.37EI a 5.5E-I (Ti r -

O\O

B-12 Table B-3 Radiometric Counting Results For Sensors Removed From Cycle 12 (cont.) Cavity Dosimetry Set 1S-1 Capsules A, B, C, D, E, and F Nd d0 Luo 0 0D Q 0 0 0B 0:

0 7 10 0

O*0 u.a 0 o - A 0.

o o .0 -t

0. 0 0-4> '0

-. 7.0 o'0 CD C

C S 7

'0 0 7 7 7 C 0.7 0 Sn 7*1 0 0. 0~

Z.

-a 7 0 0 0 0 Sn 0 .0 06 N,

C C. 0 o o 0.

N, ,9 N 0 N N 0 .-

Ny N o o 04 .0 C.

x 0 0 0 0 , N,6 cco o -. 0 CC U

-a 0 0 0 0 0 0 '0 0 0 0 0 0 0 - - - - - - - - -. - - -

0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0. 0 N N N N N NN NN NN N N (10 00 0w) 0 0000 444J 0 0 0 0 0 0 0 0 0 0 0 0 0 0

C N

N 0.

N N

0.

N NN 0.

N 0.

N N

0.

N N

0.

N N

0.

N N

0.

N N

0.

N NN 0.

N 0.

N N

0.

N N

0. N~ ~~0 NNNNN N C N 0

o 0 0 0 N,

0 N,

C4 N, *. N, Sn Sn Sn Sn Sn Sn Sn Sn Sn Sn Sn Sn Sn Sn .7 N, N - 0 0 - N N, .7 Sn .0 0 09 44, 0 0ý

-a CE 0. 0. >0,0.0ý 0. 0. 0- 0.0. 0.0.00. 0- 0.0. 0.0. 0. .ý0-01 0o 0 0ý 0.

0

0. O0,0.0> 0.0000000 0.00.0.0.00.00 . Q. 0 0.0 July 1999 Appendix B July 1999 WCAP-15253, Rev. 0

ML020920048

Text

Subject:

SUMMARY

2.1 DESCRIPTION

2.2 DESCRIPTION

SUMMARY

2.1 DESCRIPTION

2.2 DESCRIPTION

References:

Dear Mr. Torok:

Dear Mr. Becker:

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