ML020870532

Reactor Cavity Neutron Measurement Program

ML020870532
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-15334, Rev. 0
Download: ML020870532 (141)

Text

Duke Duke Energy Corporation kEnergy.

McGuire Nuclear Station 12700 Hagers Ferry Road Huntersville, NC 28078-9340 H. B. Barron (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 1 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.

H. B. Barron Attachments

U. S. Nuclear Regulatory Commission Document Control Desk March 19, 2002 Page 2 cc:

R.E. Martin U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D.C. 20555 Rani L. Franovich U. S. Nuclear Regulatory Commission License Renewal and Environmental Impacts Program Office of Nuclear Reactor Regulation Washington, D.C. 20555 Luis A. Reyes U. S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 Scott Shaeffer Senior Resident Inspector McGuire Nuclear Station

U. S. Nuclear Regulatory Commission Document Control Desk March 19, 2002 Page 3 bxc:

ELL RGC File Master File Bob Gill

Westinghouse Non-Proprietary Class 3 Duke Power Company Reactor Cavity Neutron Measurement Program for William B. McGuire Unit 2 Cycle 12 Westinghouse Electric Company LLC

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15334 Duke Power Company Reactor Cavity Neutron Measurement Program for William B. McGuire Unit 2 Cycle 12 Arnold H. Fero Radiation Engineering and Analysis November 1999 Approved:

G. A. Brassart, 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

©2000 Westinghouse Electric Company LLC All Rights Reserved

iii TABLE OF CONTENTS LIST O F TA BL ES........................................................................................................................................

v L IST O F FIG U R ES.....................................................................................................................................

xi EXECU TIV E SU M M A RY......................................................................................................................

xrii 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......................................................

2-5 3

NEUTRON TRANSPORT AND DOSIMETRY EVALUATION METHODOLOGIES..... 3-1 3.1 Neutron Transport Analysis Methods.......................................................................

3-1 3.2 Neutron Dosimetry Evaluation Methodology.........................................................

3-6 3.3 Determination of Best Estimate Reactor Vessel Exposure.....................................

3-11 4

RESULTS OF NEUTRON TRANSPORT CALCULATIONS...............................................

4-1 4.1 Reference Forward Calculation..................................................................................

4-1 4.2 Fuel Cycle Specific Adjoint Calculations.................................................................

4-11 5

EVALUATION OF SURVEILLANCE CAPSULE DOSIMETRY.........................................

5-1 5.1 Measured Reaction Rates........................................................................................

5-1 5.2 Results of the Least Squares Adjustment Procedure..........................................

5-1 6

EVALUATIONS OF REACTOR CAVITY DOSIMETRY................................................

6-1 6.1 C ycle 12 R esults............................................................................................................

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 Reaction Rates.......................................................................................................

7-1 8

BEST ESTIMATE NEUTRON EXPOSURE OF REACTOR 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

RE FERE N C ES............................................................................................................................

9-1 APPENDIX A - SURVEILLANCE CAPSULE DOSIMETRY DATA..........................................

A-1 APPENDIX B - REACTOR CAVITY DOSIMETRY DATA.........................................................

B-1 WCAP-15334, Rev. 0 November 1999 November 1999 WCAP-15334, Rev. 0

V Table 4.1-1 Table 4.1-2 Table 4.1-3 Table 4.1-4 Table 4.1-5 Table 4.1-6 Table 4.1-7 Table 4.1-8 Table 4.2-1 Table 4.2-2 Table 4.2-3 Table 4.2-4 Table 4.2-5 Table 4.2-6 WCAP-15334, Rev. 0 November 1999 LIST OF TABLES Calculated Reference Neutron Energy Spectra at Cavity Sensor Set Locations - 3411 M w t; Fa = 1.2....................................................................................

4-3 Reference Neutron Sensor Reaction Rates and Exposure Parameters at the Cavity Sensor Set Locations - 3411 Mwt; F, = 1.20..................................................

4-4 Calculated Reference Neutron Energy Spectra at Surveillance Capsule Locations - 3411 Mwt; F = 1.2....................................................................................

4-5 Reference Neutron Sensor Reaction Rates and Exposure Parameters at the Center of Surveillance Capsules - 3411 Mwt; F. = 1.20..........................................

4-6 Summary of Exposure Rates at the Reactor Vessel Clad/Base M etal Interface......................................................................................................

4-7 Relative Radial Distribution of Neutron Flux (E > 1.0 MeV) Within the R eactor Vessel W all......................................................................................................

4-8 Relative Radial Distribution of Neutron Flux (E > 0.1 MeV) Within the R eactor Vessel W all......................................................................................................

4-9 Relative Radial Distribution of Iron Displacement Rate (dpa) Within the R eactor Vessel W all....................................................................................................

4-10 Calculated Fast Neutron Flux (E > 1.0 MeV) at the Center of Reactor Vessel Surveillance C apsules................................................................................................

4-12 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Center of Reactor Vessel Surveillance Capsules....................................................................................

4-13 Calculated Fast Neutron Flux (E > 1.0 MeV) at the Reactor Vessel Clad/Base M etal Interface........................................................................................

4-14 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Reactor Vessel Clad/Base Metal Interface........................................................................................

4-15 Calculated Fast Neutron Flux (E > 1.0 MeV) at the Cavity Sensor Set L ocations...............................................................................................................

4-16 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Cavity Sensor Set L ocation s...............................................................................................................

4-17 November 1999 WCAP-15334, Rev. 0

vi LIST OF TABLES (Continued)

Table 4.2-7 Calculated Fast Neutron Flux (E > 0.1 MeV) at the Center of Reactor Vessel Surveillance Capsules................................................................................................

4-18 Table 4.2-8 Calculated Fast Neutron Fluence (E > 0.1 MeV) at the Center of Reactor Vessel Surveillance Capsules....................................................................................

4-19 Table 4.2-9 Calculated Fast Neutron Flux (E > 0.1 MeV) at the Reactor Vessel Clad/Base M etal Interface........................................................................................

4-20 Table 4.2-10 Calculated Fast Neutron Fluence (E > 0.1 MeV) at the Reactor Vessel Clad/Base M etal Interface........................................................................................

4-21 Table 4.2-11 Calculated Fast Neutron Flux (E > 0.1 MeV) at the Cavity Sensor Set Locations..........................................................................................................

4-22 Table 4.2-12 Calculated Fast Neutron Fluence (E > 0.1 MeV) at the Cavity Sensor Set Locations...............................................................................................................

4-23 Table 4.2-13 Calculated Iron Displacement Rate at the Center of Reactor Vessel Surveillance Capsules................................................................................................

4-24 Table 4.2-14 Calculated Iron Displacements at the Center of Reactor Vessel Surveillance C ap sules.......................................................................................................................

4-25 Table 4.2-15 Calculated Iron Displacement Rate at the Reactor Vessel Clad/Base Metal Interface.......................................................................................................................

4-26 Table 4.2-16 Calculated Iron Displacements at the Reactor Vessel Clad/Base M etal Interface............................................................................................................

4-27 Table 4.2-17 Calculated Iron Displacement Rate at the Cavity Sensor Set Locations............ 4-28 Table 4.2-18 Calculated Iron Displacements at the Cavity Sensor Set Locations.................... 4-29 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 V 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 WCAP-15334, Rev. 0 November 1999 WCAP-15334, Rev. 0 November 1999

vii LIST OF TABLES (Continued)

Table 5.2-3 Derived Exposure Rates from the Surveillance Capsule U Dosimetry W ithdrawn at the End of Fuel Cycle 7......................................................................

5-6 Table 5.2-4 Derived Exposure Rates from the Surveillance Capsule Y Dosimetry W ithdrawn at the End of Fuel Cycle 8......................................................................

5-7 Table 5.2-5 Derived Exposure Rates from the Surveillance Capsule Z Dosimetry W ithdrawn at the End of Fuel Cycle 8......................................................................

5-8 Table 5.2-65 Derived Exposure Rates from the Surveillance Capsule W Dosimetry W ithdrawn at the End of Fuel Cycle 10....................................................................

5-9 Table 6.1-1 Summary of Reaction Rates Derived from Multiple Foil Sensor Sets C ycle 12 Irradiation......................................................................................................

6-3 Table 6.1-2

-Fe(np), 58Ni(np), and 'Co(ny) Reaction Rates Derived from the Stainless Steel Gradient Chain at 0.50 - Cycle 12 Irradiation...........................................................

6-4 Table 6.1-3

'Fe(np), 58Ni(np), and -Co(ny) Reaction Rates Derived from the Stainless Steel Gradient Chain at 14.50 - Cycle 12 Irradiation.........................................................

6-5 Table 6.1-4 4Fe(np), 'Ni(np), and "SCo(ny) Reaction Rates Derived from the Stainless Steel Gradient Chain at 29.50 - Cycle 12 Irradiation.........................................................

6-6 Table 6.1-5

'Fe(np), 'Ni(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 G Dosimetry Evaluation 0.5 Degree Azimuth - Core Midplane - Cycle 12 Irradiation...............................

6-10 Table 6.1-7 Best Estimate Exposure Rates From the Capsule H Dosimetry Evaluation 14.5 Degree Azimuth - Core Midplane - Cycle 12 Irradiation..............................

6-11 Table 6.1-8 Best Estimate Exposure Rates from the Capsule I Dosimetry Evaluation 29.5 Degree Azimuth - Core Midplane - Cycle 12 Irradiation.............................

6-12 Table 6.1-9 Best Estimate Exposure Rates from the Capsule K Dosimetry Evaluation 44.5 Degree Azimuth - Core Midplane - Cycle 12 Irradiation.............................

6-13 Table 6.1-10 Best Estimate Exposure Rates From The Capsule J Dosimetry Evaluation 44.5 Degree Azimuth - Top of Core - Cycle 12 Irradiation...................................

6-14 Table 6.1-11 Best Estimate Exposure Rates from the Capsule L Dosimetry Evaluation 45.0 Degree Azimuth - Bottom of Core - Cycle 12 Irradiation.............................

6-15 WCAP-15334, Rev. 0 November 1999

viii Table 7.1-1 Table 7.2-1 Table 8.1-1 Table 8.1-2 Table 8.1-3 Table 8.1-4 Table 8.1-5 Table 8.1-6 Table 8.1-7 Table 8.1-8 Table 8.1-9 Table 8.1-10 LIST OF TABLES (Continued)

Comparison of Best Estimate and Calculated Exposure Rates from Surveillance Capsule and Cavity Dosimetry Irradiations......................................

7-2 Comparison of Measured and Calculated Neutron Sensor Reaction Rates from Surveillance Capsule and Cavity Dosimetry Irradiations...........................

7-5 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beitline Region of the McGuire Unit 2 Reactor Vessel 0 D egree A zim uthal A ngle..........................................................................................

8-3 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 15 D egree A zim uthal A ngle........................................................................................

8-4 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltlne Region of the McGuire Unit 2 Reactor Vessel 30 D egree A zim uthal A ngle........................................................................................

8-5 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 45 D egree A zim uthal A ngle........................................................................................

8-6 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltlne Region of the McGuire Unit 2 Reactor Vessel 0 D egree A zim uthal A ngle..........................................................................................

8-7 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 15 D egree A zim uthal A ngle........................................................................................

8-8 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 30 D egree A zim uthal A ngle........................................................................................

8-9 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 45 D egree A zim uthal A ngle......................................................................................

8-10 Summary of Best Estimate Iron Atom Displacement [dpa] Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 0 D egree A zim uthal A ngle.........................................................................................

8-11 Summary of Best Estimate Iron Atom Displacement [dpa] Projections for the Beitline Region of the McGuire Unit 2 Reactor Vessel 15 D egree A zim uthal Angle......................................................................................

8-12 WCAP-15334, Rev. 0 November 1999 WCAP-15334, Rev. 0 November 1999

ix Table 8.1-11 Table 8.1-12 Table 8.2-1 Table 8.2-2 Table 8.2-3 Table A-1 Table A-2 Table A-3 Table A-4 Table A-5 Table A-6 Table A-7 Table B-1 Table B-2 Table B-3 LIST OF TABLES (Continued)

Summary of Best Estimate Iron Atom Displacement [dpa] Projections for the Beitline Region of the McGuire Unit 2 Reactor Vessel 30 D egree A zim uthal A ngle......................................................................................

8-13 Summary of Best Estimate Iron Atom Displacement [dpal projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 45 D egree A zim uthal A ngle......................................................................................

8-14 Fast Neutron Fluence (E > 1.0 MeV) at Key Forging and Weld Locations of M cG uire U nit 2...........................................................................................................

8-16 Fast Neutron Fluence (E > 0.1 MeV) at Key Forging and Weld Locations of M cG uire U nit 2...........................................................................................................

8-17 Iron Atom Displacements (dpa) at Key Forging and Weld Locations of M cG uire U nit 2...........................................................................................................

8-18 McGuire Unit 2 Operating History - Cycles 1 Through 12...................................

A-2 Radiometric Counting Results from Sensors Removed from Capsule V............ A-5 Radiometric Counting Results from Sensors Removed from Capsule X............ A-6 Radiometric Counting Results from Sensors Removed from Capsule U........... A-7 Radiometric Counting Results from Sensors Removed from Capsule Y............ A-8 Radiometric Counting Results from Sensors Removed from Capsule Z............ A-9 Radiometric Counting Results from Sensors Removed from Capsule W......... A-10 McGuire Unit 2 Operating History - Cycle 12.........................................................

B-2 McGuire Unit 2 Dosimeter Capsule Contents for Cycle 12....................................

B-3 Radiometric Counting Results from Sensors Removed from Cycle 12 Cavity Dosimetry Set 2S-1 Capsules G, H, I, J, K, and L........................................

B-4 WCAP-15334, Rev. 0 November 1999 WCAP-15334, Rev. 0 November 1999

xi Figure 1-1 Figure 2.1-1 Figure 2.1-2 Figure 2.2-1 Figure 3.1-1 Figure 3.1-2 Figure 6.1-1 Figure 6.1-2 Figure 6.1-3 Figure 6.1-4 LIST OF FIGURES Description of Reactor Vessel Beitline Materials.....................................................

1-3 Axial Location of Multiple Foil Sensor Sets.............................................................

2-3 Irradiation Capsule for Cavity Sensor Sets...............................................................

2-4 Neutron Sensor Locations within Internal Surveillance Capsules....................... 2-6 Reactor Geometry Showing A 450 RE Sector..........................................................

3-4 Internal Surveillance Capsule Geometry..................................................................

3-5 54Fe (np) 'Mn Reaction Rates Derived from Stainless Steel Gradient Chain at 0.5 Degrees in the Reactor Cavity - Cycle 12 Irradiation....................................

6-8

'Fe (np) SIMn Reaction Rates Derived from Stainless Steel Gradient Chain at 14.5 Degrees in the Reactor Cavity - Cycle 12 Irradiation..................................

6-8

'Fe (np) 'Mn Reaction Rates Derived from Stainless Steel Gradient Chain at 29.5 Degrees in the Reactor Cavity - Cycle 12 Irradiation..................................

6-9

'Fe (np) 14Mn Reaction Rates Derived from Stainless Steel Gradient Chain at 44.5 Degrees in the Reactor Cavity - Cycle 12 Irradiation..................................

6-9 WCAP-15334, Rev. 0 November 1999 November 1999 WCAP-15334, Rev. 0

xiii EXECUTIVE

SUMMARY

All of the calculations and dosimetry evaluations presented in this report have been based on the BUGLE-96 nuclear cross-section data library derived from ENDF/B-VI.

The analysis presented in this report is consistent with the NRC approved methodology detailed in WCAP-14040-NP-A"1].

During Cycle 12 of reactor operation, a reactor cavity measurement program was instituted at McGuire Unit 2 to provide 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 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 six internal surveillance capsules withdrawn following Cycle 1, 5, 7, 8 and 10, resulting in the following projected best estimate fast neutron fluence levels at the inner radius of the reactor vessel wall:

MIDPLANE/MAXIMUM 4) (E > 1.0 MeV) [n/cm']

Azimuth 00 150 300 450 EOC 10 3.32E+18 4.96E+18 4.67E+18 5.28E+18 EOC 12 3.97E+18 5.93E+18 5.64E+18 6.36E+18 The maximum exposure location occurs at the axial core midplane. Since the intermediate and lower shell forgings are joined by a circumferential weld that is located at essentially the same location, all of these materials see approximately the same neutron exposure.

Based on the continued use of an average (Cycles 10-12) low leakage fuel loading pattern, the projected maximum fast neutron exposure of the vessel beltline materials at 21, 34, and 51 effective full power years of operation is summarized as follows:

4) (E > 1.0 MeV) [rn/cm2 Reactor Vessel Beltline Material Intermediate Shell Forging - 526840 Circumferential Weld - W05 Lower Shell Forging - 411337 Circumferential Weld - W04 Lower Transition Shell Forging - 527428 21 EFPY 34 EFPY 1.07E+19 1.68E+19 51 EFPY 2.47E+19 4.37E+18 6.86E+18 1.01E+19 WCAP-15334, Rev. 0 November 1999 November 1999 WCAP-15334, Rev. 0

xiv As further data are accumulated from subsequent irradiations, the neutron environment in the vicinity of the Unit 2 reactor 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.

WCAP-15334, Rev. 0 November 1999 WCAP-15334, Rev. 0 November 1999

1.-i 1

OVERVIEW OF THE PROGRAM The Reactor Cavity Neutron Measurement Program initiated at McGuire Unit 2 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'21, 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 reactor 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 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 2 reactor vessel is provided in Figure 1-1.

In this case, the beltline region is constructed of three ring forgings and two circumferential welds.

Each of these five materials must be considered in the overall embrittlement assessments of the reactor vessel.

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

These locations are commonly referred to as the 1/4T and 3AT 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.

In the determination of values of RT* 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 belthlne materials are required. These maximum levels will, of course, occur at the vessel inner radius.

WCAP-15334, Rev. 0, November 1"t Overview of the Program

1-2 In the event that a probabilistic fracture mechanics evaluation of the reactor vessel is performed, or if an evaluation of thermal annealing and subsequent material re-embrittlement is undertaken, a complete embrittlement profile is required for the entire volume of the reactor vessel beltline. The determination of this embrittlement profile would, in turn, necessitate the evaluation of neutron exposure gradients throughout the entire beltline.

The methodology used to provide these required best estimate neutron exposure evaluations for the McGuire Unit 2 reactor vessel is based on the underlying philosophy that, in order to minimize the uncertainties associated with vessel exposure projections, plant specific neutron transport calculations must be supported by benchmarking of the analytical approach, comparison with industry wide power reactor data bases of surveillance capsule and reactor cavity dosimetry, and, ultimately, by validation with plant specific 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 reactor vessel wall.

2.

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

3.

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

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 reactor vessel wall is described and used to project the exposure of the vessel itself.

Results of exposure evaluations from surveillance capsule dosimetry withdrawn at the end of Cycles 1, 5, 7, 8, and 10 as well as cavity dosimetry results from Cycle 12 are incorporated to provide the integrated exposure of the reactor 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 reactor 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 forgings comprising the intermediate, lower, and transition shells.

Overview of the Program WCAP-15334, Rev. 0, November 1999

1-3 All of the calculations and dosimetry evaluations presented in this report have been based on the BUGLE-96 nuclear cross-section data library derived from ENDF/B-VI.

Overview of the Program WCAP-15334, 1(ev. U, November 1'

WCAP-15334, R~ev. 0, November 1999 Overview of the Program

1-4 Intermediate Shell Forging 526840 Circumferential Weld W05 Core Lower Shell Forging 411337 o

Circumferential Weld W04 Transition Shell Forging 527428 Figure 1-1 Description of Reactor Vessel Beltline Materials Overview of the Program WCAP-15334, Rev. 0, November 1999

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 reactor 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 reactor 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 44.50 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 Bravo Loop hot leg nozzle. The support bar was suspended from support plates welded to the liner plate at El. 746'+10'". 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 clamps attached to the mounting plates bolted to the bioshield wall in the sump area below the reactor vessel. The design of the dosimetry support bar along with the WCAP-15334, Rev. 0, November 1999 Description of the Measurement Program

2-2 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 Type 6061 aluminum 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 1 and 2 was such that cadmium shields inserted into position 2 did not introduce perturbations in the thermal flux in position 1. Type 6061 aluminum 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 S4Fe(np)54Mn, ' 8Ni(n,p)5 8Co, and S9Co(n,,y)6Co 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 are used to confirm the composition of the 18-8 stainless steel gradient chain.

Vescription ot the Measurement Program WCAP-15334, Rev. 0, November 1999 Description of the Measurement Program WCAP-15334, Rev. 0, November 1999

2-3 Core Midplane 0.50 14.50 29.50 44.50 Figure 2.1-1 Axial Location of Multiple Foil Sensor Sets Description of the Measurement Program WCAP-15334, Rev. 0, November 1999 0

Multiple Foil Set

-- Gradient Chain Description of the Measurement Program 2-3 WCAP-15334, Rev. 0, November 1999

2-4 Figure 2.1-2 Irradiation Capsule for Cavity Sensor Sets Vescription ot the Measurement Program WCAP-15334, Rev. 0, November 1999 Description of the Measurement Program WCAP-15334, Rev. 0, November 1999

2-5

2.2 DESCRIPTION

OF SURVEILLANCE CAPSULE DOSIMETRY At the conclusion of the first fuel cycle at McGuire Unit 2, the first material surveillance capsule was withdrawn from its position between the neutron pad and the reactor vessel. The second internal surveillance capsule was withdrawn at the conclusion of Cycle 5, a third capsule was withdrawn at the end of Cycle 7, two more were withdrawn following Cycle 8, and the sixth and final capsule was withdrawn after Cycle 10. The neutron dosimetry contained within these capsules provided a measure of the integral exposure received by the capsules during their respective irradiation periods; i.e., Cycle 1, Cycles 1 through 5, Cycles 1 through 7, Cycles 1 through 8, and Cycles 1 through 10.

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

Relative to Figure 2.2-1, copper, iron, nickel, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within each capsule.

The cadmium-shielded uranium and neptunium fission monitors were accommodated within a dosimeter block located near the center of the capsule. Specific information pertinent to the individual sensor sets included in Capsules V, X, U, Y, Z, and W is provided in the appendices to this report.

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2-6 C.

Figure 2.2-1 Neutron Sensor Locations within Internal Surveillance Capsules Description of the Measurement Program WCAP-15334, Rev. 0, November 1999 Description of the Measurement Program WCAP-15334, Rev. 0, November 1999

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 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 reactor 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 vessel.

4.

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

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 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 Neutron Transport and Dosimetry Evaluation Methodologies WCAP-15334, Rev. 0, November 1999 WCAP-15334, Rev. 0, November 1999 Neutron Transport and Dosirnetry Evaluation Methodologies

3-2 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 carried out in rO geometry using the DORT two-dimensional discrete ordinates transport theory code[31 and the BUGLE-96 cross-section library"41. The BUGLE-96 library is a 47-neutron energy-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 S8 order of angular quadrature.

The reference forward calculation was normalized to a core midplane power density characteristic of operation at a thermal power level of 3411 MWt. The 3411 MWt power level represents the rated operating power for the McGuire Unit 2 reactor.

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

Due to the use of this bounding spatial power distribution, the results from the reference forward calculation establish conservative exposure projections for reactors of this design operating at the 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 +2o 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-96 library. Adjoint source locations were chosen at each of the azimuthal locations containing cavity dosimetry as well as at several key azimuths on the reactor vessel inner radius. In addition, adjoint calculations were carried out 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, 0 (E > 1.0 MeV).

WCAP-15334, Rev. 0, November 1999 Neutron Transport and Dosimetry Evaluation Methodologies

3-3 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,9) = f f f I(r,O,E)S(r,0,E)rdrdOdE r

0 E

where:

R(r,0) =

p (E > 1.0 MeV) at radius r and azimuthal angle 0.

I(r,0,E)=

Adjoint source importance function at radius r, azimuthal angle 0, and neutron source energy E.

S(r,0,E)=

Neutron source strength at core location r,0 and energy E.

It is important to note that the cycle specific neutron source distributions, S(r,0,E), utilized with the adjoint importance functions, I(r,0,E), permitted the use not only of fuel cycle specific spatial variations of fission rates within the reactor core, but, also allowed for the inclusion of the effects of the differing neutron yield per fission and the variation in fission 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] / [0 (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 atom 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] / [0 (E > 1.0 MeV)]

and [0 (E > 0.1 MeV)] / [ 0 (E > 1.0 MeV)] ratios from the reference forward analysis in conjunction with the cycle specific 0 (E > 1.0 MeV) solutions from the individual adjoint evaluations.

In particular, after defining the following exposure rate ratios,

[dpa/ sec ]

O(E >1.0MeV)

R= (E>O.1MeV)

O(E > 1.0 MeV)

WCAP-15334, Rev. 0, November 1999 Neutron Transport and Dosimetry Evaluation Methodologies

3-4 the corresponding fuel cycle specific exposure rates at the adjoint source locations were computed from the following relations:

dpa/sec (E > 1.0 MeV)] Ri O(E > 0.1 MeV)= [ O(E > 1.0 MeV)] R 2 All absolute calculations were also normalized to the current rated power level for McGuire Unit 2, 3411 MWt.

Neutron Transport and Dosimetry Evaluation Methodologies WCAP-15334, Rev. 0, November 1999 Neutron Transport and Dosirnetry Evaluation Methodologies WCAP-15334, Rev. 0, November 1999

3-5 400 0 BARREL~

Figure 3.1-1 Reactor Geometry Showing A 450 RE Sector Neutron Transport and Dosimetry Evaluation Methodologies WCAP-15334, Rev. 0, November 1999 WCAP-15334, Rev. 0, November 1999 Neutron Transport and Dosimetry Evaluation Methodologies

3-6 Figure 3.1-2 Internal Surveillance Capsule Geometry Neutron Transport and Dosimetry Evaluation Methodologies WCAP-15334, Rev. 0, November 1999 NETO PAD

\\

.6N

1. I-\\\\N WCAP-15334, Rev. 0, November 1999 Neutron Transport and Dosimetry Evaluation Methodologies

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 2, the measurement of specific activities was 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 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 t-ough 16]. 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 'U 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 used in the cavity irradiations, individual sensors were obtained by cutting the chains into a series of segments to provide data points at one foot intervals over an axial span encompassing +/- 6 feet relative to the reactor core midplane.

Neutron Transport and Dosimetry Evaluation Methodologies WCAP-15334, Rev. 0, November 1999

3-8 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 by utility personnel"'*.

For the sensor sets utilized in surveillance capsule and reactor cavity irradiations, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations.

Having the measured specific activities, the operating history of the reactor, and the physical characteristics of the sensors, reaction rates referenced to full power operation at 3411 MWt were determined from the following equation:

A P.

NOFYX-- Cj[1-e-Xtf]e-Xtd 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 P*f (rps/nucleus).

No

=

number of target element atoms per gram of sensor.

F

=

weight fraction of the target isotope in the sensor material.

Y number of product atoms produced per reaction.

Pi

=

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

Pf

=

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

C1

=

calculated ratio of 4) (E > 1.0 MeV) during irradiation period j to the time weighted average 0 (E > 1.0 MeV) over the entire irradiation period.

k

=

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 P /Pf 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 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.

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3-9 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 '6U 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 2 1U in the 'U fission sensors, corrections were also made to both the 'U and *TNp sensor reaction rates to account for gamma ray induced fission reactions occurring over the course of the irradiation. These photofission corrections were, likewise, location dependent and were based on the reference transport calculations described in Section 3.1.1.

For the reactor cavity fission monitors, corrections were also 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 by the vanadium tubing. These correction factors were determined by J. M. Adams, et al and are reported in a recent paperE' 81. For the three fission product nuclides measured the correction factors are:

Daughter Correction 95 Z1.048 103 1.073 137 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"191. 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.

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.

In general, the measured values, f, are linearly related to the flux, 0, by some response matrix, A:

"fS

=

A) g Neutron Transport and Dosimetry Evaluation Methodologies WCAP-15334, Rev. 0, November 1999 Neutron Transport and Dosimetry Evaluation Methodologies WCAP-15334, Rev. 0, November 1999

3-10 where i indexes the measured values belonging to a single data set s, g designates the energy group, and (x delineates spectra that may be simultaneously adjusted. For example, R =

0 ig 0

g relates a set of measured reaction rates, Ri, to a single spectrum, 0g, by the multigroup reaction cross-section, oGg. The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with large assigned uncertainties.

In the least squares adjustment, the continuous quantities (i.e., neutron spectra and cross sections) were approximated in a multi-group format consisting of 53 energy groups. The trial input spectrum was converted to the FERRET 53 group structure using the SAND-II code'201.

This procedure was carried out by first expanding the 47 group calculated spectrum into the SAND-Il 620 group structure using a SPLINE interpolation procedure in regions where group boundaries do not coincide.

The 620-point spectrum was then re-collapsed into the group structure used in FERRET.

The sensor set reaction cross-sections, obtained from the ENDF/B-VI dosimetry file 211, were also collapsed into the 53-energy group structure using the SAND-II code. In this instance, the trial spectrum, as expanded to 620 groups, was employed as a weighting function in the cross section collapsing procedure. Reaction cross-section uncertainties in the form of a 53 x 53 covariance matrix for each sensor reaction were also constructed from the information contained on the ENDF/B-VI data files. These matrices included energy group to energy group uncertainty correlations for each of the individual reactions. However, correlations between cross-sections for different sensor reactions were not included. The omission of this additional uncertainty information does not significantly impact the results of the adjustment.

Due to the importance of providing a trial spectrum that exhibits a relative energy distribution close to the actual spectrum at the sensor set locations, the neutron spectrum input to the FERRET evaluation was 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 Rg, Pgg, where Ra 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 gg = [

+ OeH where:

Neutron Transport and Dosimetry Evaluation Methodologies WCAP-15334, Rev. 0, November 1999 Neutron Transport and Dosimetry Evaluation Methodologies WCAP-15334, Rev. 0, November 1999

3-11 H

(g - g')2 2r 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'2 1. 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 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.

WCA.P-15334, Rev. 0, November 1999 Neutron Transport and Dosimetry Evaluation Methodologies

3-12 These input uncertainty assignments were based on prior experience in using the FERRET least squares adjustment approach in the analysis of neutron dosimetry from surveillance capsule, reactor cavity, and benchmark irradiations. The values are liberal enough to permit adjustment of the input spectrum to fit the measured data for all practical applications.

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

In particular, the best estimate vessel exposure is obtained from the following relationship:

(IBest Est.

= K (DCalc.

where:

(D B-tE 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 reactor vessel wall, additional uncertainty is incurred due to the analytically determined relative ratios among the various measurement points and locations within the reactor vessel wall.

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.

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3-13 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 2 reactor. The uncertainties in the exposure ratios relating dosimetry results to positions within the vessel wall were based on analytical sensitivity studies of the vessel thickness tolerance for the cavity data and on downcomer water density variations and vessel inner radius tolerance for the surveillance capsule measurements.

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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 reactor 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 () (E > 1.0 MeV), ()

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

The reference reaction rates, exposure rates, and exposure rate ratios were used in conjunction with fuel cycle specific adjoint transport calculations from Section 4.2 to provide calculated sensor set reaction rates and to project sensor set exposures in terms of () (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 the spectra from Table 4.1-3 are provided along with the calculated exposure rates in terms of of 4D (E > 1.0 MeV), 4) (E < 0.1 MeV) and dpa/sec.

Again, these data are applicable to the WCAP-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations

4-2 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 Reactor Vessel Wall Data from the reference forward calculation pertinent to the reactor 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 reactor 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 (D (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 [cD (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 reactor vessel inner radius.

Radial gradient information for 'D(E > 1.0 MeV), c1(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.

Kesults or Neutron transport Calculations WCAP 15334, Rev. 0, November 1999 FKesuits of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

Calculated Reference Neutron Energy Spectra at Cavity Sensor Set Locations 3411 MWt, F.=1.20 Lower Energ

[MeVI 1.419E+01 1.221E+01 1.OOOE+01 8.607E+00 7.408E+00 6.065E+00 4.966E+00 3.679E+00 3.012E+00 2.725E+00 2.466E+00 2.365E+00 2.346E+00 2.231E+00 1.921E+00 1.653E+00 1.353E+00 1.003E+00 8.209E-01 7.427E-01 6.081E-01 4.979E-01 3.688E-01 2.972E-01 1.832E-01 1.111E-01 6.738E-02 4.087E-02 3.183E-02 2.606E-02 2.418E-02 2.188E-02 1.503E-02 7.102E-03 3.355E-03 1.585E-03 4.540E-04 2.144E-04 1.013E-04 3.727E-05 1.068E-05 5.043E-06 1.855E-06 8.764E-07 4.140E-07 1.OOOE-07 O.OOOE+00 0.50 3.33E+05 9.OOE+05 3.27E+06 5.76E+06 8.29E+06 1.70E+07 2.23E+07 3.80E+07 2.96E+07 2.30E+07 2.86E+07 1.48E+07 4.39E+06 2.24E+07 6.16E+07 8.65E+07 1.50E+08 3.64E+08 3.60E+08 1.74E+08 8.53E+08 7.59E+08 8.34E+08 1.25E+09 1.53E+09 1.66E+09 1.06E+09 7.86E+08 2.73E+08 1.75E+08 4.36E+08 2.65E+08 5.06E+08 7.44E+08 7.38E+08 6.51E+08 1.03E+08 5.43E+08 5.48E+08 6.93E+08 8.06E+08 4.47E+08 5.54E+08 3.85E+08 3.42E+08 4.63E+08 2.87E+09 14.50 3.84E+05 1.06E+06 3.95E+06 7.07E+06 1.05E+07 2.18E+07 2.98E+07 5.31E+07 4.19E+07 3.29E+07 4.14E+07 2.13E+07 6.30E+06 3.19E+07 8.74E+07 1.23E+08 2.19E+08 5.29E+08 5.25E+08 2.79E+08 1.22E+09 1.14E+09 1.27E+09 1.83E+09 2.46E+09 2.57E+09 1.68E+09 1.26E+09 4.54E+08 3.02E+08 5.80E+08 3.90E+08 8.39E+08 1.22E+09 1.20E+09 1.06E+09 1.64E+09 8.69E+08 8.59E+08 1.08E+09 1.24E+09 6.87E+08 8.44E+08 5.81E+08 5.13E+08 6.72E+08 3.30E+09 29.50 3.38E+05 9.41E+05 3.58E+06 6.45E+06 9.64E+06 2.OOE+07 2.79E+07 5.15E+07 4.16E+07 3.32E+07 4.22E+07 2.19E+07 6.60E+06 3.35E+07 9.21E+07 1.32E+08 2.37E+08 5.85E+08 5.90E+08 3.13E+08 1.40E+09 1.32E+09 1.48E+09 2.14E+09 2.88E+09 3.04E+09 1.99E+09 1.49E+09 5.34E+08 3.56E+08 7.01E+08 4.72E+08 1.01E+09 1.45E+09 1.42E+09 1.26E+09 1.94E+09 1.03E+09 1.02E+09 1.28E+09 1.47E+09 8.12E+08 9.94E+08 6.84E+08 6.02E+08 7.88E+08 3.86E+09 44.50 2.59E+05 7.28E+05 2.84E+06 5.13E+06 7.73E+06 1.57E+07 2.22E+07 4.25E+07 3.52E+07 2.88E+07 3.66E+07 1.94E+07 6.16E+06 3.12E+07 8.58E+07 1.26E+08 2.23E+08 5.60E+08 5.71E+08 2.66E+08 1.39E+09 1.24E+09 1.35E+09 2.07E+09 2.46E+09 2.70E+09 1.70E+09 1.24E+09 4.22E+08 2.69E+08 7.31E+08 4.49E+08 8.25E+08 1.16E+09 1.16E+09 1.01E+09 1.57E+09 8.24E+08 8.28E+08 1.04E+09 1.21E+09 6.69E+08 8.22E+08 5.68E+08 5.04E+08 6.81E+08 4.24E+09 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999 Table 4.1-1 4-3 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

Reference Neutron Sensor Reaction Rates and Exposure Parameters at the Cavity Sensor Set Locations - 3411 Mwt; F = 1.20 Reaction Rate 0.50 14.50

'Cu (n,(x) (Cd) 7.44E-19 9.34E-19

'Ti (n,p) (Cd) 1.01E-17 1.31E-17 "54Fe (n,p) (Cd) 5.47E-17 7.41E-17 "58Ni (n,p) (Cd) 7.64E-17 1.04E-16 2U (n,f) (Cd) 2.74E-16 3.86E-16 237Np (n,f) (Cd) 3.94E-15 5.79E-15 59Co (n,7) 1.48E-13 1.90E-13 "59Co (n,y) (Cd) 5.09E-14 7.93E-14 238U (y,f) 1.34E-17 1.58E-17 237Np ('*,fl 3.77E-17 4.46E-17 Exposure Parameter 0p (E > 1.0 MeV)

(0 (E > 0.1 MeV)

Iron Atom Displacement Rate S(E > 0.1 MeV) / 0 (E > 1.0 MeV)

[dpa/sec] / 4) (E > 1.0 MeV) 2U (yf) / 2U (nf) 237Np (y, f) / 237Np (n,f) 9.02E+08 8.52E+09 2.97E-12 9.45 3.30E-21 0.0488 0.0096 Neutron Flux 1.30E+09 1.29E+10 29.50 8.55E-19 1.21E-17 7.13E-17 1.02E-16 3.98E-16 6.49E-15 2.23E-13 9.37E-14 1.57E-17 4.42E-17

[n/cm 2-s]

1.39E+09 1.49E+10 dpa/sec 4.46E-12 5.05E-12 9.96 10.77 3.44E-21 3.65E-21 0.0410 0.0393 0.0077 0.0068 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999 Table 4.1-2 4-4 44.50 6.79E-19 9.68E-18 5.92E-17 8.59E-17 3.58E-16 6.06E-15 2.20E-13 7.67E-14 1.38E-17 3.89E-17 1.29E+09 1.36E+10 4.58E-12 10.61 3.56E-21 0.0385 0.0064 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

4-5 Table 4.1-3 Calculated Reference Neutron Energy Spectra at Surveillance Capsule Locations - 3411 Mwt; F. = 1.2

• (E) [n/cm2-s]

q*(E) [n/cmZ-s]

Lower Single Lower Single Energy Dual Capsule Capsule Ener Dual Capsule Capsule

[MeVI 31.50 34.00 34.00

[MeV]

31.50 34.00 34.00 1.419E+01 1.221E+01 1.OOOE+01 8.607E+00 7.408E+00 6.065E+00 4.966E+00 3.679E+00 3.012E+00 2.725E+00 2.466E+00 2.365E+00 2.346E+00 2.231E+00 1.921E+00 1.653E+00 1.353E+00 1.003E+00 8.209E-01 7.427E-01 6.081E-01 4.979E-01 3.688E-01 2.972E-01 1.52E+07 4.92E+07 2.20E+08 4.37E+08 7.80E+08 1.91E+09 3.02E+09 6.43E+09 5.56E+09 4.40E+09 5.45E+09 2.75E+09 7.69E+08 3.93E+09 1.15E+10 1.44E+10 2.35E+10 5.18E+10 4.02E+10 1.87E+10 6.15E+10 5.81E+10 5.92E+10 6.36E+10 1.59E+07 5.17E+07 2.34E+08 4.66E+08 8.39E+08 2.07E+09 3.31E+09 7.20E+09 6.36E+09 5.04E+09 6.27E+09 3.16E+09 8.85E+08 4.53E+09 1.33E+10 1.68E+10 2.76E+10 6.18E+10 4.84E+10 2.25E+10 7.47E+10 7.10E+10 7.20E+10 7.83E+10 1.56E+07 5.10E+07 2.31E+08 4.60E+08 8.28E+08 2.04E+09 3.26E+09 7.10E+09 6.30E+09 4.99E+09 6.22E+09 3.14E+09 8.79E+08 4.50E+09 1.33E+10 1.69E+10 2.79E+10 6.41E+10 5.06E+10 2.33E+10 7.89E+10 7.58E+10 7.70E+10 8.59E+10 1.832E-01 1.111E-01 6.738E-02 4.087E-02 3.183E-02 2.606E-02 2.418E-02 2.188E-02 1.503E-02 7.102E-03 3.355E-03 1.585E-03 4.540E-04 2.144E-04 1.013E-04 3.727E-05 1.068E-05 5.043E-06 1.855E-06 8.764E-07 4.140E-07 1.OOOE-07 7.71E+10 8.07E+10 5.22E+10 4.93E+10 1.59E+10 6.15E+09 1.78E+10 1.20E+10 2.06E+10 4.55E+10 5.09E+10 4.74E+10 7.96E+10 3.90E+10 4.64E+10 6.07E+10 7.17E+10 3.90E+10 4.55E+10 2.86E+10 1.94E+10 2.33E+10 9.37E+10 9.88E+10 6.35E+10 6.01E+10 1.93E+10 7.44E+09 2.19E+10 1.48E+10 2.48E+10 5.48E+10 6.13E+10 5.72E+10 9.64E+10 4.70E+10 5.61E+10 7.35E+10 8.69E+10 4.72E+10 5.49E+10 3.44E+10 2.32E+10 2.78E+10 9.79E+10 1.06E+11 6.54E+10 6.28E+10 1.96E+10 7.47E+09 2.27E+10 1.59E+10 2.45E+10 5.44E+10 6.08E+10 5.71E+10 9.59E+10 4.60E+10 5.49E+10 7.16E+10 8.38E+10 4.50E+10 5.10E+10 3.11E+10 2.03E+10 2.34E+10 O.OOOE+00 6.31E+10 7.49E+10 6.33E+10 Results of Neutron Transport Calculations WCAI-'-15334, Key. U, N ovemter 1''

WCAP-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations

Reference Neutron Sensor Reaction Rates and Exposure Parameters at the Center of the Surveillance Capsules - 3411 Mwt; F. = 1.20 Dual Capsule Reaction Rate

'Cu (n,a) 6OCo

'Fe (n,p) T4Mn 58Ni (np) 58Co

'U (n,f) F.P. (Cd) 237Np (n,f) F.P. (Cd) 59Co (n,y) 6Co

'9Co (n,7) 6°Co (Cd) 23U (yf) F.P.

1 Np (yf) F.P.

Exposure Parameter p(E > 1.0 MeV) p(E > 0.1 MeV)

Iron Atom Displacement Rate S(E > 0.1 MeV) / 0 (E > 1.0 MeV)

[dpa/sec] / p (E > 1.0 MeV) 21U (y,f) / 2U (n,f) 237Np (y,f) / 37Np (n,f) 31.50 6.90E-17 7.95E-15 1.12E-14 4.34E-14 4.24E-13 6.12E-12 4.25E-12 1.58E-15 4.40E-15 34.00 7.44E-17 8.89E-15 1.26E-14 5.00E-14 5.05E-13 7.36E-12 5.13E-12 1.85E-15 5.17E-15 Single Capsule 34.00 7.34E-17 8.79E-15 1.25E-14 5.OOE-14 5.21E-13 7.06E-12 4.98E-12 1.65E-15 4.59E-15 Neutron Flux [n/cm 2-s]

1.38E+11 1.61E+11 1.63E+11 6.07E+11 7.33E+11 7.71E+11 2.64E-10 4.41 1.92E-21 3.64E-02 1.04E-02 dpa/sec 3.14E-10 Ratios 4.55 1.95E-21 3.71E-02 1.02E-02 3.25E-10 4.73 1.99E-21 3.30E-02 8.80E-03 Results of Neutron Transport Calculations Table 4.1-4 4-6 WCAP-15334, Rev. 0, November 1999

4-7 Summary of Exposure Rates at the Reactor Vessel Clad/Base Metal Interface (150 Neutron Pad Octant)

Neutron Flux [n/cm'-secl (E > 1.0 MeV) 1.76E+10 1.88E+10 2.19E+10 2.72E+10 3.14E+10 3.05E+10 2.59E+10 2.58E+10 2.95E+10 3.12E+10 (E > 0.1 MeV) 3.73E+10 3.99E+10 4.67E+10 5.80E+10 6.71E+10 6.61E+10 6.OOE+10 6.47E+10 7.53E+10 7.90E+10 dpa/sec 2.73E-11 2.91E-11 3.39E-11 4.17E-11 4.80E-11 4.68E-11 4.04E-11 4.09E-11 4.68E-11 4.94E-11 E > 0.1 MeV E > 1.0 MeV 2.12 2.12 2.13 2.14 2.14 2.17 2.32 2.51 2.55 2.53 dpa/sec E > 1.0 MeV 1.55E-21 1.55E-21 1.55E-21 1.54E-21 1.53E-21 1.53E-21 1.56E-21 1.59E-21 1.59E-21 1.58E-21 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999 Table 4.1-5 0 (deg) 0.25 5.25 10.00 15.25 20.25 25.25 30.25 35.10 40.25 44.75 WCA-P-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations

Relative Radial Distribution of Neutron Flux (E > 1.0 MeV) Within the Reactor Vessel Wall (150 Neutron Pad Octant)

Base Metal Inner Radius =

Base Metal 1/4AT =

Base Metal 1/2/2T =

Base Metal 3/T =

Base Metal Outer Radius -

Radius

[cm]

220.35 221.00 222.30 223.60 224.89 225.87 227.01 228.63 230.09 231.39 232.68 234.14 235.76 236.90 237.88 239.18 240.47 241.77 242.42 220.35 cm.

225.87 cm.

231.39 cm.

236.90 cm.

242.42 cm.

Results of Neutron Transport Calculations Table 4.1-6 4-8 0.00 1.000 0.959 0.852 0.741 0.636 0.564 0.489 0.398 0.329 0.276 0.232 0.190 0.152 0.130 0.113 0.093 0.077 0.063 0.060 Azimuthal Angle 15.00 30.00 1.000 1.000 0.958 0.960 0.851 0.852 0.738 0.739 0.632 0.635 0.559 0.562 0.484 0.488 0.393 0.396 0.324 0.327 0.272 0.276 0.228 0.231 0.187 0.190 0.149 0.152 0.127 0.130 0.110 0.113 0.091 0.093 0.074 0.077 0.060 0.062 0.057 0.060 45.00 1.000 0.957 0.847 0.732 0.625 0.551 0.475 0.384 0.315 0.263 0.220 0.179 0.142 0.120 0.104 0.085 0.069 0.055 0.053 Note:

WCAP-15334, Rev. 0, November 1999

Table 4.1-7 Relative Radial Distribution of Neutron Flux (E > 0.1 MeV) Within the Reactor Vessel Wall (150 Neutron Pad Octant)

Radius

[CM]

220.35 221.00 222.30 223.60 224.89 225.87 227.01 228.63 230.09 231.39 232.68 234.14 235.76 236.90 237.88 239.18 240.47 241.77 242.42 0.00 1.000 1.012 0.997 0.960 0.914 0.876 0.830 0.765 0.707 0.655 0.605 0.550 0.491 0.450 0.415 0.370 0.326 0.284 0.276 Note:

Base Metal Inner Radius =

Base Metal 1/4T =

Base Metal 1/2T =

Base Metal 34T =

Base Metal Outer Radius =

Azimuthal Angle 15.00 30.00 1.000 1.000 1.010 1.013 0.991 0.997 0.950 0.960 0.901 0.914 0.861 0.876 0.814 0.831 0.748 0.767 0.689 0.710 0.637 0.659 0.587 0.609 0.532 0.555 0.475 0.497 0.435 0.457 0.401 0.423 0.357 0.378 0.314 0.334 0.269 0.288 0.260 0.279 220.35 cm.

225.87 cm.

231.39 cm.

236.90 cm.

242.42 cm.

Results of Neutron Transport Calculations WCAP-15334, Rev. U, November 19

4-9 45.00 1.000 1.007 0.983 0.939 0.886 0.844 0.794 0.725 0.664 0.611 0.559 0.503 0.444 0.404 0.369 0.324 0.280 0.236 0.227 WCAP-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations

4-10 Table 4.1-8 Relative Radial Distribution of Iron Displacement Rate (dpa) Within the Reactor Vessel Wall (150 Neutron Pad Octant)

Radius Azimuthal Angle

[cm]

0.00 15.00 30.00 45.00 220.35 1.000 1.000 1.000 1.000 221.00 0.964 0.964 0.968 0.965 222.30 0.876 0.875 0.882 0.878 223.60 0.784 0.782 0.793 0.788 224.89 0.698 0.695 0.709 0.703 225.87 0.638 0.634 0.651 0.643 227.01 0.575 0.571 0.589 0.580 228.63 0.496 0.491 0.511 0.501 230.09 0.434 0.429 0.450 0.439 231.39 0.385 0.380 0.402 0.389 232.68 0.342 0.337 0.358 0.345 234.14 0.299 0.294 0.315 0.301 235.76 0.257 0.252 0.272 0.257 236.90 0.230 0.226 0.245 0.229 237.88 0.208 0.204 0.223 0.207 239.18 0.182 0.179 0.196 0.179 240.47 0.158 0.154 0.171 0.153 241.77 0.136 0.131 0.146 0.128 242.42 0.132 0.127 0.142 0.123 Note:

Base Metal Inner Radius =

220.35 cm.

Base Metal 1/4T =

225.87 cm.

Base Metal 11/22T =

231.39 cm.

Base Metal 34T =

236.90 cm.

Base Metal Outer Radius 242.42 cm.

Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

4-11 4.2 FUEL CYCLE SPECIFIC ADJOINT TRANSPORT CALCULATIONS Results of the fuel cycle specific adjoint transport calculations for the first 12 cycles of operation at McGuire Unit 2 are summarized in Tables 4.2-1 through 4.2-18. The data listed in these tables establish the means for absolute comparison of analysis and measurement for the Cycle 12 reactor cavity dosimetry irradiation as well as for the six 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 reactor 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 2.I23 dugh31 The data extracted from the fuel cycle design reports represented cycle averaged relative fuel assembly powers and burnups as well as cycle averaged relative axial distributions. Therefore, the results of the adjoint evaluation provided data in terms of fuel cycle averaged neutron flux which, when multiplied by the appropriate fuel cycle length, produced the incremental fast neutron exposure for the fuel cycle.

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 reactor 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 or fluence (E > 0.1 MeV) in Tables 4.2-7 through 4.2-12 and iron atom displacements in Tables 4.2-13 through 4.2-18.

Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

4-12 Table 4.2-1 Calculated Fast Neutron Flux (E > 1.0 MeV) at the Center of Reactor Vessel Surveillance Capsules Neutron Flux [n/cm2-sec]

Dual Capsule Cycle No.

1 2

3 4

5 6

7 8

9 10 11 12 31.50 9.96E+10 1.15E+11 1.01E+11 8.89E+10 8.56E+10 8.28E+10 8.55E+10 8.43E+10 8.39E+10 7.59E+10 7.66E+10 7.48E+10 340 1.15E+11 1.34E+11 1.18E+11 1.02E+11 9.67E+10 9.28E+10 9.63E+10 9.55E+10 9.65E+10 8.67E+10 8.72E+10 8.49E+10 Single Capsule 340 1.17E+11 1.36E+11 1.20E+11 1.03E+11 9.80E+10 9.41E+10 9.76E+10 9.68E+10 9.78E+10 8.79E+10 8.84E+10 8.61E+10 J*esults 01 Neutron transport Calculations WCAP 15334, Rev. 0, November 1999 R~esults of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

4-13 Table 4.2-2 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Center of Reactor Vessel Surveillance Capsules Cumulative Fluence [r/cm2]

Cycle Length EFPS 3.239E+07 2.158E+07 2.311E+07 2.664E+07 2.769E+07 2.891E+07 3.071E+07 3.541E+07 3.446E+07 3.701E+07 3.538E+07 12 3.794E+07 Dual Capsule 31.50 3.23E+18 5.72E+18 8.04E+18 1.04E+19 1.28E+19 1.52E+19 1.78E+19 2.08E+19 2.37E+19 2.65E+19 2.92E+19 3.20E+19 340 3.74E+18 6.64E+18 9.36E+18 1.21E+19 1.47E+19 1.74E+19 2.04E+19 2.38E+19 2.71E+19 3.03E+19 3.34E+19 3.66E+19 Results of Neutron Transport Calculations WLA1-1b334, Key. U, Novemoer 1'''

Cycle 1

2 3

4 5

6 7

8 9

10 11 Single Capsule 340 3.79E+18 6.73E+18 9.50E+18 1.22E+19 1.50E+19 1.77E+19 2.07E+19 2.41E+19 2.75E+19 3.07E+19 3.39E+19 3.71E+19 WCAP-15334, R~ev. 0, rNovember h99 Results of Neutron Transport Calculations

4-14 Table 4.2-3 Calculated Fast Neutron Flux (E > 1.0 MeV) at the Reactor Vessel Clad/Base Metal Interface (150 Neutron Pad Octant)

Neutron Flux Cycle 1

2 3

4 5

6 7

8 9

10 11 12 00 1.27E+10 1.50E+10 1.22E+10 1.25E+10 1.21E+10 1.33E+10 1.22E+10 1.25E+10 1.16E+10 1.01E+10 1.02E+10 9.39E+09 150 1.92E+10 2.28E+10 1.89E+10 1.87E+10 1.85E+10 1.93E+10 1.87E+10 1.84E+10 1.68E+10 1.48E+10 1.50E+10 1.40E+10

[n/cm2-secl 300 1.90E+10 2.21E+10 1.91E+10 1.74E+10 1.68E+10 1.66E+10 1.69E+10 1.66E+10 1.62E+10 1.47E+10 1.48E+10 1.44E+10 450 2.25E+10 2.59E+10 2.35E+10 1.95E+10 1.82E+10 1.75E+10 1.81E+10 1.82E+10 1.87E+10 1.65E+10 1.65E+10 1.60E+10 Results of Neutron Transport Calculations WCAP 15334, Rev. 0, November 1999 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

Table 4.2-4 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Reactor Vessel Clad/Base Metal Interface (15' Neutron Pad Octant)

Cycle Length EFPS 3.239E+07 2.158E+07 2.311E+07 2.664E+07 2.769E+07 2.891E+07 3.071E+07 3.541E+07 3.446E+07 3.701E+07 3.538E+07 3.794E+07 Cumulative Fluence [nrcm2l 00 4.12E+17 7.35E+17 1.02E+18 1.35E+18 1.69E+18 2.07E+18 2.45E+18 2.89E+18 3.29E+18 3.66E+18 4.02E+18 4.38E+18 150 6.22E+17 1.11E+18 1.55E+18 2.05E+18 2.56E+18 3.12E+18 3.69E+18 4.35E+18 4.93E+18 5.47E+18 6.OOE+18 6.54E+18 300 6.16E+17 1.09E+18 1.54E+18 2.OOE+18 2.46E+18 2.94E+18 3.46E+18 4.05E+18 4.61E+18 5.15E+18 5.68E+18 6.22E+18 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999 4-15 Cycle 1

2 3

4 5

6 7

8 9

10 11 12 450 7.27E+17 1.29E+18 1.83E+18 2.35E+18 2.86E+18 3.36E+18 3.92E+18 4.56E+18 5.21E+18 5.82E+18 6.41E+18 7.01E+18 WCAP-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations

4-16 Table 4.2-5 Calculated Fast Neutron Flux (E > 1.0 MeV) at the Cavity Sensor Set Locations (150 Neutron Pad Octant)

Neutron Flux [n/cm 2-secl Cycle 1

2 3

4 5

6 7

8 9

10 11 12 0.50 6.53E+08 7.70E+08 6.31E+08 6.42E+08 6.24E+08 6.75E+08 6.31E+08 6.37E+08 5.87E+08 5.15E+08 5.22E+08 4.82E+08 14.50 9.37E+08 1.11E+09 9.20E+08 9.07E+08 8.91E+08 9.31E+08 8.99E+08 8.92E+08 8.21E+08 7.24E+08 7.35E+08 6.88E+08 29.50 1.OOE+09 1.17E+09 1.01E+09 9.21E+08 8.91E+08 8.87E+08 8.96E+08 8.84E+08 8.57E+08 7.69E+08 7.76E+08 7.47E+08 44.50 9.30E+08 1.08E+09 9.65E+08 8.20E+08 7.73E+08 7.45E+08 7.69E+08 7.67E+08 7.81E+08 6.94E+08 6.95E+08 6.73E+08 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

Table 4.2-6 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Cavity Sensor Set Locations (150 Neutron Pad Octant)

Cycle Length EFPS 3.239E+07 2.158E+07 2.311E+07 2.664E+07 2.769E+07 2.891E+07 3.071E+07 3.541E+07 3.446E+07 3.701E+07 3.538E+07 3.794E+07 Cumulative Fluence [n/cm2]

0.50 2.12E+16 3.78E+16 5.24E+16 6.95E+16 8.68E+16 1.06E+17 1.26E+17 1.48E+17 1.68E+17 1.88E+17 2.06E+17 2.24E+17 14.50 3.03E+16 5.42E+16 7.55E+16 9.96E+16 1.24E+17 1.51E+17 1.79E+17 2.10E+17 2.39E+17 2.66E+17 2.92E+17 3.18E+17 29.50 3.25E+16 5.77E+16 8.10E+16 1.06E+17 1.30E+17 1.56E+17 1.83E+17 2.15E+17 2.44E+17 2.73E+17 3.OOE+17 3.29E+17 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999 4-17 Cycle 1

2 3

4 5

6 7

8 9

10 11 12 44.50 3.01E+16 5.34E+16 7.57E+16 9.75E+16 1.19E+17 1.40E+17 1.64E+17 1.91E+17 2.18E+17 2.44E+17 2.68E+17 2.94E+17 WCAP-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations

4-18 Calculated Fast Neutron Flux (E > 0.1 MeV) at the Center of Reactor Vessel Surveillance Capsules Neutron Flux [n/cm 2-secl Dual Capsule Cycle No.

1 2

3 4

5 6

7 8

9 10 11 12 31.50 4.39E+11 5.10E+11 4.44E+11 3.93E+11 3.78E+11 3.66E+11 3.77E+11 3.72E+11 3.70E+11 3.35E+11 3.38E+11 3.30E+11 340 5.25E+11 6.12E+11 5.37E+11 4.63E+11 4.40E+11 4.23E+11 4.39E+11 4.35E+11 4.39E+11 3.95E+11 3.97E+11 3.87E+11 Single Capsule 340 5.33E+11 6.20E+11 5.45E+11 4.69E+11 4.46E+11 4.28E+11 4.45E+11 4.41E+11 4.45E+11 4.00E+11 4.02E+11 3.92E+11 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999 Table 4.2-7 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

4-19 Calculated Fast Neutron Fluence (E > 0.1 MeV) at the Center of Reactor Vessel Surveillance Capsules Cumulative Fluence [n/cm2]

Cycle Length EFPS 3.239E+07 2.158E+07 2.311E+07 2.664E+07 2.769E+07 2.891E+07 3.071E+07 3.541E+07 3.446E+07 3.701E+07 3.538E+07 3.794E+07 Dual Capsule 31.50 1.42E+19 2.52E+19 3.55E+19 4.60E+19 5.64E+19 6.70E+19 7.86E+19 9.17E+19 1.05E+20 1.17E+20 1.29E+20 1.41E+20 340 1.70E+19 3.02E+19 4.26E+19 5.49E+19 6.71E+19 7.93E+19 9.28E+19 1.08E+20 1.23E+20 1.38E+20 1.52E+20 1.67E+20 Results of Neutron Transport Calculations WCAP-153'34, Key. U, November 

Table 4.2-8 Cycle 1

2 3

4 5

6 7

8 9

10 11 12 Single Capsule 340 1.73E+19 3.06E+19 4.32E+19 5.57E+19 6.81E+19 8.05E+19 9.41E+19 1.10E+20 1.25E+20 1.40E+20 1.54E+20 1.69E+20 WCAP-15334, Rev. 0, November 199-Results of Neutron Transport Calculations

4-20 Table 4.2-9 Calculated Fast Neutron Flux (E > 0.1 MeV) at the Reactor Vessel Clad/Base Metal Interface (150 Neutron Pad Octant)

Neutron Flux [n/cm 2-secl Cycle 1

2 3

4 5

6 7

8 9

10 11 12 00 2.64E+10 3.10E+10 2.52E+10 2.60E+10 2.50E+10 2.75E+10 2.54E+10 2.59E+10 2.39E+10 2.10E+10 2.12E+10 1.94E+10 15' 4.02E+10 4.78E+10 3.95E+10 3.92E+10 3.87E+10 4.05E+10 3.91E+10 3.86E+10 3.52E+10 3.09E+10 3.14E+10 2.93E+10 300 4.32E+10 5.01E+10 4.34E+10 3.94E+10 3.82E+10 3.76E+10 3.83E+10 3.77E+10 3.68E+10 3.33E+10 3.37E+10 3.27E+10 450 5.58E+10 6.44E+10 5.84E+10 4.86E+10 4.53E+10 4.35E+10 4.50E+10 4.51E+10 4.66E+10 4.11E+10 4.11E+10 3.97E+10 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

4-21 Table 4.2-10 Calculated Fast Neutron Fluence (E > 0.1 MeV) at the Reactor Vessel Clad/Base Metal Interface (150 Neutron Pad Octant)

Cycle Length EFPS 3.239E+07 2.158E+07 2.311E+07 2.664E+07 2.769E+07 2.891E+07 3.071E+07 3.541E+07 3.446E+07 3.701E+07 3.538E+07 3.794E+07 Cumulative Fluence [n/cm 2]

00 1.81E+18 3.20E+18 4.55E+18 5.84E+18 7.10E+18 8.35E+18 9.74E+18 1.13E+19 1.29E+19 1.45E+19 1.59E+19 1.74E+19 150 1.40E+18 2.48E+18 3.48E+18 4.53E+18 5.59E+18 6.68E+18 7.86E+18 9.19E+18 1.05E+19 1.17E+I19 1.29E+19 1.41E+19 300 1.30E+18 2.33E+18 3.25E+18 4.29E+18 5.36E+18 6.53E+18 7.73E+18 9.10E+18 1.03E+19 1.15E+19 1.26E+19 1.37E+19 450 8.54E+17 1.52E+18 2.11E+18 2.80E+18 3.49E+18 4.29E+18 5.07E+18 5.98E+18 6.81E+18 7.59E+18 8.34E+18 9.08E+18 Results of Neutron Transport Calculations WCAP-15334, Key. U, November 1i

Cycle 1

2 3

4 5

6 7

8 9

10 11 12 WCAP-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations

4-22 Table 4.2-11 Calculated Fast Neutron Flux (E > 0.1 MeV) at the Cavity Sensor Set Locations (150 Neutron Pad Octant)

Neutron Flux Cycle 1

2 3

4 5

6 7

8 9

10 11 12 0.50 6.18E+09 7.28E+09 5.96E+09 6.07E+09 5.90E+09 6.37E+09 5.97E+09 6.02E+09 5.55E+09 4.87E+09 4.93E+09 4.56E+09 14.50 9.33E+09 1.10E+10 9.16E+09 9.03E+09 8.87E+09 9.28E+09 8.96E+09 8.88E+09 8.18E+09 7.21E+09 7.32E+09 6.86E+09

[n/cm'-secl 29.50 1.08E+10 1.26E+10 1.09E+10 9.92E+09 9.60E+09 9.56E+09 9.65E+09 9.52E+09 9.23E+09 8.28E+09 8.36E+09 8.05E+09 44.50 9.87E+09 1.14E+10 1.02E+10 8.69E+09 8.19E+09 7.90E+09 8.16E+09 8.13E+09 8.29E+09 7.36E+09 7.37E+09 7.14E+09 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

Table 4.2-12 Calculated Fast Neutron Fluence (E > 0.1 MeV) at the Cavity Sensor Set Locations (15' Neutron Pad Octant)

Cycle Length EFPS 3.239E+07 2.158E+07 2.311E+07 2.664E+07 2.769E+07 2.891E+07 3.071E+07 3.541E+07 3.446E+07 3.701E+07 3.538E+07 3.794E+07 Cumulative Fluence [n/cm 2]

0.50 2.OOE+17 3.57E+17 4.95E+17 6.57E+17 8.20E+17 1.OOE+18 1.19E+18 1.40E+18 1.59E+18 1.77E+18 1.95E+18 2.12E+18 14.50 3.02E+17 5.40E+17 7.52E+17 9.93E+17 1.24E+18 1.51E+18 1.78E+18 2.10E+18 2.38E+18 2.65E+18 2.90E+18 3.16E+18 29.50 3.50E+17 6.22E+17 8.73E+17 1.14E+18 1.40E+18 1.68E+18 1.98E+18 2.31E+18 2.63E+18 2.94E+18 3.23E+18 3.54E+18 Results of Neutron Transport Calculations WCAP-15334, Rev. U, November 19

4-23 Cycle 1

2 3

4 5

6 7

8 9

10 11 12 44.50 3.20E+17 5.66E+17 8.03E+17 1.03E+18 1.26E+18 1.49E+18 1.74E+18 2.03E+18 2.31E+18 2.59E+18 2.85E+18 3.12E+18 WCAP-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations

4-24 Table 4.2-13 Calculated Iron Displacement Rate at the Center of Reactor Vessel Surveillance Capsules Displacement Rate Dual Capsule 31.50 1.91E-10 2.22E-10 1.93E-10 1.71E-10 1.64E-10 1.59E-10 1.64E-10 1.62E-10 1.61E-10 1.46E-10 1.47E-10 1.43E-10 340 2.25E-10 2.62E-10 2.30E-10 1.98E-10 1.88E-10 1.81E-10 1.88E-10 1.86E-10 1.88E-10 1.69E-10 1.70E-10 1.66E-10

[dpa/secl Single Capsule 340 2.28E-10 2.66E-10 2.33E-10 2.01E-10 1.91E-10 1.83E-10 1.90E-10 1.89E-10 1.91E-10 1.71E-10 1.72E-10 1.68E-10 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999 Cycle No.

1 2

3 4

5 6

7 8

9 10 11 12 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

4-25 Table 4.2-14 Calculated Iron Displacements at the Center of Reactor Vessel Surveillance Capsules Cumulative Displacements [dpal Single Cycle Length Dual Capsule Capsule CCycle EFPS 31.50 340 340 1

3.239E+07 6.19E-03 7.29E-03 7.39E-03 2

2.158E+07 1.1OE-02 1.29E-02 1.31E-02 3

2.311E+07 1.54E-02 1.82E-02 1.85E-02 4

2.664E+07 2.OOE-02 2.35E-02 2.39E-02 5

2.769E+07 2.45E-02 2.87E-02 2.91E-02 6

2.891E+07 2.91E-02 3.40E-02 3.44E-02 7

3.071E+07 3.42E-02 3.97E-02 4.03E-02 8

3.541E+07 3.99E-02 4.63E-02 4.70E-02 9

3.446E+07 4.54E-02 5.28E-02 5.35E-02 10 3.701E+07 5.08E-02 5.91E-02 5.99E-02 11 3.538E+07 5.60E-02 6.51E-02 6.60E-02 12 3.794E+07 6.15E-02 7.14E-02 7.23E-02 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

4-26 Table 4.2-15 Calculated Iron Displacement Rate at the Reactor Vessel Clad/Base Metal Interface (150 Neutron Pad Octant)

Displacement Rate [dpa/secl Cycle 00 150 300 450 1

1.95E-11 2.92E-11 2.93E-11 3.51E-11 2

2.30E-11 3.46E-11 3.40E-11 4.05E-11 3

1.86E-11 2.86E-11 2.95E-11 3.67E-11 4

1.92E-11 2.84E-11 2.67E-11 3.06E-11 5

1.85E-11 2.81E-11 2.59E-11 2.85E-11 6

2.04E-11 2.93E-11 2.55E-11 2.74E-11 7

1.88E-11 2.83E-11 2.60E-11 2.83E-11 8

1.92E-11 2.80E-11 2.56E-11 2.84E-11 9

1.77E-11 2.55E-11 2.50E-11 2.93E-11 10 1.56E-11 2.24E-11 2.26E-11 2.59E-11 11 1.57E-11 2.28E-11 2.28E-11 2.58E-11 12 1.44E-11 2.13E-11 2.22E-11 2.50E-11 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

Table 4.2-16 Calculated Iron Displacements at the Reactor Vessel Clad/Base Metal Interface (150 Neutron Pad Octant)

Cycle Length EFPS 3.239E+07 2.158E+07 2.311E+07 2.664E+07 2.769E+07 2.891E+07 3.071E+07 3.541E+07 3.446E+07 3.701E+07 3.538E+07 3.794E+07 Cumulative Displacements [dpal 00 6.32E-04 1.13E-03 1.56E-03 2.07E-03 2.58E-03 3.17E-03 3.75E-03 4.43E-03 5.04E-03 5.61E-03 6.17E-03 6.71E-03 150 9.44E-04 1.69E-03 2.35E-03 3.11E-03 3.89E-03 4.73E-03 5.60E-03 6.60E-03 7.47E-03 8.30E-03 9.11E-03 9.92E-03 300 9.49E-04 1.68E-03 2.36E-03 3.08E-03 3.79E-03 4.53E-03 5.33E-03 6.23E-03 7.09E-03 7.93E-03 8.74E-03 9.58E-03 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999 4-27 Cycle 1

2 3

4 5

6 7

8 9

10 11 12 450 1.14E-03 2.01E-03 2.86E-03 3.67E-03 4.46E-03 5.26E-03 6.13E-03 7.13E-03 8.14E-03 9.10E-03 1.OOE-02 1.1OE-02 WCA.P-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations

4-28 Table 4.2-17 Calculated Iron Displacement Rate at the Cavity Sensor Set Locations (150 Neutron Pad Octant)

Displacement Rate [dpa/secl Cycle 0.50 14.50 29.50 44.50 1

2.16E-12 3.22E-12 3.66E-12 3.31E-12 2

2.54E-12 3.81E-12 4.26E-12 3.84E-12 3

2.08E-12 3.16E-12 3.68E-12 3.44E-12 4

2.12E-12 3.12E-12 3.36E-12 2.92E-12 5

2.06E-12 3.06E-12 3.25E-12 2.75E-12 6

2.22E-12 3.20E-12 3.23E-12 2.65E-12 7

2.08E-12 3.09E-12 3.27E-12 2.74E-12 8

2.10E-12 3.07E-12 3.22E-12 2.73E-12 9

1.94E-12 2.82E-12 3.12E-12 2.78E-12 10 1.70E-12 2.49E-12 2.80E-12 2.47E-12 11 1.72E-12 2.53E-12 2.83E-12 2.48E-12 12 1.59E-12 2.37E-12 2.72E-12 2.40E-12 Results of Neutron Transport Calculations WCAP 15334, Rev. 0, November 1999 Results of Neutron Transport Calculations WCAP-15334, Rev. 0, November 1999

Table 4.2-18 Calculated Iron Displacements at the Cavity Sensor Set Locations (150 Neutron Pad Octant)

Cycle Length EFPS 3.239E+07 2.158E+07 2.311E+07 2.664E+07 2.769E+07 2.891E+07 3.071E+07 3.541E+07 3.446E+07 3.701E+07 3.538E+07 Cumulative Displacements [dpal 0.50 6.98E-05 1.25E-04 1.73E-04 2.29E-04 2.86E-04 3.50E-04 4.14E-04 4.89E-04 5.56E-04 6.18E-04 6.79E-04 Cycle 1

2 3

4 5

6 7

8 9

10 11 12 14.50 1.04E-04 1.86E-04 2.59E-04 3.43E-04 4.27E-04 5.20E-04 6.15E-04 7.23E-04 8.21E-04 9.13E-04 1.OOE-03 1.09E-03 29.50 1.18E-04 2.1OE-04 2.95E-04 3.85E-04 4.75E-04 5.68E-04 6.69E-04 7.83E-04 8.90E-04 9.94E-04 1.09E-03 1.20E-03 Results of Neutron Transport Calculations WCAV-15334, Rev. U, November 

4-29 3.794E+07 7.40E-04 44.50 1.07E-04 1.90E-04 2.70E-04 3.47E-04 4.24E-04 5.OOE-04 5.85E-04 6.81E-04 7.77E-04 8.69E-04 9.56E-04 1.05E-03 WCAP-15334, Rev. 0, November 1999 Results of Neutron Transport Calculations

5-1 5

EVALUATION OF SURVEILLANCE CAPSULE DOSIMETRY In this section, the results of the evaluations of the six neutron sensor sets withdrawn as a part of the McGuire Unit 2 Reactor Vessel Materials Surveillance Program[3 41 are presented. The capsule designation, location within the reactor, and time of withdrawal of each of these dosimetry sets were as follows:

Azimuthal Withdrawal Irradiation Capsule ID Location Time Time (EFPS)

V 31.50 - Dual End Of Cycle I 3.239E+07 X

34.00 - Dual End Of Cycle 5 1.314E+08 U

34.0' - Dual End Of Cycle 7 1.910E+08 Y

31.5' - Dual End Of Cycle 8 2.264E+08 Z

34.00 - Single End Of Cycle 8 2.264E+08 W

34.0' - Single End Of Cycle 10 2.979E+08 5.1 MEASURED REACTION RATES The radiometric counting of each of these capsule dosimetry data sets was accomplished by Westinghouse 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 2 reactor during the first ten 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 personnelI"' 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 V, X, U, Y, Z, and W. The computed reaction rates for the multiple foil sensor sets from Capsules V, X, U, Y, Z, and W 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 'U 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 WCA1'-15334, Rev. 0, November 139 WCAP-15334, Rev. 0, November 1999 Evaluation of Surveillance Capsule Dosimetry

5-2 5.2 RESULTS OF THE LEAST SQUARES ADJUSTMENT PROCEDURE The results of the application of the least squares adjustment procedure to the six 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 IG uncertainties associated with each of the derived exposure rates.

In Tables 5.2-1 through 5.2-5, 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 V, Cycles 1 through 5 for Capsule X, Cycles I through 7 for Capsule U, Cycles 1 through 8 for Capsules Y and Z, and Cycles 1 though 10 for Capsule W) 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.

Evaluation of Surveillance Capsule Dosimetry WCAP-15334, Rev. 0, November 1999 WCAP-15334, Rev. 0, November 1999 Evaluation of Surveillance Capsule Dosimetry

5-3 Summary of Reaction Rates Derived from Multiple Foil Sensor Sets Withdrawn from Internal Surveillance Capsules Reaction Rate [rps/nucleus]

Reaction

-Cu (n,a) -Co

'Fe (np) 54Mn SNi (n~p) 'Co

'U (n,f) 137Cs (Cd)

"Z7Np (nf) '"Cs (Cd)

"59Co (nY) 60Co "59Co (ny) 'Co (Cd)

Capsule V 5.21E-17 5.23E-15 7.04E-15 3.27E-14 3.30E-13 5.11E-12 2.78E-12 Capsule X 5.43E-17 5.48E-15 7.69E-15 3.62E-14 3.27E-13 5.45E-12 3.11E-12 Capsule U 5.15E-17 5.01E-15 6.98E-15 3.43E-14 3.20E-13 4.72E-12 2.85E-12 Capsule Y 4.80E-17 4.28E-15 6.34E-15 3.OOE-14 2.79E-13 3.88E-12 2.23E-12 Capsule Z 5.OOE-17 4.83E-15 6.94E-15 3.63E-14 3.28E-13 4.68E-12 2.69E-12 Capsule W 4.98E-17 4.85E-15 7.08E-15 3.40E-14 3.13E-13 4.69E-12 2.65E-12 Evaluation of Surveillance Capsule Dosimetry WCAP-15334, Rev. U, November 1i

Table 5.1-1 WCAP-15334, Rev. 0, November 1999 Evaluation of Surveillance Capsule Dosimetry

Best Estimate Exposure Rates from Surveillance Capsule V Dosimetry Withdrawn at the End of Fuel Cycle 1 Reaction Rate (rps/nucleus)

Reaction "3Cu (n,c) 'Co SFe (n,p) `4Mn "58Ni (n,p) 58Co 238U (n,f) 137Cs (Cd) 237Np (n,f) 137Cs (Cd) 59Co (n,y) "Co

'9Co (n,y) 'Co (Cd)

Measured 5.21E-17 5.23E-15 7.04E-15 3.27E-14 3.30E-13 5.11E-12 2.78E-12 Exposure Rate

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

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

4) (E < 0.414 eV) [n/cm2-sec]

dpa/sec Calculated 5.00E-17 5.76E-15 8.09E-15 3.14E-14 3.07E-13 4.43E-12 3.08E-12 Calculated 9.96E+10 4.40E+11 6.18E+10 1.91E-10 Best Estimate 5.00E-17 5.38E-15 7.45E-15 2.97E-14 3.14E-13 5.02E-12 2.82E-12 BE / Meas 0.96 1.03 1.06 0.91 0.95 0.98 1.01 Best Estimate 9.50E+10 4.41E+11 9.78E+10 1.88E-10 BE / Calc 1.00 0.93 0.92 0.95 1.02 1.13 0.92 BE / Calc 0.95 1.00 1.58 0.99 Meas / Calc 1.04 0.91 0.87 1.04 1.07 1.15 0.90 103 Uncertainty 6%

10%

15%

8%

Evaluation of Surveillance Capsule Dosimetry WCAP-15334, Rev. 0, November 1999 Table 5.2.1 5-4 Evaluation of Surveillance Capsule Dosimetry WCAP-15334, Rev. 0, November 1999

Table 5.2-2 Best Estimate Exposure Rates from the Surveillance Capsule X Dosimetry Withdrawn at the End of Fuel Cycle 5 Reaction

'Cu (n,cx) 6°Co

'Fe (np) 'Mn 8Ni (np) 58Co

'U (nf) 13"Cs (Cd)

"27Np (nf) `3 Cs (Cd) 59Co (n,y) 60Co 51Co (n,Y) 'Co (Cd)

Reaction Rate (rps/nucleus)

Best Measured Calculated Estimate 5.43E-17 5.19E-17 5.19E-17 5.48E-15 6.20E-15 5.70E-15 7.69E-15 8.77E-15 8.01E-15 3.62E-14 3.49E-14 3.21E-14 3.27E-13 3.52E-13 3.27E-13 5.45E-12 5.13E-12 5.38E-12 3.11E-12 3.58E-12 3.16E-12 BE / Meas 0.96 1.04 1.04 0.89 1.00 0.99 1.02 BE / Calc 1.00 0.92 0.91 0.92 0.93 1.05 0.88 Meas / Calc 1.05 0.88 0.88 1.04 0.93 1.06 0.87 Exposure Rate 0 (E > 1.0 MeV) [n/cm2-sec]

4 (E > 0.1 MeV) [n/cmr-sec]

0 (E < 0.414 eV) [n/cmr2-secl dpa/sec Calculated 1.12E+11 5.11E+11 7.07E+10 2.19E-10 Best Estimate 1.03E+11 4.84E+11 9.92E+10 2.06E-10 BE / Calc 0.92 0.95 1.40 0.94 Uncertainty 6%

10%

16%

8%

Evaluation of Surveillance Capsule Dosimetry WCAP-15334, Rev. 0, November 1999 5-5 Evaluation of Surveillance Capsule Dosimetry WCAP-15334, Rev. 0, November 1999

Best Estimate Exposure Rates from the Surveillance Capsule U Dosimetry Withdrawn at the End of Fuel Cycle 7 Reaction

'Cu (n,(X) '°Co 54Fe (n,p) 5Mn 58Ni (n,p) 58Co 2U (n,f) 137Cs (Cd) 237Np (n,f)'37Cs (Cd) 59Co (n Y) 60Co 59Co (ny) *Co (Cd)

Reaction Rate (rps/nucleus)

Best Measured Calculated Estimate 5.15E-17 4.85E-17 4.87E-17 5.01E-15 5.80E-15 5.27E-15 6.98E-15 8.20E-15 7.38E-15 3.43E-14 3.26E-14 2.98E-14 3.20E-13 3.29E-13 3.13E-13 4.72E-12 4.80E-12 4.68E-12 2.85E-12 3.35E-12 2.88E-12 BE / Meas 0.95 1.05 1.06 0.87 0.98 0.99 1.01 BE / Calc 1.00 0.91 0.90 0.91 0.95 0.98 0.86 Meas / Calc 1.06 0.86 0.85 1.05 0.97 0.98 0.85 Exposure Rate 0 (E > 1.0 MeV) [n/cm2-sec]

ý (E > 0.1 MeV) [n/cm 2-sec]

p (E < 0.414 eV) [n/cm2-sec]

dpa/sec Calculated 1.05E+11 4.78E+11 6.61E+10 2.05E-10 Best Estimate 9.60E+10 4.56E+11 8.10E+10 1.93E-10 BE / Calc 0.91 0.95 1.23 0.94 143 Uncertainty 6%

10%

17%

8%

Evaluation of Surveillance Capsule Dosimetry WCAP-15334, Rev. 0, November 1999 Table 5.2-3 5-6 Evaluation of Surveillance Capsule Dosimetry WCAP-15334, Rev. 0, November 1999

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

Reaction

'Cu (na) °Co

'Fe (np) 5Mn

'Ni (np) "8Co IU (nf 137Cs (Cd)

' 7Np (nf) 1'Cs (Cd) 59Co (n,7) 60Co 59Co (ny) -Co (Cd)

Measured 4.80E-17 4.28E-15 6.34E-15 3.OOE-14 2.79E-13 3.88E-12 2.23E-12 Exposure Rate q (E > 1.0 MeV) [n/cm2-secl d1 (E > 0.1 MeV) [n/cm2 -sec]

• (E < 0.414 eV) [n/cm -sec]

dpa/sec Calculated 4.61E-17 5.31E-15 7.46E-15 2.90E-14 2.83E-13 4.09E-12 2.84E-12 Calculated 9.18E+10 4.05E+11 5.69E+10 1.76E-10 Best Estimate 4.51E-17 4.64E-15 6.58E-15 2.59E-14 2.69E-13 3.84E-12 2.27E-12 BE / Meas 0.94 1.08 1.04 0.86 0.96 0.99 1.02 Best Estimate 8.25E+10 3.84E+11 7.10E+10 1.64E-10 BE / Calc 0.98 0.87 0.88 0.89 0.95 0.94 0.80 BE / Calc 0.90 0.95 1.25 0.93 Meas / Calc 1.04 0.81 0.85 1.03 0.99 0.95 0.79 1(y Uncertainty 6%

10%

16%

8%

Evaluation of Surveillance Capsule Dosimetry WCAP-15334, Rev. (1, November 

Table 5.2-4 WCAP-15334, Rev. 0, November 1999 Evaluation of Surveillance Capsule Dosimetry

Best Estimate Exposure Rates from the Surveillance Capsule Z Dosimetry Withdrawn at the End of Fuel Cycle 8 Reaction

'Cu (n,Q) 'Co 5Fe (n,p) 54Mn 58Ni (n,p) "8Co 238U (n,f) 137Cs (Cd) 237Np (n,f) 137Cs (Cd) 5'Co (n,y) 6°Co 51Co (n,y) 'Co (Cd)

Reaction Rate (rps/nucleus)

Best Measured Calculated Estimate 5.00E-17 4.78E-17 4.73E-17 4.83E-15 5.73E-15 5.16E-15 6.94E-15 8.13E-15 7.30E-15 3.63E-14 3.26E-14 3.OOE-14 3.28E-13 3.40E-13 3.24E-13 4.68E-12 4.49E-12 4.60E-12 2.69E-12 3.25E-12 2.73E-12 Exposure Rate p (E > 1.0 MeV) [n/cm2-sec]

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

0 (E < 0.414 eV) [n/cm2-sec]

dpa/sec Calculated 1.06E+11 5.03E+11 5.58E+10 2.12E-10 BE / Meas 0.95 1.07 1.05 0.83 0.99 0.98 1.01 Best Estimate 9.85E+10 4.88E+11 8.30E+10 2.02E-10 BE / Calc 0.99 0.90 0.90 0.92 0.95 1.02 0.84 BE / Calc 0.93 0.97 1.49 0.95 Meas / Calc 1.05 0.84 0.85 1.11 0.96 1.04 0.83 1(y Uncertainty 6%

10%

16%

8%

Evaluation of Surveillance Capsule Dosimetry WCAP 15334, Rev. 0, November 1999 Table 5.2-5 5-8 Evaluation of Surveillance Capsule Dosimetry WCAP-15334, Rev. 0, November 1999

5-9 Best Estimate Exposure Rates from the Surveillance Capsule W Dosimetry Withdrawn at the End of Fuel Cycle 10 Reaction Rate (rps/nucleus)

Reaction

'Cu (na) 'Co

'Fe (np) 54Mn

'Ni (np) 58Co

'U (n,f) "'Cs (Cd) 2 7 Np (nf) "37Cs (Cd)

"5 Co (ny) 'Co 59Co (n,y) '°Co (Cd)

Measured 4.98E-17 4.85E-15 7.08E-15 3.40E-14 3.13E-13 4.69E-12 2.65E-12 Exposure Rate Ct (E > 1.0 MeV) [n/cm2-sec]

p (E > 0.1 MeV) [n/cm2 -secl C* (E < 0.414 eV) [n/crn2 -sec]

dpa/sec Calculated 4.64E-17 5.56E-15 7.88E-15 3.16E-14 3.29E-13 4.35E-12 3.15E-12 Calculated 1.03E+11 4.87E+11 5.40E+10 2.05E-10 Best Estimate 4.74E-17 5.15E-15 7.32E-15 2.95E-14 3.12E-13 4.61E-12 2.70E-12 BE / Meas 0.95 1.06 1.03 0.87 1.00 0.98 1.02 Best Estimate 9.64E+10 4.71E+11 8.46E+10 1.96E-10 BE / Calc 1.02 0.93 0.93 0.93 0.95 1.06 0.86 BE / Calc 0.93 0.97 1.56 0.96 Meas / Calc 1.07 0.87 0.90 1.08 0.95 1.08 0.84 1__

Uncertainty 6%

10%

15%

8%

Evaluation of Surveillance Capsule Dosimetry WCAP-15334, Rev. 0, November 1999 Table 5.2-6 WCAP-15334, Rev. 0, November 1999 Evaluation of Surveillance Capsule Dosimetry

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 2, 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 database 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 Top o Core Bottom of (Degrees)

Core Midplane Core 0.5 G

14.5 H

29.5 I

44.5 J

K L

The contents of each of these irradiation capsules is specified in Appendix B to this report.

The irradiation history of the McGuire Unit 2 reactor during Cycle 12 is listed in Appendix B.

The irradiation history was obtained from plant personnel"'

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) -4Mn, **Ni (np) 'Co, and 59Co (ny) 'Co 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 0.50, 14.50, 29.50, and 44.5' chains, respectively. The -Fe (np)

'Mn reaction rates are also shown plotted in Figures 6.1-1 through 6.1-4 In regard to the data listed in Table 6.1-1, the 'Fe (np) 'Mn reaction rates represent an average of the bare and cadmium covered measurements for each capsule. The 'U (nf) F.P. reaction rates include corrections for 3U impurities in the 2U sensors as well as corrections for photofission reactions in both the 'U and "7Np sensors. The 'U and "7Np sensors also include Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999 WCAP-15334, Rev. 0, November 1999 Evaluation of Reactor Cavity Dosirnetry

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

6.1.2 Axial Position of Reactor Cavity Dosimetry An examination of the axial reaction rate distributions shown in Figures 6.1-1 through 6.1-4 shows an unrealistic asymmetry relative to the core midplane. The data suggests that the dosimetry support bar is located at a position lower that the as designed position. In order to estimate the amount of axial mispositioning, these axial distributions were compared to those measured during Cycle 12 at McGuire Unit 1361.

The data comparison indicates that the dosimetry support bar may be located approximately 12 to 18 inches lower than the intended installation point.

When correctly installed, there is approximately 19 inches of excess support chain between the threaded chain connector that attaches the support chain to the spring hook on the local attachment plate at plant elevation 7460+10.50 (refer to Westinghouse Drawing 6452E95, Revision 1).

If, instead of the intended threaded chain connector, a link of the support chain itself were attached to the spring hook on the local attachment plate, this would place the dosimetry support bar lower in the reactor cavity by up to 19 inches. While it was noted that there was difficulty in raising the dosimetry bar into position, there was no mention of anything unusual during the support chain attachment phase of the installation. It is recommended that a visual examination of the support chain attachment configuration be performed during the Cycle 13/14 refueling outage (Fall 2000).

While it appears that the dosimetry support bar is located at a position lower that the as designed position, this has no impact on the interpretation of the midplane dosimetry capsules presented in this report.

This is due to the fact that the cycle average axial core power distribution is flat in the central portion of the core.

6.1.3 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 la uncertainties associated with each of the derived exposure rates.

In Tables 6.1-6 through 6.1-11, 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.

Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999 Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999

Summary of Reaction Rates Derived from Multiple Foil Sensor Sets Cycle 12 Irradiation Reaction

'Cu (na) 'Co (Cd)

'Ti (np) 'Sc (Cd) 4Fe (np) "Mn (Cd)

"58Ni (np) 58Co (Cd)

'U (nf).37Cs (Cd)

"Z7Np (nf) 137Cs (Cd) 59Co (n,y) 61Co "Co (ny) 'Co (Cd)

Capsule G Capsule H 3.37E-19 4.90E-18 2.42E-17 3.41E-17 1.25E-16 2.01E-15 3.63E-14 1.68E-14 4.34E-19 6.42E-18 3.81E-17 5.04E-17 1.82E-16 3.05E-15 4.63E-14 2.77E-14 eaction Rate [rps/nucleu Capsule I Capsule K 4.20E-19 6.35E-18 3.41E-17 4.89E-17 1.96E-16 3.70E-15 6.87E-14 3.53E-14 3.38E-19 5.06E-18 2.83E-17 4.13E-17 1.92E-16 1.50E-15 4.27E-14 2.80E-14 Capsule J Capsule L 2.30E-19 3.85E-18 2.06E-17 3.32E-17 1.61E-16 2.86E-15 2.80E-14 1.81E-14 3.80E-20 6.37E-19 3.60E-18 5.48E-18 2.33E-17 4.41E-16 1.12E-14 6.95E-15 Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999 Table 6.1-1 6-3 Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999

T Fe (n,p), -Ni (n,p) and 59Co (n,y) Reaction Rates Derived from the RCND Stainless Steel Gradient Chain at 0.5' - Cycle 12 Irradiation Distance From Core Midplane (ft) 5.5 4.5 3.5 2.5 1.5 0.5

-0.5

-1.5

-2.5

-3.5

-4.5

-5.5

-6.5 TMFe (n,p) 1.88E-17 2.23E-17 2.39E-17 2.56E-17 2.50E-17 2.45E-17 2.42E-17 2.40E-17 2.17E-17 1.88E-17 1.11E-17 4.15E-18 Reaction Rate

[rps/nucleus]

58Ni (np)

'9Co (n,y) 2.77E-17 1.88E-14 3.34E-17 2.24E-14 3.62E-17 2.55E-14 3.57E-17 2.74E-14 3.54E-17 3.51E-14 3.47E-17 3.64E-14 3.39E-17 3.55E-14 3.36E-17 3.44E-14 3.23E-17 3.09E-14 2.78E-17 1.99E-14 1.72E-17 1.35E-14 6.43E-18 8.76E-15 1.21E-18 2.07E-18 3.94E-15 Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999 Table 6.1-2 6-4 Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999

"Fe (np), 'Ni (np) and "9Co (ny) Reaction Rates Derived from the RCND Stainless Steel Gradient Chain at 14.50 - Cycle 12 Irradiation Distance From Core Midplane (ft) 5.5 4.5 3.5 2.5 1.5 0.5

-0.5

-1.5

-2.5

-3.5

-4.5

-5.5

-6.5

'Fe (nop) 2.73E-17 3.14E-17 3.35E-17 3.59E-17 3.45E-17 3.43E-17 3.37E-17 3.24E-17 3.11E-17 2.74E-17 1.80E-17 7.23E-18 2.10E-18 Reaction Rate

[rps/nucleus]

"8Ni (n,p) 59Co (nM 4.13E-17 3.87E-14 4.79E-17 4.75E-14 4.95E-17 5.36E-14 4.91E-17 5.26E-14 4.87E-17 4.69E-14 4.87E-17 4.61E-14 4.79E-17 4.61E-14 4.79E-17 4.41E-14 4.62E-17 4.08E-14 4.04E-17 3.71E-14 2.73E-17 2.67E-14 1.13E-17 1.55E-14 3.29E-18 7.26E-15 Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999 Table 6.1-3 6-5 WCAP-15334, Rev. 0, November 1999 Evaluation of Reactor Cavity Dosimetry

Table 6.1-4 TMFe (n,p), 58Ni (np) and -9Co (n,y) Reaction Rates Derived from the RCND Stainless Steel Gradient Chain at 29.50 - Cycle 12 Irradiation Distance From Core Midplane (ft) 5.5 4.5 3.5 2.5 1.5 0.5

-0.5

-1.5

-2.5

-3.5

-4.5

-5.5

-6.5 TMFe (n,p) 2.70E-17 3.17E-17 3.48E-17 3.60E-17 3.47E-17 3.31E-17 3.34E-17 3.35E-17 3.18E-17 2.66E-17 1.76E-17 7.91E-18 2.35E-18 Reaction Rate

[rps/nucleus]

-Ni (n,p) 59Co (n,y) 4.30E-17 5.43E-14 4.75E-17 6.82E-14 4.99E-17 7.65E-14 4.99E-17 7.76E-14 4.95E-17 7.37E-14 4.95E-17 6.98E-14 4.75E-17 6.43E-14 4.87E-17 6.65E-14 4.71E-17 5.93E-14 4.06E-17 5.40E-14 2.72E-17 4.00E-14 1.21E-17 2.42E-14 3.77E-18 1.02E-14 Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999 6-6 Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999

'Fe (np), 'Ni (np) and -Co (n,y) Reaction Rates Derived from the RCND Stainless Steel Gradient Chain at 44.5' - Cycle 12 Irradiation Distance From Core Midplane (ft) 5.5 4.5 3.5 2.5 1.5 0.5

-0.5

-1.5

-2.5

-3.5

-4.5

-5.5

-6.5 5Fe (nop) 2.35E-17 2.97E-17 3.06E-17 3.OOE-17 3.10E-17 2.93E-17 2.85E-17 2.89E-17 2.68E-17 2.25E-17 1.39E-17 5.93E-18 5.54E-19 Reaction Rate

[rps/nucleus]

'Ni (n,p)

'Co (n,)

3.76E-17 3.36E-14 4.46E-17 4.07E-14 4.54E-17 4.52E-14 4.58E-17 4.77E-14 4.58E-17 4.69E-14 4.38E-17 4.50E-14 4.26E-17 4.48E-14 4.22E-17 4.32E-14 4.06E-17 3.90E-14 3.49E-17 3.33E-14 2.23E-17 2.27E-14 9.45E-18 1.59E-14 9.09E-19 1.86E-15 Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 19i9 Table 6.1-5 6-7 WCAP-15334, Rev. 0, November 1999 Evaluation of Reactor Cavity Dosimetry

6-8 Figure 6.1-1 -Fe (np) 54Mn Reaction Rate Derived from the Stainless Steel Gradient Chain at 0.5 Degrees in the Reactor Cavity - Cycle 12 Irradiation 54Fe (n,p) 54Mn Axial Traverse at 0.5 Deg.

cc a

0 4U a) 1.OE-16 1.OE-17 1.OE-18

-7

-5

-3

-1 1

3 5

7 Axial Position (feet)

Figure 6.1-2 -Fe (n,p) TMMn Reaction Rate Derived from the Stainless Steel Gradient Chain at 14.5 Degrees in the Reactor Cavity - Cycle 12 Irradiation 54Fe (np) 54 Mn Axial Traverse at 14.5 Deg.

4-'

O) 0m 4

rU 1.0E-16 1.0E-17 1.OE-18

-7

-5

-3

-1 1

3 5

7 Axial Position (feet)

Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999 Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999

6-9 Figure 6.1-3 *Fe (n,p) 4Mn Reaction Rate Derived from the Stainless Steel Gradient Chain at 29.5 Degrees in the Reactor Cavity - Cycle 12 Irradiation Figure 6.1-4 TMFe (n,p) -Mn Reaction Rate Derived from the Stainless Steel Gradient Chain at 44.5 Degrees in the Reactor Cavity - Cycle 12 Irradiation Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999 54Fe (np) 54Mn Axial Traverse at 29.5 Deg.

1.OE-16 o 1.OE-17 A cc 1.OE-18

-7

-5

-3

-1 1

3 5

7 Axial Position (feet) 54Fe (n,p) 54Mn Axial Traverse at 44.5 Deg.

1.0E-16 0

1.OE-18 e-1 1.OE-19

-7

-5

-3

-1 1

3 5

7 Axial Position (feet)

Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999

6-10 Best Estimate Exposure Rates from the Capsule G Dosimetry Evaluation 0.5' Azimuth - Core Midplane - Cycle 12 Irradiation Reaction Rate (rps/nucleus)

Reaction 63Cu (n,a) 'Co (cd) 46Ti (n,p) 4Sc (Cd) 54Fe (n,p) '4Mn (Cd)

"58Ni (n,p) 51Co (Cd) 28U (n,f) 1B7Cs (Cd) 237Np (n,f) 137Cs (Cd) 59Co (n,y) 6°Co 59Co (n,y) 6Co (Cd)

Measured 3.37E-19 4.90E-18 2.42E-17 3.41E-17 1.25E-16 2.01E-15 3.63E-14 1.68E-14 Exposure Rate 0 (E > 1.0 MeV) [n/cm2-sec]

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

(p (E < 0.414 eV) [n/cm 2-sec]

dpa/sec Calculated 3.98E-19 5.42E-18 2.93E-17 4.08E-17 1.46E-16 2.11E-15 7.91E-14 2.72E-14 Calculated 4.82E+08 4.56E+09 1.77E+09 1.59E-12 Best Estimate 3.40E-19 4.72E-18 2.48E-17 3.46E-17 1.25E-16 1.91E-15 3.69E-14 1.69E-14 BE / Meas BE / Calc Meas / Calc 1.01 0.96 1.02 1.01 1.00 0.95 1.02 1.01 Best Estimate 4.14E+08 4.01E+09 7.27E+08 1.38E-12 0.85 0.87 0.85 0.85 0.86 0.91 0.47 0.62 BE / Calc 0.86 0.88 0.41 0.87 0.85 0.90 0.83 0.84 0.86 0.95 0.46 0.62 1(T Uncertainty 6%

11%

12%

9%

Evaluation ot Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999 Table 6.1-6 Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999

6-11 Table 6.1-7 Best Estimate Exposure Rates From the Capsule H Dosimetry Evaluation 14.50 Azimuth - Core Midplane - Cycle 12 Irradiation Reaction Rate (rps/nucleus)

Reaction

'Cu (n,ox) 'Co (cd)

'Ti (np) 'Sc (Cd)

T4Fe (np) M4Mn (Cd)

"58Ni (np) 'Co (Cd)

'U (nf) `37Cs (Cd)

' 7Np (nf) 137Cs (Cd)

"59Co (n,y) 60Co "59Co (ny) *'Co (Cd)

Measured 4.34E-19 6.42E-18 3.81E-17 5.04E-17 1.82E-16 3.05E-15 4.63E-14 2.77E-14 Exposure Rate 4 (E > 1.0 MeV) [n/cm2 -sec]

q (E > 0.1 MeV) [n/cm2-secl 4*

(E < 0.414 eV) [n/cm2-sec]

dpa/sec Calculated 4.96E-19 6.95E-18 3.94E-17 5.53E-17 2.05E-16 3.08E-15 1.01E-13 4.21E-14 Calculated 6.88E+08 6.86E+09 2.10E+09 2.37E-12 Best Estimate 4.43E-19 6.36E-18 3.68E-17 5.10E-17 1.91E-16 2.94E-15 4.77E-14 2.75E-14 BE / Meas 1.02 0.99 0.97 1.01 1.05 0.96 1.03 0.99 Best Estimate 6.45E+08 6.30E+09 7.83E+08 2.17E-12 BE / Calc Meas / Calc 0.89 0.92 0.93 0.92 0.93 0.95 0.47 0.65 BE / Calc 0.94 0.92 0.37 0.92 0.88 0.92 0.97 0.91 0.89 0.99 0.46 0.66 Uncertainty 6%

11%

16%

9%

Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999 WCAP-15334, Rev. 0, November 1999 Evaluation of Reactor Cavity Dosimetry

6-12 Best Estimate Exposure Rates From the Capsule I Dosimetry Evaluation 29.5' Azimuth - Core Midplane - Cycle 12 Irradiation Reaction Rate (rps/nucleus)

Reaction 63CU (n,(x) 6°Co (cd) 46Ti (n,p) "Sc (Cd)

TMFe (n,p) 1 Mn (Cd)

"Ni (n,p) 51Co (Cd) 2U (n,f) 137Cs (Cd) 237Np (n,f) 137Cs (Cd) 5 9

Co (n,'y) 6 0 CO 5

9CO (n,y) '°Co (Cd)

Exposure Rate p (E > 1.0 MeV) [n/cm 2-sec]

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

0 (E < 0.414 eV) [n/cm2-sec]

dpa/sec Measured 4.20E-19 6.35E-18 3.41E-17 4.89E-17 1.96E-16 3.70E-15 6.87E-14 Calculated 4.61E-19 6.53E-18 3.85E-17 5.49E-17 2.15E-16 3.50E-15 1.20E-13 Best Estimate 4.23E-19 6.11E-18 3.49E-17 4.98E-17 1.97E-16 3.47E-15 6.98E-14 3.53E-14 5.06E-14 3.54E-14 Calculated 7.47E+08 8.05E+09 2.49E+09 2.72E-12 BE / Meas 1.01 0.96 1.02 1.02 1.01 0.94 1.02 1.00 Best Estimate 6.91E+08 7.64E+09 1.30E+09 2.56E-12 BE / Calc Meas / Calc 0.92 0.94 0.91 0.91 0.92 0.99 0.58 0.70 BE / Calc 0.92 0.95 0.52 0.94 0.91 0.97 0.89 0.89 0.91 1.06 0.57 0.70 1a Uncertainty 6%

11%

14%

9%

Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999 Table 6.1-8 Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999

6-13 Best Estimate Exposure Rates From the Capsule K Dosimetry Evaluation 44.5' Azimuth - Core Midplane - Cycle 12 Irradiation Reaction Rate (rps/nucleus)

Reaction

'Cu (na) 'Co (cd)

'Ti (np) 4Sc (Cd)

T4Fe (np) TMn (Cd)

'Ni (np) 58Co (Cd) 1U (n,f) 137Cs (Cd) 2 7Np (n,f) 137Cs (Cd)

"59Co (ny)6OCo "59Co (ny) Co (Cd)

Measured 3.38E-19 5.06E-18 2.83E-17 4.13E-17 1.92E-16 1.50E-15 4.27E-14 2.80E-14 Exposure Rate t (E > 1.0 MeV) [n/cm2-secl t(E > 0.1 MeV) [n/cm2-sec]

qt (E < 0.414 eV) [n/cm2-sec]

dpa/sec Calculated 3.56E-19 5.07E-18 3.10E-17 4.50E-17 1.87E-16 3.17E-15 1.15E-13 4.01E-14 Calculated 6.74E+08 7.14E+09 2.56E+09 2.40E-12 Best Estimate 3.41E-19 4.92E-18 2.87E-17 4.11E-17 1.62E-16 1.95E-15 4.46E-14 2.74E-14 BE / Meas 1.01 0.97 1.01 1.00 0.84 1.30 1.04 0.98 Best Estimate 5.49E+08 5.07E+09 6.44E+08 1.75E-12 BE / Calc Meas / Calc 0.96 0.97 0.93 0.91 0.87 0.62 0.39 0.68 BE / Calc 0.82 0.71 0.25 0.73 0.95 1.00 0.91 0.92 1.03 0.47 0.37 0.70 Uncertainty 6%

11%

16%

9%

Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999 Table 6.1-9 WCAP-15334, Rev. 0, November 1999 Evaluation of Reactor Cavity Dosimetry

6-14 Table 6.1-10 Best Estimate Exposure Rates From the Capsule J Dosimetry Evaluation 44.5' Azimuth - Top of Core - Cycle 12 Irradiation Reaction Rate (rns/nuc1eu)

Reaction

'Cu (n,cL) 'Co (cd)

'Ti (n,p) '6Sc (Cd)

-4Fe (n,p) 5TMn (Cd) 58Ni (n,p) 58Co (Cd)

'U (nf) 137Cs (Cd) 237Np (n,f) 137Cs (Cd)

-Co (n,7) 'Co

' 9Co (n,y) 'Co (Cd)

Measured 2.30E-19 3.85E-18 2.06E-17 3.32E-17 1.61E-16 2.86E-15 2.80E-14 1.81E-14 Exposure Rate

4) (E > 1.0 MeV) [n/cm 2-sec]

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

(p (E < 0.414 eV) [n/cm2-sec]

dpa/sec Calculated 2.59E-19 3.69E-18 2.26E-17 3.27E-17 1.36E-16 2.31E-15 8.38E-14 2.92E-14 Calculated 4.90E+08 5.20E+09 1.87E+09 1.75E-12 Best Estimate 2.38E-19 3.62E-18 2.20E-17 3.30E-17 1.43E-16 2.66E-15 2.94E-14 1.78E-14 BE / Meas 1.03 0.94 1.07 0.99 0.89 0.93 1.05 0.98 Best Estimate 5.27E+08 5.54E+09 4.37E+08 1.84E-12 BE / Calc 0.92 0.98 0.97 1.01 1.05 1.15 0.35 0.61 BE / Calc 1.08 1.07 0.23 1.05 Meas / Calc 0.89 1.04 0.91 1.02 1.18 1.24 0.33 0.62 Uncertainty 6%

11%

16%

9%

Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999 Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999

6-15 Table 6.1-11 Best Estimate Exposure Rates From the Capsule L Dosimetry Evaluation 44.50 Azimuth - Bottom of Core - Cycle 12 Irradiation Reaction Rate (rps/nucleus)

Reaction

'Cu (na) 'Co, (cd)

'Ti (n~p) 'Sc (Cd)

'Fe (np) "Mn (Cd)

'Ni (np) 'Co (Cd) 2U (n,f) `37Cs (Cd)

' 7Np (nf) 137Cs (Cd) 59Co (n,Y) 60Co "59Co (ny) 'Co (Cd)

Measured 3.80E-20 6.37E-19 3.60E-18 5.48E-18 2.33E-17 4.41E-16 1.12E-14 6.95E-15 Exposure Rate 4 (E > 1.0 MeV) In/cm2-secl (E > 0.1 MeV) [n/cm2-sec]

p (E < 0.414 eV) [n/cm 2-sec]

dpa/sec Calculated 4.52E-20 6.45E-19 3.94E-18 5.72E-18 2.38E-17 4.04E-16 1.46E-14 5.11E-15 Calculated 8.57E+07 9.08E+08 3.26E+08 3.05E-13 Best Estimate 4.OOE-20 6.06E-19 3.69E-18 5.44E-18 2.33E-17 4.26E-16 1.15E-14 6.79E-15 BE / Meas 1.05 0.95 1.03 0.99 1.00 0.97 1.03 0.98 Best Estimate 8.52E+07 9.32E+08 1.68E+08 3.09E-13 BE / Calc 0.88 0.94 0.94 0.95 0.98 1.05 0.79 1.33 BE / Calc 1.00 1.03 0.51 1.01 Meas / Calc 0.84 0.99 0.91 0.96 0.98 1.09 0.77 1.36 la Uncertainty 6%

11%

15%

9%

Evaluation of Reactor Cavity Dosimetry WCAP-15334, Rev. 0, November 1999 WCAP-15334, Rev. 0, November 1999 Evaluation of Reactor Cavity Dosimetry

7-1 7

COMPARISON OF CALCULATIONS WITH MEASUREMENTS As described in Section 3.3, the best estimate neutron exposure projections for the McGuire Unit 2 reactor vessel were based on a combination of plant specific neutron transport calculations and plant specific measurements. Direct comparisons of the transport calculations with the McGuire Unit 2 measurement data base were used to quantify the biases that may exist due to the transport methodology, reactor modeling, and/or reactor operating characteristics over the respective irradiation periods.

In this section, comparisons of the measurement results from surveillance capsule and reactor cavity dosimetry with corresponding analytical predictions at the measurement locations are presented. These comparisons are provided on two levels. In the first instance, predictions of fast neutron exposure rates in terms of 0 (E > 1.0 MeV), 0 (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 to calculation (BE/C) comparisons yield an accurate plant specific bias factor that can be applied to neutron transport calculations performed for the McGuire Unit 2 reactor to produce best estimate exposure projections for the reactor 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 six 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 0.91, 0.93, and 0.92 for (D (E > 1.0 MeV), 0 (E > 0.1 MeV), and dpa/sec, respectively.

The standard deviations associated with each of the ten-sample-data-sets were 4.6%, 9.4%, and 8.1%, respectively.

7.2 COMPARISONS OF MEASURED AND CALCULATED SENSOR REACTION RATES In Table 7.2-1, measurement to 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 3.2% to 18.1% on an individual reaction basis; whereas, the overall average M/C ratio for the entire data set has an associated 1o standard deviation of 11.6%. Furthermore, the average M/C bias WCAP-15334, Rev. 0, November 1999 Comparison of Calculations with Measurements

7-2 of 0.94 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.

Comparison of Calculations with Measurements WCAP-15334, Rev. 0, November 1999 Comparison of Calculations with Measurements WCAP-15334, Rev. 0, November 1999

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]

Surveillance Capsules Capsule V Capsule X Capsule U Capsule Y Capsule Z Capsule W 0.50 Cavity Cycle 12 14.50 Cavity Cycle 12 29.50 Cavity Cycle 12 44.5-Cavity Cycle 12 Average BE/C Bias Factor

% Standard Deviation (1a)

Calculated 3.23E+18 1.47E+19 2.04E+19 2.08E+19 2.41E+19 3.07E+19 1.83E+16 2.61E+16 2.84E+16 2.55E+16 Best Estimate 3.08E+18 1.35E+19 1.83E+19 1.87E+19 2.23E+19 2.87E+19 1.57E+16 2.45E+16 2.62E+16 2.08E+16 Note: The R; 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 reactor vessel exposure projections as described in Section 8.3.

7-3 BE/C 0.95 0.92 0.90 0.90 0.93 0.93 0.86 0.94 0.92 0.82 0.91 4.6%

WCAP-15334, Rev. 0, November 1999 Comparison of Calculations with Measurements

Comparison of Best Estimate and Calculated Exposure Rates from Surveillance Capsule and Cavity Dosimetry Irradiations Neutron Fluence (E > 0.1 MeV) [n/cm 2]

Surveillance Capsules Capsule V Capsule X Capsule U Capsule Y Capsule Z Capsule W 0.50 Cavity Cycle 12 14.50 Cavi Cycle 12 29.50 Cavi Cycle 12 44.50 Cavi Cycle 12 Average BE/C Bias Factor

% Standard Deviation (l(Y)

Calculated 1.42E+19 6.71E+19 9.28E+19 9.17E+19 1.10E+20 1.40E+20 1.73E+17 2.60E+17 3.05E+17 2.71E+17 Best Estimate 1.43E+19 6.36E+19 8.71E+19 8.70E+19 1.10E+20 1.40E+20 1.52E+17 2.39E+17 2.90E+17 1.92E+17 Note: The lcy 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 reactor vessel exposure projections as described in Section 8.3.

Table 7.1-1 (cont.)

7-4 BE/C 1.00 0.95 0.94 0.95 1.01 1.00 0.88 0.92 0.95 0.71 0.93 9.4%

Comparison of Calculations with Measurements WCAP-15334, Rev. 0, November 1999

Table 7.1-1 (cont.)

Comparison of Best Estimate and Calculated Exposure Rates from Surveillance Capsule and Cavity Dosimetry Irradiations Iron Displacements [dpal Calculated Surveillance Capsules Capsule V Capsule X Capsule U Capsule Y Capsule Z Capsule W 0.50 Cavity Cycle 12 14.50 Cavity Cycle 12 29.50 Cavity Cycle 12 44.5' Cavity Cycle 12 Average BE/C Bias Factor

% Standard Deviation (l1()

6.19E-03 2.87E-02 3.97E-02 3.99E-02 4.70E-02 5.99E-02 6.03E-05 8.98E-05 1.03E-04 9.10E-05 Best Estimate 6.10E-03 2.70E-02 3.68E-02 3.71E-02 4.58E-02 5.85E-02 5.25E-05 8.24E-05 9.71E-05 6.65E-05 Note: The 1y 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 reactor vessel exposure projections as described in Section 8.3.

7-5 BE/C 0.99 0.94 0.93 0.93 0.97 0.98 0.87 0.92 0.94 0.73 0.92 8.1%

Comparison of Calculations with Measurements WCAP-15334, Rev. 0, November 1999

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

'Cu (n,(x) 46Ti (np)

-Fe (np)

• Ni (np) 2U (n,f) 37Np (n,f)

Surveillance Capsules Capsule V 1.04 0.91 0.87 1.04 1.07 Capsule X 1.05 0.88 0.88 1.04 0.93 Capsule U 1.06 0.86 0.85 1.05 0.97 Capsule Y 1.04 0.81 0.85 1.03 0.99 Capsule Z 1.05 0.84 0.85 1.11 0.96 Capsule W 1.07 0.87 0.90 1.08 0.95 0.50 Cavi Cycle 12 0.85 0.90 0.83 0.84 0.86 0.95 14.5' Cavi Cycle 12 0.88 0.92 0.97 0.91 0.89 0.99 29.5' Cavi Cycle 12 0.91 0.97 0.89 0.89 0.91 1.06 44.5' Cavi Cycle 12 0.95 1.00 0.91 0.92 1.03 0.47 Average 0.99 0.95 0.88 0.88 1.00 0.93

% Std. Dev. (17) 8.5%

4.8%

5.3%

3.2%

8.5%

18.1%

Overall M/C Average 0.94

% Std. Dev. (1a) 11.0%

Comparison of Calculations with Measurements WCAP-15334, Rev. 0, November 1999 Comparison of Calculations with Measurements WCAP-15334, Rev. 0, November 1999

8-1 8

BEST ESTIMATE NEUTRON EXPOSURE OF REACTOR 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 2 reactor vessel through the completion of Cycle 12. Based on the continued use of the core power distributions producing the Cycles 10-12 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.=

Caic.

where:

4 Es

= 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.

Ocac.

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:

() (E > 1.0 MeV) 0.91 + 4.6%

() (E > 0.1 MeV) 0.93 + 9.4%

dpa 0.92 + 8.1%

These bias factors were based on the results of the continuous monitoring program at McGuire Unit 2 that has provided measured data from six 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 1cy level.

Additional uncertainties associated with the evaluation of the best estimate vessel exposure are discussed in Section 8.3.

Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999

8-2 8.1.1 Exposure Accumulated 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 composite axial core power distributions were then used 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 accumulated by the reactor vessel through the end of Cycles 10 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 2 will expand resulting in reduced uncertainties and an improved accuracy in the assessment of vessel condition.

8.1.2 Projection of Future Vessel Exposure At the end of Cycle 12, the McGuire Unit 2 reactor had accumulated 11.76 effective full power years (EFPY) of operation. In order to establish a framework for the assessment of the future reactor 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 Cycle 10-12 irradiations were representative of all future fuel cycles. That is, that future fuel designs would incorporate the low leakage fuel management concept employed during Cycles 10 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 reactor vessel exposure management. The validity of these projections for future operation will be confirmed via the continued reactor cavity neutron monitoring program.

Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 WCAP-15334, Rev. 0, November 1999 Best Estimate Neutron Exposure of Pressure Vessel Materials

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 2 Reactor Vessel 0' Azimuthal Angle Distance from Core Midplane 6.00 5.00 4.00 3.00 2.00 1.00 0.00

-1.00

-2.00

-3.00

-4.00

-5.00

-6.00 Cycle 10 9.44 EFPY 1.38E+18 2.57E+18 3.02E+18 3.21E+18 3.28E+18 3.30E+18 3.32E+18 3.32E+18 3.28E+18 3.22E+18 3.07E+18 2.66E+18 1.35E+18 Cycle 12 11.76 EFPY 1.66E+18 3.06E+18 3.61E+18 3.84E+18 3.91E+18 3.94E+18 3.97E+18 3.98E+18 3.93E+18 3.86E+18 3.70E+18 3.23E+18 1.63E+18 21 EFPY 2.75E+18 5.08E+18 5.99E+18 6.37E+18 6.50E+18 6.54E+18 6.59E+18 6.60E+18 6.53E+18 6.41E+18 6.15E+18 5.36E+18 2.70E+18 Projected Exposures 34 EFPY 4.29E+18 7.93E+18 9.34E+18 9.93E+18 1.01E+19 1.02E+19 1.03E+19 1.03E+19 1.02E+19 1.OOE+19 9.59E+18 8.36E+18 4.21E+18 Best Estimate Neutron Exposure of Pressure Vessel Materials 51 EFPY 6.30E+18 1.16E+19 1.37E+19 1.46E+19 1.49E+19 1.50E+19 1.51E+19 1.51E+19 1.50E+19 1.47E+19 1.41E+19 1.23E+19 6.19E+18 WCAP-15334, Rev. 0, November 1999

8-4 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 150 Azimuthal Angle Distance from Core Midplane 6.00 5.00 4.00 3.00 2.00 1.00 0.00

-1.00

-2.00

-3.00

-4.00

-5.00

-6.00 Cycle 10 9.44 EFPY 2.07E+18 3.84E+18 4.51E+18 4.80E+18 4.89E+18 4.93E+18 4.96E+18 4.96E+18 4.90E+18 4.80E+18 4.59E+18 3.98E+18 2.02E+18 Cycle 12 11.76 EFPY 2.47E+18 4.57E+18 5.38E+18 5.72E+18 5.84E+18 5.88E+18 5.93E+18 5.93E+18 5.86E+18 5.76E+18 5.53E+18 4.82E+18 2.43E+18 21 EFPY 4.08E+18 7.54E+18 8.89E+18 9.45E+18 9.64E+18 9.71E+18 9.78E+18 9.80E+18 9.68E+18 9.51E+18 9.13E+18 7.96E+18 4.OOE+18 Projected Exposures 34 EFPY 6.35E+18 1.17E+19 1.38E+19 1.47E+19 1.50E+19 1.51E+19 1.52E+19 1.52E+19 1.51E+19 1.48E+19 1.42E+19 1.24E+19 6.23E+18 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev, 0, November 1999 Table 8.1-2 51 EFPY 9.31E+18 1.72E+19 2.03E+19 2.16E+19 2.20E+19 2.21E+19 2.23E+19 2.23E+19 2.21E+19 2.17E+19 2.08E+19 1.81E+19 9.13E+18 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999

8-5 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 300 Azimuthal Angle Distance from Core Midplane 6.00 5.00 4.00 3.00 2.00 1.00 0.00

-1.00

-2.00

-3.00

-4.00

-5.00

-6.00 Cycle 10 9.44 EFPY 1.94E+18 3.61E+18 4.25E+18 4.52E+18 4.61E+18 4.64E+18 4.67E+18 4.67E+18 4.61E+18 4.52E+18 4.32E+18 3.75E+18 1.90E+18 Cycle 12 11.76 EFPY 2.35E+18 4.35E+18 5.12E+18 5.45E+18 5.56E+18 5.60E+18 5.64E+18 5.65E+18 5.58E+18 5.48E+18 5.26E+18 4.59E+18 2.31E+18 21 EFPY 3.97E+18 7.33E+18 8.64E+18 9.18E+18 9.37E+18 9.44E+18 9.51E+18 9.52E+18 9.41E+18 9.25E+18 8.87E+18 7.73E+18 3.89E+18 Projected Exposures 34 EFPY 6.24E+18 1.15E+19 1.36E+19 1.44E+19 1.47E+19 1.48E+19 1.50E+19 1.50E+19 1.48E+19 1.45E+19 1.39E+19 1.22E+19 6.12E+18 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 Table 8.1-3 51 EFPY 9.21E+18 1.70E+19 2.01E+19 2.13E+19 2.17E+19 2.19E+19 2.21E+19 2.21E+19 2.18E+19 2.15E+19 2.06E+19 1.79E+19 9.04E+18 WCAP-15334, Rev. 0, November 1999 Best Estimate Neutron Exposure of Pressure Vessel Materials

8-6 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 45' Azimuthal Angle Distance from Core Midplane 6.00 5.00 4.00 3.00 2.00 1.00 0.00

-1.00

-2.00

-3.00

-4.00

-5.00

-6.00 Cycle 10 9.44 EFPY 2.20E+18 4.08E+18 4.80E+18 5.10E+18 5.21E+18 5.25E+18 5.28E+18 5.28E+18 5.21E+18 5.11E+18 4.88E+18 4.23E+18 2.15E+18 Cycle 12 11.76 EFPY 2.65E+18 4.90E+18 5.78E+18 6.14E+18 6.26E+18 6.31E+18 6.36E+18 6.37E+18 6.29E+18 6.18E+18 5.93E+18 5.17E+18 2.60E+18 Projected Exposures 21 EFPY 4.45E+18 8.23E+18 9.70E+18 1.03E+19 1.05E+19 1.06E+19 1.07E+19 1.07E+19 1.06E+19 1.04E+19 9.96E+18 8.68E+18 4.37E+18 34 EFPY 6.99E+18 1.29E+19 1.52E+19 1.62E+19 1.65E+19 1.66E+19 1.68E+19 1.68E+19 1.66E+19 1.63E+19 1.56E+19 1.36E+19 6.86E+18 Best Estimate Neutron Exposure of Pressure Vessel Materials Table 8.1-4 51 EFPY 1.03E+19 1.91E+19 2.25E+19 2.39E+19 2.43E+19 2.45E+19 2.47E+19 2.47E+19 2.45E+19 2.40E+19 2.31E+19 2.01E+19 1.01E+19 WCAP-15334, Rev. 0, November 1999

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 2 Reactor Vessel 0' Azimuthal Angle Distance from Core Midplane 6.00 5.00 4.00 3.00 2.00 1.00 0.00

-1.00

-2.00

-3.00

-4.00

-5.00

-6.00 Cycle 10 9.44 EFPY 2.94E+18 5.46E+18 6.43E+18 6.83E+18 6.97E+18 7.02E+18 7.06E+18 7.06E+18 6.97E+18 6.84E+18 6.54E+18 5.67E+18 2.88E+18 Cycle 12 11.76 EFPY 3.52E+18 6.51E+18 7.67E+18 8.16E+18 8.32E+18 8.38E+18 8.45E+18 8.46E+18 8.36E+18 8.21E+18 7.88E+18 6.87E+18 3.46E+18 Projected Exposures 21 EFPY 5.85E+18 1.08E+19 1.27E+19 1.35E+19 1.38E+19 1.39E+19 1.40E+19 1.40E+19 1.39E+19 1.36E+19 1.31E+19 1.14E+19 5.74E+18 34 EFPY 9.13E+18 1.69E+19 1.99E+19 2.11E+19 2.15E+19 2.17E+19 2.19E+19 2.19E+19 2.16E+19 2.13E+19 2.04E+19 1.78E+19 8.96E+18 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 51 EFPY 1.34E+19 2.48E+19 2.92E+19 3.10E+19 3.17E+19 3.19E+19 3.21E+19 3.22E+19 3.18E+19 3.13E+19 3.OOE+19 2.61E+19 1.32E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999

8-8 Table 8.1-6 Distance from Core Midplane 6.00 5.00 4.00 3.00 2.00 1.00 0.00

-1.00

-2.00

-3.00

-4.00

-5.00

-6.00 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 150 Azimuthal Angle Cycle 10 9.44 EFPY 4.44E+18 8.24E+18 9.70E+18 1.03E+19 1.05E+19 1.06E+19 1.07E+19 1.07E+19 1.05E+19 1.03E+19 9.87E+18 8.56E+18 4.35E+18 Cycle 12 11.76 EFPY 5.31E+18 9.82E+18 1.16E+19 1.23E+19 1.25E+19 1.26E+19 1.27E+19 1.28E+19 1.26E+19 1.24E+19 1.19E+19 1.04E+19 5.21E+18 21 EFPY 8.77E+18 1.62E+19 1.91E+19 2.03E+19 2.07E+19 2.09E+19 2.10E+19 2.11E+19 2.08E+19 2.04E+19 1.96E+19 1.71E+19 8.61E+18 Projected Exposures 34 EFPY 1.36E+19 2.52E+19 2.97E+19 3.16E+19 3.22E+19 3.25E+19 3.27E+19 3.27E+19 3.24E+19 3.18E+19 3.05E+19 2.66E+19 1.34E+19 51 EFPY 2.OOE+19 3.70E+19 4.36E+19 4.63E+19 4.72E+19 4.76E+19 4.80E+19 4.80E+19 4.75E+19 4.66E+19 4.47E+19 3.90E+19 1.96E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999

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 2 Reactor Vessel 300 Azimuthal Angle Distance from Core Midplane 6.00 5.00 4.00 3.00 2.00 1.00 0.00

-1.00

-2.00

-3.00

-4.00

-5.00

-6.00 Cycle 10 9.44 EFPY 4.53E+18 8.41E+18 9.90E+18 1.05E+19 1.07E+19 1.08E+19 1.09E+19 1.09E+19 1.07E+19 1.05E+19 1.01E+19 8.73E+18 4.44E+18 Cycle 12 11.76 EFPY 5.48E+18 1.01E+19 1.19E+19 1.27E+19 1.29E+19 1.30E+19 1.31E+19 1.32E+19 1.30E+19 1.28E+19 1.23E+19 1.07E+19 5.38E+18 Projected Exposures 21 EFPY 9.24E+18 1.71E+19 2.01E+19 2.14E+19 2.18E+19 2.20E+19 2.22E+19 2.22E+19 2.19E+19 2.15E+19 2.07E+19 1.80E+19 9.07E+18 34 EFPY 1.45E+19 2.69E+19 3.16E+19 3.36E+19 3.43E+19 3.46E+19 3.48E+19 3.49E+19 3.45E+19 3.39E+19 3.25E+19 2.83E+19 1.43E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 51 EFPY 2.15E+19 3.96E+19 4.67E+19 4.97E+19 5.07E+19 5.10E+19 5.14E+19 5.15E+19 5.09E+19 5.OOE+19 4.80E+19 4.18E+19 2.11E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999

8-10 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) Exposure Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 450 Azimuthal Angle Distance from Core Midplane 6.00 5.00 4.00 3.00 2.00 1.00 0.00

-1.00

-2.00

-3.00

-4.00

-5.00

-6.00 Cycle 10 9.44 EFPY 5.60E+18 1.04E+19 1.22E+19 1.30E+19 1.33E+19 1.34E+19 1.35E+19 1.35E+19 1.33E+19 1.30E+19 1.25E+19 1.08E+19 5.49E+18 Cycle 12 11.76 EFPY 6.76E+18 1.25E+19 1.47E+19 1.57E+19 1.60E+19 1.61E+19 1.62E+19 1.62E+19 1.60E+19 1.58E+19 1.51E+19 1.32E+19 6.64E+18 Projected Exposures 21 EFPY 1.14E+19 2.10E+19 2.47E+19 2.63E+19 2.68E+19 2.70E+19 2.72E+19 2.73E+19 2.70E+19 2.65E+19 2.54E+19 2.21E+19 1.11E+19 34 EFPY 1.78E+19 3.30E+19 3.88E+19 4.13E+19 4.21E+19 4.24E+19 4.28E+19 4.28E+19 4.23E+19 4.16E+19 3.99E+19 3.48E+19 1.75E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 Table 8.1-8 51 EFPY 2.63E+19 4.86E+19 5.73E+19 6.09E+19 6.21E+19 6.26E+19 6.30E+19 6.31E+19 6.24E+19 6.13E+19 5.88E+19 5.13E+19 2.58E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999

8-11 Summary of Best Estimate Iron Atom Displacement [dpa] Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 0' Azimuthal Angle Distance from Core Midplane 6.00 5.00 4.00 3.00 2.00 1.00 0.00

-1.00

-2.00

-3.00

-4.00

-5.00

-6.00 Cycle 10 9.44 EFPY 2.15E-03 3.99E-03 4.69E-03 4.99E-03 5.09E-03 5.13E-03 5.16E-03 5.16E-03 5.09E-03 4.99E-03 4.77E-03 4.14E-03 2.10E-03 Cycle 12 11.76 EFPY 2.57E-03 4.76E-03 5.60E-03 5.96E-03 6.08E-03 6.12E-03 6.17E-03 6.18E-03 6.11E-03 6.OOE-03 5.75E-03 5.02E-03 2.53E-03 21 EFPY 4.27E-03 7.90E-03 9.31E-03 9.89E-03 1.01E-02 1.02E-02 1.02E-02 1.03E-02 1.01E-02 9.96E-03 9.55E-03 8.33E-03 4.19E-03 Projected Exposures 34 EFPY 6.66E-03 1.23E-02 1.45E-02 1.54E-02 1.57E-02 1.59E-02 1.60E-02 1.60E-02 1.58E-02 1.55E-02 1.49E-02 1.30E-02 6.54E-03 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 Table 8.1-9 51 EFPY 9.79E-03 1.81E-02 2.13E-02 2.27E-02 2.31E-02 2.33E-02 2.35E-02 2.35E-02 2.32E-02 2.28E-02 2.19E-02 1.91E-02 9.61E-03 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999

8-12 Table 8.1-10 Summary of Best Estimate Iron Atom Displacement [dpa] Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 150 Azimuthal Angle Distance from Core Midplane 6.00 5.00 4.00 3.00 2.00 1.00 0.00

-1.00

-2.00

-3.00

-4.00

-5.00

-6.00 Cycle 10 9.44 EFPY 3.18E-03 5.90E-03 6.94E-03 7.38E-03 7.53E-03 7.59E-03 7.63E-03 7.63E-03 7.53E-03 7.39E-03 7.06E-03 6.12E-03 3.11E-03 Cycle 12 11.76 EFPY 3.80E-03 7.02E-03 8.28E-03 8.80E-03 8.97E-03 9.04E-03 9.11E-03 9.12E-03 9.02E-03 8.86E-03 8.50E-03 7.41E-03 3.73E-03 21 EFPY 6.28E-03 1.16E-02 1.37E-02 1.45E-02 1.48E-02 1.49E-02 1.50E-02 1.51E-02 1.49E-02 1.46E-02 1.40E-02 1.22E-02 6.16E-03 Projected Exposures 34 EFPY 9.76E-03 1.80E-02 2.13E-02 2.26E-02 2.30E-02 2.32E-02 2.34E-02 2.34E-02 2.32E-02 2.27E-02 2.18E-02 1.90E-02 9.58E-03 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 51 EFPY 1.43E-02 2.65E-02 3.12E-02 3.31E-02 3.38E-02 3.41E-02 3.43E-02 3.44E-02 3.40E-02 3.34E-02 3.20E-02 2.79E-02 1.40E-02 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999

8-13 Table 8.1-11 Summary of Best Estimate Iron Atom Displacement [dpa] Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 300 Azimuthal Angle Distance from Core Midplane 6.00 5.00 4.00 3.00 2.00 1.00 0.00

-1.00

-2.00

-3.00

-4.00

-5.00

-6.00 Cycle 10 9.44 EFPY 3.03E-03 5.63E-03 6.63E-03 7.05E-03 7.19E-03 7.25E-03 7.29E-03 7.29E-03 7.20E-03 7.06E-03 6.74E-03 5.85E-03 2.97E-03 Cycle 12 11.76 EFPY 3.67E-03 6.79E-03 8.00E-03 8.50E-03 8.67E-03 8.74E-03 8.80E-03 8.81E-03 8.71E-03 8.56E-03 8.21E-03 7.16E-03 3.60E-03 Projected Exposures 21 EFPY 6.19E-03 1.14E-02 1.35E-02 1.43E-02 1.46E-02 1.47E-02 1.48E-02 1.49E-02 1.47E-02 1.44E-02 1.38E-02 1.21E-02 6.07E-03 34 EFPY 9.73E-03 1.80E-02 2.12E-02 2.25E-02 2.30E-02 2.32E-02 2.33E-02 2.34E-02 2.31E-02 2.27E-02 2.18E-02 1.90E-02 9.55E-03 Best Estimate Neutron Exposure of Pressure Vessel Materials 51 EFPY 1.44E-02 2.65E-02 3.13E-02 3.33E-02 3.39E-02 3.42E-02 3.44E-02 3.45E-02 3.41E-02 3.35E-02 3.21E-02 2.80E-02 1.41E-02 WCAP-15334, Rev. 0, November 1999

8-14 Table 8.1-12 Summary of Best Estimate Iron Atom Displacement [dpal Projections for the Beltline Region of the McGuire Unit 2 Reactor Vessel 450 Azimuthal Angle Distance from Core Midplane 6.00 5.00 4.00 3.00 2.00 1.00 0.00

-1.00

-2.00

-3.00

-4.00

-5.00

-6.00 Cycle 10 9.44 EFPY 3.48E-03 6.46E-03 7.61E-03 8.08E-03 8.25E-03 8.31E-03 8.36E-03 8.36E-03 8.25E-03 8.09E-03 7.74E-03 6.71E-03 3.41E-03 Cycle 12 11.76 EFPY 4.20E-03 7.76E-03 9.15E-03 9.73E-03 9.92E-03 9.99E-03 1.O0E-02 1.01E-02 9.96E-03 9.79E-03 9.39E-03 8.19E-03 4.12E-03 Projected Exposures 21 EFPY 7.06E-03 1.30E-02 1.54E-02 1.63E-02 1.67E-02 1.68E-02 1.69E-02 1.69E-02 1.67E-02 1.64E-02 1.58E-02 1.38E-02 6.92E-03 34 EFPY 1.11E-02 2.05E-02 2.41E-02 2.56E-02 2.62E-02 2.64E-02 2.66E-02 2.66E-02 2.63E-02 2.58E-02 2.48E-02 2.16E-02 1.09E-02 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 51 EFPY 1.63E-02 3.02E-02 3.56E-02 3.78E-02 3.86E-02 3.89E-02 3.92E-02 3.92E-02 3.87E-02 3.81E-02 3.65E-02 3.18E-02 1.60E-02 WCAP-15334, Rev. 0, November 1999 Best Estimate Neutron Exposure of Pressure Vessel Materials

8-15 8.2 EXPOSURE OF SPECIFIC BELTLINE MATERIALS As shown in Figure 1.1-1, the beltline region of the McGuire Unit 2 reactor vessel is comprised of three ring forgings (one intermediate shell, one lower shell, and one lower transition shell) and two circumferential welds joining the three shells.

The uppermost of the two circumferential welds (W05) is centered just above (-6.94 inches) the axial midplane of the active core at the axial location of the maximum vessel exposure. The intermediate shell extends upward to an elevation well above the active fuel and the lower shell extends downward to an elevation just below (-6.97 inches) the bottom of the active fuel.

The lowermost of the two circumferential welds (W04) is located at this point, joining the lower transition shell to the lower shell. 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 region materials are listed in Table 8.2-1 through 8.2-3. In these tables, the weld and forging exposure is expressed in terms of (D (E > 1.0 MeV), D (E > 0.1 MeV), and dpa. Data are also provided at the end of Cycle 10 for ease of comparison to the data in Reference 34.

The peak axial fluence occurs at the 450 azimuth behind the neutron pad throughout the service life of the unit on all three beltline region materials.

Due to the asymmetry of the neutron pads there is variation in the neutron exposure of the reactor vessel from octant to octant. With no surveillance capsule holder present, the neutron pad span ranges from 300 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 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 reactor vessel neutron exposure at all azimuthal locations.

Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 Best Estimate Neutron Exposure of Pressure Vessel Materials WCA-P-15334, Rev. 0, November 1999

8-16 Fast Neutron Fluence (E > 1.0 MeV) at Key Forging and Weld Locations of McGuire Unit 2 Best Estimate 1D(E>1.0 MeV) [n/cm2]

Cycle 10 Cycle 12 Location 9.44 EFPY 11.76 EFPY Intermediate Shell Forging - 526840 Circumferential Weld - W05 Lower Shell Forging - 411337 00 3.32E+18 3.97E+18 150 4.96E+18 5.93E+18 300 4.67E+18 5.64E+18 450 5.28E+18 6.36E+18 Circumferential Weld - W04 Lower Transition Shell Forging - 527428 00 1.35E+18 1.63E+18 150 2.02E+18 2.43E+18 300 1.90E+18 2.31E+18 450 2.15E+18 2.60E+18 21 EFPY 6.59E+18 9.78E+18 9.51E+18 1.07E+19 2.70E+18 4.OOE+18 3.89E+18 4.37E+18 Projected Exposures 34 EFPY 1.03E+19 1.52E+19 1.50E+19 1.68E+19 4.21E+18 6.23E+18 6.12E+18 6.86E+18 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 Table 8.2-1 51 EFPY 1.51E+19 2.23E+19 2.21E+19 2.47E+19 6.19E+18 9.13E+18 9.04E+18 1.01E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999

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

Cycle 10 Cycle 12 Location 9.44 EFPY 11.76 EFPY Intermediate Shell Forging - 526840 Circumferential Weld - W05 Lower Shell Forging - 411337 00 7.06E+18 8.45E+18 150 1.07E+19 1.27E+19 300 1.09E+19 1.31E+19 450 1.35E+19 1.62E+19 Circumferential Weld - W04 Lower Transition Shell Forging - 527428 00 2.88E+18 3.46E+18 150 4.35E+18 5.21E+18 300 4.44E+18 5.38E+18 450 5.49E+18 6.64E+18 21 EFPY 1.40E+19 2.10E+19 2.22E+19 2.72E+19 5.74E+18 8.61E+18 9.07E+18 1.11E+19 Projected Exposures 34 EFPY 2.19E+19 3.27E+19 3.48E+19 4.28E+19 8.96E+18 1.34E+19 1.43E+19 1.75E+19 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 Table 8.2-2 51 EFPY 3.21E+19 4.80E+19 5.14E+19 6.30E+19 1.32E+19 1.96E+19 2.11E+19 2.58E+19 WCAP-15334, Rev. 0, November 1999 Best Estimate Neutron Exposure of Pressure Vessel Materials

Iron Atom Displacements [dpa] at Key Plate and Weld Locations of McGuire Unit 2 Best Estim Cycle 10 Location 9.44 EFPY Intermediate Shell Forging - 526840 Circumferential Weld - W05 Lower Shell Forging - 411337 00 5.16E-03 150 7.63E-03 300 7.29E-03 450 8.36E-03 Circumferential Weld - W04 Lower Transition Shell Forging - 527428 00 2.10E-03 150 3.11E-03 300 2.97E-03 450 3.41E-03 ate Iron Atom Cycle 12 1.76 EFPY 6.17E-03 9.11E-03 8.80E-03 1.01E-02 2.53E-03 3.73E-03 3.60E-03 4.12E-03 Displacements [dval 21 EFPY 1.02E-02 1.50E-02 1.48E-02 1.69E-02 4.19E-03 6.16E-03 6.07E-03 6.92E-03 Projected Exposures 34 EFPY 1.60E-02 2.34E-02 2.33E-02 2.66E-02 6.54E-03 9.58E-03 9.55E-03 1.09E-02 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 Table 8.2-3 8-18 51 EFPY 2.35E-02 3.43E-02 3.44E-02 3.92E-02 9.61E-03 1.40E-02 1.41E-02 1.60E-02 DisLae nt

[-r

-a Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999

8-19 8.3 UNCERTAINTIES IN EXPOSURE PROJECTIONS The overall uncertainty in the best estimate exposure projections within the reactor vessel wall arises 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 reactor 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 reactor vessel and the exposure at the measurement locations is based on the vessel thickness tolerance relative to the cavity data and on the downcomer water density variations and the reactor vessel inner radius tolerance relative to the surveillance capsule data.

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

(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 reactor vessel locations has been obtained from benchmarking studies using the Westinghouse neutron transport methodology"*1 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, reactor 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)

(D (E > 0.1 MeV) dpa Bias Factor 4.6%

9.4%

8.1%

Analytical 6.0%

6.0%

6.0%

Combined 7.5%

11.2%

10.1%

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

Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999 Best Estimate Neutron Exposure of Pressure Vessel Materials WCAP-15334, Rev. 0, November 1999

9-1 9

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1.

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2.

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7.

ASTM Designation E853-87 (Re-approved 1995), "Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results," in 1999 Annual Book of ASTM Standards, Volume 12.02, ASTM, West Conshohocken, PA, 1999.

8.

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9.

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10.

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11.

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12.

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ASTM Designation E1005-97, "Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance," in 1999 Annual Book of ASTM Standards, Volume 12.02, ASTM, West Conshohocken, PA, 1999.

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RSICC 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," Proceedings of the 4th ASTMiEURATOM Symposium on Reactor Dosimetry, NUREG/CP-0029, NRC, Washington, D.C., July 1982.

23.

"The Nuclear Design and Core Physics Characteristics of the W. B. McGuire Unit 2 Nuclear Power Plant Cycle 1," WCAP-10182, September 1982. [Westinghouse Proprietary Class 2]

24.

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

WCAP-10747, March 1985. [Westinghouse Proprietary Class 2]

25.

"The Nuclear Design of the W. B. McGuire Unit 2 Nuclear Power Plant Cycle 3", WCAP 11048, March 1986. [Westinghouse Proprietary Class 2]

26.

"The Nuclear Design of the W. B. McGuire Unit 2 Nuclear Power Plant Cycle 4", WCAP 11530, June 1987. [Westinghouse Proprietary Class 2]

27.

"The Nuclear Design of the W. B. McGuire Unit 2 Nuclear Power Plant Cycle 5", WCAP 11891, July 1988. [Westinghouse Proprietary Class 2]

28.

"The Nuclear Design of the W. B. McGuire Unit 2 Nuclear Power Plant Cycle 6", WCAP 12316, August 1989. [Westinghouse Proprietary Class 2]

29.

"The Nuclear Design of the W. B. McGuire Unit 2 Nuclear Power Plant Cycle 7", WCAP 12736, July 1990. [Westinghouse Proprietary Class 2]

References WCA.P-15334, Rev. 0, November 1999

9-3

30.

K. Naugle, "Transmittal of the McGuire Unit 2 Cycle 8 Core Inventory, Average Assembly Burnups, and Average Axial Power Conditions, October 10, 1996. [DPC Proprietary Information from DPC Calc. Files: MCC-1553.05-00-0083, Rev. 3, June 1993 and MCC-1553.05-00-0155, February 1994]

31.

K. Naugle, "Transmittal of the McGuire Unit 2 Cycle 9 Core Inventory, Average Assembly Burnups, and Average Axial Power Conditions, October 10, 1996. [DPC Proprietary Information from DPC Calc. Files: MCC-1553.05-00-0123, Rev. 3, June 1993 and MCC-1553.05-00-0183, December 1994]

32.

K. Naugle, "Transmittal of the McGuire Unit 2 Cycle 10 Core Inventory, Average Assembly Burnups, and Average Axial Power Conditions, October 10, 1996. [DPC Proprietary Information from DPC Calc. Files: MCC-1553.05-00-0154, December 1993 and MCC-1553.05-00-0185, Rev. 1, April 1996]

33.

T. E. Foley, "Core Power Distribution Information for McGuire Unit 2 Neutron Dosimetry Analysis," November 11, 1999. [DPC Proprietary Information]

34.

"Analysis of Capsule W from the Duke Power Company McGuire Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-14799, March 1997.

35.

"Westinghouse Fast Neutron Exposure Methodology for Pressure Vessel Fluence Determination and Dosimetry Evaluation," WCAP-13362, May 1992. [Westinghouse Proprietary Class 21.

36.

"Duke Power Company Reactor Cavity Neutron Measurement program for William B.

McGuire Unit 1 Cycle 12," WCAP-15253, July 1999.

References WCAP-15334, Rev. 0, November 1999

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

The startup and shutdown dates for each fuel cycle comprising the irradiation history of the surveillance capsules are listed below. Data for the two fuel cycles following the withdrawal of Capsule W are also included for ease of reference.

Cycle 1

2 3

4 5

6 7

8 9

10 11 12 05/07/83 05/08/85 06/27/86 07/05/87 07/27/88 09/19/89 12/29/90 03/17/92 09/14/93 01/12/95 05/14/96 12/18/97 Shutdown 01/25/85 03/14/86 05/01/87 05/27/88 07/05/89 09/01/90 01/09/92 07/01/93 11/24/94 04/05/96 10/03/97 03/12/99 Capsule V Withdrawn Capsule X Withdrawn Capsule U Withdrawn Capsules Y & Z Withdrawn Capsule W Withdrawn The detailed operating history of the reactor over the course of these twelve fuel cycles is provided in Table A-1. Note that the reference full power for McGuire Unit 2 is 3411 MWt.

The measured specific activities of the monitors removed from Capsules V, X, U, Y, Z, and W are provided in Tables A-2 through A-7, respectively. The locations of the various monitors within the surveillance capsules may be obtained from Reference 2.

Appendix A WCAP-15334, Rev. U, November 1999 WCAP-15334, Rev. 0, November 1999 Appendix A

McGuire Unit 2 Operating History - Cycles 1 Through 12 Date May-s Jun-8 Jul-83 Aug-8 Sep-8 Oct-&

Nov-8 Dec-&

Jan-8z Feb-8z Mar-&

Apr-&

May-8 Jun-84 Jul-84 Aug-8 Sep-84 Oct-84 Nov-S Dec-8&

Jan-85 MWt-h EFPS Cycle 1 MWt-hr 3

29,804 3

281,343 503 3

713,998 3

1,283,803 3

1,407,029 3

1,677,739

• 3 1,637,288 430,628 4

1,908,578

4 2,306,794

4 2,301,882 4

2,131,577 4

2,382,476 1,558,900

• 4 614,384 4

2,272,550 2,260,438

4 1,822,772 4

1,719,551 1,948,246 Lr Cycle 2 Cycle 3 Date MWt-hr Date MWt-hr May-85 1,613,921 Jun-86 98,860 Jun-85 2,080,510 Jul-86 2,419,839 Jul-85 949,838 Aug-86 2,295,334 Aug-85 1,698,280 Sep-86 2,455,840 Sep-85 2,451,943 Oct-86 2,268,723 Oct-85 2,342,207 Nov-86 834,670 Nov-85 2,366,518 Dec-86 2,539,926 Dec-85 1,328,243 Jan-87 1,942,205 Jan-86 2,284,248 Feb-87 2,094,930 Feb-86 2,262,594 Mar-87 2,537,082 Mar-86 1,070,655 Apr-87 2,391,293 May-87 20,798 30,690,283 3.239E+07 20,448,957 2.158E+07 21,899,500 2.311E+07 Cycle 4 Date MWt-hr Jul-87 1,968,987 Aug-87 2,304,777 Sep-87 2,209,664 Oct-87 2,534,464 Nov-87 2,091,145 Dec-87 2,449,205 Jan-88 2,375,936 Feb-88 2,372,171 Mar-88 2,528,325 Apr-88 2,446,326 May-88 1,965,065 25,246,065 2.664E+07 Appendix A WCAP-15334, Rev. 0, November 1999 Table A-1 A-2 Appendix A WCAP-15334, Rev. 0, November 1999

McGuire Unit 2 Operating History - Cycles 1 Through 12 (continued)

Cycle 5 Date MWt-hr Jul-88 182,814 Aug-88 2,312,603 Sep-88 2,447,331 Oct-88 2,518,180 Nov-88 2,395,542 Dec-88 2,528,858 Jan-89 2,475,080 Feb-89 2,280,475 Mar-89 2,248,654 Apr-89 1,929,466 May-89 2,259,109 Jun-89 2,365,375 Jul-89 290,705 MWt-hr 26,234,192 EFPS 2.769E+07 Cycle 6 Cycle 7 Cycle 8 Date MWt-hr Date MWt-hr Date MWt-hr Sep-89 745,948 Dec-90 55,107 Mar-92 846,822 Oct-89 2,483,669 Jan-91 2,425,825 Apr-92 2,321,028 Nov-89 2,400,111 Feb-91 2,231,028 May-92 1,624,848 Dec-89 2,469,574 Mar-91 2,538,082 Jun-92 403,108 Jan-90 2,458,034 Apr-91 2,444,210 Jul-92 2,529,061 Feb-90 2,283,648 May-91 2,523,168 Aug-92 2,074,181 Mar-90 2,469,938 Jun-91 2,372,591 Sep-92 2,449,036 Apr-90 2,446,697 Jul-91 2,191,309 Oct-92 2,531,961 May-90 2,474,847 Aug-91 2,528,077 Nov-92 2,449,669 Jun-90 2,359,097 Sep-91 2,315,536 Dec-92 2,533,105 Jul-90 2,492,907 Oct-91 2,156,131 Jan-93 2,525,260 Aug-90 2,298,172 Nov-91 2,185,092 Feb-93 2,016,910 Sep-90 5,074 Dec-91 2,500,590 Mar-93 2,369,376 Jan-92 628,307 Apr-93 2,452,571 May-93 2,465,261 Jun-93 1,947,721 Jul-93 7,767 27,387,716 2.891E+07 29,095,053 3.071E+07 33,547,685 3.541E+07 Appendix A WCAP-15334, Rev. 0, November 1999 Table A-1 A-3 WCAP-15334, Rev. 0, November 1999 Appendix A

McGuire Unit 2 Operating History - Cycles 1 Through 12 (continued)

Cycle 10 Cycle 11 Cycle 12 MWt-hr Date Date Sep-93 Oct-93 Nov-93 Dec-93 Jan-94 Feb-94 Mar-94 Apr-94 May-94 Jun-94 Jul-94 Aug-94 Sep-94 Oct-94 Nov-94 914,446 Jan-95 1,265,251 Feb-95 2,422,534 Mar-95 2,077,334 Apr-95 2,018,852 May-95 2,283,841 Jun-95 2,533,507 Jul-95 2,445,941 Aug-95 2,529,621 Sep-95 2,452,082 Oct-95 2,535,315 Nov-95 2,469,538 Dec-95 2,444,624 Jan-96 2,533,244 Feb-96 1,726,862 Mar-96 Apr-96 Date MWt-hr MWt-hr 1,510,651 2,286,904 2,445,829 1,953,193 2,534,654 2,404,622 2,534,676 2,516,579 2,451,456 2,536,146 2,452,222 1,767,303 2,526,107 2,366,757 2,452,599 325,181 35,064,879 3.701E+07 Appendix A WCAP-15334, Rev. 0, November 1999 Table A-1 Cycle 9 A-4 May-96 413,906 Jun-96 0

Jul-96 2,302,509 Aug-96 2,530,734 Sep-96 2,451,632 Oct-96 2,477,019 Nov-96 1,487,425 Dec-96 2,529,393 Jan-97 2,521,053 Feb-97 2,288,023 Mar-97 2,070,129 Apr-97 2,356,756 May-97 2,532,084 Jun-97 1,141,152 Jul-97 1,562,319 Aug-97 2,526,832 Sep-97 2,165,325 Oct-97 161,797 33,518,088 3.538E+07 Date Dec-97 Jan-98 Feb-98 Mar-98 Apr-98 May-98 Jun-98 Jul-98 Aug-98 Sep-98 Oct-98 Nov-98 Dec-98 Jan-99 Feb-99 Mar-99 MWt-hr 815,605 2,531,723 2,123,326 2,486,372 2,445,127 2,534,880 2,452,802 2,535,620 2,534,604 2,445,546 2,538,199 2,451,324 2,535,549 2,527,869 2,190,213 797,901 35,946,660 3.794E+07 MWt-hr 32,652,992 EFPS 3.446E+07 Appendix A WCAP-15334, Rev. 0, November 1999

A-5 Table A-2 Radiomtetric Counting Results For Sensors Removed From Capsule V CHEMICAL ANALYSIS REPORT WESTINGHOUSE ADVANCED ENERGY SYSTEMS DIVISIONANL SVREUTM.

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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 12/18/97 03/12/99 RCND Set 2S-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 2 is 3411 MWt.

The capsule loading table for the RCND Set 2S-1 irradiated during Cycle 12 is provided in Table B-2. The measured specific activities of the monitors removed from RCND Set 2S-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.

Appendix B WCAP-15334, Rev. 0, November 1999 WCAP-15334, Rev. 0, November 1999 Appendix B

B-2 Table B-i McGuire Unit 2 Operating History - Cycle 12 Cycle 12 Date MWt-hr Dec-97 815,605 Jan-98 2,531,723 Feb-98 2,123,326 Mar-98 2,486,372 Apr-98 2,445,127 May-98 2,534,880 Jun-98 2,452,802 Jul-98 2,535,620 Aug-98 2,534,604 Sep-98 2,445,546 Oct-98 2,538,199 Nov-98 2,451,324 Dec-98 2,535,549 Jan-99 2,527,869 Feb-99 2,190,213 Mar-99 797,901 MWt-hr 35,946,660 EFPS 3.794E+07 Appendix B WCAP-15334, Rev. 0, November 1999 Appendix B WCAP-15334, Rev. 0, November 1999

B-3 Table B-2 McGuire Unit 2 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 G-1 B

G G

G-2 Cd CG G

G G

CG G-3 Cd 46 42 H-1 B

H H

H-2 Cd CH H

H H

CH H-3 Cd 47 43 I-1 B

I I

1-2 Cd C

I I

I C

1-3 Cd 48 44 J-1 B

i J

J-2 Cd CJ i

i i

Ci J-3 Cd 49 45 K-1 B

AA AA K-2 Cd DA AA AA AA DA K-3 Cd 50 46 L-1 B

AB AB L-2 Cd DB AB AB AB DB L-3 Cd 51 47 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 37Np oxide detector, and two cadmium covers.

WCAP-15334, Rev. 0, November 1999 Appendix B

Table B-3 Appropriate methods were used and are i Appropriate quality assurance/quality c accordance with Antech Ltd. 's Quality Appendix B.

If you have any questions, ry J. Grohregin Supervisor EJG:rks Enclosures ndicated accordingly on the data tables.

ontrol measures were performed in Assurance Plan and 10 CFR, part 50 please call me at 724-722-5219.

Brian M. Carson QA/QC Coordinator Appendix B WCAP-15334, Rev. 0, November 1999 Radiometric Counting Results For Sensors Removed From Cycle 12 Cavity Dosimetry Set 2S-1 Capsules G, H, I, J, K, and L t

Antech Ltd.

Waltz Mill Site - P.O.

Box 158 - Madison, PA 15663-0158 - Phone: (724) 722-5214, Fax: (724) 722-5208 July 8, 1999 Certificate of Conformance Mr.

Larry Becker Westinghouse Electric Company CD&ME Department F Building, MS 60

Madison, PA 15663 Dosimetry Characterization; Purchase Order No. DIPP-7500 McGuire Unit #2 cycle #12; Duke Power; Nuclear Service Division Antech Ltd. Project No.

99-0326W

Dear Mr. Becker:

Enclosed are analytical results for samples submitted by Westinghouse Electric Company.

Samples were received and logged in for analysis on May 28, 1999.

B4 Appendix B WCAP-15334, Rev. 0, November 1999

Radiometric Counting Results For Sensors Removed From Cycle 12 Cavity Dosimetry Set 2S-1 Capsules G, H, I, J, K, and L (continued)

ANTECH LTD.

CASE NARRATIVE I.

PROJECT LOGIN INFORMATIONt A:

PROJECT NUMBERS; ANTECH LTD.:

CLIENT:

99-0326W Purchase Order Number:

DIPP-7500 B:

SAMPLE IDEINTIFICATIONSz Antech ID Client ID 9905-0279W G-Fe Bare 9905-0281W G-Ni Cd 9905-0283W G Ti Cd 9905-0285W CG CoAl Cd 9905-0287W NP-237 42 Cd 9905-0289W CH-Fe Cd 9905-0291W H-Cu Cd 9905-0293W H-CoAl Bare 9905-0295W U-238 47 Cd 9905-0297W I-Fe Bare 9905-0299W I-Ni Cd 9905-0301W I-Ti Cd 9905-0303W CI-CoAl Cd 9905-0305W Np-237 44 cd 9905-0307W CJ-Fe Cd 9905-0309W J-Cu Cd 9905-0311W J-CoAl Bare 9905-0313W U-238 49 Cd 9905-0315W AA-Fe Bare 9905-0317W AA-Ni Cd 9905-0319W AA-Ti Cd 9905-0321W PA-CoAl cd 9905-0323W NP-237 46 Cd 9905-0325W DB-Fe Cd 9905-0327W AB-Cu Cd 9905-0329W AB-CoAl Bare 9905-0331W U-238 51 Cd 9905-0333W CHAIN 2S-1(0) 9905-0335W

+4.5(0) 9905-0337W

+2.5(0) 9905-0339W

+0.5(0) 9905-0341W

-1.5(0) 9905-0343W

-3.5(0) 9905-0345W

-5.5(0) 9905-0347W CHAIN 2S-1 (15) 9905-0349W

+4.5 (15) 9905-0351W

+2.5 (15)

Antech ID Client ID 9905-0280W 9905-0282W 9905-0284W 9905-0286W 9905-0288W 9905-0290W 9905-0292W 9905-0294W 9905 0296W 9905-0298W 9905-0300W 9905-0302W 9905 0304W 9905-0306W 9905-0308W 9905-0310W 9905-0312W 9905-0314W 9905-0316W 9905-0318W 9905-0320W 9905-0322W 9905-0324W 9905-0326W 9905-0328W 9905-0330W 9905-0332W 9905-0334W 9905-0336W 9905-0338W 9905-0340W 9905-0342W 9905-0344W 9905-0346W 9905-0348W 9905-0350W 9905-0352W GC -Fe Cd G Cu Cd G-CoAl Bare U-238 46 Cd H-Fe Bare H-Ni Cd H-Ti Cd CH-CoAI Cd NP-237-43 Cd CI-Fe Cd I-Cu Cd I-CoAl Bare U-238 48 Cd J-Fe Bare J-Ni Cd J-Ti Cd CJ-CoAl Cd NP-237 45 Cd DA-Fe Cd AA-Cu Cd AA-CoAI Bare U-238 50 Cd AB-Fe Bare AB-Ni Cd AB-Ti Cd DB-CoAl Cd NP-237 47 Cd

+5.5(0)

+3.5(0)

+1.5(0)

-0.5(0)

-2.5(0)

-4.5(0)

-6.5 (0)

+5.5 (15)

+3.5 (15)

+1.5 (15)

Appendix B WCAP-15334, Rev. 0, November 1999 Table B-3 B-5 WCAP-15334, Rev. 0, November 1999 Appendix B

Radiometric Counting Results For Sensors Removed From Cycle 12 Cavity Dosimetry Set 2S-1 Capsules G, H, I, J, K, and L (continued)

ANTECH LTD.

CASE NARRATIVE (Continued)

+0.5 (15)

-1.5 (15)

-3.5 (15)

-5.5 (15)

CHAIN 2 S 1 (30)

+4.5 (30)

+2.5 (30)

+0.5 (30)

-i.5 (30)

-3.5 (30)

-5.5 (30)

CHAIN 2S-1 (45)

+4.5 (45)

+2.5 (45)

+0.5 (45) 5 (45)

-3.5 (45)

-5.5 (45)

Nist 101E 9905-0354W 9905-0356W 9905-0358W 9905-0360W 9905-0362W 9905-0364W 9905-0366W 9905-0368W 9905-0370W 9905-0372W 9905-0374W 9905-0376W 9905-0378W 9905-0380W 9905-0382W 9905-0384W 9905-0386W 9905-0388W 9905-0391W

-0.5

-2.5

-4.5

-6.5

-5.5

+3.5

+1.5

-0.5

-2.5

-4.5

-6.5

+5.5

+3.5

+1.5

-0.5

-2.5

-4.5

-6.5 True (15)

(15)

(15)

(15)

(30)

(30)

(30)

(30)

(30)

(30)

(30)

(45)

(45)

(45)

(45)

(45)

(45)

(45)

Value NIST 101E C.

SHIPPING/RECEIVING COMMENTS; Final Report:

07/07/99 II.

PREPARATION/ANALYSIS COMMENTSt At METALSs NONE B:

RADIOLOGICAL:

Decay correct date for gamma analysis is 6/11/99.

III.

GENERAL COMMENT

Sz Trailing zeroes and decimal places appearing on the data should not be interpreted as precision of the analytical procedure, but rather as a result of reporting format.

Appendix B WCAP-15334, Rev. Q. November 1999 Table B-3 B-6 9905-0353W 9905-0355W 9905-0357W 9905-0359W 9905-0361W 9905 0363W 9905-0365W 9905-0367W 9905-0369W 9905-0371W 9905-0373W 9905-0375W 9905-0377W 9905-0379W 9905-0381W 9905-0383W 9905-0385W 9905-0387W 9905-0390W Appendix B WCAP-15334, Rev. 0, November 1999

Radiometric Counting Results For Sensors Removed From Cycle 12 Cavity Dosimetry Set 2S-1 Capsules G, H, I, J, K, and L (continued) .4 44 C

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Radiometric Counting Results For Sensors Removed From Cycle 12 Cavity Dosimetry Set 2S-1 Capsules G, H, I, J, K, and L (continued)

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Radiometric Counting Results For Sensors Removed From Cycle 12 Cavity Dosimetry Set 2S-1 Capsules G, H, I, J, K, and L (continued)

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\\0 Page 1 of 2 Parameter Identification Mn-54 Co-58 Co-60 Antech Client A-524 A-524 A-524 Sawple I0 Sample 10 dips/ir dps/waj dps/mg 9905-0333W CHAIN 2S-1(0)

.(1) 9905-0334w

+5.5(0) 3.89M0 t 2.8E-1 6.77E0 +/- 5.0E-1 3.39EI t 7.8E-1 9905-0335W

-4.5(0) 4.62E0 +/- 3.35-1 8.15E0 +/- 5.9E-1 4.04E1 +/- 9.2E-1 9905-0336W

+3.5(0) 4.96E0 +/- 2.6E-1 8.84E0 t 3.5E-1 4.60E1 t 9.9E-1 9905-0337W

+2.5(0) 5.31E0 +/- 2.7E-1 8.73E0 t 3.6E-1 4.95E1 +/- 1.1E0 9905-033wM

+1.5(0) 5.1950 a 2.8E-1 8.64E0 +/- 3.6E-1 6.34E1 + 1.3E0 9905-0339W

+0.5(0) 5.08E0 +/- 2.7E-1 8.47E0

• 3.7E-1 6.56E1 +/- 1.4E0 9905-034*0W

-0.5(0) 5.01E0 t 2.7E-1 8.28E0 t 3.6E-1 6.4051 +/- 1.4E0 9905-0341Y

-1.5(0) 4.97E0 +/- 2.6E-I 8.22EO +/- 3.6E-1 6.21E1 +/- 1.350 9905-0342W

-2.5(0) 4.5IEO +/- 2.5E-1 7.88E0 +/- 3.5E-1 5.57E1 +/- 1.20 9905-0343W

-3.5(0) 3.90E0 +/- 2.1E-1 6.79E0 t 3.0E-1 3.59E1 +/- 7.8E-1 9905-0344w

-4.5(0) 2.3050 +/- 1.2E-1 4.20E0 a 1.6E-1 2.44E1 +/- 5.2F-1 9905-0345W

-5.5(0) 8.60E-1 +/- 7.1E-2 1.57E0 +/- 1.2E-1 1.58E1 +/- 3.7E-1 9905-0346W

-6.5(0) 2.50E-1 t 3.6E-2 5.05E-1 +/- 5.3E-2 7.1150 t 1.7E-1 9905-0347W CHAIN 2S-1 (15) 9905-0348W

+5.5 (15) 5.67E0 +/- 2.7E-1 1.0151 _+/- 3.9E-1 6.99E1 +/- 1.47E0 9905-0349W

+4.5 (15) 6.51E0 +/- 3.4E-1 1.17E1 +/- 4.5E-1 8.57E1 t 1.850 9905-0350W

+3.5 (15) 6.95E0 +/- 3.5E-1 1.21E1 +/- 5.1E-1 9.67E1 +/- 2.0EO 9905-0351W

+2.5 (15) 7.45E0 +/- 3.5E-1 1.20E1 +/- 4.8E-1 9.49E1 +/- 2.0EO 9905-0352W

+1.5 (15) 7.16E0 +/- 3.2E-1 1.19E1 +/- 4.8E-1 8.47E1

+/- 1.8E0 9905-0353W

+0.5 (15) 7.12E0 a 3.3E-1 1.19t1 +/- 4.7E-1 8.31E1 +/- 1.BEO 9905-0354w

-0.5 (15) 7.0OEO +/- 3.3E-1 1.17E1 +/- 4.4E-1 8.31E1 +/- 1.7E0 9905-0355W

-1.5 (15) 6.71E0 +/- 3.0E-1 1.17E1 t 4.4E-1 7.95E1 t 1.7E0 9905-0356W

-2.5 (15) 6.45E0 +/- 2.8E-1 1.13E1 +/- 4.2E-1 7.36E1 i 1.6EO 9905-0357W

-3.5 (15) 5.69E0 t 2.9E-1 9.8750 +/- 4.0E-1 6.70E1 a 1.4E0 9905-0358W

-4.5 (15) 3.73O t 2.0E-1 6.66E0 +/- 3.0E-1 4.81E1 +/- 1.0EO 9905-0359W

-5.5 (15) 1.50E0 +/- 1.0E-1 2.77E0 +/- 1.4E-1 2.80EI t 6.1E-1 9905-036*0W

-6.5 (15) 4.36E-1 +/- 5.2E-2 8.05E-1 +/- 6.6E-2 1.31E1 +/- 2.9E-1 See footnotes at end of table.

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Table 2 General Data Table Westinghouse Electric Corqpany Antech Ltd. Project No. 99-0326W Dosimetry Characterization; McGuire Unit #2 Cycle #12 Charge Order No. DIPP-7500; Duke Power etc fl)

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Table 2 (Continued)

Page 2 of 2 Parameter Identification Mn-54 Co-58 Co-60 Antech CL lent A-524 A-524 A-524 SaWte ID Sample ID ctslm

• s/rs dps/*m 9905-0361W CHAIN 2 S-i (30) 9905-03621"

+5.5 (30) 5.60E0 t 2.8E-1 1.05E1 +/- 4.3E-1 9.8001 +/- 2.100 9905-0363W

+4.5 (30) 6.58E0 +/- 3.4E-1 1.16G1 t 4.9E-1 1.23E2 +/- 2.6E0 9905-0364W

+3.5 (30) 7.22E0 t 3.9E-1 1,2201 +/- 5.2E-1 1.38E2 t 2.9E0 9905-0365W

+2.5 (30) 7.4760 +/- 3.9E-1 1.22E1 t 5.2E-1 1.40M2

• 2.900 9905-0366W

+1.5 (30) 7.1900 +/- 3.80-1 1.2101

+/- 4.8E-1 1.33E2 +/- 2.800 9905-0367W

+0.5 (30) 6.8700 +/- 3.7E-1 1.21E1 +/- 5.06-1 1.26E2 +/- 2.600 9905-0368W

-0.5 (30) 6.9260 +/- 4.0E-1 1.16E1

• 4.8E-1 1.16E2 +/- 2.460 9905-03691W

-1.5 (30) 6.9400 +/- 3.7E-1 1.1901 +/- 4.9E-1 1.20E2 i 2.500 9905-0370W

-2.5 (30) 6.5900 +/- 3.6E-1 1.1561 +/- 4.8E-1 1.07E2 +/- 2.3E0 9905-0371W

-3.5 (30) 5.51E0 +/- 2.8E-1 9.9100 +/- 4.4E-1 9.7561 a 2,1E0 9905-0372N

-4.5 (30) 3.6600 +/- 2.6E-1 6.65E0 t 3.3E-1 7.2201 a 1.5E0 9905-0373W4

-5.5 (30) 1.64E0 +/- 1.6E-1 2.9600 t 1.9E-1 4.3601 a 9.3E-1 9905-037414

-6.5 (30) 4.88M-1 +/- 5.70-2 9.Z00-1 t 8.6F-2 1.5461 +/- 4.3E-1 9905-0375W CHAIN 2S-1 (45) 9905-0376W

+5.5 (45) 4.8760 +/- 3.5E-1 9.1800 +/- 6.7E-1 6.06E1 t 1.4E0 9905-0377W

+435 (45) 6.1600 +/- 3.1E-1 1.0901 +/- 4.4E-1 7.34E1 +/- 1.600 9905-0378W

+3.5 (45) 6.35E0 +/- 3.0E-1 1.11E1 +/- 4.4E-1 8.1601 +/- 1.700 9905-0379W

+2.5 (45) 6.23E0 +/- 3.16-1 1.12E1 t 4.6E-1 8.60E1 +/- 1.800 9905-0380W

+1.5 (45) 6.43E0 +/- 3.3E-1 1.12E1 t 4.5E-1 8.46E1

• 1.800 9905-0381W

+0.5 (45) 6.07E0 +/- 3.2E-1 1.07E1 +/- 4.4E-1 8.12V1 +/- 1.700 9905-0382W

-0.5 (45) 5.9160 t 3.3E-1 1.04E1 t 4.2E-1 8.09E1 +/- 1.760 9905-0383W

-1.5 (45) 5.9900 +/- 3.1E-1 1.03E1 +/- 4.2E-1 7.8A01 t 1.0EO 9905-0384W

-2.5 (45) 5.5600 +/- 2.8E-1 9.93E0 +/- 3.9E-1 7.0361 1 1.5E0 9905-0385W

-3.5 (45) 4.67M0 +/- 2.6E-1 8.5300 +/- 3.7E-1 6.01E1 +/- 1.3E0 9905-0386M

-4.5 (45) 2.8800 i 2.16-1 5.44E0 +/- 4.06-1 4,0901

_ 9.4E-1 9905-0387W

-5.5 (45) 1.23E0 +/- 1.0E-1 2.3160

• 1.8E-1 2.8701 +/- 6.6E-1 9905-0388M

-6.5 (45) 1.156-1 t 1.7E-2 2.22E-1 t 2.6t-2 3.35E0 +/- 2.3E-I (O)Dash denotes not analyzed.

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Table 4 General Data Table Westinghouse Electric Company Antech Ltd. Project No.

99-0326W Dosimetry Characterization; McGuire Unit #2 Cycle #12 Charge Order No. DIPP-7500; Duke Power Parameter Identification Cobalt (Total)

Iron (Total)

Nickel (Total)

Antech Client Date 6010(1) 6010())

6010(i)

Sample ID Sample ID Collected mg/kg mo/ko mg/kg 9905-0333W CHAIN 2S-1(0) 1600 720000 97000 9905-0347W CHAIN 2S-1 (15) 1600 750000 100000 9905-0361W CHAIN 2 S-I (30) 1700 720000 98000 9905-0375W CHAIN 2S-1 (45) 1700 750000 100000 9905-0389W Method Blank

<2.0

<10.00

<10.00 9905-0390W Nist 101E 2000 708300 96360 9905-0391W True Value NIST 101E 1800 690000 94800

( 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.

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