ML072490211

WCAP-16093-NP, Revision 2, Analysis of Capsule U from the South Texas Project Nuclear Operating Company, South Texas Unit 2 Reactor Vessel Radiation Surveillance Program

ML072490211
Person / Time
Site:	South Texas
Issue date:	07/31/2007
From:	Burgos B Westinghouse
To:	Office of Nuclear Reactor Regulation
References
WCAP-16093-NP, Rev 2
Download: ML072490211 (174)

Text

Westinghouse Non-Proprietary Class 3 WCAP-16093-NP July 2007 Revision 2 Analysis of Capsule U from the South Texas Project Nuclear Operating Company, South Texas Unit 2 Reactor Vessel Radiation Surveillance Program 0%)Westinghouse 7

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16093-NP, Revision 2 Analysis of Capsule U from the South Texas Project Nuclear Operating Company, South Texas Unit 2 Reactor Vessel Radiation Surveillance Program B. N. Burgos*

July 2007 Approved by:

J. S. Carlson, Manager*

Primary Component Asset Management

• Electronically approved records are authenticated in the Electronic Document Management System.

Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355

@2007 Westinghouse Electric Company LLC All Rights Reserved

ili TABLE OF CONTENTS L IS T O F T A B L E S..................................................................................................................................

iv L IS T O F F IG U R E S...............................................................................................................................

vi P R E F A C E..........................................................................................................................................

v iii EX EC U T IV E SU M M A R Y...........................................

ix I

SUMMARY

OF RESULTS...................................................

1-1 2

IN T R O D U C T IO N...................................................................................................................

2-1 3

B A C K G R O U N D.....................................................................................................................

3-1 4

D ESCRIPTION O F PROG RA M.............................................................................................

4-1 5

TESTING OF SPECIMENS FROM CAPSULE U..................................................................

5-1 5.1 O V E R V IE W................................................................................................................

5-1 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS...............................

5-3 5.3 TEN SILE TEST RESULTS........................................................................................

5-5 5.4 I/2T COMPACT TENSION SPECIMEN TESTS........................................................

5-5 6

RADIATION ANALYSIS AND NEUTRON DOSIMETRY...................................................

6-1 6.1 IN TR O D U CT IO N................................................................................................

6-1 6.2 DISCRETE ORDINATES ANALYSIS........................................................................

6-2 6.3 NEUTRON DOSIMETRY........................................................................................

6-5 6.4 CALCULATIONAL UNCERTAINTIES......................................................................

6-6 7

SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE...............................................

7-1 8

R E FE R E N C E S........................................................................................................................

8-1 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS..............................................

A-0 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS....................... B-0 APPENDIX C CHARPY V-NOTCH PLOTS FOR CAPSULE U USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FHTI'VNG METHOD................

C-0 APPENDIX D SOUTH TEXAS UNIT 2 SURVEILLANCE PROGRAM CREDIBILITY EVA LU ATIO N...............

D -0 WCAP-16093-NP, Rev. 2 July 2007

iv LIST OF TABLES Table 4-1 Chemical Composition (wt %) of the South Texas Unit 2 Reactor Vessel Surveillance M aterials (U nirradiated)..................................................

........................................... 4-3 Table 4-2 Heat Treatment History of the South Texas Unit 2 Reactor Vessel Surveillance Materials.........................................................

4-4 Table 5-1 Charpy V-Notch Data for the South Texas Unit 2 Intermediate Shell Plate R2507-1 Irradiated to a Fluence of 2.40 x 10'9 n/cm2 (E > 1.0 MeV) (Longitudinal Orientation)..5-6 Table 5-2 Charpy V-Notch Data for the South Texas Unit 2 Intermediate Shell Plate R2507-1 Irradiated to a Fluence of 2.40 x 1019 r/cm2 (E > 1.0 MeV) (Transverse Orientation)..... 5-7 Table 5-3 Charpy V-notch Data for the South Texas Unit 2 Surveillance Weld Material Irradiated to a Fluence of 2.40 x 10'9 n/cm2 (E> 1.0 MeV)............................................

5-8 Table 5-4 Charpy V-notch Data for the South Texas Unit 2 Heat-Affected-Zone (HAZ)

Material Irradiated to a Fluence of 2.40 x I019 n/cm2 (F> 1.0 MeV).............................. 5-9 Table 5-5 Instrumented Charpy Impact Test Results for the South Texas Unit 2 Intermediate Shell Plate R2507-1 Irradiated to a Fluence of 2.40 x 10'9 n/cm2 (E> 1.0 MeV)

(Longitudinal O rientation).........................................................................................

5-10 Table 5-6 Instrumented Charpy Impact Test Results for the South Texas Unit 2 Intermediate Shell Plate R2507-1 Irradiated to a Fluence of 2.40 x l0'9 n/cm2 (E> 1.0 MeV)

(T ransverse O rientation).............................................................................................

5-11 Table 5-7 Instrumented Charpy Impact Test Results for the South Texas Unit 2 Surveillance Weld Metal Irradiated to a Fluence of 2.40 x 10'9 n/cm 2 (E> 1.0 MeV)....................... 5-12 Table 5-8 Instrumented Charpy Impact Test Results for the South Texas Unit 2 Heat-Affected-Zone (HAZ) Metal Irradiated to a Fluence of 2.40 x 10'9 n/m 2 (>

1.0MeV)............. 5-13 Table 5-9 Effect of Irradiation to 2.40 x 10'9 n/cm 2 (E> 1.0 MeV) on the Capsule U Notch Toughness Properties of the South Texas Unit 2 Reactor Vessel Surveillance M aterials....................................................................................................................

5-14 Table 5-10 Comparison of the South Texas Unit 2 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, R evision 2, Predictions...............................................................................................

5-15 WCAP-16093-NP, Rev. 2 July 2007

v LIST OF TABLES (Cont.)

Table 5-11 Tensile Properties of the South Texas Unit 2 Capsule U Reactor Vessel Surveillance Materials Irradiated to 2-40 x 1019 n/cm2 (E> 1.0MeV)..............................................

5-16 Table 6-1 Calculated Neutron Exposure Rates and Integrated Exposures At The Surveillance C apsule C enter.....................................................................................

6-13 Table 6-2 Calculated Azimuthal Variation of Maximum Exposure Rates and Integrated Exposures at the Reactor Vessel Clad/Base Metal Interface.........................................

6-17 Table 6-3 Calculated Integrated Exposures for Key Vessel Plate and Weld Materials At The Reactor Vessel Clad/Base M etal Interface...................................................................

6-21 Table 6-4 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV) Within The R eactor V essel W all..................................................................................................

6-23 Table 6-5 Relative Radial Distribution Of Iron Atom Displacements (dpa) Within The R eactor V essel W all....................................................................................................

6-23 Table 6-6 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from South Texas Project U nit 2.........................................................................................

6-24 Table 6-7 Calculated Surveillance Capsule Lead Factors............................................................

6-24 Table 7-1 Recommended Surveillance Capsule Withdrawal Schedule...........................................

7-1 WCAP-16093-NP, Rev. 2 July 2007

vi LIST OF FIGURES Figure 4-1 Arrangement of Surveillance Capsules in the South Texas Unit 2 Reactor Vessel........ 4-5 Figure 4-2 Capsule U Diagram Showing the Location of Specimens, Thermal Monitors, and D osim eters...............................................................................................................

4-6 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Longitudinal Orientation)..........................

5-17 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Longitudinal Orientation)............. 5-18 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Longitudinal Orientation).........................

5-19 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Transverse Orientation)..............................

5-20 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Transverse Orientation)..............................

5-21 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Transverse Orientation).............................

5-22 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for South Texas Unit 2 Reactor Vessel W eld M etal.......................................................................................................

5-23 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature. for South Texas Unit 2 Reactor Vessel W eld M etal...........................................................................................

5-24 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for South Texas Unit 2 Reactor Vessel W eld M etal.................................................................................................................

5-25 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for South Texas Unit 2 Reactor Vessel Heat-Affected-Zone M aterial.............................................................................

5-26 Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for South Texas Unit 2 Reactor Vessel Heat-Affected-Zone M aterial.............................................................................

5-27 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for South Texas Unit 2 Reactor Vessel Heat-Affected-Zone M aterial.............................................................................

5-28 Figure 5-13 Charpy Impact Specimen Fracture Surfaces for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Longitudinal Orientation)......................................

5-29 WCAP-16093-NP, Rev. 2 July 2007

vii LIST OF FIGURES (Cont.)

Figure 5-14 Charpy Impact Specimen Fracture Surfaces for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Transverse Orientation).........................................

5-30 Figure 5-15 Charpy Impact Specimen Fracture Surfaces for South Texas Unit 2 Reactor Vessel W eld M eta l................................................................................................................

5 -3 1 Figure 5-16 Charpy Impact Specimen Fracture Surfaces for South Texas Unit 2 Reactor Vessel H eat-A ffected-Zone M etal.........................................................................................

5-32 Figure 5-17 Tensile Properties for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Longitudinal O rientation)...........................................................................

5-33 Figure 5-18 Tensile Properties for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R 2507-1 (Transverse O rientation)..............................................................................

5-34 Figure 5-19 Tensile Properties for South Texas Unit 2 Reactor Vessel Weld Metal.......................... 5-35 Figure 5-20 Fractured Tensile Specimens from South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Longitudinal Orientation).......................................

5-36 Figure 5-21 Fractured Tensile Specimens from South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Transverse Orientation).........................................

5-37 Figure 5-22 Fractured Tensile Specimens from South Texas Unit 2 Reactor Vessel Weld Metal...... 5-38 Figure 5-23 Engineering Stress-Strain Curves for South Texas Unit 2 Intermediate Shell Plate R2507-1 Tensile Specimens HL-1, HL-2 and HL-3 (Longitudinal Orientation)............ 5-39 Figure 5-24 Engineering Stress-Strain Curves for South Texas Unit 2 Intermediate Shell Plate R2507-1 Tensile Specimens HT-1, HT-2 and HT-3 (Transverse Orientation)............... 5-41 Figure 5-25 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens HW-2, HW-2 and fH W -3......................................

5-4 3 Figure 6-1 South Texas Project Unit 2 rO Reactor Geometry 12.50 Neutron Pad at the Core M idplane..........................................................

6-8 20.00 Neutron Pad at the Core Midplane.....

.......................................... 6-9 22.5' Neutron Pad at the Core Midplane........................................................

6-10 Figure 6-2 South Texas Project Unit 2 r,z Reactor Geometry w ith N eutron Pad.................................................................................

6-1l w ithout N eutron Pad..................................................................................

6-12 WCAP-16093-NP, Rev. 2 July 2007

viii PREFACE Revision 2 has been tecntically reviewed by:

Reviewer (Revision 2):

F.C. Gift*

RECORD OF REVISIONS Revision 0:

Revision 1:

Revision 2:

Original Issue Appendix D text was updated to reflect the correct plant name. This change was editorial in nature only.

Additional formatting changes were made to be consistent with the current standards of EDMS. These changes were editorial in nature only.

• Electronically approved records are authenticated in the Electronic Document Management System.

WCAP-16093-NP, Rev. 2 July 2007

ix EXECUTIVE

SUMMARY

The purpose of this report is to document the results of the testing of surveillance Capsule U from South Texas Unit 2. Capsule U was removed at 10.31 EFPY and post irradiation mechanical tests of the Charpy V-notch and tensile specimens were performed. A fluence evaluation utilizing the recently released neutron transport and dosimetry cross-section libraries was derived from the ENDF/B-VI data-base. Capsule U received a fluence of 2.40 x 10' 9 n/cm2 (E > 1.0 MeV) after irradiation to 10.31 EFPY. The peak clad/base metal interface vessel fluence after 10.31 EFPY of plant operation was 7.52 x 10"s n/cm2 (E > 1.0 MeV).

This evaluation lead to the following conclusions: 1) The measured 30 ft-lb shift in transition temperature values of the intermediate shell plate R2507-1 contained in capsule U (longitudinal) is less than the Regulatory Guide 1.99, Revision 21'3, predictions. 2) The measured 30 ft-lb shift in transition temperature values of the intermediate shell plate R2507-1 contained in capsule U (transverse) is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift value is less than two sigma allowance by Regulatory Guide 1.99, Revision 2. 3) The measured 30 fl-lb shift in transition temperature values of the weld metal contained in capsule U is less than the Regulatory Guide 1.99, Revision 2, predictions-

4) The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules U contained in the South Texas Unit 2 surveillance program are less than the Regulatory Guide 1.99, Revision 2 predictions. 5) All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the current license (34 EFPY) as required by IOCFR50, Appendix G 121. 6) The South Texas Unit 2 surveillance data was found to be credible. This evaluation can be found in Appendix D.

Lastly, a brief summary of the Charpy V-notch testing can be found in Section 1. All Charpy V-notch data was plotted using a symmetric hyperbolic tangent curve fitting program.

WCAP-16093-NP, Rev. 2 July 2007

1-1 I

SUMMARY

OF RESULTS The analysis of the reactor vessel materials contained in surveillance Capsule U, the third capsule removed and tested from the South Texas Unit 2 reactor pressure vessel, led to the following conclusions:

The Charpy V-notch data presented in WCAP-14978131 were based on a re-plot of all capsule data from WCAP-9967141 and WCAP-131821s" using CVGRAPH, Version 4.1, which is a symmetric hyperbolic tangent curve-fitting program. The results presented herein only for the Capsule U test results, which are also based on using CVGRAPH, Version 4.1. This report also shows the composite plots that show the results from the previous capsules. Appendix C presents the CVGRAPH, Version 4.1, Charpy V-notch plots and the program input data.

Capsule U received an average fast neutron fluence (E> 1.0 MeV) of 2.40 x 1019 n/cm 2 after 10.31 effective full power years (EFPY) of plant operation.

Irradiation of the reactor vessel intermediate shell plate R2507-1 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation), resulted in an irradiated 30 ft-lb transition temperature of-10.49'F and an irradiated 50 ft-lb transition temperature of 21.74'F. This results in a 30 ft-lb transition temperature increase of 27.48°F and a 50 ft-lb transition temperature increase of 29.02'F for the longitudinal oriented specimens.

Irradiation of the reactor vessel intermediate shell plate R2507-I Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 22.23'F and an irradiated 50 fl-lb transition temperature of 65.3°E This results in a 30 ft-lb transition temperature increase of 40.18'F and a 50 ft-lb transition temperature increase of 47.2'F for the longitudinal oriented specimens.

Irradiation of the weld metal (heat number 90209) Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of 5.88gF and an irradiated 50 ft-lb transition temperature of 33.53'F. This results in a 30 ft-lb transition temperature increase of 20.64°F and a 50 ft-lb transition temperature increase of 21.471F.

Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of-95.05°F and an irradiated 50 ft-lb transition temperature of-50.42'F. This results in a 30 ft-lb transition temperature increase of 21.84'F and a 50 ft-lb transition temperature increase of 33.77'R The average upper shelf energy of the intermediate shell plate R2507-1 (longitudinal orientation) resulted in no energy decrease after irradiation. This results in an irradiated average upper shelf energy of 138 ft-lb for the longitudinal oriented specimens.

Introduction WCAP-16093-NP, Rev. 2 July 2007

1-2 The average upper shelf energy of the Intermediate Shell Plate R2507-1 (transverse orientation) resulted in no energy decrease after irradiation. This results in an irradiated average upper shelf energy of 98 fi-lb for the longitudinal oriented specimens.

The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of 1 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 97 ft-lb for the weld metal specimens.

The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an average energy decrease of 26 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 129 ft-lb for the weld HAZ metal.

A comparison, as presented in Table 5-10, of the South Texas Unit 2 reactor vessel surveillance material test results with the Regulatory Guide 1.99, Revision 2[1] predictions led to the following conclusions:

The measured 30 ft-lb shift in transition temperature value of the intermediate shell plate R2507-1 contained in capsule U (longitudinal) is less than the Regulatory Guide 1.99, Revision 2, predictions.

The measured 30 ft-lb shift in transition temperature value of the intermediate shell plate R2507-1 contained in capsule U (transverse) is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift value is less than the two sigma allowance by Regulatory Guide 1.99, Revision 2.

The measured 30 ft-lb shift in transition temperature value of the weld metal contained in capsule U is less than the Regulatory Guide 1.99, Revision 2, predictions.

The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules U contained in the South Texas Unit 2 surveillance program are less than the Regulatory Guide 1.99, Revision 2 predictions.

All beitline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the end of the current license (34 EFPY) as required by IOCFR50, Appendix G j2].

The calculated end-of-license (34 EFPY) neutron fluence (E> 1.0 MeV) at the core midplane for the South Texas Unit 2 reactor vessel using the Regulatory Guide 1.99, Revision 2 attenuation formula (i.e., Equation #3 in the guide) are as follows:

Calculated:

Vessel inner radius* = 2.36 x 1019 n/cm 2 Vessel 1/4 thickness = 1.41 x 10' 9n/cm2 Vessel 3/4 thickness = 4.99 x 1011 n/Cm2

• Clad/base metal interface. (Interpolated From Table 6-2)

Introduction WCAP-16093-NP, Rev. 2 July 2007

2-1 2

INTRODUCTION This report presents the results of the examination of Capsule U, the third capsule removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the South Texas Project Nuclear Operating Company, South Texas Unit 2 reactor pressure vessel materials under actual operating conditions.

The surveillance program for the South Texas Project Nuclear Operating Company South Texas Unit 2 reactor pressure vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the pre-irradiation mechanical properties of the reactor vessel materials are presented in WCAP-9967, "Houston Lighting & Power Company South Texas Project Unit No. 2 Reactor Vessel Radiation Surveillance Program"' 41. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-79, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels.' "I Capsule U was removed from the reactor after 10.31 EFPY of exposure and shipped to the Westinghouse Science and Technology Department Hot Cell Facility, where the post-irradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.

This report summarizes the testing of and the post-irradiation data obtained from surveillance capsule U removed from the South Texas Project Nuclear Operating Company South Texas Unit 2 reactor vessel and discusses the analysis of the data.

Introduction WCAP-16093-NP, Rev. 2 July 2007

3-1 3

BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because-it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA533 Grade B Class I (base material of the South Texas Unit 2 reactor pressure vessel belhine) are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness during high-energy irradiation.

A method for ensuring the integrity of reactor pressure vessels has been presented in "Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code 7. The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RTNrDT).

RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-2086) or the temperature 60'F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented perpendicular (transverse) to the major working direction of the plate. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (K1, curve) which appears in Appendix G to the ASME CodePT. The Ki, curve is a lower bound of static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the KI, curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined using these allowable stress intensity factors.

RTmTD and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The changes in mechanical properties of a given reactor pressure vessel steel, due to irradiation, can be monitored by a reactor vessel surveillance program, such as the South Texas Unit 2 reactor vessel radiation surveillance programn4, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested.

The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNDrT) due to irradiation is added to the initial RTNDT, along with a margin (M) to cover uncertainties, to adjust the RTNDr (ART) for radiation embrittlement. This ART (RTNDT initial + M + ARTNDT) is used to index the material to the K1, curve and, in turn, to set operating limits for the nuclear power plant that take into account the effects of irradiation on the reactor vessel materials.

Background

WCAP-16093-NP, Rev. 2 July 2007

4-1 4

DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the South Texas Unit 2 reactor pressure vessel core region (beltline) materials were inserted in the reactor vessel prior to initial plant start-up. The six capsules were positioned in the reactor vessel between the neutron pads and the vessel wall as shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core. The capsules contain specimens made from intermediate shell plate R2507-1, weld metal fabricated with weld wire Type B4, Heat Number 90209 and Linde Type 124 flux, Lot Number 1061, which is identical to that used in the actual fabrication of the intermediate to lower shell circumferential weld seam and lower shell longitudinal weld seams. The surveillance weld was fabricated with the same heat of weld wire as all beltline region welds and is therefore representative of all of the reactor vessel beltline region welds.

Capsule U was removed after 10.31 effective full power years (EFPY) of plant operation. This capsule contained Charpy V-notch, tensile, and 1/2T-CT fracture mechanics specimens made from Intermediate Shell Plate R2507-1 and submerged arc weld metal representative of all the reactor vessel beltline region weld seams. In addition, this capsule contained Charpy V-notch specimens from the weld Heat-Affected-Zone (HAZ) metal of plate R2507-1.

Test material obtained from the intermediate shell course plate (after thermal heat treatment and forming of the plate) was taken at least one plate thickness from the quenched edges of the plate. All test specimens were machined from the 1/44 thickness location of the plate after performing a simulated post-weld stress-relieved treatment on the test material and also from weld and heat-affected-zone metal of a stress-relieved weldment joining intermediate shell plate R2507-1 and adjacent intermediate shell plate R2507-2. All heat-affected-zone specimens were obtained from the weld heat-affected-zone of the Intermediate Shell Plate R2507-1.

Charpy V-notch impact specimens from intermediate shell plate R2507-1 were machined in the longitudinal orientation (longitudinal axis of the specimen parallel to the major working direction) and also in the transverse orientation (longitudinal axis of the specimen perpendicular to the major working direction).

The core region weld Charpy impact specimens were machined from the weldment such that the long dimension of each Charpy specimen was perpendicular to the weld direction. The notch of the weld metal Charpy specimens was machined such that the direction of crack propagation in the specimen was in the welding direction.

Tensile specimens from intermediate shell plate R2507-1 were machined in both the longitudinal and transverse orientations. Tensile specimens from the weld metal were oriented with the long dimension of the specimen perpendicular to the weld direction.

Compact tension test specimens from intermediate shell plate R2507-1 were machined in the longitudinal and transverse orientations. Compact tension test specimens from the weld metal were machined perpendicular to the weld direction with the notch oriented in the direction of welding. All specimens were fatigue pre-cracked according to ASTM E399.

The chemical composition and heat treatment of the unirradiated surveillance materials are presented in Tables 4-1 and 4-2, respectively. The data in Table 4-1 and 4-2 was obtained from the unirradiated surveillance program report, WCAP-9967, Appendix A.

Description of Program WCAP-16093-NP, Rev-2 July 2007

4-2 Capsule U contained dosimeter wires of pure iron, copper, nickel, and aluminum-0. 15 weight percent cobalt (cadmium-shielded and unshielded). In addition, cadmium shielded dosimeters of neptunium (Np27) and uranium (U238) were placed in the capsule to measure the integrated flux at specific neutron energy levels.

The capsule contained thermal monitors made from two low-melting-point eutectic alloys and sealed in Pyrex tubes. These thermal monitors were used to define the maximum temperature attained by the test specimens during irradiation. The composition of the two eutectic alloys and their melting points are as follows:

2.5% Ag, 97.5% Pb 1.5% Ag, 1.0% Sn, 97.5% Pb Melting Point: 5790F (3040C)

Melting Point: 590TF (3 100C)

The arrangement of the various mechanical specimens, dosimeters and thermal monitors contained in Capsule U is shown in Figure 4-2.

Description of Program WCAP-16093-NP, Rev. 2 July 2007

4-3 iT~le4-~3 3 3 henea Cmpsimo wt} f heSo~utht Tea nt" 3.3 l..-emt ?

'r t

33

a.

3 W

M C

0.220 0.120 Mn 1.550 1.450 P

0.006 0.010 S

0.012 0.011 Si 0.210 0.380 Ni 0.650 0.150 Mo 0.560 0.530 Cr 0.050 0.120 Cu 0.040 0_010 A]

0.021 0.012 Co 0.011 0.013 Pb

<0.001 0.001 W

<0.010 0.005 Ti

<0.010

<0.010 Zr 0.001

<0.001 V

0.002

<0.002 Sn

<0.001

<0.001 As 0.014 0.003 Cb

<0.010

<0.001 N2 0.009 0.004 B

<0.001

<0.001 Notes:

(a)

Data obtained from WCAP-9967 and duplicated herein for completeness.

(b)

Weld wire Type B4, Heat Number 90209, Flux Type Linde 124, and Flux Lot Number 1061.

Surveillance weldment is from a weld between the intermediate shell plates R2507-1 and R2507-2 and is identical to the intermediate to lower shell circumferential weld seam and the lower shell longitudinal weld seams.

Description of Program WCAP-16093-NP, Rev. 2 July 2007

4-4 Notes:

(a)

(b)

This table was taken from WCAP-9967 t.

The stress relief heat treatment received by the surveillance test plate and weldment have been simulated.

Description of Program WCAP-16093-NP, Rev. 2 July 2007

4-5 O"

CAPSULE U

- V 270"

-90*

w REACTOR VESSEL 180" PLAN VIEW ELEVATION ViEW Figure 4-1 Arrangement of Surveillance Capsules in the South Texas Unit 2 Reactor Vessel Description of Program WCAP-16093-NP, Rev. 2 July 2007

5-1 5

TESTING OF SPECIMENS FROM CAPSULE U 5.1 OVERVIEW The post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens was performed in the Remote Metallographic Facility (RMF) at the Westinghouse Research and Technology Park. Testing was performed in accordance with IOCFR50, Appendices G and H121, ASTM Specification E 185-821'1, and Westinghouse Procedure RMF 8402191, Revision 2 as modified by Westinghouse RMF Procedures 8102110], Revision i, and 81031"], Revision 1.

Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-9967 41. No discrepancies were found.

Examination of the two low-melting point 579°F (304'C) and 590°F (310°C) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximum temperature to which the test specimens were exposed was less than 579'F (304'C).

The Charpy impact tests were performed per ASTM Specification E23-98 11 21 and RMF Procedure 8103, Rev. 1, on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy impact test machine is instrumented with a GRC 930-1 instrumentation system, feeding information into an IBM compatible computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED). From the load-time curve (Appendix B), the load of general yielding (Poy), the time to general yielding (try), the maximum load (PM), and the time to maximum load (tM) can be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (PF), and the load at which fast fracture terminated is identified as the arrest load (PA).

The energy at maximum load (Em) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (EF,) is the difference between the total energy to fracture (ED) and the energy at maximum load (Em).

The yield stress (ay) was calculated from the three-point bend formula having the following expression:

ar = (Por

• L) / [B *(W a)2

• C]

(1) where:.

L

=

distance between the specimen supports in the impact machine B

=

the width of the specimen measured parallel to the notch W

height of the specimen, measured perpendicularly.to the notch a

notch depth The constant C is dependent on the notch flank angle (4)), notch root radius (p) and the type of loading (i.e.,

pure bending or three-point bending). In three-point bending, for a Charpy specimen in which ÷ 45' and p = 0.010 inch, Equation 1 is valid with C = 1.21. Therefore, (for L = 4W),

Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-2 cr, = (Poy

• L) /B

• (W_ a)2 *1.21] = (3.305 *Poy

• W) I[ B *(W-a) 2]

(2)

For the Charpy specimen, B 0.394 inch, W = 0.394 inch and a = 0.079 inch. Equation 2 then reduces to:

ar = 33.3

• Per (3) where a*, is in units of psi and Pry is in units oflbs. The flow stress was calculated from the average of the yield and maximum loads, also using the three-point bend formula.

The symbol A in columns 4, 5, and 6 of Tables 5-5 through 5-8 is the cross-section area under the notch of the Charpy specimens:

A= B * (W-a)= 0.1241 sq.in.

(4)

Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM Specification E23-98 and A370-97at" 1. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.

Tensile tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specification E8-99114 ) and E21-92 (1998)1"51, and Procedure RMF 8102, Rev. 1. All pull rods, grips, and pins were made of Inconel 718. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.

Extension measurements were made with a linear variable displacement transducer extensometer. The extensometer knife-edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length was 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-931' 61.

Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air. Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperatures. Chromel-Alumel thermocouples were positioned at the center and at each end of the gage section of a dummy specimen and in each tensile machine griper. In the test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower tensile machine griper and controller temperatures was developed over the range from room temperature to 550°F. During the actual testing, the grip temperatures were used to obtain desired specimen temperatures. Experiments have indicated that this method is accurate to +2°F The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined.

from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.

Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-3 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS The results of the Charpy V-notch impact tests performed on the various materials contained in Capsule U, which received a fluence of 2.40 x 1019 n/cm2(E> 1.0 MeV) in 10.31 EFPY of operation, are presented in Tables 5-1 through 5-8 and are compared with unirradiated results141 as shown in Figures 5-1 through 5-12.

The transition temperature increases and upper shelf energy decreases for the Capsule U materials are summarized in Table 5-9 and led to the following results:

Irradiation of the reactor vessel intermediate shell plate R2507-1 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction (longitudinal orientation), resulted in an irradiated 30 ft-lb transition temperature of - 10.49'F and an irradiated 50 ft-lb transition temperature of 21.74*F. This results in a 30 ft-lb transition temperature increase of 27.48°F and a 50 fl-lb transition temperature increase of 29.02'F for the longitudinal oriented specimens.

Irradiation of the reactor vessel intermediate shell plate R2507-I Charpy specimens, oriented with the longitudinal axis of the specimen perpendicular to the major working direction (transverse orientation), resulted in an irradiated 30 ft-lb transition temperature of 22.23°F and an irradiated 50 ft-lb transition temperature of 65.3F. This results in a 30 ft-lb transition temperature increase of 40. 18'F and a 50 ft-lb transition temperature increase of 47.2*F for the longitudinal oriented specimens.

Irradiation of the weld metal (heat number 90209) Charpy specimens resulted in an irradiated 30 fl-lb transition temperature of 5.88F and an irradiated 50 ft-lb transition temperature of 33.53*F. This results in a 30 ft-lb transition temperature increase of 20.64°F and a 50 ft-lb transition temperature increase of 21.47'F.

Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens resulted in an irradiated 30 ft-lb transition temperature of -95.05°F and an irradiated 50 ft-lb transition temperature of-50.42"F. This results in a 30 ft-lb transition temperature increase of 21.84°F and a 50 ft-lb transition temperature increase of 33.77F.

The average upper shelf energy of the intermediate shell plate R2507-1 (longitudinal orientation) resulted in no energy decrease after irradiation. This results in an irradiated average upper shelf energy of 138 ft-lb for the longitudinal oriented specimens.

The average upper shelf energy of the Intermediate Shell Plate R2507-1 (transverse orientation) resulted in no energy decrease after irradiation. This results in an irradiated average upper shelf energy of 98 ft-lb for the longitudinal oriented specimens.

The average upper shelf energy of the weld metal Charpy specimens resulted in an average energy decrease of I ft-lb after irradiation. This results in an irradiated average upper shelf energy of 97 ft-lb for the weld metal specimens.

Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-4 The average upper shelf energy of the weld HAZ metal Charpy specimens resulted in an average energy decrease of 26 ft-lb after irradiation. This results in an irradiated average upper shelf energy of 129 ft-lb for the weld HAZ metal.

A comparison, as presented in Table 5-10, of the South Texas Unit 2 reactor vessel surveillance material test results with the Regulatory Guide 1.99, Revision 2111 predictions led to the following conclusions:

The measured 30 ft-lb shift in transition temperature of the intermediate shell plate R2507-1 contained in capsule U (longitudinal) is less than the Regulatory Guide 1.99, Revision 2, predictions.

The measured 30 ft-lb shift in transition temperature of the intermediate shell plate R2507-1 contained in capsule U (transverse) is greater than the Regulatory Guide 1.99, Revision 2, predictions. However, the shift value is less than two sigma allowance by Regulatory Guide 1.99, Revision 2.

The measured 30 ft-lb shift in transition temperature of the weld metal contained in capsule U is less than the Regulatory Guide 1.99, Revision 2, predictions.

The measured percent decrease in upper shelf energy for all the surveillance materials of Capsules U contained in the South Texas Unit 2 surveillance program are less than the Regulatory Guide 1.99, Revision 2 predictions.

All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are predicted to maintain an upper shelf energy greater than 50 ft-lb throughout the end of the current license (34 EFPY) as required by I0CFR50, Appendix G [21.

The fracture appearance of each irradiated Charpy specimen from the various surveillance Capsule U materials is shown in Figures 5-13 through 5-16 and shows an increasingly ductile or tougher appearance with increasing test temperature.

The load-time records for individual instrumented Charpy specimen tests are shown in Appendix B.

The Charpy V-notch data presented in WCAP-14978 131 were based on a re-plot of all capsule data from WCAP-9967 (l and WCAP-13182151 using CVGRAPH, Version 4.1, which is a symmetric hyperbolic tangent curve-fitting program. The results presented herein only for the Capsule U test results, which are also based on using CVGRAPH, Version 4.1. This report also shows the composite plots that show the results from the previous capsules. Appendix C presents the CVGRAPH, Version 4.1, Charpy V-notch plots and the program input data for Capsule U.

Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-5 5.3 TENSILE TEST RESULTS The results of the tensile tests performed on the various materials contained in Capsule U irradiated to 2.40 x 101 n/cm 2 (E> 1.0 MeV) are presented in Table 5-1i and are compared with unirradiated results?41 as shown in Figures 5-17 and 5-19.

The results of the tensile tests performed on the intermediate shell plate R2507-1 (longitudinal orientation) indicated that irradiation to 2.40 x 10'9 n/cm 2 (E> 1.0 MeV) caused approximately a 6 to 10 ksi increase in the 0.2 percent offset yield strength and approximately a 7 ksi increase in the ultimate tensile strength when compared to unirradiated datat41. See Figure 5-17.

The results of the tensile tests performed on the intermediate shell plate R2507-1 (transverse orientation) indicated that irradiation to 2.40 x 1019 n/cm2 (E> 1.0 MeV) caused approximately a 4 to 6 ksi increase in the 0.2 percent offset yield strength and approximately a 4 to 6 ksi increase in the ultimate tensile strength when compared to unirradiated data1 41. See Figure 5-18.

The results of the tensile tests performed on the surveillance weld metal indicated that irradiation to 2.40 x 109 Dn/cm 2 (E> 1.0 MeV) caused approximately a 2 ksi increase in the 0.2 percent offset yield strength and approximately a 2 to 3 ksi increase in the ultimate tensile strength when compared to unirradiated data141. See Figure 5-19.

The fractured tensile specimens for the intermediate shell plate R2507-1 material are shown in Figures 5-20 and 5-21, while the fractured tensile specimens for the surveillance weld metal are shown in Figure 5-22.

The engineering stress-strain curves for the tensile tests are shown in Figures 5-23 through 5-25.

5.4 1/2T COMPACT TENSION SPECIMEN TESTS Per the surveillance capsule testing contract, the 1/2T Compact Tension Specimens were not tested and are being stored at the Westinghouse Research and Technology Park Hot Cell facility.

Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-6 Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-7 Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

,5-8 Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-9 MHH3e

-2

-c

.Tfne 129 6

80

.00ax 2

HH2

-4JW 150 10-I00 n&Hr2

-150-k

-117 0

00 HH4

-t125

-87 13 18 4

0.0 5

HH7?

-108

-78 30 41 II 0.28 10 HHI4

-75

-59 52 71 18 0.46 15 HH5

-50

-46 36 49 15 0.38 35 HHI 1

-25

-32 56 76 26 0.66 55 HH 10 0

-1 8

90 122 52 1.32 80 HH8 25

-4 i130 176 75 1.91 100 HHI5 50 10 61 83 35 0.89 55 HH9 75 24 96 130 53 1.35 70 HHI2 125 52 115 156 65 1.65 100 HH6 150 66 1!4 155 53 1.35 100 HH13

!75 79 144 195 73 1.85 100 HH1 200 93 143 194 76 1.93 100 Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5.10 0

C-z

( D 4 r%

Testin~g of Specimens from Capsule U

5-11 Table.5.6t Results for the South* Texas Urni t2 In terme~dlate Shell Plate P.2507=.: "

1 2raireAo~

Fune 01Sui x.1,ncm (E>I,0 MeV),

i.

-rans~verse

.O nato)....

, :. c,

• , r-.*q.'.'*

Test es o

a

~

~

Yed.Fow Sqpe T

M

-,~G,, "

Xt' "am"".e"

~

a...

...2. "L"". '....

.....*

• L, d.

,.. L* oa

.L

6.

i..'d

.re" s.,

HT..1

-75 2

16 9

8 1098 0.09 1190 0.10 1173 0

37 38 HT3

-50 10 81 44 37 3847 0.15 4034 0.17 4034 0

128 131 HT4

-25 24 193 155 38 3474 0.14 4384 0.37 4379 0

116 131 HT6 0

22 177 66 111 3384 0.14 4259 0.22 4194 0

113 127 HT7 25 32 258 193 65 3496 0.14 4472 0.44 4472 0

116 133 HT8 50 35 282 186 96 3313 0.14 4287 0.44 4273 591 110 127 HTI0 60 50 403 226 177 3345 0.14 4387 0.52 4358 425 111 129 HT12 75 59 475 325 150 3402 0.15 4518 0.69 4433 389 113 132 HT9 100 64 516 309 207 3323 0..15 4403 0.67 4373 1204 111 129 HT2 125 79 637 304 333 3184 0.15 4355 0.67 4180 2169 106 126 HTl4 150 79 637 299 337 3151 0.14 4262 0.67 3915 1313 105 123 HTI 190 101 814 284 530 3007 0.14 4072 0.66 n/a n/a 100 118 HTI3 225 102 822 290 532 3041 0.14 4171 0.66 r/a n/a 101 120 HT15 250 108 870 284

• 586 2934 0.14 4124 0.67 n/a n/a 98 118 HT5 275 104 838 288 550 2862 0.17 4004 0.70 n/a n/a 95 114 N)CD 0

C:)

Testing of Specimens from Capsule U

5-12 Cz c:0<

0*

Testing of Specimens from Capsule U

5.13 T

5-T HH6 150 114 91V HHI-13 175 144 116 HH1 200 143 115 0

o Testing of Specimens from Capsule U

5-14 TaI~3 Efet'fraWai4t

.40 0dc E>1.,O,MeVY') tiv e~a~i~

oc~~

ughneýss opr th SouthTexs. Unit, 2 Rt

, Vessel Surveillance Mater,~s'.

."3 i

s n e Shrage 30 27t.48)~""'~e 32.3 33.33

-727 21,74 a

l.

Average Energy 38so14

___ong._

)e Mtra Tr.ito eatr (0 p,

~

ansionTem36peratulre (0F)toi.eper eý 1 3,,ATIý Un-ait~radit

,AT,Unrhated'

Imrdiattd.

, AT,3,,,4Jifidiatedý,`

Irraditd In ternediate Shell

-37.97

-10.49 27.48

-0.9 32.43 33.33

-7.27 21.74 29.02 138 140 Plate R2507-1 (Long')

Intermediate Shell

-17.95 22.23 40.18 25.62 81.98 56.35 18.1 65.3 47.2 98 104 Plate R2507-1 (Trans.)

Weld Metal

-14.76 5.88 20.64 9.76 35,43 25.66 12.06 33.53 21.47 98 97 (Heat # 90209)

HAZ Metal

-116.9

-95.05 21.84

-60.75

-21.39 39.36

-84.2

-50,42 33.77 155 129

a.

"Average" is defined as the value read from the curve fit through the data points of the Charpy tests (see Figures 5-1, 54, 5-7 and 5-10).

b.

"Average" is defined as the value read from the curve fit through the data points of the Charpy tests (see Figures 5-2, 5-5, 5-8 and 5-11).

0.

":3 )

Testing of Specimens.from Capsule U

5-15

'~

peatre ShIflMnj~jie hl e~ir ec es i

a 9

vsn2,Predijc6~s

~ "4-

~

'.05 I "...r.,'

K.

1....M r

r a..)e M

fmppeir ShelfvEbergyi.

d(x)

EAI 0eV) z;dctd

-?,

kfam I OW measu (b

W4 PIMeaSUredW Intermediate Shell V

0.23 15.60 16.39 14 0

Plate R2507-1 Y

1.21 27.30 33.96 20 4

(Longitudinal)

U 2.40 32.24 27.48 23 0

Intermediate Shell V

0.23 15.60 11.86 14 0

Plate R2507-1 Y

1.21 27.30 35.26 20 0

(Transverse)

U 2.40 32.24 40.18 23 0

Surveillance V

0.23 14.70 0d) 14 5

Program Y

1.21 25.73 4.08 20 0

Weld Metal U

2.40 30.38 20.64 23 1

Heat Affected Zone V

0.23 d0e 12 Material Y

1.21 54.2 12 U

2.40 21.84 17 Notes:

(a)

Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material.

(b)

Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1 (See Appendix C)

(c)

Values are based on the definition of upper shelf energy given in ASTM El185-82.

(d)

The fluence values presented here are the calculated values, not the best estimate values.

(e)

Due to the scatter in the capsule V weld and HAZ Charpy test results, a true Hyperbolic Tangent Curve fit resulted in AT3o values of -7.60 F and -0.76°F, respecfively, when compared to unirradiated Charpy test data.

Physically this should not happen. Hence, based on engineering judgement a value of 0°F will be used in RTNDT calculations.

Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-16 Tewil ip Ii;Is r'e 11;0 1

1

/

Table:S.il

.esie.Proprte hSutezaI ni 2.Capsule

ýU!ReactorVessel qSurneianceMati**

l A

Maera ~ 'S~'

iITit 02%

Utiniate - Fr acue drcure' Fracture' ý" U hi~f 0rm ffi Ota1P.

ýRed~ci 4,~~!~ixuer emxp.

tengtb' t~od Sress" Eogtow lnaio nAe Inter. Shell Plate HL 2 75 75.5 96.3 3.1 1674 62.1 11,3 24.9 63 R2507-1(Long.)

HLl1 300 72.7 89.8 2.9 171.4 58.9 9.0 19.0 66 HL 3 550 66,0 93.5 3.2 166.3 64.2 11.0 22.1 61 Inter. Shell Plate HT 3 75 74.4 94.8 3.2 173.5 64.4 10.5 22.0 63 R2507-1 (Trans.)

HT 2 300 67,5 86.7 3.3 157.6 66.8 9.7 19.5 58 H-T 1 550 65.4 91.1 4.0 139.4 81.1 9.0 14.0 42 Weld Metal HW 1 75 77.8 91.8 2.8 186.0 57.2 11.8 24.5 69 I-W 3 300 72.3 87.0 2.9 192.8 58.1 9.0 20.8 70 HW 2 550 71.6 91.6 3.2 163.0 64.4 1

10.2 21.4 61

',*Y:

• ! =:ii:!!ii

!

f*-*'

i i

i Testing of Specimens from Capsule U

5-17 INTERMEDIATE SHELL PLATE R2507-1 (LONG)

CVGRAPH 41 Hyperbolic Tangent Curve Printed at 11'26.04 on 04-M-2003 Results Curve Ruence ISE d-LSF USE d-USE T o 30 d-T

• 30 T o 50 d-T

• 50 2

3 4

0 0

0 0

2.19 219 2.19 2J9 0

0 0

0 1381 140 139 133 0

2

-37N7

-11149

-4 0

Z1.48 1l69 33M6

-727 21.74 627 3004 0

13.5 37.32 C,)

-3w0

-200

-100 0

100 200 300 400 500 600 Temperature in Degrees F CuTre Legend 1 1-3e Data Set(s) Plotted Material Curve Plant Capsule OCI Heat) 1 2

3 4

S12 S12 512 UNIRR U

V Y

PLATE SA533BI PLATE SA33W1 PLATE SA533MB PLATE SA533W1 LT NR 62 097-1 LT NR 62 067-1 LT NR 62 067-1 LT NR 62 067-1 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Longitudinal Orientation)

Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-18 r-4 4-4

).-4 INTERMEDIATE SHELL PLATE R2507-1 (LONG)

DGRAPII 4.1 Hyperbolic Tangent Curve Printed at 1lZ?7T1 on 04-M-M Results Curve Fluence USE d-1S1 T

• LE35 d-T o LE35 1

0 MBM*

0

-.9 0

2 0

795

-401 3Z43 3

0 853 1.48 61 7

4 0

8

-5 37M 3.78 1507 1007-0 0

V A/-.

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Degrees F Curve Legend jO-20---

40 Data Set(s) Plotted Material Curve Plant Capsule Orn teat 1

2 3

4 S12 s12 ST2 UNIRR U

V Y

PLATE SAIXIBI PLATE SA5,.BI PLATE SAOM3BI PLATE SA533BI LT NR GZ 067-1 LT NR 6i2 067-1 LT NR 62 067-1 LT NR 62 067-1 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Longitudinal Orientation)

Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-19 INTERMEDIATE SHELL PLATE R2507-1 (LONG)

CVYRAPH 41 Hyperbolic Tangent Curve Printed at H1A48M4 on 04--2-2M03 Results Curve Fluence T 50W Shear d-T a 1i1' Shear d-T c W/ Shear 1

2 3

4 0

0 0

0 Z7J8 5577 3a73 45 0

2R59 1156 17M 0) a4

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Degrees F Curve Legend 1I1-30 4

Data "et(s) Plotted Material Curve Plant Capsule Oft Heatl 2

3 4

ST2 glT2 ST2 UN1R U

V Y

KATE SA533O PLATE SA533BI PLATE SA533BI PIATE SA5%3W LT LT LT LT NR 62 067-1 NR 62 067-1 NR 62 067-1 NR 62 067-1 Figure 5-3

.Charpy V-Notch Percent Shear vs. Temperature for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Longitudinal Orientation)

.Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-20 INTERMEDIATE SHELL PLATE R2507-1 (TRANS.)

C*HRAPH 4.1 Hyperbolic Tangent Curve Printed at IZM26 on W)4-02-20M Reults Curve Fluence ISE d-LSE USE d-USE T o 30 d-T o 30 T o 50 d-T o 50 1

2 3

4 0

0 0

0 219 2_19 2.19 2,19 0

0 0

0 98 104 105 105 0

6 7

7

-1735

--6.08 17.31 0

40.1 11.88 3526 181 653 45IM 56.39 0

472 Mi91 w82 U)

Q0 f.-r 4Z a)

-300

-200

-100 0

100 200 300 4W0 5w0 600 Temperature in Degrees F Curve Legend 10D 20-4-

Data Set(s) Plotted Material Curve Plant Capsule ot Beati 1

2 3

4 S12 S12 ST2 UNIRR U

V Y

PLATE SA533BI PLATE SA533Bi PLATE SA533B1 PLATE SA533B1 TL NR 62 067-1 nI NR C2 067-1 TL NR 62 O7-I n n 6(R2 o6m Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Transverse Orientation)

Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-21 INTERMEDIATE SHELL R-2507-1 (TRANS.)

CVGRAPH 41 Hyperbolic Tangent Curve Printed at 12_309 on 04-M-20-30 Results Curve Fluence USE d-USE To LE35 d-T o LE35 1

2 3

4 0

0 0

0 669 68.17 7856 6536 0

126 1166

-1.53 2a.62 819 4785 67.03 0

5635 2=L 41.4 co U],)

-300

-200

-100 0

100 200 300 4400 500 600 Temperature in Degrees F Clive Legend 30 1n-20---

4 Data Set(s) Plotted Material Curve Plant Capsule Ori Beat#

2 3

4 SM2 S12 UNIRE U

V Y

PLATE SA533B1 PLATE SA533BI PLATE SA533R1 PLATE SA5MBi It NR 1 067-1 11 NR 62 067-1 TL HR 62 067-1 TL NR 62 067-Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Transverse Orientation)

Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007.

5-22 INTERMEDIATE SHELL PLATE R2507-1 (TRANS.)

CVGRAPH 41 Hyperbolic Tangent Curve Printed at IaM on 04-M)-2003 Rullts Curve Fluence T o T/. Shear d-T o 5 Shear 3

4 0

0 0

0 21112 89.35 67.03 5710 0

6123 29.06 (U

Q)

<Uj 0-,

-300

-200

-100 0

100 200 300 400 500 600 Temperature in Degrees F Curve Lagend ID-20-3O-4 a-Data Set~s) Plotted Material Curve Plant Capsule Ori.

Heatd 1

2 3

4 I21 212 212 UM V

V Y

PIRATE SA,1IA PLATE SA533BI PLATE SAW53I1 PLATE SA533RB TL NR 62 067-1 TL NR 62 067-1 TL NR 62 067-1 TL NR 62 067-1 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Transverse Orientation)

Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-23 SURVEILLANCE WELD METAL CYGRAPH 41 Hyperbolic Tangent Curve Printed at 13911045 on 04-M2-2003 eesults Curve Fluence ISE d-ISE USE d-USE T o 30 d-T o 30 T o 50 d-T 9 50 2

4 0

0 0

0 219 0

98 2w9 0

97 2.19 0

93 2.19 0

101 0

-14.76

-1 5A

-5

-227 3

-1061 0

12.A 20-64 3353 6 1824 4.M 3242 0

21.47 618 2036 0) 4-1 0--)

z U

-300

-200

-100 0

100 200 300 400 500 6w0 Temperature in Degrees F Curve lgend 3

1.-

20--

4-Data Set"s) Plotted Material Curve Plant aisule 0ri.

Hteat Curve Plant CODSUle Material Ori.

Heat I 1

2 3

4 M12 UNIRR M

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90209 9Mo Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for South Texas Unit 2 Reactor Vessel Weld Metal Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-24 SURVEILLANCE WELD METAL CVGRAPH 41 Hyperbolic Tangent Curve Printed at 13JI107 on 04-02-2003 Results Curve Fluence USE d-USE T o LE35 d-T a LEM P-4 f4) 1 0

72.06 0

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100 200 300 400 500 500 Temperature in Degrees F Curve Legend I1--

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Data Set(s) Plotted Material Curve Plant Caisule 0n Ileati 2

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$12 ST2 512 5T2 UNIP*

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Y WELD WELD WELD maL 90209 9M29 90209 90209 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for South Texas Unit 2 Reactor Vessel Weld Metal Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-25 SURVEILLANCE WELD METAL CVCGRPII 41 Hyperbolic Tangent Curve Printed at 131353 on 04-02-2103 Results Curve Fluence T a 50,/ Shear d-T 0 50z. Shear d-T 0 59/. Shear 1

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V Y

TIm1 WELD WELD MDII 9M 9009 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for South Texas Unit 2 Reactor Vessel Weld Metal Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-26 HEAT AFFECTED ZONE MATERIAL CV(HRAP1I 41 Hyperbolic Tangent Curve Printed at 13'*M3I on N-W-20M Results Curve Fluence LSE d-LSE USE d-USE T o 30 d-T o 30 7 o 50 d-T o 50 2

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HEAT AFYF ZONE ST' V

HEAT AFWD ZONE Y

HEAT AFYD ZONE Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for South Texas Unit 2 Reactor Vessel Heat-Affected-Zone Material Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-27 HEAT AFFECTED ZONE CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 14138 on 04-M-2003 Results Curve Fluence USE d-USE T o U35 d-T o L5 2

3 4

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HEAT AFFU ZONE sr2 V

HEAT AFFD ZONE ST2 Y

HEAT AFFD ONE Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for South Texas Unit 2 Reactor Vessel Heat-Affected-Zone Material Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-29 HEAT AFFECTED ZONE MATERIAL CVGRAPH 4U Hyperbolic Tangent Curve Printed at 14"1a16 on 04-03-2003 Results Curve Fluence T o 5b'z Shear d-T a 5()'z Shear T o W/ Shear d-T 0 %/ Shear 2

3 4

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4 Data Set(s) Plotted Material Curve Plant Capsule Orn Heati 1

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4 ST2 ST2 S12 UNIRR U

V Y

HEAT AFFD ZONE HEAT AFFD ZONE HEAT AFFl ZONE HEAT AFl) ZONE Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for South Texas Unit 2 Reactor Vessel Heat-Affected-Zone Material Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

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WCAP-.16093-NP, Rev-2 July 2007>

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Interinediate Shel I'IPat R25(07-, ("Ir n*ierse Orit-niati

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Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

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100 200 300 400 500 600 70'0 TEMPERATURE (F)

Figure 5-18 Tensile Properties for South Texas Unit 2 Reactor Vessel Intermediate Shell Plate R2507-1 (Transverse Orientation)

Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-35 (0C) 0 50 100 150 200 250 300 350 110 100 90 S80 70 S60 50 40 I

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Figure 5-19 Tensile Properties for South Texas Unit 2 Reactor Vessel Weld Metal Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2.

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SOUTH TEXAS PROJECT UNIT 2 "U" CAPSULE HL-1 300'F 0

0.05 0.1 0.15, STRAIN, IN/IN 0.2 0.25 0.3 Figure 5-23 Engineering Stress-Strain Curves for South Texas Unit 2 Intermediate Shell Plate R2507-1 Tensile Specimens HL-I, HL-2 and HL-3 (Longitudinal Orientation)

Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-40 SOUTH TEXAS PROJECT UNIT-2

'UrCAPSULE 100 90 80 70 c

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~40 30 20 10 0

HL-3 550TF 0

0.05 0.1 0.15 0.2 0.25 STRAIN, IN/lN 0.3 Figure 5-23 Continued Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-41 U)

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SOUTH TEXAS PROJECT UNIT 2 "U` CAPSULE HT-3 75 F 0

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90 80 70 60 50 40 30 20 10 0

SOUTH TEXAS PROJECT UNIT 2 "U' CAPSULE HT-2 300 F

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Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-42 w

100 90 80 70 60 50 40 30 20 10 0

SOUTH TEXAS PROJECT UNIT 2 "U'CAPSULE HT-1 5*0 F 0

0.05 0.1 0.15 STRAIN, WiN 02 0.25 0.3 Figure 5-24 Continued Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-43 SOUTH TEXAS PROJECT UNIT 2 U* CAPSULE u) tJJ u) 100 90 80 70 60 50 40 30 20 10 0

HW-I 75 'F 0

0.05 0.1 0.15 STRAIN. IN/IN 0.2 0.25 0.3 f5 co)

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SOUTH TEXAS PROJECT UNIT 2

'V CAPSULE HW-3 300'F 0

0.05 0.1 0.15 STRAIN. IN/IN 0.2 0.25 0.3 Figure 5-25 Engineering Stress-Strain Curves for Weld Metal Tensile Specimens HW-1, HW-2 and HW-3 Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

5-44 SOUTH TEXAS UNIT 2

'U' CAPSULE Uj 1100 90 80 70 60 50 40 30 20 10 0

HW-2 550"F 0

0.05 0.1 0.15 0.2 0.25 STRAIN, IN/1N 0.3 Figure 5-25 Continued Testing of Specimens from Capsule U WCAP-16093-NP, Rev. 2 July 2007

6-1 6

RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6.1 INTRODUCTION

This section describes a discrete ordinates S. transport analysis performed for the South Texas Project Unit 2 reactor to determine the neutron radiation environment within the reactor pressure vessel and surveillance capsules. In this analysis, fast neutron exposure parameters in terms of fast neutron fluence (E > 1.0 MeV) and iron atom displacements (dpa) were established on a plant and fuel cycle specific basis. An evaluation of the most recent dosimetry sensor set from Capsule U, withdrawn at the end of the ninth plant operating cycle, is provided. In addition, to provide an up-to-date data base applicable to the South Texas Project Unit 2 reactor, sensor sets from previously withdrawn capsules (V and Y) were re-analyzed using the current dosimetry evaluation methodology. These dosimetry updates are presented in Appendix A of this report-Comparisons of the results from these dosimetry evaluations with the analytical predictions served to validate the plant specific neutron transport calculations. These validated calculations subsequently formed the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 54 Effective Full Power Years (EPPY). These projections also account for a plant uprating, from 3800 MWt to 3853 MWt, which began at the onset of the tenth operating cycle.

The use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for the development of damage trend curves as well as for the implementation of trend curve data to assess the condition of the vessel. In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves and improved accuracy in the evaluation of damage gradients through the reactor vessel wall.

Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, "Analysis and Interpretation of Light-Water Reactor Surveillance Results," recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a database for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, "Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials."

All of the calculations and dosimetry evaluations described in this section and in Appendix A were based on the latest available nuclear cross-section data derived from ENDF/B-VI and made use of the latest available calculational tools. Furthermore, the neutron transport and dosimetry evaluation methodologies follow the guidance and meet the requirements of Regulatory Guide 1. 190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."11 91 Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC approved methodology described in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 1996."0' The specific calculational methods applied are also consistent with those described in WCAP-1 5557, "Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology.' 421]

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-2 6.2 DISCRETE ORDINATES ANALYSIS A plan view of the South Texas Project Unit 2 reactor geometry at the core midplane is shown in Figure 4-

1. Six irradiation capsules attached to the neutron pad are included in the reactor design that constitutes the reactor vessel surveillance program. The capsules are located at azimuthal angles of 58.50, 610, 121.50, 238.50, 2410, and 301.50 as shown in Figure 4-1. These full-core positions correspond to the following octant symmetric capsule locations represented in Figure 6-1: 290 from the core cardinal axis (for the 61' and 2410 dual surveillance capsule holder locations) and 31.5' from the core cardinal axes (for the 121.50 and 301.5' single surveillance capsule holder locations, and for the 58.5' and 238.50 dual surveillance capsule holder locations). The stainless steel specimen containers are 1. 182-inch by I-inch and are approximately 56 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central 5 fed of the 14-foot high reactor core.

From a neutronic standpoint, the surveillance capsules and associated support structures are significant.

The presence of these materials has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus between the neutron pads and the reactor vessel. In order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model.

In performing the fast neutron exposure evaluations for the South Texas Project Unit 2 reactor vessel and surveillance capsules, a series of fuel cycle specific forward transport calculations were carried out using the following three-dimensional flux synthesis technique:

0(r, 0, z) = 0(r, 0)

• 0(r, z) 0(r) where 4(rO,z) is the synthesized three-dimensional neutron flux distribution, ý(rO) is the transport solution in rO geometry, ý(rz) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and 4r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the r,0 two-dimensional calculation. This synthesis procedure was carried out for each operating cycle at South Texas Project Unit 2.

For the South Texas Project Unit 2 transport calculations, the rO models depicted in Figure 6-1 were utilized since, with the'exception of the neutron pads, the reactor is octant symmetric. These rO models include the core, the reactor internals, the neutron pads - including explicit representations of octants not containing surveillance capsules and octants with surveillance capsules at 290 and 31.50, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. These models formed the basis for the calculated results and enabled making comparisons to the surveillance capsule dosimetry evaluations. In developing these analytical models, nominal design dimensions were employed for the various structural components. Likewise, water temperatures, and hence, coolant densities in the reactor core and downeomer regions of the reactor were taken to be representative of full power operating conditions. The coolant densities were treated on a fuel cycle specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, et cetera. The geometric mesh description of the r,0 reactor models consisted of 183 radial by 99 azimuthal intervals. Mesh sizes were Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-3 description of the r,O reactor models consisted of 183 radial by 99 azimuthal intervals. Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the r,O calculations was set at a value of 0.001.

The r,z models used for the South Texas Project Unit 2 calculations are shown in Figure 6-2 and extend radially from the centerline of the reactor core out to a location interior to the primary biological shield and over an axial span from an elevation 3-feet below the active fuel to approximately 5-feet above the active fuel. (Note that the only difference between these r,z models, and the corresponding r models, is the inclusion or exclusion of the neutron pads. R,Z / R synthesis factors with the neutron pad were used for the capsule locations and the 450 vessel location, whereas the corresponding synthesis factors without the neutron pad were used for the 0', 150, and 300 azimuthal locations of the pressure vessel.) As in the case of the rO models, nominal design dimensions and full power coolant densities were employed in the calculations. In this case, the homogenous core region was treated as an equivalent cylinder with a volume equal to that of the active core zone. The stainless steel former plates located between the core baffle and core barrel regions were also explicitly included in the model. The r,z geometric mesh description of these reactor models consisted of 137 radial by 191 axial intervals. As in the case of the r,0 calculations, mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the r,z calculations was also set at a value of 0.001.

The one-dimensional radial models used in the synthesis procedure consisted of the same 137 radial mesh intervals included in the r,z models. Thus, radial synthesis factors could be determined on a meshwise basis throughout the entire geometry.

The core power distributions used in the plant specific transport analysis were taken from the appropriate South Texas Project Unit 2 fuel cycle design reports. The data extracted from the design reports represented cycle dependent fuel assembly enrichments, bumups, and axial power distributions. This information was used to develop spatial and energy dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle averaged neutron flux, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of individual fuel assemblies.

From these assembly dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

All of the transport calculations supporting this analysis were carried out using the DORT discrete ordinates code Version 3.1 t221 and the BUGLE-96 cross-section library.t231 The BUGLE-96 library provides a 67 group coupled neutron-gamma ray cross-section data set produced specifically for light water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P5 legendre expansion and angular discretization was modeled with an S16 order of angular quadrature.

Energy and space dependent core power distributions, as well as system operating temperatures, were treated on a fuel cycle specific basis.

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2

.July 2007

6-4 Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-7. In Table 6-1, the calculated exposure rates and integrated exposures, expressed in terms of both neutron fluence (E > 1.0 MeV) and dpa, are given at the radial and azimuthal center of the octant symmetric surveillance capsule positions, i.e., for the 290 dual capsule, 31.5' dual capsule, and 31.5' single capsule. These results, representative of the axial mnidplane of the active core, establish the calculated exposure of the surveillance capsules withdrawn to date as well as projected into the future. Similar information is provided in Tables 6-2 and 6-3 for the reactor vessel inner radius. The vessel data given in Table 6-2 are representative of the axial location of the maximum neutron exposure at each of the four azimuthal locations. Maximum integrated neutron exposure results for key vessel plate and weld materials are subsequently given in Table 6-3 for the end of Cycle 9 and beyond. It is also important to note that the data for the vessel inner radius were taken at the clad/base metal interface, and thus, represent the maximum calculated exposure levels of the vessel plates and welds.

Both calculated fluence (E > 1.0 MeV) and dpa data are provided in Table 6-1 through Table 6-3. These data tabulations include both plant and fuel cycle specific calculated neutron exposures at the end of the ninth operating fuel cycle as well as projections for the current operating fuel cycle, i.e., Cycle 10, and future projections to 16, 32, 48, and 54 effective full power years (EFPY). The projections were based on the assumption that the core power distributions and associated plant operating characteristics from Cycle 10 were representative of future plant operation. The future projections are also based on the current reactor power level of 3853 MWt.

Radial gradient information applicable to fast (E > 1.0 MeV) neutron fluence and dpa are given in Tables 6-4 and 6-5, respectively. The data, based on the cumulative integrated exposures from Cycles I through 10, are presented on a relative basis for each exposure parameter at several azimuthal locations.

Exposure distributions through the vessel wall may be obtained by multiplying the calculated exposure at the vessel inner radius by the gradient data listed in Tables 6-4 and 6-5.

The calculated fast neutron exposures for the three surveillance capsules withdrawn from the South Texas Project Unit 2 reactor are provided in Table 6-6. These assigned neutron exposure levels are based on the plant and fuel cycle specific neutron transport calculations performed for the South Texas Project Unit 2 reactor.

Updated lead factors for the South Texas Project Unit 2 surveillance capsules are provided in Table 6-7.

The capsule lead factor is defined as the ratio of the calculated fluence (E > 1.0 MeV) at the geometric center of the surveillance capsule to the corresponding maximum calculated fluence at thepressure vessel clad/base metal interface. In Table 6-7, the lead factors for capsules that have been withdrawn from the reactor (V, Y, and U) were based on the calculated fluence values for the irradiation period corresponding to the time of withdrawal for the individual capsules. For the capsules remaining in the reactor (W, X, and Z), the lead factors correspond to the calculated fluence values at the end of Cycle 10, the current operating fuel cycle for South Texas Project Unit 2.

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP; Rev. 2 July 2007

6-5 6.3 NEUTRON DOSIMETRY The validity of the calculated neutron exposures previously reported in Section 6.2 is demonstrated by a direct comparison against the measured sensor reaction rates and via a least squares evaluation performed for each of the capsule dosimetry sets. However, since the neutron dosimetry measurement data merely serves to validate the calculated results, only the direct comparison of measured-to-calculated results for the most recent surveillance capsule removed from service is provided in this section of the report. For completeness, the assessment of all measured dosimetry removed to date, based on both direct and least squares evaluation comparisons, is documented in Appendix A.

The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsule U, that was withdrawn from South Texas Project Unit 2 at the end of the ninth fuel cycle, is summarized below.

ReactionRates art'~

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____a____e__

R__t 63Cu(n,CE)WCo 4.32E-17 3.98E-17 1.09 "4Fe(n,p)-Mn 4.30E-15 4.40E-15 0.98 58Ni(n,p)58Co 6.13E-15 6.17E-15 0.99 238U(n,p)137Cs (Cd) 2.77E-14 2.36E-14 1.17 237Np(n,f)' 37Cs (Cd) 2.49E-13 2.30E-13 1.08 Average:

1.06

% Standard Deviation:

7.5 The measured-to-calculated (M/C) reaction rate ratios for the Capsule U threshold reactions range from 0.98 to 1.17, and the average M/C ratio is 1.06+/- 7.5% (I(Y). This direct comparison falls well within the

+/- 20% criterion specified in Regulatory Guide 1.190; furthermore, it is consistent with the full set of comparisons given in Appendix A for all measured dosimetry removed to date from the South Texas Project Unit 2 reactor. These comparisons validate the current analytical results described in Section 6.2; therefore, the calculations are deemed applicable for South Texas Project Unit 2.

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-6 6.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the South Texas Project Unit 2 surveillance capsule and reactor pressure vessel is based on the recommended approach provided in Regulatory Guide 1. 190. In particular, the qualification of the methodology was carried out in the following four stages:

I -

Comparison of calculations with benchmark measurements from the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL).

2 -

Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H. B. Robinson power reactor benchmark experiment.

3 -

An analytical sensitivity study addressing the uncertainty components resulting from important input parameters applicable to the plant specific transport calculations used in the neutron exposure assessments.

4 -

Comparisons of the plant specific calculations with all available dosimetry results from the South Texas Project Unit 2 surveillance program.

The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operational or geometric variables that impact power reactor calculations. The second phase of the qualification (H. B. Robinson comparisons) addressed uncertainties in these additional areas that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations. The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant specific input parameters. The overall calculational uncertainty applicable to the South Texas Project Unit 2 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with South Texas Project Unit 2 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule and pressure vessel neutron exposures previously described in Section 6.2. As such, the validation of the South Texas Project Unit 2 analytical model based on the measured plant dosimetry is completely described in Appendix A.

The following summarizes.the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference 21.

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-7

_5' ý'V

ý A-1-31--sNle-PCA Comparisons 3%

3%

H. B. Robinson Comparisons 3%

3%

Analytical Sensitivity Studies 10%

11%

Additional Uncertainty for Factors not Explicitly Evaluated 5%

5%

Net Calculational Uncertainty 12%

13%

The net calculational uncertainty was determined by combining the individual components in quadrature.

Therefore, the resultant uncertainty was treated as random and no systematic bias was applied to the analytical results.

The plant specific measurement comparisons described in Appendix A support these uncertainty assessments for South Texas Project Unit 2.

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-8 Figure 6-1 South Texas Project Unit 2 r,0 Reactor Geometry with a 12.5' Neutron Pad at the Core Midplane E

R Axis (cm)

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-9 Figure 6-1 (continued)

South Texas Project Unit 2 rO Reactor Geometry with a 20.0' Neutron Pad at the Core Midplane R Axis (cm)

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-10 Figure 6-1 (continued)

South Texas Project Unit 2 rO Reactor Geometry with a 22.5' Neutron Pad at the Core Midplane r:

R Axis (cm)

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-11 Figure 6-2 South Texas Project Unit 2 r,z Reactor Geometry with Neutron Pad 0)-

(-'I 0)r L

X'j-LI QD

0) [

"I I I.

0 50 100 150.

200 250 300 R Axis (cm)

Note: This model was used in The assessment of the surveillance capsule neutron exposures and reactor vessel neutron exposures at the 450 azimuthal location only.

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-12 Figure 6-2 (continued)

South Texas Project Unit 2 r,z Reactor Geometry without Neutron Pad E -

0)

(-

il'-,

100 150 200 250 300 R Axis (cm)

Note: This model was used in the assessment of the reactor vessel neutron exposures at the 00, 15'. and 300 azimuthal locations only.

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-13 Table 6-1 Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Neutrons (E > 1.0 MeV)

Cumulative Cumulative Neutron Flux (E > 1.0 MeV)

Cycle Irradiation Irradiation

[n/cm2-s]

Length Time Time Dual Dual Single Cycle

[EFPS]

[EFPS]

IEFPY]

290 31.50 31.50 2.73E+07 2.73E+07 0.87 8.58E+ 10 9.30E+10 9.20E+ 10 2

2.25E+07 4.98E+07 1.58 6.49E+10 6.84E+10 6.76E+ i0 3

3.28E+07 8.26E+07 2.62 7.48E+10 7.93E+!10 7.83E+10 4

4.13E+07 1.24E+08 3.92 7.91E+10 8.51E+10 8.42E+10 5

3.80E+07 1.62E+08 5.13 6.83E+10 7.35E+10 7.27E+10 6

4.89E+07 2.1]E+08 6.68 5.35E+10 5.63E+10 5.56E+10 7

2.95E+07 2.40E+08 7.61 6.97E+10 8.18E+!10 8.11 E+10 8

3.96E+07 2.80E+08 8.87 6.56E+10 7.20E+ 10 7.13E+ 10 9

4.55E+07 3.25E+08 10.31 6.30E+10 6.71E+10 6.63E+10 I O(Prj.)

4.19E+07 3.67E+08 11.64 6.79E+10 7.31E+10 7.23E+ 10 Future 1.38E+08 5.05E+08 16.00 6.79E+ 10 7.31 E+10 7.23E+10 Future 5.05E+08 1.01 E+09 32.00 6.79E+10 7.31 E+ 10 7.23E+ 10 Future 5.05E+08 1.51 E+09 48.00 6.79E+10 7.31E3+10 7.23E+ 10 Future 1.89E+08 1.70E+09 54.00 6.79E+ 10 7.31E+10 7.23E+ 10 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-14 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Neutrons (E > 1.0 MeV)

Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation

[n/cm2]

Length Time Time Dual Dual Single Cycle

[EFPS]

[EFPS]

[EFPY]

290 31.50 31.50 I

2.73E+07 2.73E+07 0.87 2.34E+18 2.54E+18 2.51E+18 2

2.25E+07 4.98E+07 1.58 3.80E+ 18 4.08E+ 18 4.03E+ 18 3

3.28E+07 8.26E+07 2.62 6.25E+18 6.68E+18 6.60E+ 18 4

4.13E+07 1.24E+08 3.92 9.52E+ 18 1.02E+ 19 1.01 E+ 19 5

3.80E+07 1.62E+08 5.13 1.21E+19 1.30E+19 1.28E+19 6

4.89E+07 2.II E+08 6.68 1.47E+ 19 1.57E+ 19 1.56E+ 19 7

2.95E+07 2,40E+08 7.61 1.68E+19 1.81E+19 1.79E+19 8

3.96E+07 2.80E+08 8.87 1.94E+ 19 2.10E+19 2.08E+19 9

4.55E+07 3.25E+08 10.31 2.23E+ 19 2.40E+ 19 2.38E+ 19 10(Prj.)

4.19E+07 3.67E+08 11.64 2.51E+19 2.71E+19 2.68E+19 Future 1.38E+08 5.05E+08 16.00 3.44E+ 19 3.72E+ 19 3.68E+ 19 Future 5.05E+08 1.01E+09 32.00 6.87E+19 7.41E+19 7.33E+19 Future 5.05E+08 1.51E+09 48.00 1.03E+20

1. 1 IE+20

1. 1 0E+20 Future 1.89E+08 1.70E+09 54.00

1. 16E+20 1.25E+20 I.24E+20 Nole: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-15 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Iron Atom Displacements Cumulative Cumulative Displacement Rate Cycle Irradiation Irradiation

[dpa/sl Length Time Time Dual Dual Single Cycle

[EFPS]

IEFPSJ

[EFPY]

290 31.50 31-50 1

2.73E+07 2.73E+07 0.87 1.67E-10 1.81E-10 1.79E-10 2

2.25E+07 4.98E+07 1.58 1.26E-10 1.33E-10 1.31E-10 3

3.28E+07 8.26E+07 2.62 1.45E-10 1.54E-10 1.51E-10 4

4.13E+07 1.24E+08 3.92 1.54E-10 1.65E-10 1.63E-10 5

3.80E+07 1.62E+08 5.13 1.32E-10 1.42E-10 1.41E-10 6

4.89E+07 2.1I1 E+08 6.68 1.03E-10 1.09E-10 1.07E-10 7

2.95E+07 2.40E+08 7.61 1.35E-10 1.59E-10 1.57E-10 8

3.96E+07 2.80E+08 8.87 1.27E-10 1.39E-10 1-37E-10 9

4.55E+07 3.25E+08 10.31 1.22E-10 1.29E-10 1.28E-10 10(Prj.)

4.19E+07 3.67E+08 11.64 1.31E-10 I.41E-10 1.39E-10 Future 1.38E+08 5.05E+08 16.00 1.31E-10 1.41E-10 1.39E-10 Future 5.05E+08 1.0 1 E+09 32.00 1.311E-10 1.41 E-10 1.39E-10 Future 5.05E+08 1.51E+09 48.00 1.31E-10 1.41E-10 1.39E-10 Future 1.89E+08 1.70E+09 54.00 1.31E-10 1.41E-10 1.39E-10 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-16 Table 6-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Iron Atom Displacements Cumulative Cumulative Displacements Cycle Irradiation Irradiation Idpal Length Time Time Dual Dual Single Cycle

[EFPS]

[EFPS]

[EFPY]

290 31.50 31.50 1

2.73E+07 2.73E+07 0.87 4.57E-03 4.95E-03 4.89E-03 2

2.25E+07 4.98E+07 1.58 7.4 1E-03 7.94E-03 7.83E-03 3

3.28E+07 8.26E+07 2.62 1.22E-02 1.30E-02 1.28E-02 4

4.13E+07 1.24E+08 3.92 1.85E-02 1.98E-02 1.95E-02 5

3.80E+07 1.62E+08 5.13 2.35E-02 2.52E-02 2.49E-02 6

4.89E+07 2.11 E+08 6.68 2.86E-02 3.05E-02 3.01 E-02 7

2.95E+07 2.40E+08 7.61 3.26E-02 3.52E-02 3.47E-02 8

3.96E+07 2.80E+08 8.87 3.76E-02 4.07E-02 4.02E-02 9

4.55E+07 3.25E+08 10.31 4.32E-02 4.66E-02 4.60E-02 I0(Prj.)

4.19E+07 3.67E+08 11.64 4.87E-02 5.25E-02 5.18E-02 Future 1.38E+08 5.05E+08 16.00 6.68E-02 7.20E-02

7. I OE-02 Future 5.05E+08 1.01E+09 32.00 1.33E-01 1.43E-01 1.41E-0I Future 5.05E+08 1.51E+09 48.00 1.99E-01 2.15E-01 2.12E-01 Future 1.89E+08 1.70E+09 54.00 2.24E-01 2.41E-OI 2-38E-01 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-17 Table 6-2 Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Neutron Flux (E > 1.0 MeV)

Cycle Irradiation Irradiation

[n/cm 2-s]

Length Time Time Cycle

[EFPS]

[EFPS]

[EFPY]

00 150 300 450 1

2.73E+07 2.73E+07 0.87 1.26E+10 1.87E+10 2.14E+ 10 2.77E+ 10 2

2.25E+07 4.98E+07 1.58 1.26E+10 1.75E+10 1.71E+10 2.17E+I0 3

3.28E+07 8.26E+07 2.62 1.24E+10 1.81E+10 1.85E+i0 2.21E+10 4

4.13E+07 1.24E+08 3.92 1.23E+ 10 1.79E+ 10 1.94E+ 10 2.62E+10 5

3.80E+07 1.62E+08 5.13 1.03E+10 1.50E+ 10 1.68E+1i0 2.24E+ 10 6

4.89E+07 2.11E+08 6.68 9.00E+09 1.28E+10 1.34E+10 1.85E+10 7

2.95E+07 2.40E+08 7.61 7.64E+09 1.16E+ 10 1.74E+ I0 3.02E+ 10 8

3.96E+07 2.80E+08 8-87 8.29E+09 1.26E+10 1.64E+10 2.36E+ 10 9

4.55E+07 3-25E+08 10.31 8.22E+09 1.24E+10 1.58E+I0 1.95E+10 10(Prj.)

4.19E+07 3.67E+08 11.64 1.01E+10 1.50E+I0 1.72E+i10 2.16E+-10 Future 1.38E+08 5.05E+08 16.00

!.011E+10 1.50E+10 1.72E+ 10 2.16E+ 10 Future 5.05E+08 1.01E+09 32.00 1.01E+10 1.50E+I0 1.72E+10 2.16E+10 Future 5.05E+08 1.51E+09 48.00 1.01E+10 1.50E+l0 1.72E+10 2.16E+10 Future 1.89E+08 1.70E+09 54.00 1.01E+10 i.50E+ 10 1.72E+10 2.16E+ 10 Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-18 Table 6-2 cont'd Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Neutron Fluence (E > 1.0 MeV)

Cycle Irradiation Irradiation

[n/cm2]

Length Time Time Cycle IEFPS]

[EFPS]

[EFPY]

00 150 300 450 1

2.73E+07 2.73E+07 0.87 3.43E+17 5.12E+17 5.85E+17 7.58E+17 2

2.25E+07 4.98E+07 1.58 6.04E+17 8.74E+ 17 9.39E+ 17 1.24E+18 3

3.28E+07 8.26E+07 2.62 1.011E+18 1.47E+ 18 1.54E+ 18 1.97E+ 18 4

4.13E+07 1.24E+08 3.92 1.52E+18 2.20E+ 18 2.34E+ 18 3.05E+ 18 5

3.80E+07 1.62E+08 5.13 1.91 E+18 2.77E+18 2.98E+18 3.90E+ 18 6

4.89E+07 2.11 E+08 6.68 2.34E+18 3.38E+18 3.62E+1 8 4.81E+18 7

2.95E+07 2.40E+08 7.61 2.56E+ 18 3.73E+ 18 4.13E+ 18 5.70E+ 18 8

3.96E+07 2.80E+08 8.87 2.89E+ 18 4.22E+ 18 4.78E+ 18 6.63E+ 18 9

4-55E+07 3.25E+08 10.31 3.26E+18 4.78E+18 5.49E+ 18 7.52E+18 10(Prj.)

4.19E+07 3.67E+08 11.64 3.68E+18 5.41 E+18 6.21E+18 8.42E+118 Future 1.38E+08 5.05E+08 16.00 5.07E+.18 7.48E+18 8.58E+ 18

1. 14E+ 19 Future 5.05E+08 1.01 E+09 32.00 1.02E+ 19 1.51 E+19 1.73E+19 2.22E+ 19 Future 5.05E+08 1.51 E+09 48.00 1.52E+19 2.26E+19 2.59E+19 3.31E+19 Future 1.89E+08 1.70E+09 54.00 1.7] E+19 2.55E+19 2.92E+ 19 3.72E+ 19 Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-19 Table 6-2 cont'd Calculated Azimuthal Variation Of Fast Neutron Exposure Rates And Iron Atom Displacement Rates At The Reactor Vessel CladlBase Metal Interface Cumulative Cumulative Iron Atom Displacement Rate Cycle Irradiation Irradiation

[dpa/s)

Length Time Time Cycle

[EFPS]

[EFPS]

[EFPY]

00 150 300 450 1

2.73E+07 2.73E+07 0.87 1.95E-I I 2.88E-I I 3.30E-1 4.27E-11 2

2.25E+07 4.98E+07 1.58 1.96E-I1 2.69E-I1 2.64E-11 3.33E-11 3

3.28E+07 8.26E+07 2.62 1.92E-11 2.78E-I I 2.85E-I 1 3.40E-11 4

4.13E+07 1.24E+08 3.92 1.91E-II 2.76E-I 1 2.99E-11 4.04E-11 5

3.80E+07 1.62E+08 5.13 1.60E-I I 2.30E-I I 2.59E-1 I 3.45E-11 6

4.89E+07 2.11 E+08 6.68 1.40E-I 1.97E-I I 2.07E-Il 2.85E-I1 7

2.95E+07 2.40E+08 7.61

1. 19E-II 1.79E-11 2.69E-Il 4.64E-I1 8

3.96E+07 2.80E+08 8.87 1.29E-1 !

1.94E-I l 2.53E-1I 3.63E-11 9

4.55E+07 3.25E+08 10.31 1.28E1-I 1.91E-II 2.43E1-I1 3.01E-11 I 0(Prj.)

4.19E+07 3.67E+08 11.64 1.56E-i !

2.3 1E-I I 2.65E-11 3.32E-I1 Future 1.38E+08 5.05E+08 16.00 1.56E-11 2.31E-II 2.65E-11 3.32E-1I Future 5.05E+08 1.01E+09 32.00 1.56E-1l 2.31 E-I I 2.65E-11 3.32E-11 Future 5.05E+08 1.51E+09 48.00 1.56E-11 2.311E-iI 2.65E-1I 3.32E-I 1 Future 1.89E+08 1.70E+09 54.00 1.56E-11 2.311E-I1 2.65E-II 3.32E-II Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-20 Table 6-2 cont'd Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Cumulative Cumulative Iron Atom Displacements Cycle Irradiation Irradiation

[dpa]

Length Time Time Cycle

[EFPS]

IEFPS]

[EFPY]

0o 050 300 450 I

2.73E+07 2.73E+07 0.87 5.32E-04 7.86E-04 9.01E-04 1.17E-03 2

2.25E+07 4.98E+07 1.58 9.37E-04 1.34E-03 1-45E-03 1.91E-03 3

3.28E+07 8.26E+07 2.62 1.57E-03 2.25E-03 2.38E-03 3.03E-03 4

4.13E+07 1.24E+08 3.92 2.35E-03 3.38E-03 3.61E-03 4.69E-03 5

3-80E+07 1.62E+08 5.13 2.96E-03 4.25E-03 4.59E-03 6.OOE-03 6

4.89E+07 2.1 IE+08 6.68 3.63E-03 5.20E-03.

5.58E-03 7.39E-03 7

2.95E+07 2.40E+08 7-61 3.98E-03 5.73E-03 6.37E-03 8.76E-03 8

3.96E+07 2.80E+08 8.87 4.49E-03 6.49E-03 7.37E-03 1.02E-02 9

4.55E+07 3.25E+08 10.31 5.07E-03 7.35E-03 8.47E-03 1.16E-02 10(Prj.)

4.1 9E+07 3.67E+08 11.64 5.72E-03 8-32E-03 9.58E-03 1.30E-02 Future 1.38E+08 5.05E+08 16.00 7.87E-03

1. 15E-02 1.32E-02 1.75E-02 Future 5.05E+08 I.OIE+09 32.00 1.58E-02 2.32E-02 2.66E-02 3.43E-02 Future 5.05E+08 1.51 E+09 48.00 2.37E-02 3.48E-02 4.OOE-02

5. I OE-02 Future 1.89E+08 1.70E+09 54.00 2.66E-02 3.92E-02 4.50E-02 5.73E-02 Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-21 Table 6-3 Calculated Integrated Exposures for Key Vessel Plate and Weld Materials At The Reactor Vessel Clad/Base Metal Interface Neutron Fluence (E > 1.0 MeV), n/cm 2 Material Location 10.31 EFPY 11.64 EFPY 16 EFPY 32 EFPY 48 EFPY 54 EFPY Upper-to-Intermediate Shell Circumferential Weld Intermediate Shell Basemetal Intermediate Shell Longitudinal Weld 0° Azimuth Intermediate-to-Lower Shell Circumferential Weld Lower Shell Basemetal Lower Shell Longitudinal Weld 900 Azimuth Lower Shell Longitudinal Weld 2100 and 3300 Azimuth 12.5' Neutron Pad Lower Shell-to-Lower Head Circumferential Weld 450 8.03E+16 8.78E+16 1.13E+17 2.03E+ 17 2.94E+ 17 3.28E+ 17 450 5.56E+18 6.27E+18 8.57E+18 1.70E+19 2.55E+19 2.86E+19 00 3.20E+18 3.61E+18 4.97E+18 9-95E+18 1.49E+19 1.68E+19 450 5.54E+18 6.24E+18 8.52E+18 1.69E+19 2.52E+19 2.84E+19 450 7.52E+18 8.42E+18 1.14E+19 2.22E+19 3.31E+19 3.72E+19 00 3.26E+18 3.68E+18 5.07E+18 1-02E+ 19 1.52E+19 1.71E+19 3C0 7.30E+18 8.21E+18 1.12E+19 2.22E+19 3.32E+19 3.73E+19 450 8.47E+16 9.26E+16 1.18E+17 2.13E+17 3.08E+17 3.43E+17 Notes:

(1)

The Intermediate Shell Longitudinal Weld, 120' Azimuth, 200 Neutron Pad results at the 300 location are bounded by the Intermediate Shell Basermetal values at the 450 location.

(2)

The Intermediate Shell Longitudinal Weld, 2400. Azimuth, 22.50 Neutron Pad results at the 300 location are bounded by the Intermediate Shell Basemetal values at the 45' location.

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-22 Table 6-3 cont'd Calculated Integrated Exposures for KeyVessel Plate and Weld Materials At The Reactor Vessel CladlBase Metal Interface Iron Atom Displacements, dpa Material Location 10.31 EFPY 11.64 EFPY 16 EFPY 32 EFPY 48EFPY 54 EFPY Upper-to-Intermediale Shell 450 1.39E-04 1.53E-04 1.96E-04 3.55E-04 5.14E-04 Circumferential Weld Intermediate Shell 450 8.80E-03 9.91E-03 1.36E-02 2.69E-02 4.03E-02 Basemetal Intermediate Shell 00 4.96E-03 5.61E-03 7.72E-03 1.55E-02 2.32E-02 Longitudinal Weld 00 Azimuth Intermediate-to-Lower Shell 450 8.78E-03 9.88E-03 1.35E-02 2.68E-02 4.OOE-02 Circumferential Weld Lower Shell

45) 1.166E-02 1.30E-02 1.75E-02 3.43E-02 5.1 OE-02 Basemetal Lower Shell 00 5.07E-03 5.72E-03 7.87E-03 1.58E-02 2.37E-02 Longitudinal Weld 900 Azimuth Lower Shell 300 1.OE-02 1.23E-02 1.68E-02 3.34E-02 4.99E-02 Longitudinal Weld 2100 and 3300 Azimuth 12.50 Neutron Pad Lower Shell-to-Lower Head 450 1.511E-04 1.65E-04 2.11E-04 3.82E-04 5.52E-04 Circumferential Weld Notes:

(3)

The Intermediate Shell Longitudinal Weld, 120' Azimuth, 200 Neutron Pad results at the 30' location are bounded by the Intermediate Shell Basemetal values at the 45' location.

(4)

The Intermediate Shell Longitudinal Weld, 2400 Azimuth, 22.50 Neutron Pad results at the 30' location are bounded by the Intermediate Shell Basemetal values at the 450 location.

5.73E-04 4.53E-02 2.61 E-02 4.50E-02 5.73E-02 2.66E-02 5.61 E-02 6.16E-04 Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-23 Table 6-4 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV)

Within The Reactor Vessel Wall RADIUS AZIMUTHAL ANGLE (cm) 00 150 300 450 220.350 1.000 1.000 1.000 1.000 225.868 0.563 0.559 0.553 0.550 231.385 0.277 0.272 0.268 0.264 236.903 0.131 0.127 0.125 0.122 242.420 0.063 0.058 0.057 0.055 Note:

Base Metal Inner Radius = 220.350 cm Base Metal 1/4T

= 225.868 cm Base Metal 1/2T

= 231.385 cm Base Metal 3/4T

= 236.903 cm Base Metal Outer Radius = 242.420 cm Table 6-5 Relative Radial Distribution Of Iron Atom Displacements'(dpa)

Within The Reactor Vessel Wall RADIUS AZIMUTHAL ANGLE (cm) 00 150 300 450 220.350 1.000 1.000 1.000 1.000 225.868 0.638 0.633 0.632 0.642 231.385 0.388 0.380 0.381 0.392 236.903 0.236 0.226 0.229 0.236 242.420 0.142 0.129 0.130 0.134 Note:

Base Metal Inner Radius = 220.350 cm Base Metal 1/4T

= 225.868 cm Base Metal 1/2T

= 231.385 cm Base Metal 3/4T

= 236.903 cm Base Metal Outer Radius = 242.420 cm Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

6-24 Table 6-6 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from South Texas Project Unit 2 Irradiation Time Fluence (E > 1.0 MeV)

Iron Displacements Capsule

[EFPYi

[n/cm2]

[dpa]

V 0.87 2.34E+ 18 4.57E-03 Y

5.13 1.21 E+ 19 2.35E-02 U

10.31 j

2.40E+ 19 4.66E-02 Table 6-7 Calculated Surveillance Capsule Lead Factors Capsule ID And Location Status Lead Factor V (290 dual)

Withdrawn EOC 1 3.09 Y (29' dual)

Withdrawn EOC 5 3.11 U (31.50 dual)

Withdrawn EOC 9 3.20 X (31.5' dual)

In Reactor 3.22 W (31.5' single)

In Reactor 3.19 Z (31.50 single)

In Reactor 3.19 Note: Lead factors for capsules remaining in the reactor are based on cycle specific exposure calculations through the current operating fuel reload, i.e., Cycle 10.

Radiation Analysis and Neutron Dosimetry WCAP-16093-NP, Rev. 2 July 2007

7-I 7

SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following surveillance capsule removal schedule meets the requirements of ASTM E185-82 and is recommended for future capsules to be removed from the South Texas Unit 2 reactor vessel. This recommended removal schedule is applicable to 32 EFPY of operation.

iai71gi xq mne ux.

ýfc

~u~~jja 7a 4

~

iffi dkfl N

d N

• ~~~

S;2~ jr.

U

~

Pz*WkRNAb,

."p1a,~~t I

-a ýSm CM, I

a CAJ.......... F1.8 mpulI*....

I, o

V 610 3.09 0.87 2.34 x 10" (c)

Y 2410 3.11 5.13 1.21 x 10'9 (c)

U 58.50 3.20 10.31 2.40 x 10'9 (c)

X 238.50 3.22 Standby (d)

W 121.50 3.19 Standby (d)

Z 301.50 3.19 Standby (d)

Notes:

(a) Updated in Capsule u dosimetry analysis.

(b) Effective Full Power Years (EFPY) from plant startup.

(c) Plant specific evaluation.

(d) Section Xl.M3 1, "Reactor Vessel Surveillance," of NUREG-1801 states that any surveillance capsules that are left 'n the reactor vessel should provide meaningful metallurgical data. The NRC specifically states that anything beyond 60 years of exposure is not meaningful metallurgical data. Hence, it is recommended that Capsule "X" be removed and tested at 16 EFPY (3.65 x 10'9 n/cm2, E > 1.0 MeV, i.e., the peak 53 EFPY fluence). Capsules "'W" and "Z" should be removed and placed in storage at 16 EFPY Surveillance Capsule Removal Schedule WCAP-16093-NP, Rev. 2 July 2007

8-1 8

REFERENCES I. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May, 1988.

2.

Code of Federal Regulations, 10CFR50, Appendix G Fracture Toughness Requirements, and Appendix H, Reactor Vessel Material Surveillance Program Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C.

3.

WCAP-14978, Analysis of Capsule Yfrom the Houston Lighting and Power Company South Texas Unit 2 Reactor Vessel Radiation Surveillance Program, E. Terek, et. al., dated December 1997.

4.

WCAP-9967, Houston Lighting and Power Company South Texas Project Unit No. 2 Reactor Vessel Radiation Surveillance Program, L.R. Singer, dated January 1982.

5-WCAP-13182, Analysis of Capsule Vfrom the Houston Lighting and Power Company South Texas Unit 2 Reactor Vessel Radiation Surveillance Program, J.M. Chicots, et. al., dated February 1992.

6.

ASTM E208, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.

7.

Section Xl of the ASME Boiler and Pressure Vessel Code, Appendix, Fracture Toughness Criteria for Protection Against Failure

8.

ASTM E 185-82, Standard Practice for Conducting Surveillance Testsfor Light-Water Cooled Nuclear Power Reactor Vessels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA.

9.

Procedure RMF 8402, Surveillance Capsule Testing Program, Revision 2.

10. Procedure RMF 8102, Tensile Testing, Revision i.

11. Procedure RMF 8103, Charpy Impact Testing, Revision 1.

12. ASTM E23-98, Standard Test Method for Notched Bar Impact Testing of Metallic Materials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1998.

13. ASTM A370-97a, Standard Test Methods and Definitions for Mechanical Testing of Steel Products, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1997.

14. ASTM E8-99, Standard Test Methodsfor Tension Testing of Metallic Materials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1999.

References WCAP-16093-NP, Rev. 2 July 2007

8-2

15. ASTM E21-92 (1998), Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1998.

16. ASTM E83-93, Standard Practice for Verification.and Classification of Extensometers, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.

17. ASTM E 185-79, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels

18. WCAP-14370, Use of the Hyperbolic Tangent Function for Fitting Transition Temperature Toughness Data, T. R. Mager, et al, May 1995.

19. Regulatory Guide RG-1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

20. WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, January 1996.

21. WCAP-15557, Revision 0, Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology, August 2000.

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

23. RSIC Data Library Collection DLC-1 85, "BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDFIB-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," March 1996.

References WCAP-16093-NP, Rev. 2 July 2007

A-0 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-I A.I Neutron Dosimetry Comparisons of measured dosimetry results to both the calculated and least squares adjusted values for all surveillance capsules withdrawn from service to date at South Texas Project Unit 2 are described herein. The sensor sets from these capsules have been analyzed in accordance with the current dosimetry evaluation methodology described in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.IA-1 One of the main purposes for presenting this material is to demonstrate that the overall measurements agree with the calculated and least squares adjusted values to within +/- 20% as specified by Regulatory Guide 1.190, thus serving to validate the calculated neutron exposures previously reported in Section 6.2 of this report. This information may also be useful in the future, in particular, as least squares adjustment techniques become accepted in the regulatory environment.

A.I.1 Sensor Reaction Rate Determinations In this section, the results of the evaluations of the three neutron sensor sets withdrawn to date as part of the South Texas Project Unit 2 Reactor Vessel Materials Surveillance Program are presented. The capsule designation, location within the reactor, and time of withdrawal of each of these dosimetry sets were as follows:

z-C.apsule 1D*,. !**iAz

'imuhl

.i[ii!*. & *WidrawalI:***:'"

..,e**

,Irradmt~ionA,**(:;.:-

-Locaio m'

k.'; ~m[F

.. ~4 V

29' dual End of Cycle 1 0.87 Y

29" dual End of Cycle 5 5.13 U

31.5* dual End of Cycle 9 10.31 The azimuthal locations included in the above tabulation represent the first octant equivalent azimuthal angle of the geometric center of the respective surveillance capsules.

Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-2 The passive neutron sensors included in the evaluations of Survillance Capsules V, Y, and U are summarized as follows:

• The cobalt-aluminum measurements for this plant include both bare wire and cadmium-covered sensors.

Since all of the dosimetry monitors were accommodated within the dosimeter block centered at the radial, azimuthal, and axial center of the material test specimen array, gradient corrections were not required for these reaction rates. Pertinent physical and nuclear characteristics of the passive neutron sensors are listed in Table A-I.

The use of passive monitors such as those listed above does not yield a direct measure of the energy dependent neutron flux at the point of interest. 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 neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:

the measured specific activity of each monitor, the physical characteristics of each monitor, the operating history of the reactor, the energy response of each monitor, and the neutron energy spectrum at the monitor location.

Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-3 The radiometric counting of the neutron sensors from Capsule V was carried out by the Westinghouse Analytical Services Laboratory at the Waltz Mill Site.A'21 The radiometric counting of the sensors from Capsule Y were performed by the Antech Analytical LaboratorytA 31, and the sensors from Capsule U were counted by Pace Analytical Services, Inc., also located at the Waltz Mill Site. In all cases, the radiometric counting followed established ASTM procedures. Following sample preparation and weighing, the specific activity of each sensor was determined by means of a high-resolution gamma spectrometer. For the copper, iron, nickel, and cobalt-aluminum sensors, these analyses were performed by direct counting of each of the individual samples. In the case of the uranium and neptunium fission sensors, the analyses were carried out by direct counting preceded by dissolution and chemical separation of cesium from the sensor material.

The irradiation history of the reactor over the irradiation periods experienced by Capsules V, Y, and U was based on the reported monthly power generation of South Texas Project Unit 2 from initial reactor startup through the end of the dosimetry evaluation period. For the sensor sets utilized in the surveillance capsules, 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. The irradiation history applicable to Capsules V, Y, and U is given in Table A-2.

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

A R=

No FYY Y Cf [1-e

[e-']

where:

R

=

Reaction rate averaged over the irradiation period and referenced to operation at a core power level of Prf (rps/nucleus).

A

=

Measured specific activity (dps/gm).

No

=

Number of target element atoms per gram of sensor.

F

=

Weight fraction of the target isotope in the sensor material.

Y

=

Number of product atoms produced per reaction.

Pj

=

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

P,,=

Maximum or reference power level of the reactor (MW).

=

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

X

=

Decay constant of the product isotope (1/sec).

tj

=

Length of irradiation period j (sec).

t

=

Decay time following irradiation period j (sec).

Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-4 and the summation is carried out over the total number of monthly intervals comprising the irradiation period.

In the equation describing the reaction rate calculation, the ratio [P1]/[Pf accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio Cj, which was calculated for each fuel cycle using the transport methodology discussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in flux level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, C, is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those employing low leakage fuel management, the additional Ci term should be employed. The impact of changing flux levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from non-low leakage to low leakage fuel management or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another.

The fuel cycle specific neutron flux values along with the computed values for Cj are listed in Table A-3.

These flux values represent the cycle dependent results at the radial and azimuthal center of the respective capsules at the axial elevation of the active fuel midplane.

Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 238U measurements to account for the presence of 233U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.

Corrections were also made to the 3U and 237Np sensor reaction rates to account for gamma ray induced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the South Texas Project Unit 2 fission sensor reaction rates are summarized as follows:

.Cot rectio

'5 iep sýi~ U 23U Impurity/Pu Build-in 0.873 0.837 0.796 238U(y,f) 0.969 0.969 0.969 Net 23SU Correction 0.846 0.811 0.771 237Np(,,f) 0.991 0.991 0.991 These factors were applied in a multiplicative fashion to the decay corrected uranium and neptunium fission sensor reaction rates.

Results of the sensor reaction rate determinations for Capsules V, Y, and U are given in Table A-4. In Table A-4, the measured specific activities, decay corrected saturated specific activities, and computed reaction rates for each sensor indexed to the radial center of the capsule are listed. The fission sensor reaction rates are listed both with and without the applied corrections for 2 38U impurities, plutonium build-in, and gamma ray induced fission effects.

Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-5 A.1.2 Least Squares Evaluation of Sensor Sets Least squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations resulting in a Best Estimate neutron energy spectrum with associated uncertainties. Best Estimates for key exposure parameters such as (E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least squares methods, as applied to surveillance capsule dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrum within their respective uncertainties. For example, g

relates a set of measured reaction rates, Ri, to a single neutron spectrum, ýg, through the multigroup dosimeter reaction cross-section, aig, each with an uncertainty S. The primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the least squares evaluation of the South Texas Project Unit 2 surveillance capsule dosimetry, the FERRET code1A41 was employed to combine the results of the plant specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters (4(E > 1.0 MeV) and dpa) along with associated uncertainties for the three in-vessel capsules withdrawn to date.

The application of the least squares methodology requires the following input:

I - The calculated neutron energy spectrum and associated uncertainties at the measurement location.

2 - The measured reaction rates and associated uncertainty for each sensor contained in the multiple foil set.

3 - The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set.

For the South Texas Project Unit 2 application, the calculated neutron spectrum was obtained from the results of plant specific neutron transport calculations described in Section 6.2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section A.I.I. The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section libraryt^S. The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard EO 18, "Application of ASTM Evaluated Cross-Section Data File, Matrix E 706 (11B)".

The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum were input to the least squares procedure in the form of variances and covariances.

The assignment of the input uncertainties followed the guidance provided in ASTM Standard E 944, "Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance."

Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-6 The following provides a summary of the uncertainties associated with the least squares evaluation of the South Texas Project Unit 2 surveillance capsule sensor sets.

Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, irradiation history corrections, and corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM National Consensus Standards for reaction rate determinations for each sensor type.

After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least squares evaluation:

63Cu(n,a)6°Co 5%

54Fe(n,p) 54Mn 5%

5"Ni(np)58Co 5%

238U(n,f) 137Cs 10%

23 7Np(n,f)137Cs 10%

59Co(n,-y)6Co 5%

These uncertainties are given at the I ( level.

Dosimetry Cross-Section Uncertainties The reaction rate cross-sections used in the least squares evaluations were taken from the SNLRML library. This data library provides reaction cross-sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross-sections and uncertainties are provided in a fine multigroup structure for use in least squares adjustment applications. These cross-sections were compiled from the most recent cross-section evaluations and they have been tested with respect to their accuracy and consistency for least squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as in the fluence and energy characterization of 14 MeV neutron sources.

For sensors included in the South Texas Project Unit 2 surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.

Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-7

'2 Reaction~.

W

,i, k i ii,

I 63CU(n'CE)6CO 54Fe(n,p)54Mn

-58Ni(n~pV58CO 238u(n,O']37cS 237 Np(n,O' 137cS 59Co(n,y)'Co 4.08-4.16%

3.05-3.11%

4.49-4.56%

0.54-0.64%

10.32-10.97%

0.79-3.59%

These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in LWR irradiations.

Calculated Neutron Spectrum The neutrori spectra input to the least squares adjustment procedure were obtained directly from the results of plant specific transport calculations for each surveillance capsule irradiation period and location. The spectrum for each capsule was input in an absolute sense (rather than as simply a relative spectral shape). Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements.

While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks and the dosimetry cross-section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:

MF,

=R 2 +R *R.*P.

where R. specifies an overall fractional normalization uncertainty and the fractional uncertainties R. and R8, specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:

Ps"=ll

]S. + 0 eu" 89JJ 1

.9e 1

where (g - g, 27 2y Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-8 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range y (0 specifies the strength of the latter term). The value of 8 is 1.0 when g = g', and is 0.0 otherwise.

The set of parameters defining the input covariance matrix for the South Texas Project Unit 2 calculated spectra was as follows:

Flux Normalization Uncertainty (R,) 15%

Flux Group Uncertainties (Rg, Rg,)

(E > 0.0055 MeV)

(0.68 eV < E < 0.0055 MeV)

(E < 0.68 eV) 15%

29%

52%

Short Range Correlation (0)

(E > 0.0055 MeV)

(0.68 eV < E < 0.0055 MeV)

(E < 0.68 eV) 0.9 0.5 0.5 Flux Group Correlation Range (y)

(E > 0.0055 MeV)

(0.68 eV < E < 0.0055 MeV)

(E < 0.68 eV) 6 3

2 Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-9 A.I.3 Comparisons of Measurements and Calculations Results of the least squares evaluations of the dosimetry from the South Texas Project Unit 2 surveillance capsules withdrawn to date are provided in Tables A-5 and A-6. In Table A-5, measured, calculated, and best-estimate values for sensor reaction rates are given for each capsule. Also provided in this tabulation are ratios of the measured reaction rates to both the calculated and least squares adjusted reaction rates.

These ratios of M/C and M/BE illustrate the consistency of the fit of the calculated neutron energy spectra to the measured reaction rates both before and after adjustment. In Table A-6, comparison of the calculated and best estimate values of neutron flux (E > 1.0 MeV) and iron atom displacement rate are tabulated along with the BE/C ratios observed for each of the capsules.

The data comparisons provided in Tables A-5 and A-6 show that the adjustments to the calculated spectra are relatively small and well within the assigned uncertainties for the calculated spectra, measured sensor reaction rates, and dosimetry reaction cross-sections. Further, these results indicate that the use of the least squares evaluation results in a reduction in the uncertainties associated with the exposure of the surveillance capsules. From Section 6.4 of this report, it may be noted that the uncertainty associated with the unadjusted calculation of neutron fluence (E > 1.0 MeV) and iron atom displacements at the surveillance capsule locations is specified as 12% at the 1y level. From Table A-6, it is noted that the corresponding uncertainties associated with the least squares adjusted exposure parameters have been reduced to 6% for neutron flux (E > 1.0 MeV) and 8% for iron atom displacement rate. Again, the uncertainties from the least squares evaluation are at the I level.

Further comparisons of the measurement results with calculations are given in Tables A-7 and A-8.

These comparisons are given on two levels. In Table A-7, calculations of individual threshold sensor reaction rates are compared directly with the corresponding measurements. These threshold reaction rate comparisons provide a good evaluation of the accuracy of the fast neutron portion of the calculated energy spectra. In Table A-8, calculations of fast neutron exposure rates in terms of (E > 1.0 MeV) and dpa/s are compared with the best estimate results obtained from the least squares evaluation of the capsule dosimetry results. These two levels of comparison yield consistent and similar results with all measurement-to-calculation comparisons falling well within the 20% limits specified as the acceptance criteria in Regulatory Guide 1.190.

In the case of the direct comparison of measured and calculated sensor reaction rates, the M/C comparisons for fast neutron reactions range from 0.90-1.21 for the 15 samples included in the data set.

The overall average M/C ratio for the entire set of South Texas Project Unit 2 data is 1.04 with an associated standard deviation of 10.5%.

In the comparisons of best estimate and calculated fast neutron exposure parameters, the corresponding BE/C comparisons for the capsule data sets range from 0.95-1.06 for neutron flux (E > 1.0 MeV) and from 0.95 to 1.07 for iron atom displacement rate. The overall average BE/C ratios for neutron flux (E > 1.0 MeV) and iron atom displacement rate are 1.02 with a standard deviation of 5.6% and 1.02 with a standard deviation of 6.3%, respectively.

Based on these comparisons, it is concluded that the calculated fast neutron exposures provided in Section 6.2 of this report are validated for use in the assessment of the condition of the materials comprising the beltline region of the South Texas Project Unit 2 reactor pressure vessel.

Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-10 Table A-I Nuclear Parameters Used In The Evaluation Of Neutron Sensors Note: The 90% response range is defined such that, in the neutron spectrum characteristic of the South Texas Project Unit 2 surveillance capsules, approximately 90% of the sensor response is due to neutrons in the energy range specified with approximately 5% of the total response due to neutrons with energies below the lower limit and 5% of the total response due to neutrons with energies above the tipper limit.

Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-Il Table A-2 Monthly Thermal Generation During The First Nine Fuel Cycles Of The South Texas Project Unit 2 Reactor (Reactor Power of 3800 MWt through the end of Cycle 9)

Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-12 Table A-2 cont'd Monthly Thermal Generation During The First Nine Fuel Cycles Of The South Texas Project Unit 2 Reactor (Reactor Power of 3800 MWt through the end of Cycle 9)

Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-13 Table A-3 Calculated C1 Factors at the Surveillance Capsule Center Core Midplane Elevation

_4 Vsl N

~

~

sI

~

~

I 8.58E+10 8.58E+10 9.30E+10 1.000 1.146 1.258 2

6.49E+10 6.84E+10 0.867 0.925 3

7.48E+10 7.93E+10 0.999 1.073 4

7.91E+10 8.51E+10 1.057 1.151 5

6.83E+10 7.35E+10 0.912 0.994 6

5.63E+10 0.761 7

8.18E+10 1.106 8

7.20E+ 10 0.975 9

6.71 E+ 10 0.908 Average 8.58E+10 7.49E+10I 7.39E+ 10 1.000 1.000 1.000 Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-14 Table A-4 Measured Sensor Activities And Reaction Rates Surveillance Capsule V 12 R

y

.YP.

,djusteored;_,

&Raton (d&oaio

~~g

~

~

~

6 3CU (n,t) 60Co Top 3.50E+04 3.54E+05 3.54E+05 5.39E-17 Middle 3.38E+04 3.42E+05 3.42E+05 5.2 1E-17 Bottom 3.33E+04 3.36E+05 3.36E+05 5.13E-17 Average 5.25E-17

-Fe (n,p) 54Mn Top 9.97E+05 3.21E+06 3.21E+06 5.08E-15 Middle 9_73E+05 3.13E+06 3.13E÷06 4.9613-15 Bottom 9.43E+05 3.03E+06 3.03E+06 4.81 E-15 Average 4.95E-15 58Ni (n,p) 58Co Top 7.13E+06 4.60E+07 4.60E+07 6.59E-15 Middle 6.89E+06 4.45E+07 4.45E+07 6.37E-15 Bottom 6.60E+06.

4.26E+07 4.26E+07 6.10E-115 Average 6.35E-1S 238U (n,f) 137Cs (Cd)

Middle 1.15E+05 5.91E+06 5.91E+06 3.88E-14 238U (n,f) '37Cs (Cd)

Including 235U, 239Pu, and yfission corrections:

3.28E-14 237Np (n,f) 137Cs (Cd)

Middle 9.86E+05 5.07E+07 5.07E+07 3.23E-13 237Np (n,0) 1 3 7 Cs (Cd)

Including,fission correction:

3.20E-13 "9Co (n,T) 'Co Top 7.25E+06 7.32E+07 7.32E+0"7 4.78E-12 Middle 7.47E+06 7.55E+07 7.55E+07 4.92E-12 Bottom 7.80E+06 7.88E+07 T788E+07 5.14E-12 Average 4.95E-12

'9Co (n,y) 6°Co (Cd)

Middle 4.09E+06 4.13E+07 4.13E+07 2.70E-12 Bottom 4.24E+06 4.28E+07 4.28E+07 2.80E-12 Average 2.75E-12 Notes: 1) Measured specific activities are indexed to a counting date of March 15, 1991.

2) The average 23BU (n,t) reaction rate of 3.28E-14 includes a correction factor of 0.873 to account for plutonium build-in and an additional factor of 0.969 to account for photo-fission effects in the sensor.

3) The average 73 7Np (n,f) reaction rate of 3.20E-13 includes a correction factor of 0.991 to account for photo-fission effects in the sensor.

Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-15 Table A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule Y

..:.(...;'./,

-'k"

". "*.A *i*.rt-d.'X "A.j,*..d-Mes,e d W)~

dwh.it,,

I (g

C H U p !

§*v I m,

N,

.P.O. ;..

6 3CU (n,a) Coco Top 1.19E+05 2.85E+05 2.85E+05 4.34E-17 Middle 1.16E+05 2.78E+05 2.78E+05 4.23E-17 Bottom

1. 14E+05 2.73E+05 2.73E+05 4.16E-17 Average 4.25E-17 54Fe (n,p) '4Mn Top 1.63E+06 2-64E+06 2.64E+06 4.19E-15 Middle 1.59E+06 2.58E+06 2.58E+06 4.09E-15 Bottom 1.55E+06 2.51E+06 2.51E+06 3.99E-15 Average 4.09E-15 58Ni (n,p) 58Co Top 1.03E+07 4.16E+07 4.16E+07 5.96E-15 Middle L.OIE+07 4.08E+07 4.08E+07 5.84E-15 Bottom 9.70E+06 3.92E+07 3.92E+07 5.61 E-15 Average 5.80E-15

'3'U (n,f) '3 Cs (Cd)

Middle 5.06E+05 4.69E+06 4.69E+06 3.08E-14 23 8 U (n,f) 13 7Cs (Cd)

Including 23 5 U, 2 39 Pu, and ',fission corrections:

2.50E-14 237 1137CS-,-

3 Np (n,f) 137Cs (Cd)

Middle 3.56E+06 3.30E+07

_3.30E+07 2.11E-13 7Np (n,f) 7Cs (Cd)

Including y,fission correction:

2.09E-13

-9Co (ny) 6°Co Top 2.33E+07 5.58E+07 5.58E+07 3.64E-12 Middle 2.41E+07 5.77E+07 5.77E+07 3.76E-12 Bottom 2.45E+07 5.86E+07 5.86E+07 3.82E-12 Average 3.74E-12 59Co (n,'y) 6"Co. (Cd)

Top 1.26E+07 3.02E+07 3.02E+07 1.97E-12 Middle 1.29E+07 3.09E+07 3.09E+07 2.01 E-12 Bottom I.32E+07 3.16E+07 3.16E+07 2.06E-12 Average 2.01E-12 Notes: I) Measured specific activities are indexed to a counting date of June 19, 1997.

2) The average 238U (n,f) reaction rate of 2.50E-14 includes a correction factor of 0.837 to account for plutonium build-in and an additional factor of 0.969 to account for photo-fission effects in the sensor.

3) The average 237Np (n,f) reaction rate of 2.09E-1.3 includes a correction factor of 0.991 to account for photo-fission effects in the sensor.

Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-16 Table A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule U I

f Notes: 1) Measured specific activities are indexed to a counting date of February 1, 2003.

2) The average 238U (n,f) reaction rate of. 2.77E-14 includes a correction factor of 0.796 to account for plutonium build-in and an additional factor of 0.969 to account for photo-fission effects in the sensor.

3) The average 237Np (n,f) reaction rate of 2.49E-13 includes a correction factor of 0.991 to account for photo-fission effects in the sensor.

Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-17 Table A-5 Comparison of Measured, Calculated, and Best Estimate Reaction Rates At The Surveillance Capsule Center Capsule V Capsule Y

... * **"'.*:..Rac!on

, Meaii*.u" c.u..!tmted* !*-,Egtimiate' :-.

M/C

"/BE

....i 63Cu(n,a)6°Co 4.25E-17 3.99E-17 4.11 E-17 1.07 1.03 54Fe(np)-AMn 4.09E-15 4.42E-15 4.24E-15 0.93 0.96 5"Ni(n,p)- 8Co 5.80E-15 6.21E-15 5.93E-15 0.93 0.98 238U(n,f)' 37Cs (Cd) 2.50E-14 2.39E-14 2.27E-14 1.05 1.10 237Np(n,f)' 37Cs (Cd) 2.09E-1 3 2.33E-13 2.15E-13 0.90 0.97

' 9Co(ny) 60Co 3.74E-12 3.19E-12 3.67E-12 1.17 1.02 59Co(n,T)6°Co (Cd) 2.01E-12 2.25E-12 2.04E-12 0.89 0.99 Capsule U

__a "La___________

W...

63Cu(n,ox)"OCo 4.32E-17 3.98E-17 4.20E-17 1.09 1.03 54Fe(n,p)-4Mn 4.30E-15 4.40E-15 4.47E-15 0.98 0.96 5 8Ni(n,p)18Co 6.13E-15 6.17E--15 6.28E-15 0.99 0.98 283U(n,f)"37Cs (Cd) 2.77E-14 2.36E-14 2.45E-14 1.17 1.13 237Np(n,f)137Cs (Cd) 2.49E-13 2.30E-13 2.45E-13 1.08 1.02 59Co(n,'y)6Co 3.77E-12 3.17E-12 3.70E-12 1.19 1.02 59Co(ny)°Co (Cd) 2.02E-12 2.23E-12 2.05E-12 0.91 0.99 Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-18 Table A-6 Comparison of Calculated and Best Estimate Exposure Rates At The Surveillance Capsule Center Note: Calculated results are based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period.

Capul e ain~t."'U/C V

1.67E-10 1.79E-l10 8%

1.07 Y

1.46E-10 1.38E-10 8%

0.95 U

!..43E-10 1.511E-10 8%

1.05 Note: Calculated results are based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period.

Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-19 Table A-7 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions 63Cu(n,a)6Co 1.19 1.07 1.09 54Fe(n,p) 54mn 0.99 0.93 0.98

-5 Ni(np)5 8Co 0.90 0.93 0.99 23 8U(n,p)137Cs (Cd) 1.21 1.05 1.17 27Np(n, f I 37Cs (Cd) 1.19 0.90 1.08 Average 1.10 0.97 1.06

% Standard Deviation 12.7 7.9 7.5 Note: The overall average M/C ratio for the set of 15 sensor measurements is 1.04 with an associated standard deviation of 10.5%.

Table A-8 Comparison of Best Estimate/Calculated (BE/C) Exposure Rate Ratios V

1.05 1.07 Y

0.95 0.95 U

1.06 1.05 Average 1.02 1.02

% Standard Deviation 5.6 6.3 Appendix A WCAP-16093-NP, Rev. 2 July 2007

A-20 Appendix A References A-I.

Regulatory Guide RG-1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

A-2.

WCAP-13182, Revision 0, "Analysis of Capsule V from the Houston Lighting and Power Company South Texas Unit 2 Reactor Vessel Radiation Surveillance Program," February 1992.

A-3.

WCAP-14978, Revision 0, "Analysis of Capsule Y from the Houston Lighting and Power Company South Texas Unit 2 Reactor Vessel Radiation Surveillance Program," December 1997.

A-4.

A. Schmittroth, FERRETData Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

A-5.

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

Appendix A WCAP-16093-NP, Rev. 2 July 2007

B-0 APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS Specimen prefix "HL" denotes Lower Plate, Longitudinal Orientation Specimen prefix "HT" denotes Lower Plate, Transverse Orientation Specimen prefix "HW" denotes Weld Material Specimen prefix "HH-" denotes Heat-Affected Zone material Load (1) is in units of lbs Time (1) is in units of milli seconds Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-I 5000.00 4000-00 2000.00.

1000o.0o nil IL +.--

0.00 1.00 2.00 3.00 Time-I (ms)

HL14, -50°F 4.00 5.00 6.00 5000-00 4000.00 2000.00 1000.00 0.00 0.0

.............. I........

00 1.00 2.00 3.00 Time-I (ms)

HL7, -25°F 4.00 5.00 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-2

• 30M0000 o,

0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-I (ms)

HL6, 0°F 5 0 130.0 0............*............ *..............

4 000.00

,e*

.... I:

4000.00

~30 M0.00 ca 0

-j 2000.00 1 0 0 0.0 0 0.00**

0.00.

1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

HLU, 10F Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-3 5000.00 4000.00 3000.00

-J 2000.00 1000.00 a

a F

n m 0.00 10.0 2.00 3.00 Tune-I (ms)

HL8, 20°F 4.00 5.00 6.00 5000.00 4000.00 2000.00 1000.00 l11111 n rtn I 0.100 1.00 2.00 3.00 T1im-1 (ms)

HL9, 40°F 4.00 5.00 6.J G0 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-4 4000.00 3400.00 2000-00 1000.00 n rin 0.00 1.00 2.00 3.00 Tuw-I (ms)

HL12, 50-F 4.00 5.00 6.00 6.00 0

V 0-J

)o 3.00 TH5e-1 (ms)

HL15, 75°F Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-5 a

0

-J 5000.00 4000.00-3000.00 2000.00-1000.oo0 a

( X..

0.00 1.00 2.00 3.00 lime-1 (ms)

IL13, 100°F 4.00 5.00 600

.0

-u 0

-j 0.00 1.00 2.00 3.00 4.00 5.00 Time-1 (ms)

HL3, 125°F 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-6

" 300000

-o

-J 2000,00 0.00 0.

5OOW.O0 4000.00

• 3000.00 2000.M0 10o0.00.

00 1.00 2.00 3.00 4.00 5.00 Time-1 (ins)

HL10, 150°F 6.00 0.00 1.00 2.00 3.00 TM-I1 (ns)

L11, 175°F 4.00 5.00 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-7 3000.00.*.........

1 0 0 0.0 0......

0.00 0.00 1.00 2.00 3.00 4.00 5.00 600 Time-i (ms)

HL5, 200°F 5 0.0............".........".....................................

4000.00 3 0 0.0 0...............................

............i.............

0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-i (ms)

HL4, 225°F Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-8 a

-J 0.00 1.00 2.00 3.00 4.00 5.00 "ime-I (ms)

HL2, 250-F 6.00 5000.00 4000.00

-, 3000.00 2000.00 1000.00

...........r

...I...................

0.00 1.00 2_00 3.00 TTne-1 (ms)

[T11, -75°F 4.00.

5.00 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-9 5000.00 4000.00 3000.00 ca

-J 2000.00 1000.00 a

a..............

rim!

0.00 1.00 2.00 3.00 Tre-1 (ms) 1T3, -50°F 4.00 sf0 6.

5000.00 4000.00

.3000.00 2000.00 1000.00

.00

.(00 0.00 1.00 2.00 3.00 rTioe-1 (ms)

HT4, -25°F 4.00 5.00 6.

Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-10 500 400

'7 300 0ca 200 10 0 0 0

.0 0 00...

n nfl 0.00 1.00 2.00 3.00 Time-i (ms)

HT6, 00F 4_00 5.00 6.00 5000.00 4000.001 3000.00

-o 2000.00 11000.00 0.00

+

-~-

r r

0.00 1.00 2.00 3.00 lime-i (ms)

HT7, 25°F 4.00 5.00 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-I1 S3000.1 0

0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-I (ms)

HT8, 50°F 5000.00 3000.00 2000.00.

000.00 0,00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tine-l (ms)

HT10, 60-F Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-12 0.00 1.00 2.00 3.00 4.00 5.00 6.00 lime-I (ms)

HT12, 75°F 5 0 1*.00 50M.00 00" 2

3 M

0 0 0 0 2

0,1 7 M OD 1000.00 0.00 0.00 1.00 2.00 3.00 4.00 5,00 6.00 Time-1 (ms)

HT9, 100°F Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-13 12 0

-j 0.00 1.00 2.00 3.00 Time-I (ms)

HT2, 125°F 4.00 5.00 b.00 5000.00 4000.00

" 3000.00

-J 2000.00 1000.00 0.00 0.a

...................a..........

0 i

i 6

I i

i i*

L

=

k 1.00 2.00 3.00.

Time-i (ms)

HT14, 150°F 4.00 5.00 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-14 0.0

,i 2000.0 0 0.00 0.00 1_00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

HT1, 1900F 5 0 0 0.0 0...

5 W.0 0...

4000.00 2000.00 1000.00.

0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-I (ms)

HT13, 225°F Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-15 0.00 1.00 2.00 3.00 4.00 Trfe-1 (Ms)

HT15, 2500F 5.00 6.00 0.00 r i

i I 0.00 1.00 2.00 3.00 4.00 5.00 Time-I (ms)

HT5, 275-F 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-16

.o 0

.00 1.00 00 3.00 Time-1 (ms)

HW6, F 4.00 5.00 6.00 3000.00 4000.00

-0 3000.00 2000.00 1000.00 a

0.00 1.00 20.0 3.00 Time-1 (ms)

HWI5, -50°F 4.00 5.00 6.00 Appendix 5 WCAP-16093-NP, Rev. 2 July 2007

B-17 5000.00 4000.00) 30M.0000-20J 1000.00 0.00 1.00 2.00 3.00 Twne-I (ms)

HW9, -25°F 4.00 5.00 6.00 2000.00 1000.00-0.00 1 00 2.00 3.00 4.00 5.00 Time-I (ms)

HWIO, 0°F 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-18 4000.00 4 I D 0 01

.: 3 r.

.......... I.....................................

Li 1g 3 0 00 _0 o

200000 10 0.00 0.00 1.00 200 3.00 4.00 5.00 600 trne-1 (ms)

HW12, 10°F 5 0 0 0.0 0 4000.00 C.. 3000.00 Ur 100000 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

HW8, 30 0F Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-19

~3000.00.....

20 (

2000.00 1000.00 0.00 0.00 100 2.00 3.00 4.00 500 6.00 Tine-I (ms)

HW3, 50°F 5 00 0.00

• ."4000.00 "7

3 0 0 0.0 0

3 M0.0 0 0...

0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Tun-I (ms)

HW4, 500F Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-20 3W300000 0

-j 0.00 100 2.00 3.00 4.00 5.00 6.00 Time-1 (ms)

HW14, 750F 12

~0* 3000.00-Uo

-J 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-i (ins)

HW2, 125°F Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-2I 5000.00 400000-

• 3000.00' 2000.00-1000.00 wiiiJ[*

0.00 1.00 2.00 3.00 rime-i (ins)

HW5, 150°F 4.00 5.00 6.00 4000.1

~03000.1 0

0.00 1.00 2.00 3.00 4.00 5.00 "F7e.2 (ms)

HW7, 200°F 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-22 5 0 00.00

.4000.00...

.. a 0

(

C 2 0 0 0.0 0 1 0 0 0.0 0..

0.00 0.00 1.00 2.00 3.00 4.00 5.00 6_00 Time-I (ms)

HW13, 225-F 5000.00 2000.00 1 000.00 0.00 0.00 1.00 2.00 3.00 4.00 5.00 6.00 Time-i (m)

HW11, 250°F Appendix B WCAP-1 6093-NP, Rev. 2 July 2007

B-23 5000.00 4000.00

, 3000.00 0-J 2000.00 1000.00 r

0.00 2.00 3.00 Time-1 (ins)

HW1, 275-F 4.00 500 600

-D

-J 00 3.00 Tm#e-1 (ms)

HH3, -200°F 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-24

-6 0

_k 5000.00.

• 4000.001-3000.100 2000 00 1000.00 0.00 0.00

........... I......

a 1.00 2.00 3.00 Time-I (ms)

HH2, -150-F 4.00 5.00 6.00 5000.00 3000.00

2000.100 1000.00 0.00 1.00 2.00 3.00 Time-i (ins)

H4, -125°F 4.00 5.00 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-25

.0

'7 3 00 0.0 0 2000.00 1 0 0 0.0 0 0.00 0.00 1.00 2,00 3.00 4.00 5.00 6.00 Time-1 (ms)

HH7, -108°F 5 0 0 0.0 0 3 0 0 0.0 0 S 3000.00...........-

20 00I).0 0 I..

10oo o.0 0 0.00 0.00 1.00 200 3.00 4.00 5.00 6.00 Time-1 (ms)

HHI4, -750F Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-26 5000.00 4000.00

• 3000.00 2000.00 1000.00 a

.......................a P

0.00 1.00 2.00 3.00 Time-1 (ms)

HH5, -50°F 4.00 5.00 6.00 4000.00 t

30oo.00 200000 0.00 1.00 2.00 3.00 4.00 5.00 Time-I (ms)

HHll, -25°F 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-27 5 0 0 0 0 0 40 M0.00

.I..

1000.00 0.00 P

0.00 1.00 2,3 400 5.00 6.00 Trne-1 (ms)

HH10, O°F 5 0 0 0.0 0 4000.00

£2 3000.00 C

0.00 1.00 2.00 3.00 4.00 5.00 6.00 Thrne-1 (is)

HH8, 25°F Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-28 5000.00 3000.00

.0 0.,

2000.00 1000.00 m

lIIRI 0.00 1.00 2.00 3.00 Time-I (=)

HH15, 50D°F 4.00 5.00 6.00

!3

'713 0.00 1.00 2.00 3.00 4.00 5.00 Trne-I (ms)

HH9, 75-F 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-29 5000.00 4000.00 3000.00 2000.00-1000.00.

r a

m

...........a II IIII J*

n r

i i

0.00 1.00 2.00 3.00 Time-i (ms)

H12, 125°F 4.00 5.00 5.00

-J 0.00 1.00 2.00 3.00 4.00 5.00 lme-1 (ms)

HH6, 150°F 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

B-30 5000.00 4000.00 213000.00 200000 1000.00 m

r m

4 ---

6 i

0.00 100 2.00 3.00 Time-1 (ms)

HH13, 175-F 4.00 5.00 6.n0

.J 0.00 1.00 2.00 3.00 4.00 5.00 T'rne-I (ms)

HH1, 200-F 6.00 Appendix B WCAP-16093-NP, Rev. 2 July 2007

C-0 APPENDIX C CHARPY V-NOTCH PLOTS FOR CAPSULE U USING SYMMETRIC HYPERBOLIC TANGENT CURVE-FITTING METHOD Contained in Table C-I are the upper shelf energy values used as input for the generation of the Charpy V-notch plots using CVGRAPH, Version 4.1. The definition for Upper Shelf Energy (USE) is given in ASTM E 185-82, Section 4.18, and reads as follows:

"upper shel energy level - the average energy value for all Charpy specimens (normally three) whose test temperature is above the upper end of the transition region. For specimens tested in sets of three at each test temperature, the set having the highest average may be regarded as defining the upper shelf energy."

If there are specimens tested in set of three at each temperature Westinghouse reports the set having the highest average energy as the USE (usually unirradiated material). If the specimens were not tested in sets of three at each temperature Westinghouse reports the average of all 100% shear Charpy data as the USE.

Hence, the USE values reported in Table C-I and used to generate the Charpy V-notch curves were determined utilizing this methodology.

The lower shelf energy values were fixed at 2.2 ft-lb for all cases.

pp i

,~g.

Intermediate Shell Plate 138 139 133 140 R2507-1 (Long.)

Intermediate Shell Plate 98 105 105 104 R2507-1 (Trans.)

Weld Metal 98 93 101 97 (heat # 90209)

HAZ Material 155 137 137 129 Appendix C

/

WCAP-16093-NP, Rev. 2 July 2007

CAPSULE U CVGRAPH 41 Hyperbolic Tangent Curve Printed at l

.102.O on 04-02-2003 Page I Coefficients of Curve I S

A= 71.9 B= 68.9 C=86B2 ID =4921 Equation is: CVN = A + B I tanh((T - TO)/C)

Upper Shelf Energy: 140 Fixed Temp. at 30 ft-lbs

-10.4 Temp. at 50 ft-lbs 21.7 Lower Shelf Energy-219 Fixed Material: PLATE SA533BI Heat Number. NR 62 067-1 Orientation: LT Capsule: U Total Fluence:

9)

Q) z U>

-300

-2W0

-100 0

100 200 300 Temperature in Degrees Data Set(s) Plotted Plant-Sf2 Cap-- U Material PLATE SA533BI Ori_ LT Heat Charpy V-Notch Data Lure Input CVN Energy Computed CIN Energy 400 F

5W0 600 NR 62 067-1 Tempera Differential

-50 010 20 40 50 75 100 7

28 22 46 58 60 68 107 105 14.2 2331 35.74 4L93 48.75 63Bt 71.71 9027 10735

-7.92 4S8

-1a74 924

-3.81

-3171 16.02

-2.35 Data continued on next pakge WCAP-16093-NP, Rev. 2 July 2007 C-I

C-2 WCAP-16093-NP, Rev. 2 C-2

- *July 2007

CAPSULE U CVGRAPH 41 Hyperbolic Tangent Curve Printed at 1(Y2413 on 04-02-2003 Page I Coefficients of Curve I A =40.37 l = 39,37 C=71 TO = 418 Upper Shelf 1&- 79.75 Material: PLATE S Equation is: LE.

A + B

• I tanh((T - TO)/C) I Temperature at L. 35:

32.4 Lo

'A533Bl Heat Number NR 62 067-1 Capsule: U Total Fluence wer Shelf ILX I Fixed Orientation: LT U)

X 4)

-300

-200

-100 0

100 200 30 Temperature in Degrees Data Set(s) Plotted Plant ST2 Cap-U Material PLATE SA5331B1 OrL LT Heat Charpy V-Notch Data ture Input Lateral Expansion Computed LK 4W0 F

5w0 600 Ij NR 62 067-Tempera Differential

-50

-25 0

10 20 40 50 75 100 I

15 Ii 28 35 37 40 63 65 6.46 1131 19.39 23165 2B.45 3916 44.69 57.37 66.83

-5W6 3BB

-9.39 434 654

-2.16

-4.69 5.62 483 Data continued on next. page C-3 WCAP-16093-NP, Rev. 2 July 2007

Material:

CAPSULE U Page 2 PLATE SA533BI Heat Number NR 62 067-1 C

Capsule: U Total Fluence Charpy V-Notch Data (Continued)

Input Lateral Expansion Computed LE.

70

?Z78 75 7614 78 7722 82 78.83 81 7929 78 7952 lrientation: LT Temperature 125 150 175 200 250 Differential

-2.78

-_14

.07 316 17

-452 SUM of RESIDUALS = -2.87 C-4 WCAP-16093-NP, Rev. 2 July 2007

CAPSULE U CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 10,252 on 04-02-2003 Page 1 Coefficients of Curve I A=50 B=50 C = 56.2 TO = 5.77 Equation is Shear/. = A + B

• I tanh((T - TD)/C) I Temperature at 5R/. Shear 56.7 Material: PLATE SA533BI Heat Number.- NR 62 067-1 Capsule: U Total Fluence Orientation: LT (D

4-.)

0--

0U

-300

-200

-100 0

100 200 300 400 Temperature in Degrees F Data Set(s) Plotted Plant-ST Cap.- U Material-PLATE SA533B1 OrL LT Heat f NR 62 Charpy V-Notch Data ture Input Percent Shear Computed Percent Shear 500 600 067-1 Tempera

-50

-25 0

to 20 40 50 75 100 2

10 15 15 25 30 40 75 85 V31 5.42 1a.19 16.52 21.99 3639 449 6637 8V7 Differential

-.31 4.57 2B

-152 3

-6.39

-49 8.2 229 SData continued on next page C-5 WCAP-16093-NP, Rev. 2 July 2007

Material: P CAPSULE U Page 2 IATE SA533BI Heat Number. NR 62 067-1 Orient Capsule: U Total Fluence Charpy V-Notch Data (Continued)

Input Percent Shear Computed Percent Shear 85 92.05 100 96.5 90 98-54 100 99.39 100 99.74 100 99B39 ationr IX Temperature 125 150 175 200 225 250 Differential

-7.05 W4

-854 11

.1 M of RESIDUALS = -303 Sul C-6 WCAP-16093-NP, Rev. 2 July 2007

CAPSULE U CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 10-2719 on 04-02-2003 Page I Coefficients of Curve 1 A = 53.09 B = 50.9 C = 10051 TO =71.43 Upper Shelf Energy: 104 Fixed Material:

Equation is CVN A

A + B* I tanh((T - '0)/C) I Temp. at 30 ft-lbs 222 Temp. at 50 ft-lbs 653 Lower Shelf Energy. 219 Fixed PLATE SA533BI Heat Number NR 62 067-1 Orientation: TL Capsule: U Total Fluenee:

Cr) n S

15(

10(

r54 Tempera

-75

-50

-25 0

25 50 60 75 100 Li Y

-300

-200

-100 0

100 200 300 Temperature in Degrees Data Set(s) Plotted Plant SM2 Cap: U Material: PLATE SA533BI Ori: TL Heat Charpy V-Notch Data ture Input CVN Energy Computed CVN Energy 2

7.44 10 10.54 24 1522 22 2L99 32 31J2 35 4Z4 50 47.33 59 549 64 6718 4W0 F

500 600 NR 62 067-1 Differential

-5.44

-.54 1177 0ff7

-7.4 266 4.09

-318 Data continued on next page C-7 WCAP-16093-NP, Rev. 2 July 2007

Material: PL CAPSULE U Page 2 ATE SA533B1 HeaL Number. NR 62 067-1 Orien Capsule: U Total Fluence Charpy V-Notch Data (Continued)

Input CVN Energy Computed CVN Energy 79 77.91 79 8626 101 952 102 99.42 108 10116 104 10225 SI Temperature 125 150 190 225 250 tation: TL Differential LOB

-7.6 5.79 2.57 6B]3 L74 JM of R-1DUALS = 10.49 WCAP-16093-NP, Rev. 2 July 2007 C-8

CAPSULE U CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 10-2124 on 04-02-2003 Page 1 Coefficients of Curve 1 A = 34.58 B =3358 C= 91.43 IN = 80.5 I

Upper Shelf LE. 68.17 Material: PLATE S Equation is LF = A - B *I tanh((T - TO)/C) I Temperature at LK 35:

81.9 Lo A533B1 Heat Number. NR 62 067-1 Capsule: U Total Fluence:

wer Shelf L& I Fixed Orientation: TL 27-U) r-F-4 r-r--

(D 1W0 1007 5(F0 U

-300

-200

-100 0

100 200 300 Temperature in Degrees Data Set(s) Plotted Plant S'T2 Cap: U Material PLATE SA533BI OriL TL Heat Charpy V-Notch Data Temperature Input Lateral Expansion Computed L.

400 F

,500 600 tI NR 62 067-1 Differential

-75

-50

-25 0

25 50 60 75 100 0

1 10 11 18 20 30 31 44 3.15 4.63 7.03 10.78 1628 23.66 27.05

~3?43 41.51

-3.15 63

%96 21 1.71

-3166 294

-1.43

&-48 Dat~a continued on next page C-9 WCAP-16093-NP, Rev. 2 July 2007

Material:

CAPSULE U Page 2 PLATE SA533BI Heat Number. NR 62 067-1 Capsule U Total Fluence.

Charpy V-Notch Data (Continued)

Input Lateral Expansion Computed LR 51 49.64 50 56.04 67 251 63 65.41 66 6655 69 6722

)rientation: TL Temperature 125 150 190 225 250 275 Differential 1.35

-6.04 4.48

-2.41

-.55 1.77 SUM of RESIDUALS = -?95 C-10 WCAP-16093-NP, Rev. 2 July 2007

CAPSULE U CYGRAPH 4.1 Hyperbolic Tangent Curve Printed at 10R29.54 on 04-02-2003 Page 1 Coefficients of Curve I

.I A=50 B=50 C = 62.16 TO = 0935 I

Equation is: Shear/ = A + B

• I tanh((T - TO)/C) I Temperature at 5W-/. Shear.

89.3 Material: PLATE SA533B1 Heat Number. NR 62 067-1 Capsule: U Total Fluence Orientation: TL 04)

C..)

IL 0-

-300

-200

-100 0

100 200 300 Temperature in Degrees Data Set(s) Plotted Plant: SI2 Cap-- U Material PLATE SA533B1 Ori"- TL Heat *.

Charpy V-Notch Data ture Input Percent Shear Computed Percent Shear 400 F

500 600 NR 62 067-1 Tempera

-75

-W0

-25 0

25 50 60 75 100 2

5 10 10 15 20 25 35 60

.5 2.46 5.34 I[19 2t19 271.99 38.65 58.47 Differential

[49 3B68 7.53 4U5 3.8

-199

-_Z99

-3.65

[52 Data continued on next page WCAP-16093-NP, Rev. 2 July 2007 C-1I

Material: P CAPSULE U Page 2 LATE SA533BI Heat Number. NR 62 067-1

Orient, Capsule: U Tow Fluence Charpy V-Notch Data (Continued)

Input Percent Shear Computed Percent Shear 80 75A89 85 7.55 100 9622 100 98.74 100 99.43 100 99.74 Ltion: TL Temperature 125 150 190 225 250 275 Differential 41

-2.55 3.77 125 56 25 4 of RESIDUAIS = 2L65 SUI WCAP-16093-NP, Rev. 2 July 2007 C-12

CAPSULE U CVGRAPH 41 Hyperbolic Tangent Curve Printed at 10.3258 on 04--02-2003 Page 1 Coefficients of Curve I A = 4959 B = 47.4 C = 61.7 TO = 33.01 Equation is CVN = A + B* I tanh((T - TO)/C) I Upper Shelf Energy: 97 Fixed Tempi at 30 ft-lbs 5.8 Temp. at 50 ft-lbs 335 Lower Shelf Energy. 219 Fixed Material: WELD Heat Number. 90209 Capsule: U Total Fluence:

Orientation':

I) 0.

C-)

-300

-200

-100 0

100 200 300 Temperature in Degrees Data Set(s) Plotted Plant-M2 CapN U Material WELD OrL Heat.

400 50W 6W0 F

90209 Temperature

-75

-50

-25 0

10 30 50 50 75 Charpy V-Notch Data Input CVN Energy Computed CVN Energy 4

4.97 10 822 17 14.74 28 26.4 28 32.69 41 4728 64 62.32 72 62.32 74 77.64 Differential L77 225 L59

-4.69

-628 1.67 9j67

-3.64 Data continued -on next page WCAP-16093-NP, Rev. 2 July 2007 C-13

CAPSULE U Page 2 Mater Temperature 125 150 200 225 250 275 rial: WELD Heat Number. 9M209 Orientation:

Capsule U Total Mluencm Charpy V-Notch Data (Continued)

Input CVN Energy Computed CYN Energy 94 92.42 86 94.9 91 96.57 97 96B1 96 9691 104 96.96 Differential 157

-489

-557 18

-.91 7.03 of RESIDUALS = -522 SUM WCAP-16093-NP, Rev. 2 July 2007 C-14

CAPSULE U CVGRAPH 4J Hyperbolic Tangent Curve Printed at 1M.3656 on 04-02-2003 Page 1 Coefficients of Curve 1 A = 34.04 B = 33.04 C = 57.95 T1 = 3175 Upper Shelf LE-67.08 Material:

Equation is L. = A + B * [ tanh((T - TO)/C) I Temperature at.E 35 35.4 Ic w

Heat Number 90209 C

Capsule: U Total Fluence wer Shelf LX: I Fixed Irientation:

20(

t -

9 q -

. ~~~q a

N -

-I Cr) 150'-f_

1U'

-~----

'p 9

4 r-Cb II 5__

I'i-V

-300

-200

-100 0

100 200 300 Temperature in Degrees Data Sets) Plotted Plant 812 Cap: U Material: WELD OrL Heat k.

Charpy V-Notch Data Temperature Input Lateral Expansion Computed LK 400 F

500 600 9020 Differential

-75

-50

-25 0

10 30 50 50 75 0

3 10 20 20 25 45 50 51 251 4.47 0.68 16.71 2121 31.9 43.07 43.07 5425

-2.51

-1.47 1.31 328

-121

-6.9 1.92 6.92

-325 D

Data continued on next page

• WCAP-16093-NP, Rev. 2 July 2007 C-15

M.

CAPSULE U Page 2 aterial: WELD Heat Number. 90209 Orie Capsule: U Total Fluence Charpy V-Notch Data (Continued)

Input Lateral Expansion Computed LK 64 6436 62 65.9 63 66M 67 66.99 71 67.04 70 67.06 Temperature 150 200 225 250 275 ntation:

Differential

-a.9 0

a95 2.93 SUM of RESIDUALS = -3.14 C-16 WCAP-16093-NP, Rev. 2 July 2007

CAPSULE U CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 1039:52 on 04-02-2003 Page I Coefficients of Curve I A=50 B=50 C=742 TO=1359 Equation is Shear/ = A + B

• I tanh((T - TO)/C) i Temperature at 50z, Shear 13.5 Material: WELD Heat Number 90209 Capsule: U Total Fluence:.

Orientation:

C) 4-)

Q) 0*

-300

-200

-100 0

100 200 300 400 Temperature in Degrees F Data Set(s) Plotted Plant STI2 Cap:- U MaterialRELD Ori:

Heat. 90209 Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear 500 600 Differential

[56

-29

-1.13 4.04

-758 4.13

-27 229 1.07

-75

-50

-25 0

10 30 50 50 75 10 15 25 45 40 65 70 75 85 8.43 1529 26.13 40.95 4758 60B6 72.7 7V7

&3.92 Data continued on next page WCAP-16093-NP, Rev-2 July 2007 C-1 7--m!

CAPSULE U Page 2 Temperature 125 150 205 2250 250 Material: WELD iteat Number 90209 Orientation:

Capsule: U Total Fiuence:

Charpy V-Notch Data (Continued)

Input Percent Shear Computed Percent Shear Differential 95 9525

-25 95 97.51

-251 98 9934

-1.34 98 99&

-1.6 100 99.82 17 100 9991.

.08 SUM of RESIDUALS = -412 WCAP-16093-NP, Rev. 2 July 2007 C-I 8

CAPSULE U CVGRAPH 41 Hyperbolic Tangent Curve Printed at lf0A24 on 04-02-2003 Page 1 Coefficients of Curve I A = 6559 B = 63.4 C= 11625 TO = -2122 Upper Shelf Energy: 129 Fixed Equation Lz CVN I

Temp. at 30 ft-lb&

Material: HEAT AFFD ZONE Capsule:

= A + B I tanh((T - T'0)/C) I

-95 Temp. at 50 11t-l4b

-50.4 Lower Shell Energy-219 Fixed Heat Number.

Orientatiorn U

Total Fluence:

U) 10 z

C-)

-300

-200

-100 0

100 200 300 Temperature in Degrees Data Set(s) Plotted Plant-S2 Cap: U Material-HEAT AFFD ZONE Ori:

Charpy V-Notch Data Temperature Input CVN Energy Computed CVN Energy 400 F

Heat h.

500 600

-200

-150

-125

-108

-75

--50

-25 0

25 6

7 13 30 52 36 56 90 130 7.79 14.67 20.41 25.46 382 5022 63.54 77.04 896 Differential

-179

-7.67

-7.41 4.53 13.79

-1422

-7.54 12.95 40.43 I

Data continued on next page

• WCAP-16093-NP, Rev. 2

- July 2007 C-19

CAPSULE U Page 2 Temperature 50 75 125 150 175 200 Material: HEAT AFFD ZONE Beat Number.

Orientation:

Capsule U Total Fluencw Charpy V-Notch Data (Continued)

Input CYN Energy Computed CVN Energy Differential 61 10021

-3921 96 108.66

-1266 115 119.51

-4.51 114 12266

-8.86 144 124.8 1919 143 12624 16.75 SUM of RISIDUALS = 3.95 WCAP-16093-NP, Rev. 2 July 2007 C-20

CAPSULE U CVGRAPH 41 Hyperbolic Tangent Curve Printed at 10:5321 on 04-02-2003 Page 1 Coefficients of Curve I A = 3211 B = 31.1 C = 70.44 TO = -2625 Equation is LK = A + B I tanh((T - TO)/C)

Upper Shelf L.F 6402 Temperature at LK 35: -213 Lower Shelf LE-I Fixed Material: HEAT AFlF ZONE Heat Number.

Orientation:

Capsule U Total Fluence 2007 U) 15 c00 0

LI~~-

4 I

n 2o 50F-00 U

I

-300

-200

-100 0

100 200 300 Temperature in Degrees Data Setqs) Plotted Plant S912 Cap: U Material: HEAT AFPD ZONE Ori-Charpy V-Notch Data Temperature Input Lateral Expansion Computed LE.

4O0 F

500 6O0 Heat f.

Differential

-200

-150

-125

-108

-75

--50

-25 0

25 01 411 18 15 26 52 75 IA5 4.63 13.74 2Z47 33Z37 4414 5258

-L45

-484

-133 431 425

-7.47

-7.37 7.85 22,41

'Data continued on next page 0"~

WCAP-16093-NP, Rev. 2 July 2007 C-21

CAPSULE U Page 2 Temperature 50 75 125 150 175 200 Material: HEAT AFFD ZONE Heat Number Orientation:

Capsule. U Total Fluenwe Charpy V-Notch Data (Continued)

Input Lateral Expansion Computed LE Differential 35 58107

-2307 53 6122

-8m 65 63.76 123 53 6419 19 73 64.41 858 76 6452 1L47 SUM of RIDUALS = 14 WCAP-16093-NP, Rev. 2 July 2007 C-22

CAPSULE U CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 1056:43 on 04-02-2003 Page 1 Coefficients of Curve I A

50 B=50 C =

12 MO=-278 Equation is Shear/ = A + B '

tanh((T - TD)/C)

Temperature at 50;/ Shear -271 Material: HEAT AFVD ZONE Heat Number:

Orientation:

Capsule: U Total Fluence:

cn a)

C.)

09

-300

-200

-100 0

100 200 300 400 Temperature in Degrees F Data Set(s) Plotted Plant: ST2 Capý U Material: HEAT AFl) ZONE OrL Heat Charpy V-Notch Data Temperature Input Percent Shear Computed Percent Shear 500 600

-200

-125

-108

-75

-50

-25 0

25 2

2 510 15 35 55 80 100 1.46 4.78 8.45 122 23.78 36.45 51.33 6597 78.08 Differential

.53

-228

-Z.78

-145 3.66 14.02 21.91

~'Data continued on next page WCAP-16093-NP, Rev. 2 July 2007 C-23

CAPSULE U Page 2 Temperature 50 75 125 150 175 200 Material: HEAT AFFD ZONE Capsule: U Charpy V-Notch Input Percent Shear 56 70 100 100 100 100 Heat Number Orientation:

Total Fluence:

Data (Continued)

Computed Percent Shear Differential 86.75

-3L75

-233-33 97.6 2.39 98.68 131 9927

.72 99.6

.39 SUM of RESIDUA[S =-27B6 WCAP-16093-NP, Rev. 2 July 2007 C-24

D-0 APPENDIX D SOUTH TEXAS UNIT 2 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION Appendix D WCAP-16093-NP, Rev. 2 July 2007

D-1 INTRODUCTION:

Regulatory Guide 1.99, Revision 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation enmbrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

To date there has been three surveillance capsules removed from the South Texas Unit 2 reactor vessel.

To use these surveillance data sets, they must be shown to be credible. In accordance with the discussion of Regulatory Guide 1.99, Revision 2, there are five requirements that must be met for the surveillance data to be judged credible.

The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the South Texas Unit 2 reactor vessel surveillance data and deternline if the South Texas Unit 2 surveillance data is credible.

EVALUATION:

Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The South Texas Unit 2 reactor vessel consists of the following beltline region materials:

Intermediate Shell Plates R2507-1, 2, 3 Lower Shell Plates R3022-1, 2, 3 Intennediate & Lower Shell Longitudinal Weld Seams (Heat # 90209),

Intermediate to Lower Shell Circumferential Weld Seam (Heat # 90209).

At the time when the South Texas Unit 2 surveillance program material was selected it was believed that copper and phosphorus were the elements most important to embrittlement of the reactor vessel steels.

The intermediate shell plate R2507-1 had one of the highest initial RTNDT and the lowest USE of all plate materials in the beitline region. In addition, the intermediate shell plate R2507-1 had approximately the same copper and phosphorus content as the other beltline plate materials. Therefore, based on the highest initial RTNDT and the lowest USE, the intermediate shell plate R2507-1 was chosen for the surveillance program.

Appendix D WCAP-16093-NP, Rev. 2 July 2007

D-2 The weld material in the South Texas Unit 2 surveillance program was made of the same wire as all the reactor vessel beltline welds, thus it was chosen as the surveillance weld material.

Hence, Criterion I is met for the South Texas Unit 2 reactor vessel.

Criterion 2:

Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper shelf energy unambiguously.

Based on engineering judgment, the scatter in the data presented in these plots is small enough to permit the determination of the 30 ft-lb temperature and the upper shelf energy of the South Texas Unit 2 surveillance materials unambiguously. Hence, the South Texas Unit 2 surveillance program meets this criterion.

Criterion 3:

When there are two or more sets of surveillance data from one reactor, the scatter of ARTNUJT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28*F for welds and 17°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82.

The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these ARTNDT values about this line is less than 28'F for welds and less than 17'F for the plate.

Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2. In addition, the recommended NRC methods for determining credibility will be followed. The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13, 1998. At this meeting the NRC presented five cases. Of the five cases, Case 1 ("Surveillance data available from plant but no other source') most closely represents the situation listed above for South Texas Unit 2 surveillance weld metal and plate materials.

Appendix D WCAP-16093-NP, Rev. 2 July 2007

D-3 TABLE D-I Calculation of Chemistry Factors using South Texas Unit 2 Surveillance Capsule Data Intermediate Shell V

0.23 0.60 16.39 9.83 0.36 Plate R2507-1 Y

1.21 1.05 33.96 35.66 1.10 (Longitudinal)

U 2.40 1.24 27.48 34.08 1.54 Intermediate Shell V

0.23 0.60 11.86 7.12 0.36 Plate R2507-1 Y

1.21 1.05 35.26 37.02 1.10 (Transverse)

U 2.40 1.24 40.18 49.82 1.54 SUM:

173.53 6.00 CFR257- = X(FF

• RTNDT) +

'( FF2 ) = (173.53) + (6.00) = 28.9°F Surveillance Weld V

0.23 0.60

-7.6

-4.64 0.30 Material"a)

Y 1.21 1.05 4.08 4.28 1.10 U

2.40 1.24 20.64 25.59 1.54 SUM:

25.23 3.00 CF s., we, = YX(FF

• RTNT,) + X( FF2) = (25.23)-

(3.01)= 8.4'F Notes:

(a) f= fluence. See Table 6.2-3, Ix 109 n/cm', E > 1.0 MeV].

(b)

FF = fluence factor = e0-23.-io.

(c)

ARTNDT values are the measured 30 ft-lb shift values taken from Appendix C, herein ['F].

The scatter of ARTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table D-2.

Appendix D WCAP-16093-NP, Rev. 2 July 2007

D-4 Table D-2:

South Texas Unit 2 Surveillance Capsule Data Scatter about the Best-Fit Line for Surveillance Forging Materials.

Table D-2 indicates that no data point falls outside the +/- Icr of 17'F scatter band for the intermediate shell plate R2507-1 surveillance data. In addition, no data points fall outside the +/- !i of 287F scatter band for the surveillance weld data. Therefore, the intermediate shell plate R2507-1 and the weld data is deemed credible per the third criterion.

Appendix D WCAP-16093-NP, Rev. 2 July 2007

D-5 Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25°F.

The capsule specimens are located in the reactor between the core barrel and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the neutron pad.

The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25°F. Hence, this criterion is met.

Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.

The South Texas Unit 2 surveillance program does not contain correlation monitor material. Therefore, this criterion is not applicable to the South Texas Unit 2 surveillance program.

CONCLUSION:

Based on the preceding responses to all five criteria of Regulatory Guide 1.99, Revision 2, Section B and 10 CFR 50.61, the South Texas Unit 2 surveillance plate and weld data is credible.

Appendix D WCAP-16093-NP, Rev. 2 July 2007