Text

Rev. 0 to WCAP-16346-NP, Comanche Peak Units 1 & 2 Heatup & Cooldown Limit Curves for Normal Operation.
	ML062540320
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Site:	Comanche Peak
Issue date:	10/31/2004
From:	Hayes E, Laubham T; Westinghouse
To:	; Office of Nuclear Reactor Regulation
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	WCAP-16346-NP, Rev 0
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Westinghouse Non-Proprietary Class 3 WCAP-16346-NP October 2004 Revision 0 Comanche Peak Units I and 2 Heatup and Cooldown Limit Curves for Normal Operation OWestinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16346-NP, Revision 0 Comanche Peak Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation "T.J. Laubharn E.T. Hayes NOVEMBER 2004 Manager ent Design & Analysis Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355 02004 Westinghouse Electric Company LLC All Rights Reserved

WCAP- 16346-NII ii WCAP- I6346-NI' ii PREFACE This report has been technically reviewed and verified by:

C.M. Burton _ _ _ _

RECORD OF REVISION Revision 0: Original Issue

WCAP- 16346-NP iii TABLE OF CONTENTS LIST O F TAB LE S ................................................................................................................................. iv LIST O F FIGU R ES ............................................................................................................................... vi EXECUTIVE

SUMMARY

.................................................................................................................... vii 1 IN T RO D U C T IO N ...................................................................................................................... 1 2 FRACTURE TOUGHNESS PROPERTIES ................................................................................ 2 3 RADIATION ANALYSIS AND NEUTRON DOSIMETRY (UNIT 1)..................................... 9 4 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS ............ 30 5 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE ..................................... 34 6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES .................. 42 7 REFER EN C ES ......................................................................................................................... 52 APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS ....................................................................... A-1 APPENDIX B THERMAL STRESS INTENSITY FACTORS ................................................... B-1

WCAP-16346-NP iv LIST OFTABLES Table 2-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RT -DTValues for the Comanche Peak Units I and 2 Reactor Vessel Materials ................................... 3 Table 2-2 Calculated Integrated Neutron Exposure of the Surveillance Capsules

@ Comanche Peak Units I and 2 ................................................................................ 5 Table 2-3 Calculation of CF Values using Comanche Peak Unit I Surveillance Capsule Test R esults .................................................................................................................... 6 Table 2-4 Calculation of CF Values using Comanche Peak Unit 2 Surveillance Capsule Test Results .................................................................................................................... 7 Table 2-5 Summary of the Comanche Peak Units I and 2 Reactor Vessel Beltline Material Chemistry Factors... 8 Table 3-1 Calculated Neutron Exposure and Integrated Exposures At The Surveillance C apsule C enter .................................................... ........................................................ 16 Table 3-2 Calculated Azimuthal Variation Of Maximum Exposure Rates and Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface ..................................... 20 Table 3-3 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV) Within The Reactor Vessel Wall .................................................................................................. 24 Table 3-4 Relative Radial Distribution Of Iron Atom Displacements (dpa) Within The Reactor Vessel Wall .................................................................................................. 24 Table 3-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn From Com anche Peak Unit I ............................................................................................. 25 Table 3-6 Calculated Surveillance Capsule Lead Factors ......................................................... 25 Table 5-1 Calculated Neutron Fluence Projections at the Peak Location on the Reactor Vessel Clad/Base Metal Interface (n/cm2 , E > 1.0 MeV) ..................................................... 35 Table 5-2 Summary of the Vessel Surface, 1/4T and 3/4T Fluence Values used for the Generation of the 36 EFPY Heatup/Cooldowvn Curves for Comanche Peak U nits I & 2 .................................................................................................................. 36 Table 5-3 Summary of the 1/4T and 3/4T Fluence Factor Values used for the Generation of the 36 EFPY Heatup/Cooldown Curves for Comanche Peak Units 1 & 2........................ 36

WCAP- 16346-NP V LIST OF TABLES - (continued)

Table 5-4 Calculation of the Comanche Peak Unit I ART Values for the 1/4T Location

@ 36 EFPY .................................................................................................................. 37 Table 5-5 Calculation of the Comanche Peak Unit I ART Values for the 3/4T Location

@ 36 E FPY .................................................................................................................. 38 Table 5-6 Calculation of the Comanche Peak Unit 2 ART Values for the 1/4T Location

@ 36 E PY .................................................................................................................. 39 Table 5-7 Calculation of the Comanche Peak Unit 2 ART Values for the 3/4T Location

@ 36 E FPY .................................................................................................................. 40 Table 5-8 Summary of the Limiting ART Values Used in the Generation of the Comanche Peak Units 1 and 2 Heatup/Cooldown Curves ............................................................................ 41 Table 6-1 36 EFPY Heatup Curve Data Points Using 1998 App. G Methodology (w/Kjc, w/o Flange Notch & Uncertainties for Instrumentation Errors) ............................ 48 Table 6-2 36 EFPY Cooldown Curve Data Points Using 1998 App. G Methodology (w/Klc, wv/o Flange Notch & Uncertainties for Instrumentation Errors) ............................ 49 Table 6-3 36 EFPY Heatup Curve Data Points Using 1998 App. G Methodology (w/Kic & Flange Notch, w/o Uncertainties for Instrumentation Errors) ............................ 50 Table 6-4 36 EFPY Cooldown Curve Data Points Using 1998 App. G Methodology (w/Kic & Flange Notch, w/o Uncertainties for Instrumentation Errors) ............................ 51

WCAP- 16346-NP vi LIST OF FIGURES Figure 3-1 Comanche Peak Unit I r,0 Reactor Geometry Span at the Core Midplane with a 12.50 Neutron Pad ................................................................................ 26 with a 200 Neutron Pad .................................................................................. 27 with a 22.5' N eutron Pad ................................................................................ 28 Figure 3-2 Comanche Peak Unit 1 rz Reactor Geometry with Neutron Pad ................................ 29 Figure 6-1 Comanche Peak Units I and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 20, 60 and I00 0 F/hr) Applicable for the First 36 EFPY (w/o the "Flange-Notch" & Margins for Instrumentation Errors) Using 1998 App. G Methodology (wv/K 1 ) ........................................................................................................................ 44 Figure 6-2 Comanche Peak Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to I00'F/hr) Applicable for the First 36 EFPY (wv/o the "Flange-Notch" &

Margins for Instrumentation Errors) Using 1998 App. G Methodology (wv/K1 ) ........................................................................................................................ 45 Figure 6-3 Comanche Peak Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 20, 60 and 100 0F/hr) Applicable for the First 36 EFPY (w/ the "Flange-Notch" but w/o Margins for Instrumentation Errors) Using 1998 App. G Methodology (w /Ki )) ........................................................................................................................ 46 Figure 6-4 Comanche Peak Units I and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 36 EFPY (wi/ the "Flange-Notch" but w/o Margins for Instrumentation Errors) Using 1998 App. G Methodology (w /K 1) ........................................................................................................................ 47

WCAP-16346-NP vii EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure temperature (PT) limit curves for normal operation of the Comanche Peak Units I and 2 reactor vessels.

The PT curves were generated based on the latest available reactor vessel information and updated calculated fluences. The new Comanche Peak Units I and 2 heatup and cooldown pressure-temperature limit curves were generated using the "axial flaw" methodology of 1998 ASME Code,Section XI through the 2000 Addenda, which allows the use of the K1 , methodology. The material %viththe highest adjusted reference temperature (ART) was the Unit I Intermediate Shell Plate R- 1107-1. The PT limit curves were generated for 36 EFPY using heatup rates of 20, 60 and 100°F/hr and cooldown rates of 0, 20, 40, 60 and 100WF/hr. Lastly, two sets of PT Curves are provided, one with the flange notch requirement and one without. These curves can be found in Figures 6-1 through 6-4.

WCAP-16346-NP I 1 INTRODUCTION Fleatup and cooldown limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTNDT, and adding a margin. The unirradiated RTNDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60'F.

RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT (IRTNDT). The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials."11 3 Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTNDT + margins for uncertainties) at the I/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface.

The heatup and cooldown curves documented in this report were generated using the most limiting ART values and the NRC approved methodology documented in WCAP- 14040-NP-A, Revision 4121, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

The purpose of this report is to present the calculations and the development of the Comanche Peak Units I and 2 heatup and cooldown curves for 36 EFPY. This report documents the calculated ART values and the development of the PT limit curves for normal operation. The PT curves herein were generated without instrumentation errors. The PT curves include a hydrostatic leak test limit curve from 2485 psig to 2000 psig, along with the pressure-temperature limits for the vessel flange region per the requirements of 10 CFR Part 50, Appendix G1 -1.

WCAP- 16346-NP 2 2 FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic materials in the reactor coolant pressure boundary are determined in accordance with the NRC Standard Review Plan 141.The beltline material properties of the Comanche Peak Units 1 and 2 reactor vessel are presented in Table 2-1.

Best estimate copper (Cu) and nickel (Ni) weight percent values used to calculate chemistry factors (CF) in accordance with Regulatory Guide 1.99, Revision 2, are provided in Table 2-1. Additionally, surveillance capsule data is available for two capsules already removed from both Comanche Peak reactor vessels. This surveillance capsule data was also used to calculate CF values per Position 2.1 of Regulatory Guide 1.99, Revision 2 in Tables 2-3 and 2-4. These CF values are summarized in Table 2-5.

The Regulatory Guide 1.99, Revision 2 methodology used to develop the heatup and cooldown curves documented in this report is the same as that documented in WCAP-14040, Revision 4. The chemistry factors (CFs) were calculated using Regulatory Guide 1.99 Revision 2, Positions 1.1 and 2.1. Position 1.1 uses the Tables from the Reg. Guide along with the best estimate copper and nickel weight percents, which are presented in Table 2-1. Position 2.1 uses the surveillance capsule data from all capsules withdrawn to date. The fluence values used to determine the CFs in Tables 2-3 and 2-4 are the calculated fluence values at the surveillance capsule locations. Hence, the calculated fluence values were used for all cases.

Included in Table 2-2 are the Calculated Capsule fluence values for Comanche Peak Units I and 2. All capsule fluence values were determined using ENDF/B-VI cross-sections and followed the guidance in Regulatory Guide 1.19019.

It should be noted that in the calculations of chemistry factors in Tables 2-3 and 2-4, the ratio was applied to account for chemistry differences between the vessel weld material and the surveillance weld material.

As far as temperature adjustments are concerned, the Comanche Peak Units I and 2 data does not require any adjustments since it is being applied to their own plants.

WCAP-16346-NP 3 TABLE 2-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTNtur Values for the Comanche Peak Units I and 2 Reactor Vessel Materials Material Description Cu (%),a, Ni(%)(Ba Initial RTNDr(A)

Comanche Peak Unit I Closure Head Flange R-1102-1 --- 0.77 40°F Vessel Flange R-1101-1 --- 0.72 100 F Intermediate Shell Plate R- 1107-1 (b) 0.07 0.62 100 F Intermediate Shell Plate R-I107-2(b) 0.07 0.67 -l00 F Intermediate Shell Plate R- 1107-3(bl 0.06 0.65 1O°F Lower Shell Plate R-1108_1(b) 0.08 0.65 00 F Lower Shell Plate R- 1108-2(b) 0.06 0.60 20OF Lower Shell Plate R- 110 8 -3(bl 0.08 0.65 00 F Beitline Region Weld Metal(') 0.045 0.20 -70WF Surveillance Program Weld Metal(c) 0.04 0.22 Comanche Peak Unit 2 Closure Head Flange R-3802-I --- 0.71 40°F Vessel Flange R-3801-1 --- 0.70 -100 F Intermediate Shell Plate R3807-1 0.06 0.64 -20°F Intermediate Shell Plate R3807-2 0.06 0.64 10°F Intermediate Shell Plate R3807-3 0.05 0.60 -20°F Lower Shell Plate R3816-1 0.05 0.59 -30°F Lower Shell Plate R3816-2 0.03 0.65 OF Lower Shell Plate R3816-3 0.04 0.63 -40°F Intermediate & Lower Shell Longitudinal Welds (d) 0.046 0.059 -50OF Intermediate to Lower Shell Girth Weld (d) 0.046 0.059 -60°F Comanche Peak Unit 2 Surveillance Weld Metal"'l 0.035 0.091 - - -

Notes: See Next Page

WCAP-16346-NP 4 Notes for Table 2-1:

(a) Based on Measured Data.

(b) The Cu & Ni weight percent for all intermediate and lower shell plates were calculated from the average of two data points listed on the Combustion Engineering (CE) Certified Material Test Report (CMTR), which are listed in Reference 5. Note that these are different values than those provided to the NRC from TXU in their 92-01 response (See Reference 6). However, the values listed above will produce an equal or more conservative Chemistry Factor (CF).

(c) All Unit I weld metal was fabricated with weld wire type B4, heat # 88112, flux type Linde 0091, and flux lot number 0145. The best estimate Cu & Ni for the bcltline region welds %astaken from Reference 7, which was originally documented in CE Report NPSD-10391s1.

(d) The Unit 2 surveillance weld was made with the same weld wire and flux as the intermediate to lower shell girth weld (wehl wire heat # 89833,flL" Ijpe Linle 124). The longitudinal welds seams were also made with weld heat # 89833, but with flux type Linde 0091. The best estimate Cu & Ni for the beltline region welds was taken from Reference 7, which was originally documented in CE Report NPSD-1039ts]

WCAP-16346-NP 5 TABLE 2-2 Calculated Integrated Neutron Exposure of the Surveillance Capsules @ Comanche Peak Units I and 2 Capsule I Fhlence Comanche Peak Unit 1(s)

U 3.18 x 1018 n/cm2 , (E > 1.0 MeV)

Y 1.49 x 10'9 n/cm2 , (E > 1.0 MeV)

Comanche Peak Unit 2(")

U 3.15 x 1028 n/cm 2, (E > 1.0 MeV)

X 2.20 x 10t9 n/cm 2, (E > 1.0 MeV)

NOTES:

(a) See Section 3, Table 3-5.

(b) Per WCAP-16277-NPr "].

WCAP- 16346-NP 6 TABLE 2-3 Calculation of CF Values using Comanche Peak Unit I Surveillance Capsule Test Results (a) F = Calculated Fluence (1019 n/cm 2, E > 1.0 MeV). See Table 2-2 (b) FF = Fluence Factor = F(O.28 -0.

log F (c) All available data is from Comanche Peak Unit 11"J. Therefore, no temperature adjustment is required.

(d) The measured ARTNrDrr values for the weld metal have been adjusted by a ratio of 1.04.

(e) The CVGRAPH calculated value is -14.14'F. 0.0°F was used in the calculation for conservatism.

[Note that the CFfrom the previous analysisin Reference 11 was 15.7°Ffor the surveillance lower shell plate and 10.7°Ffor the surveillanceweld. As can be seen above there is only a minor change (i.e., <17)to the CF values. Thus, the credibilityevaluation from the previous analysis remains valid.. .All Unit 1 surveillance data is credible.]

WCAP-16346-NP 7 WCAP-l 6346-NP 7 TABLE 2-4 Calculation of CF values using Comanche Peak Unit 2 Surveillance Capsule Data Capsule ea) FFWa' ARTNDT(C) FF x ARTNDT FF2 Material Inter. Shell R3807-2 U 0.315 0.683 1.6 1.093 0.466 (Longitudinal) X 2.20 1.21 1.6 1.94 1.46 Inter. Shell R3807-2 U 0.315 0.683 23.4 15.982 0.466 (Transverse) X 2.20 1.21 52.9 64.01 1.46 SUM 83.025 3.852 CFR3 8O7-2 = X( FF x ARTNDT) + Z( FF 2) = 83.025 -f- 3.852 = 21.6°F Weld Metal U 0.315 0.683 3.74(d) 2.55 0.466 (Heat # 89833) X 2.20 1.21 50.13 (d) 60.66 1.46 SUM 63.21 1.926 CFwV:LD = Y2( FF x ARTN.I,) + Z( FF2) = 63.21 + 1.926 = 32.81F Notes:

(a) F = Calculated Fluence. Units are x 1019 n/cm2 (E > 1.0 MeV). See Table 2-2.

(b) FF = Fluence Factor = fo°28'0'log0.

(c) All available data is from Comanche Peak Unit 21'°1. Therefore, no temperature adjustment is required.

(d) The measured ARTNDT values for the weld metal have been adjusted by a ratio of 1.04.

WCAP-16346-NP 8 TABLE 2-5 Summary of the Comanche Peak Units I and 2 Reactor Vessel Beltline Material Chemistry Factors Material Reg. Guide 1.99, Rev. 2 Reg. Guide 1.99, Rev. 2 Position 1.1 CF's Position 2.1 CF's Comanche Peak Unit 1 Intermediate Shell Plate R-1107-1 440 F - - -

(Heat # C4021-1)

Intermediate Shell Plate R-l 107-2 440 F ---

(Heat # B7854-1)

Intermediate Shell Plate R-l 107-3 370 F ---

(Heat # C4106-2)

Lower Shell Plate R-1 108-1 (Heat # C4464-1) 5IF -- -

Lower Shell Plate R-1108-2 (Heat # C4533-2) 370 F 16.1 OF Lower Shell Plate R-1 108-3 (Heat # C4589-1) 51OF ---

All Beltline Region Welds 46 0 F 11.5 0 F (Heat # 88112)

Comanche Peak Unit 2 Intermediate Shell Plate R3807-1 370 F (Heat # C5522-1)

Intermediate Shell Plate R3807-2 370 F 21.6 0 F (Heat # C5522-2)

Intermediate Shell Plate R3807-3 31F - - -

(Heat # B9566-1)

Lower Shell Plate R3816-1 (Heat # NR64435-1) 31°F -- -

Lower Shell Plate R3816-2 (Heat # NR64439-1) 20OF ---

Lower Shell Plate R3816-3 (Heat # NR64443-1) 26 0 F ---

Interniediate & Lower Shell Longitudinal Welds 31.5 0 F 32.8 0 F (tHeat # 89833)

Intermediate to Lower Shell Girth Weld 31.5 0 F 32.8 0 F (Heat # 89833)

WCAP- 16346-NP 9 WCAP-1 6346-NP 9 3 RADIATION ANALYSIS AND NEUTRON DOSIMETRY (UNIT 1)

3.1 INTRODUCTION

This section describes a discrete ordinates S,, transport analysis performed for the Comanche Peak Unit 1 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. In addition, neutron dosimetry sensor sets from the first two surveillance capsules withdrawn from the Comanche Peak Unit 1 reactor 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 (EFPY).

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 of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."191 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 I-leatup and Cooldown Limit Curves," May 2004.121

WCAP-16346-NP 10 3.2 DISCRETE ORDINATES ANALYSIS 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 58.5', 610, 121.50, 238.50, 2410, and 301.50. These full core positions correspond to the following octant symmetric locations represented in Figure 3-1: 29' from the core cardinal axes (for the 61' and 2410 dual surveillance capsule holder locations found in octants with a 22.50 neutron pad segment) and 31.50 from the core cardinal axes (for the 121.50 and 301.5' single surveillance capsule holder locations found in octants with a 20.0' neutron pad segment, and for the 58.50 and the 238.5' dual surveillance capsule holder locations found in octants with a 22.50 neutron pad segment). The stainless steel specimen containers are 1.1 82-inch by 1-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 feet of the 12-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 thenselves must be included in the analytical model.

In performing the fast neutron exposure evaluations for the Comanche Peak Unit 1 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:

z) 9(Pr, 0, z) = yp(r,0)

p(r, where (p(r,0,z) is the synthesized three-dimensional neutron flux distribution, q)(r,0) is the transport solution in r,O geometry, (p(rz) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and p(r) 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 Comanche Peak Unit 1.

For the Comanche Peak Unit 1 transport calculations, the r,0 models depicted in Figure 3-1 were utilized since, with the exception of the neutron pads, the reactor is octant symmetric. These r,0 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 29' 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 downcomer 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 rO reactor models consisted of 183 radial by 99 azimuthal intervals. Mesh sizes were chosen to assure that

WCAP- 16346-NP 11 proper convergence of the inner iterations was achieved on a point-wise basis. The point-wise inner iteration flux convergence criterion utilized in the r,0 calculations was set at a value of 0.001.

The rz model used for the Comanche Peak Unit I calculations is shown in Figure 3-2 and extends 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 below the lower core plate to above the upper core plate. As in the case of the r,0 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 rz geometric mesh description of these reactor models consisted of 153 radial by 188 axial intervals. As in the case of the rO calculations, mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a point-wise basis.

The point-wise inner iteration flux convergence criterion utilized in the rz calculations was also set at a value of 0.00 1.

The one-dimensional radial model used in the synthesis procedure consisted of the same 153 radial mesh intervals included in the rz model. Thus, radial synthesis factors could be determined on a mesh-wise basis throughout the entire geometry.

The core power distributions used in the plant specific transport analysis were provided by TXU Electric for each of the first ten fuel cycles at Comanche Peak Unit 1[14-181. Specifically, the data utilized included cycle dependent fuel assembly initial enrichments, bum-ups, 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 burn-up 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 E193 and the BUGLE-96 cross-section library.120'1 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 S 16 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.

Selected results from the neutron transport analyses are provided in Tables 3-1 through 3-6. In Table 3-1, 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, 31.50 dual capsule, and 31.50 single capsule. These results, representative of the axial midplane 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 Table 3-2 for the reactor vessel inner radius at five azimuthal locations. The vessel data given

WCAP- 16346-NP 12 in Table 3-2 were taken at the clad/base metal interface, and thus, represent maximum calculated exposure levels on the vessel.

From the data provided in Table 3-2 it is noted that the peak clad/base metal interface vessel fluence (E > 1.0 MeV) at the end of the tenth fuel cycle (i.e., after 11.69 EFPY of plant operation) was 7.26x 10' n/cm 2 .

Both calculated fluence (E > 1.0 MeV) and dpa data are provided in Table 3-1 and Table 3-2. These data tabulations include both plant and fuel cycle specific calculated neutron exposures at the end of the tenth fuel cycle as well as future projections to 15, 20, 25, 32, 36, 48, and 54 EFPY. The calculations for Cycle 10 account for an uprate from 3411 MWt to 3458 MWt that occurred at the onset of cycle ten. 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 3458 MWt.

Radial gradient information applicable to fast (E > 1.0 MeV) neutron fluence and dpa are given in Tables 3-3 and 3-4, respectively. The data, based on maximum cumulative integrated exposures at the end of Cycle 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 3-3 and 3-4.

The calculated fast neutron exposures for the two surveillance capsules withdrawn from the Comanche Peak Unit I reactor are provided in Table 3-5 (also shown in Table 3-1 under the "Dual 31.5°" Column for Capsule X and under the "29' Dual" Column for Capsule Y). These assigned neutron exposure levels are based on the plant and fuel cycle specific neutron transport calculations performed for the Comanche Peak Unit 1 reactor.

Updated lead factors for the Comanche Peak Unit 1 surveillance capsules are provided in Table 3-6. 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 the pressure vessel clad/base metal interface. In Table 3-6, the lead factors for capsules that have been withdrawn from the reactor (U and Y) 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 (V, NV, X, and Z), the lead factor corresponds to the calculated fluence values at the end of Cycle 10, the last completed fuel cycle for Comanche Peak Unit 1.

WCAP- 16346-NP 13 3.3 NEUTRON DOSIMETRY The validity of the calculated neutron exposures previously reported in Section 3.2 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 Y, that was withdrawn from Comanche Peak Unit I at the end of the sixth fuel cycle, is summarized below.

Reaction Rates (rps/atom) MI/C Reaction Ratio Measured Calculated 63Cu(n,a) 60Co 4.77E-17 4.166E-17 1.15 54Fe(n,p) 54 Mn 4.78E-15 4.55E-15 1.05 SSNi(np)"SCo 6.51E-15 6.37E-15 1.02 8 U(n,p)' 37Cs (Cd) 2- 2.93E-14 2.43E-14 1.21 237 Np(n,f)137Cs (Cd) 2.57E-13 2.36E-13 1.09 Average: 1.10

% Standard Deviation: 6.7 The measured-to-calculated (M/C) reaction rate ratios for the Capsule Y threshold reactions range from 1.02 to 1.21, and the average M/C ratio is 1.10 +/- 6.7% (l). 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 Comanche Peak Unit I reactor. These comparisons validate the current analytical results described in Section 3.2; therefore, the calculations are deemed applicable for Comanche Peak Unit 1.

WCAP-1 6346-NP 14 3.4 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Comanche Peak Unit I 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:

1- 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 Comanche Peak Unit 1 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 Comanche Peak Unit I analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Comanche Peak Unit 1 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 3.2. As such, the validation of the Comanche Peak Unit I 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 2.

WCAP-1 6346-NP 15 Capsule Vessel IR PCA Comparisons 3% 3%

It. 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 Comanche Peak Unit 1.

WCAP-1 6346-NP 16 TABLE 3-1 Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cycle Cycle Cumulative Irradiation Neutron Flux (E > 1.0 McV)

Length Time [n/cm2 -s]

IEFPS] [EFPSI IEFPYi Dual Dual Single 290 31.50 31.50 1 2.89E+07 2.89E+07 0.91 1.02E+11 1.1OE+1l 1.09E+11 2 2.43E+07 5.32E+07 1.68 6.54E+10 7.3113+10 7.25E+10 3 2.30E+07 7.62E+07 2.41 7.22E+10 8.07E+10 8.OOE+10 4 3.70E+07 1.13E+08 3.59 8.39E+10 9.19E+10 9.11E+10 5 4.24E+07 1.56E+08 4.93 7.03E+10 7.19E+10 7.111E+10 6 4.14E+07 1.97E+08 6.24 6.45E+10 7.19E+10 7.1313+10 7 4.41E+07 2.41E+08 7.64 6.82E+10 7.2513+10 7.18E+10 8 4.37E+07 2.85E+08 9.02 6.40E+10 6.95E+10 6.88E+10 9 4.3913+07 3.29E+08 10.42 6.98E+10 7.14E+10 7.05E+10 10 4.011E+07 3.69E+08 11.69 7.76E+10 7.98E+10 7.89E+10 Future 1.05E+08 4.73E+08 15.00 7.76E+10 7.9813+10 7.89E+10 Future 1.58E+08 6.31E+08 20.00 7.76E+10 7.9813+10 7.89E+10 Future 1.58E+08 7.89E+08 25.00 7.76E+10 7.98E+10 7.89E+10 Future 2.21E+08 1.011E+09 32.00 7.76E+10 7.98E+10 7.89E+10 Future 1.26E+08 1.14E+09 36.00 7.76E+10 7.98E+10 7.89E+10 Future 3.79E+08 1.5113+09 48.00 7.76E+10 7.9813+10 7.89E+10 Future 1.89E+08 1.70E+09 54.00 7.76E+10 7.9813+10 7.89E+10 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

WCAP- 16346-NP 17 WCAP- 16346-NP 17 TABLE 3-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cycle Cycle Cumulative Irradiation Neutron Fluence (E > 1.0 MeV)

Length Time [n/cmnl

[EFPSI [EFPSI IEFPY] Dual Dual Single 290 31.50 31.50 I 2.89E+07 2.89E+07 0.91 2.93E+18 3.18E+18 3.15E+18 2 2.43E+07 5.32E+07 1.68 4.52E+18 4.96E+18 4.9113+18 3 2.30E+07 7.62E+07 2.41 6.18E+18 6.82E+18 6.76E+18 4 3.70E+07 1.13E+08 3.59 9.28E+18 1.02E+19 1.01E+19 5 4.24E+07 1.56E+08 4.93 1.23E+19 1.33E+19 1.31E+19 6 4.14E+07 1.97E+08 6.24 1.49E+19 1.62E+19 1.6113+19 7 4.41E+07 2.411E+08 7.64 1.79E+19 1.94E+19 1.93E+19 8 4.37E+07 2.85E+08 9.02 2.07E+19 2.25E+19 2.23E+19 9 4.39E+07 3.29E+08 10.42 2.38E+19 2.56E+19 2.54E+19 10 4.01E+07 3.69E+08 11.69 2.69E+19 2.88E+19 2.85E+19 Future 1.05E+08 4.73E+08 15.00 3.50E+19 3.72E+19 3.68E+19 Future 1.58E+08 6.31E+08 20.00 4.73E+19 4.98E+19 4.92E+19 Future 1.58E+08 7.89E+08 25.00 5.951E+19 6.24E+19 6.17E+19 Future 2.211E+08 1.011E+09 32.00 7.67E+19 8.OOE+19 7.91E+19 Future 1.26E+08 1.14E+09 36.00 8.651E+19 9.01E+19 8.911E+19 Future 3.79E+08 1.51E+09 48.00 1.16E+20 1.20E+20 1.1913+20 Future 1.89E+08 1.70E+09 54.00 1.311E+20 1.35E+20 1.34E+20 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

WCAP- 16346-NP is WCAP- 16346-NP 18 TABLE 3-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cycle Cycle Cumulative Irradiation Iron Atom Displacement Rate Length Time [dpa/sl JEFPS] IEFPSI JEFPYj Dual Dual Single 290 31.50 31.50 I 2.89E+07 2.89E+07 0.91 1.99E-10 2.1613-10 2.1413-10 2 2.43E+07 5.32E+07 1.68 1.27E-10 1.42E-10 1.411E-10 3 2.30E+07 7.62E+07 2.41 1.40E-10 1.5613-10 1.55E-10 4 3.70E+07 1.1313+08 3.59 1.63E-10 1.78E-10 1.77E-10 5 4.24E+07 1.56E+08 4.93 1.37E-10 1.40E-10 1.38E-10 6 4.14E+07 1.9713+08 6.24 1.2613-10 1.40E-10 1.39E-10 7 4.411E+07 2.41E+08 7.64 1.33E-10 1.41E-10 1.39E-10 8 4.37E+07 2.85E+08 9.02 1.2413-10 1.3513-10 1.33E-10 9 4.39E+07 3.2913+08 10.42 1.3613-10 1.39E-10 1.37E-10 10 4.01E+07 3.69E+08 11.69 1.51E-10 1.55E-10 1.53E-10 Future 1.05E+08 4.73E+08 15.00 1.51E-10 1.55E-10 1.53E-10 Future 1.58E+08 6.31E+08 20.00 1.51E-10 1.55E-10 1.53E-10 Future 1.58E+08 7.89E+08 25.00 1.5113-10 1.55E-10 1.53E-10 Future 2.21E+08 1.011E+09 32.00 1.5113-10 1.55E-10 1.53E-10 Future 1.26E+08 1.1413+09 36.00 1.51E-10 1.55E-10 1.53E-10 Future 3.79E+08 1.51E+09 48.00 1.5113-10 1.5513-10 1.5313-10 Future 1.8913+08 1.70E+09 54.00 1.51E-10 1.55E-10 1.53E-10 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

WCAP-16346-NP 19 WCAP-1 6346-NP 19 TABLE 3-1 cont'd Calculated Neutron Exposure Rates And Integrated Exposures At The Surveillance Capsule Center Cycle Cycle Cumulative Irradiation Iron Atom Displacements Length Time _dpa]

JEFPSJ IEFPS] IEFPYJ Dual Dual Single 290 31.50 31.50 1 2.89E+07 2.89E+07 0.91 5.74E-03 6.24E-03 6.17E-03 2 2.43E+07 5.32E+07 1.68 8.83E-03 9.68E-03 9.59E-03 3 2.30E+07 7.62E+07 2.41 1.21 E-02 1.33E-02 1.32E-02 4 3.70E+07 1.1313+08 3.59 1.8113-02 1.99E-02 1.97E-02 5 4.24E+07 1.56E+08 4.93 2.39E-02 2.58E-02 2.55E-02 6 4.14E+07 1.97E+08 6.24 2.911E-02 3.16E-02 3.13E-02 7 4.4 11E+07 2.411E+08 7.64 3.49E-02 3.78E-02 3.7413-02 8 4.3713+07 2.85E+08 9.02 4.04E-02 4.37E-02 4.32E-02 9 4.3913+07 3.29E+08 10.42 4.63E-02 4.98E-02 4.93E-02 10 4.011E+07 3.69E+08 11.69 5.24E-02 5.60E-02 5.54E-02 Future 1.0511+08 4.73E+08 15.00 6.82E-02 7.22E-02 7.1413-02 Future 1.58E+08 6.31E+08 20.00 9.20E-02 9.67E-02 9.55E-02 Future 1.5813+08 7.8913+08 25.00 1.16E-01 1.211E-01 1.20E-01 Future 2.211E+08 1.0113+09 32.00 1.49E-01 1.55E-01 1.54E-01 Future 1.26E+08 1.14E+09 36.00 1.68E-01 1.7513-01 1.7313-01 Future 3.7913+08 1.511E+09 48.00 2.25E-01 2.34E-01 2.311E-01 Future 1.89E+08 1.70E+09 54.00 2.5413-01 2.63E-01 2.60E-01 Note: Neutron exposure values reported for the surveillance capsules are centered at the core midplane.

WCAP-16346-NP 20 WCAP- 16346-NP 20 TABLE 3-2 Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Cycle Cycle Cumulative Irradiation Neutron Flux (E > 1.0 MeV)

Length Time [n/cm 2-sl

_EFPSI IEFPSI IEFPYI 00 150 210 300 450 I 2.89E+07 2.89E+07 0.91 1.42E+10 2.1813+10 2.59E+10 2.5413+10 2.75E+10 2 2.43E+07 5.32E+07 1.68 9.96E+09 1.3813+10 1.58E+10 1.70E+10 1.81E+10 3 2.30E+07 7.62E+07 2.41 1.0513+10 1.45E+10 1.69E+10 1.84E+10 1.94E+10 4 3.70E+07 1.13E+08 3.59 1.05E+10 1.77E+10 2.07E+10 2.111E+10 2.0913+10 5 4.24E+07 1.56E+08 4.93 1.31E+10 1.95E+10 2.10E+10 1.77E+10 1.60E+10 6 4.1413+07 1.9713+08 6.24 1.14E+10 1.43E+10 1.6113+10 1.66E+10 1.85E+10 7 4.41E+07 2.41E+08 7.64 1.38E+10 1.9213+10 2.01E+10 1.76E+10 1.75E+10 8 4.37E+07 2.85E+08 9.02 1.20E+10 1.54E+10 1.70E+10 1.65E+10 1.69E+10 9 4.39E+07 3.29E+08 10.42 1.35E+10 1.9713+10 2.11E+10 1.7813+10 1.53E+10 10 4.011E+07 3.69E+08 11.69 1.29E+10 2.01E+I0 2.24E+10 1.96E+10 1.74E+10 Future 1.0513+08 4.73E+08 15.00 1.2913+10 2.0113+10 2.24E+10 1.9613+10 1.74E+10 Future 1.58E+08 6.311E+08 20.00 1.2913+10 2.0113+10 2.24E+10 1.9613+10 1.74E+10 Future 1.58E+08 7.89E+08 25.00 1.29E+10 2.0113+10 2.24E+10 1.9613+10 1.74E+10 Future 2.2113+08 1.01E+09 32.00 1.2913+10 2.0113+10 2.24E+10 1.9613+10 1.7413+10 Future 1.26E+08 1.14E+09 36.00 1.2913+10 2.011E+10 2.24E+10 1.96E+10 1.74E+10 Future 3.7913+08 1.511E+09 48.00 1.2913+10 2.0113+10 2.24E+10 1.96E+10 1.74E+10 Future 1.8913+08 1.70E+09 54.00 1.29E+10 2.011E+10 2.24E+10 1.9613+10 1.74E+10

WCAP- 16346-NP 21 WCAP-1 6346-NP 21 TABLE 3-2 cont'd Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Clad/Base Metal Interface Cycle Cycle Cumulative Irradiation Neutron Fluence (E > 1.0 MeV)

Length Time _n/cnm]

[EFPSI [EFPSI [EFPYI 00 150 210 300 450 1 2.89E+07 2.8913+07 0.91 4.10E+17 6.2813+17 7.48E+17 7.32E+17 7.93E+17 2 2.43E+07 5.32E+07 1.68 6.45E+17 9.55E+17 1.12E+18 1.1313+18 1.2213+18 3 2.30E+07 7.62E+07 2.41 8.86E+17 1.29E+18 1.51E+18 1.56E+18 1.67E+18 4 3.70E+07 1.1313+08 3.59 1.27E+18 1.94E+18 2.27E+18 2.3413+18 2.44E+18 5 4.24E+07 1.56E+08 4.93 1.82E+18 2.76E+18 3.15E+ 18 3.07E+18 3.1113+18 6 4.1413+07 1.97E+08 6.24 2.3013+18 3.3513+18 3.82E+18 3.76E+18 3.87E+18 7 4.411E+07 2.41E+08 7.64 2.8913+18 4.19E+18 4.69E+18 4.53E+18 4.64E+18 8 4.37E+07 2.85E+08 9.02 3.42E+18 4.86E+18 5.43E+18 5.25E+18 5.38E+18 9 4.39E+07 3.29E+08 10.42 4.0113+18 5.73E+18 6.36E+18 6.03E+18 6.05E+18 10 4.011E+07 3.69E+08 11.69 4.53E+18 6.53E+18 7.26E+18 6.82E+18 6.75E+18 Future 1.05E+08 4.73E+08 15.00 5.8813+18 8.63E+18 9.60E+18 8.8713+18 8.56E+18 Future 1.58E+08 6.311E+08 20.00 7.911E+18 1.1813+19 1.31E+19 1.2013+19 1.1313+19 Future 1.58E+08 7.89E+08 25.00 9.95E+18 1.50E+19 1.67E+19 1.5113+19 1.4013+19 Future 2.211E+08 1.0113+09 32.00 1.2813+19 1.94E+ 19 2.16E+ 19 1.9413+19 1.7913+19 Future 1.2613+08 1.1413+09 36.00 1.44E+19 2.20E+19 2.4513+19 2.19E+19 2.0113+19 Future 3.79E+08 1.511E+09 48.00 1.9313+19 2.96E+19 3.29E+19 2.9313+19 2.67E+19 Future 1.89E+08 1.70E+09 54.00 2.17E+19 3.34E+19 3.72E+19 3.3013+19 2.9913+19

WCAP- 16346-NP 22 WCAP-l 6346-NP 22 TABLE 3-2 cont'd Calculated Azimuthal Variation Of Fast Neutron Exposure Rates And Iron Atom Displacement Rates At The Reactor Vessel Clad/Base Metal Interface Cycle Cycle Cumulative Irradiation Iron Atom Displacement Rate Length Time Idpa/s]

IEFPSI IEFPSJ IEFPYI 00 150 210 300 450 1 2.89E+07 2.8913+07 0.91 2.2113-11 3.3413-11 3.97E--11 3.9113-11 4.35E-11 2 2.43E+07 5.32E+07 1.68 1.5513-11 2.13E-11 2.4313-11 2.6213-11 2.86E- 11 3 2.3013+07 7.62E+07 2.41 1.6313-11 2.24E-11 2.5913-11 2.83E-I1 3.07E-11 4 3.70E+07 1.1313+08 3.59 1.63E-11 2.7213-11 3.17E-11 3.26E-11 3.31E-11 5 4.24E+07 1.56E+08 4.93 2.04E-11 2.99E-11 3.2 IE-11 2.73E-11 2.53E-11 6 4.1413+07 1.9713+08 6.24 1.7613-11 2.2113-11 2.4813-11 2.5713-11 2.93 E-11 7 4.41 E+07 2.41E+08 7.64 2.14E-1 1 2.95E- 11 3.08E-I 1 2.72E-11 2.78E- 1 8 4.37E+07 2.85E+08 9.02 1.87E-11 2.3613-11 2.60E- I 1 2.55E-11 2.68E-11 9 4.3913+07 3.2913+08 10.42 2.0913-11 3.031E-11 3.23E- 11 2.74E-11 2.43E-11 10 4.01E+07 3.69E+08 11.69 2.01E-11 3.0913-11 3.43E-11 3.0213-11 2.7513-11 Future 1.0513+08 4.73E+08 15.00 2.0113-11 3.09E-11 3.43E-11 3.02E-11 2.7513-11 Future 1.58E+08 6.311E+08 20.00 2.0113-11 3.09E-11 3.43E-11 3.02E-11 2.75E- II Future 1.58E+08 7.89E+08 25.00 2.011E-11 3.09E-11 3.43E-1 1 3.02E-11 2.75E-11 Future 2.2113+08 1.01E+09 32.00 2.0113-11 3.09E-1 1 3.431E-11 3.02E--11 2.7513-11 Future 1.26E+08 1.14E+09 36.00 2.01E-I I 3.09E-1 1 3.431E-11 3.0213-11 2.75E-II Future 3.79E+08 1.511E+09 48.00 2.0113-11 3.0913-11 3.431E-11 3.02E-11 2.7513-11 Future 1.8913+08 1.7013+09 54.00 2.01E-11 3.0913-11 3.43E-11 3.02E1-11 2.7513-11

WCAP-16346-NP 23 TABLE 3-2 cont'd Calculated Azimuthal Variation Of Maximum Exposure Rates And Integrated Exposures At The Reactor Vessel Cla&dBase Metal Interface Cycle Cycle Cumulative Irradiation Iron Atom Displacements Length Time [dpal IEFPSI [EFPSI JEFPYI 00 150 210 300 450 1 2.89E+07 2.89E+07 0.91 6.3613-04 9.64E-04 1.1413-03 1.1313-03 1.2513-03 2 2.43E+07 5.32E+07 1.68 1.00E-03 1.4713-03 1.7213-03 1.7513-03 1.93E-03 3 2.30E+07 7.62E+07 2.41 1.38E-03 1.98E-03 2.32E-03 2.40E-03 2.64E-03 4 3.70E+07 1.1313+08 3.59 1.98E-03 2.99E-03 3.4913-03 3.60E-03 3.8613-03 5 4.24E+07 1.56E+08 4.93 2.8413-03 4.24E-03 4.83E-03 4.74E-03 4.911E-03 6 4.14E+07 1.9713+08 6.24 3.5713-03 5.151E-03 5.8613-03 5.81E-03 6.12E-03 7 4.411E+07 2.41E+08 7.64 4.50E-03 6.44E-03 7.20E-03 6.9913-03 7.33E-03 8 4.37E+07 2.851E+08 9.02 5.3 1 E-03 7.47E-03 8.34E-03 8.111E-03 8.511E-03 9 4.39E+07 3.2913+08 10.42 6.2313-03 8.80E-03 9.76E-03 9.3 11E-03 9.5713-03 10 4.011E+07 3.69E+08 11.69 7.0413-03 1.0013-02 1.11E-02 1.0513-02 1.0713-02 Future 1.05E+08 4.73E+08 15.00 9.1313-03 1.3313-02 1.4713-02 1.3713-02 1.3613-02 Future 1.5813+08 6.31 E+08 20.00 1.23E-02 1.81E-02 2.01 E-02 1.8513-02 1.7913-02 Future 1.58E+08 7.89E+08 25.00 1.5513-02 2.3013-02 2.5613-02 2.32E-02 2.22E-02 Future 2.2113+08 1.011E+09 32.00 1.9913-02 2.98E-02 3.31 E-02 2.9913-02 2.8313-02 Future 1.26E+08 1.1413+09 36.00 2.2413-02 3.37E-02 3.7513-02 3.37E-02 3.1813-02 Future 3.79E+08 1.511E+09 48.00 3.0013-02 4.54E-02 5.04E-02 4.5213-02 4.2213-02 Future 1.89E+08 1.7013+09 54.00 3.3813-02 5.13E-02 5.69E-02 5.0913-02 4.7413-02

WCAP-16346-NP 24 TABLE 3-3 Relative Radial Distribution Of Neutron Fluence (E > 1.0 MeV)

Within The Reactor Vessel Wall RADIUS AZIMUTIIAL ANGLE (cm) 00 150 210 300 450 220.11 1.000 1.000 1.000 1.000 1.000 225.59 0.571 0.566 0.564 0.561 0.558 231.06 0.282 0.276 0.274 0.273 0.269 236.54 0.134 0.129 0.128 0.128 0.125 242.01 0.064 0.059 0.058 0.059 0.057 Note: Base Metal Inner Radius 220.11 cm Base Metal 1/4T = 225.59 cm Base Metal 1/2T = 231.06 cm Base Metal 3/4T = 236.54 cm Base Metal Outer Radius = 242.01 cm Note: Relative radial distribution data are based on the maximum cumulative integrated exposures from Cycles I through 10.

TABLE 3-4 Relative Radial Distribution Of Iron Atom Displacements (dpa)

Within The Reactor Vessel Wall RADIUS AZIMUTIIALANGLE (cM) 00 150 210 300 450 220.11 1.000 1.000 1.000 1.000 1.000 225.59 0.644 0.636 0.635 0.637 0.646 231.06 0.392 0.381 0.379 0.384 0.395 236.54 0.239 0.227 0.225 0.231 0.239 242.01 0.144 0.129 0.126 0.133 0.136 Note: Base Metal Inner Radius = 220.11 cm Base Metal 1/4T 225.59 em Base Metal 1/2T = 231.06 cm Base Metal 3/4T = 236.54 cm Base Metal Outer Radius = 242.01 cm Note: Relative radial distribution data are based on the maximum cumulative integrated exposures from Cycles I through 10.

WCAP- 16346-NP 25

~VCAP-16346-NP 25 TABLE 3-5 Calculated Fast Neutron Exposure of Surveillance Capsules Withdrawn from Comanche Peak Unit I TABLE 3-6 Calculated Surveillance Capsule Lead Factors Capsule ID And Location Status Lead Factor(b)

U (31.50 Dual) Withdrawn EOC 1 4.01 Y (29.00 Dual) Withdrawn EOC 6 3.85 V (29.00 Dual) Withdrawn EOC 9(a) 3.74 W (31.5) Single) Withdrawn EOC 9(a) 3.99 X (31.5' Dual) In Reactor 3.97 Z (31.50 Single) In Reactor 3.93 Note:

(a) Capsules were removed during IRF09 and transferred to the spent fuel pool.

(b) Lead factors for capsules remaining in the reactor are based on cycle specific exposure calculations through the last completed fuel cycle, i.e., Cycle 10.

WCAP- 16346-NP 26 FIGURE 3-1 Comanche Peak Unit I rO Reactor Geometry with a 12.50 Neutron Pad Span at the Core Midplane 240-180-E X' 120-60-0- . I I-0 75 150 225 300 R Axis (cm)

WCAP- 16346-NP 27 WCAP-1 6346-NP 27 FIGURE 3-1 (continued)

Comanche Peak Unit I rO Reactor Geometry with a 20.00 Neutron Pad Span at the Core Midplane 240-180-E X* 120-60-0-

0 75 150 225 300 R Axis (cm)

WCAP- 16346-NP 28 WCAP- 16346-NP 28 FIGURE 3-1 (continued)

Comanche Peak Unit I rO Reactor Geometry with a 22.50 Neutron Pad Span at the Core Midplane 240-180-E

..0 x 120-60-0-

0 75 150 225 300 R Axis (cm)

WCAP-16346-NP 29 FIGURE 3-2 Comanche Peak Unit I rz Reactor Geometry with Neutron Pad U

x",

0 75 150 225 300 R Axis (cm)

WCAP- 16346-NP 30 4 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 4.1 OVERALL APPROACH The ASME approach for calculating the allowable limit curves for various heatup and cooldowvn rates specifies that the total stress intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, Kl,, for the metal temperature at that time. K1, is obtained from the reference fracture toughness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code1t 21. The K1, curve is given by the following equation:

K1,= 33.2 + 20.734*e [0.

02 (T- RTND)] (1)

where, K = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This K1, curve is based on the lower bound of static critical K, values measured as a function of temperature on specimens of SA-533 Grade B Classl, SA-508-1, SA-508-2, SA-508-3 steel.

4.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C* Kim + Kit < Kc (2)

where, Ki, = stress intensity factor caused by membrane (pressure) stress Kit = stress intensity factor caused by the thermal gradients Kl = function of temperature relative to the RTNDT of the material C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical

WCAP- 16346-NP 31 For membrane tension, the corresponding K, for the postulated defect is:

Kin, = M. X (pR; / t) (3) where, Mm for an inside surface flaw is given by:

Min = 1.85 for t < 2, Min = 0.92617t for 2<5 ft _<3A64, Mm = 3.21 for F > 3.464 Similarly, Min for an outside surface flaw is given by:

Mil 1.77 for f <2, Mm = 0.893ft- for 2* ft <3.464, Mn = 3.09 for 17 > 3.464 and p = internal pressure, Ri = vessel inner radius, and t = vessel wall thickness.

For bending stress, the corresponding K, for the postulated defect is:

Kib = Mb

Maximum Stress, where Mb is two-thirds of Mm The maximum K, produced by radial thermal gradient for the postulated inside surface defect of G-2120 is Kit = 0.953x10, 3 x CR x t2 5 , where CR is the cooldown rate in *F/hr., or for a postulated outside surface defect, Ki, = 0.753x 10." x HU x t 2.5 , where HU is the heatup rate in °F/hr.

The through-wall temperature difference associated with the maximum thermal K, can be determined from Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from Fig.

G-2214-2 for the maximum thermal K,.

(a) The maximum thermal K, relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(l) and (2).

(b) Alternatively, the K, for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 1/4-thickness inside surface defect using the relationship:

Ki, = (1.0359Co + 0.6322Ci + 0A753C2 + 0.3855C 3 ) *..J*' (4)

WCAP- 16346-NP 32 or similarly, Krr during heatup for a /4-thickness outside surface defect using the relationship:

Ki, = (1.043Co + 0.630C, + 0.481 C2 + 0.401 C3) * -4T. (5) where the coefficients Co, C1, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

a(x) = Co+ Ca(x / a)+ C2(x / a) 2 + C3(x/a) 3 (6) and x is a variable that represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.

Note, that equations 3, 4 and 5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. Therefore, the P-T curve methodology is unchanged from that described in WCAP- 14040-NP-A, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves""2 ] Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.

At any time during the heatup or cooldown transient, K1, is determined by the metal temperature at the tip of a postulated flaw at the I/4T and 3/4T location, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kit, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of Ki, at the l/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K1 , exceeds Kit, the calculated allowable pressure during cooldown will be greater than the steady-state value.

WCAP-16346-NP 33 The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a l/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K1 , for the 1/4T crack during heatup is lower than the KI, for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K1 ¢values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

4.3 CLOSURE hIEAD/VESSEL FLANGE REQUIREMENTS 10 CFR Part 50, Appendix G131 addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psi), which is 621 psig for Comanche Peak Units 1 and 2.

The limiting unirradiated RTNDT of 40'F occurs in both the closure head flanges of the Comanche Peak Units I and 2 reactor vessels, so the minimum allowable temperature of this region is 160°F at pressures greater than 621 psig. This limit is shown in Figures 6-1 through 6-4 wherever applicable.

WCAP- 16346-NP 34 5 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART = Initial RTNDT + ARTNDT + Margin (7)

Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Codet' 31. If measured values of initial RTNI)T for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

ARTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

CF

t&.°'-0.0Iogf) (8)

ARTNDT =

To calculate ARTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.

f1,dpOli x) = f.surf

C (-0.24x) (9) where x inches (vessel beltline thickness is 8.63 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 8 to calculate the ARTNDT at the specific depth.

The Westinghouse Radiation Engineering and Analysis Group evaluated the vessel fluence projections and the results of the calculated peak fluence values at various azimuthal locations on the vessel clad/base metal interface are presented in Table 5-1. The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with methods presented in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves".

Tables 5-2 and 5-3 contain the 1/4T and 3/4T calculated fluences and fluence factors, per the Regulatory Guide 1.99, Revision 2, used to calculate the 36 EFPY ART values for all beltline materials in the Comanche Peak Units I and 2 reactor vessel.

Margin is calculated as, M = 2 Vy +

The standard deviation for the initial RTNDT margin term, is 0, 00 F when the initial RTNur is a measured value, and 17'F when a generic value is available. The standard deviation for the ARTNDr margin term, 0a, is 17'F for plates or forgings, and 8.5'F for plates or forgings when surveillance data is used. For welds, oT is equal to 28'F when surveillance capsule data is not used, and is 14°F (half the value) when credible surveillance capsule data is used. 0a need not exceed 0.5 times the mean value of ARTNDT.

Contained in Tables 5-4 through 5-7 are the calculations of the Unit 1 and Unit 2 36 EFPYART values used for generation of the heatup and cooldown curves.

WCAP- 16346-NP 35 WCAP- 16346-NP 35 TABLE 5-1 Calculated Neutron Fluence Projections at the Peak Location on the Reactor Vessel CladfBase Metal Interface [n/cm2 , (E > 1.0 MeV)]

Azimuthal Location EFPY 00 150 210 300 450 Comanche Peak Unit 1 32.00 1.28E+19 1.94E+19 2.16E-+19 1.94E+19 1.7913+19 36.00 1.4413+19 2.20E÷+19 2.45E+19 2.19E+19 2.0113+19 48.00 1.93E+19 2.9612+19 3.29E+19 2.93E+19 2.67 E+19 54.00 2.17E+ 19 3.34E+19 3.7213+19 3.30E+19 2.99E+ 19 Comanche Peak Unit 2 32.00 1.42E+19 1.91E+19 N/A 1.93E+19 2.03E+ 19 36.00 1.6013+19 2.15E-+19 N/A 2.17E+19 2.29E+19 48.00 2.14E+19 2.8813+19 N/A 2.8913+19 3.07E+19 54.00 j2.411E+19 3.24E+ 19 N/A 3.26E+ 19 3.46E+ 19

WCAP- 16346-NP 36 TABLE 5-2 Summary of the Vessel Surface, 1/4T and 3/4T Fluence Values used for the Generation of the 36 EFPY Heatup/Cooldown Curves for Comanche Peak Units 1 & 2 Vesscl('a Surface(sa1T YT 2 2 2 (n/cm ,E > 1.0 Me) (n/cm ,E > 1.0 McV) (n/cm ,E > 1.0 MeV')

Comanche Peak Unit 1 2.45 x 10I9 1.46 x 10'9 0.52 x 109 Comanche Peak Unit 2 2.29 x 10'" 1.36 x 10'9 0.48 x 10'9 Notes:

(a) The peak fluence from each Unit will be used for all the beltline materials, thus listing each material individually is not necessary.

TABLE 5-3 Summary of the 1/4T and 3/4T Fluence Factor Values used for the Generation of the 36 EFPY lHeatup/Cooldown Curves for Comanche Peak Units I & 2 VesselPa) 1/4T F'b' 1/T FF Feb) /4T 3/41 FF Comanche Peak Unit 1 1.46 x 10'9 1.10 0.52 x 10" 0.82 Comanche Peak Unit 2 1.36 x 10' 9 1.09 0.48 x 10'9 0.80 Notes:

(a) The peak fluence from each Unit %%ill be used for all the beltline materials, thus listing each material individually is not necessary.

(b) Units are (n/cm 2 ,E > 1.0 MeV).

WCAP- 16346-NP 37 TABLE 5-4 Calculation of the Comanche Peak Unit 1 ART Values for the 1/4T Location @ 36 EFPY Material Reg. Guide CF~a) / T FF IRTrnDT(h) ARTNDrn c) MI(d) ARTW) 1.99 Rev. 2 (OF) (oF) (oF) (OF) (OF)

Method Intermediate Rmed0ate Shell Plate Position 1.1 44 1.10 10 48.4 34 92 R-I107-1 Intermediate Shell Plate Position 1.1 44 1.10 -10 48.4 34 72 R-1107-2 Intermediate Shell Plate Position 1.1 37 1.10 10 40.7 34 85 R-1107-3 Lower Shell Plate R-1108-1 Position 1.1 51 1.10 0 56.1 34 90 Position 1.1 37 40.7 34 95 5

Position 2.1 1. 1.10 20 17. 7 Lower Shell Plate R-1 108-2 Position 2.1 16.2 17.82 17 55 Lower Shell Plate R-1108-3 Position 1.1 51 1.10 0 56.1 34 90 All Beltline Region Welds Position 1.1 46 1.10 -70 50.6 50.6 31 (Heat # 88112) Position 2.1 11.5 12.65 12.65 -45 NOTES:

(a) Chemistry Factors taken from Table 2-5.

(b) Initial RTNor values are measured values; see Table 2-1.

(c) ARTNrl = CF

FF (d) Margin = 2*(a"2 +o,0,2)12. Note all Survcillance Data is Credible.

(c) ART = IRTN*,T + ARTN,.T + M (This value was rounded per ASTM E29, using the "Rounding Method".)

WCAP- 16346-NP 38 TABLE 5-5 Calculation of the Comanche Peak Unit I ART Values for the 3/4T Location @ 36 EFPY Material Reg. Guide Cva) Y. T FF IRTD-r(b) ARTADr(M) *1 td ART(e) 1.99 Re'. 2 (OF) (OF) (OF) (OF) (0 F)

Method Intermediate Shell Plate Position 1.1 44 0.82 10 36.08 34 80 R-1 107-1 Intermediate Shell Plate Position 1.1 44 0.82 -10 36.08 34 60 R-1 107-2 Intermediate Shell Plate Position 1.1 37 0.82 10 30.34 30.34 71 R-1 107-3 Lower Shell Plate R-1108-1 Position 1.1 51 0.82 0 41.82 34 76 Position 1.1 37 20 30.34 30.34 81 0.82 Lower Shell Plate R- 1108-2 Position 2.1 16.2 13.28 13.28 47 Lower Shell Plate R-1 108-3 Position 1.1 51 0.82 0 41.82 34 76 All Beltline Region Welds Position 1.1 46 0.82 -70 37.72 37.72 5 (Heat # 88112) Position 2.1 11.5 9.43 9.43 -51 NOTES:

(a) Chemistry Factors taken from Table 2-5.

(b) Initial RTNr values are measured values; see Table 2-1.

(c) ARTNur = CF

FF (d) Margin = 2*(o, +o,2)i*. Note all Survcillance Data is Credible.

(c) ART = IRTNDT + ARTNwr + M (This value was rounded per ASTM E29, using the "Rounding Method".)

WCAP- 16346-NP 39 TABLE 5-6 Calculation of the Comanche Peak Unit 2 ART Values for the 1/4T Location @ 36 EFPY Material Reg. Guide CW" 1/4T FF IRTDtrrIN" ARTNDT*(c) M1*ad ARVeC 1.99 Rev. 2 (OF) (OF) (oF) (OF) (OF)

Method Intermediate Shell Plate R3807-1 Position 1.1 37 1.09 -20 40.33 34 54 Position 1.1 37 10 40.33 34 84 1.09 Intermediate Shell Plate R3807-2 Position 2.1 21.6 23.54 34(f) 68 Intermediate Shell Plate R3807-3 Position 1.1 31 1.09 -20 33.79 33.79 48 Lower Shell Plate R3816-1 Position 1.1 31 1.09 -30 33.79 33.79 38 Lower Shell Plate R3816-2 Position 1.1 20 1.09 0 21.8 21.8 44 Lower Shell Plate R3816-3 Position 1.1 26 1.09 -40 28.34 28.34 17 Intermediate& Lower Shell Position 1.1 31.5 34.34 34.34 19 Longitudinal Welds Position 2.1 32.8 1.09 -50 35.75 28.0(0 14 (Heat # 89833)

Intermediate to Lower Shell Position 1.1 31.5 1.09 -60 34.34 34.34 9 Girth Weld (Heat # 89833) Position 2.1 32.8 35.75 28.0(" 4 NOTES:

(a) Chemistry Factors taken from Table 2-5.

(b) Initial RTNDT values are measured values; see Table 2-1.

(c) ARTNDT = CF

FF (d) Margin=2 *(fi2 +o2) V2.

(e) ART = IRTNDT + ARTNtrr + M (This value was rounded per ASTM E29, using the "Rounding Method'.)

(0 The surveillance plate (Itter. Shell Plate R3807-1) data is not credible, while the surveillance weld data is credibleVIt.

WCAP- 16346-NP 40 TABLE 5-7 Calculation of the Comanche Peak Unit 2 ART Values for the 3/4T Location @ 36 EFPY Material Reg. Guide CFoal - T FF IRTN".r~h) ARTNDTrc) Nj1dl ART(C) 1.99 Rev. 2 (0F) (OF) (OF) (OF) (oF)

Method Intermediate Shell Plate R3807-1 Position 1.1 37 0.80 -20 29.6 29.6 39 Position 1.1 37 29.6 29.6 69 0.80 10 Intermediate Shell Plate R3807-2 Position 2.1 21.6 17.28 34") 61 Intermediate Shell Plate R3807-3 Position 1.1 31 0.80 -20 24.8 24.8 30 Lower Shell Plate R3816-1 Position 1.1 31 0.80 -30 24.8 24.8 20 Lower Shell Plate R3816-2 Position 1.1 20 0.80 0 16.0 16.0 32 Lower Shell Plate R3816-3 Position 1.1 26 0.80 -40 20.8 20.8 2 Intermediate & Lower Shell Position 1.1 31.5 25.2 25.2 0 Longitudinal Welds Position 2.1 32.8 0.80 -50 26.24 26.24(0 2 (Heat # 89833)

Intermediate to Lower Shell Position 1.1 31.5 0.80 -60 25.2 25.2 -10 Girth Weld (Heat # 89833) Position 2.1 32.8 26.24 26.24(0 -8 NOTES:

(a) Chemistry Factors taken from Table 2-5.

(b) Initial RTNDT values are measured values; see Table 2-1.

(c) ARTNDT = CF

FF (d) Margin = 2"(o2 +os)'-.

(e) ART = IRTDr + ARTNDT + M (This value was rounded per ASTM E29, using the "Rounding Method".)

(0 The surveillance plate (litter.Shell Plate R3807-1) data is not credible, while the surveillance weld data is credibIle.

WCAP- 16346-NP 41 Based on review of the ART values for both units, the Intermediate Shell Plate R- 1107-1 from Comanche Peak Unit I is the most limiting material because it has the highest ART value. Contained in Table 5-8 is a summary of the limitingARTs from both units. The values from Unit I are to be used in generation of the Comanche Peak Units I and 2 reactor vessel PT limit curves. These limiting curves will be presented in Section 6.

TABLE 5-8 Summary of the Limiting ART Values Used in the Generation of the Comanche Peak Units I and 2 Heatup/Cooldown Curves EFPY 1/ T Limiting ART 1 T Limiting ART Comanche Peak Unit 1 36* 92 80 Comanche Peak Unit 2 36 84 69 Used in the generation of the Comanche Peak Units 1 and 2 PT Limit Curves as presented in Section 6.

For simplicity, TXU has been operating with common of PT limit curves for both Units since initial startup and TXU would prefer to continue using common curves. The reasoning behind this is two-fold. 1) From temperatures equal to 60'F to 160'F, the allowable pressure will be limited by the "flange-notch" pressure of 621 psig (from 10CFR50 Appendix G). Thus, generating PT limits with a Unit I specific and Unit 2 specific ART value would still result in the same PT limit curve. 2) From temperatures beyond 16097, the advantages of using common of PT limit curves clearly outweighs the negligible increase in the allowable pressure that is mathematically available for the Unit 2 vessel.

As a note, by comparing the approximate allowable pressure of the current PT limits (i.e., use K1 , @ 16 EFPY) for I00°F/hr heatup to the allowable pressure generated herein (@ 36 EFPY) for the same heatup rate, there is already an approximate 450 to 500 psig increase in allowable pressure that is achieve from using the newer, less restrictive, ASME Code methodology (i.e., use of K1 , at 36 EFPY). Thus, there is little added benefit in generating PT limits using ART values that are only 8 and I I *F apart.

WCAP- 16346-NP 42 6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Sections 4 and 5 of this report. This approved methodology is also presented in WCAP- 14040-NP-A, Revision 4.

Figures 6-1 and 6-3 present the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 20, 60 and 100°F/hr applicable for the first 36 EFPY with and without the "Flange-Notch" requirement. These curves were generated using the1998 ASME Code Section XI, Appendix G Figures 6-2 and 6-4 present the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of 0, 20, 40, 60 and 100°F/hr applicable for 36 EFPY with and without the "Flange-Notch" requirement. Again, these curves were generated using thel998 ASME Code Section XI, Appendix G. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit line shown in Figures 6-1 through 6-4. This is in addition to other criteria, which must be met before the reactor is made critical, as discussed below in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 6-1 and 6-3. The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in thel 998 ASME Code Section XI, Appendix G as follows:

1.5 Ki. < K1,

where, Kim is the stress intensity factor covered by membrane (pressure) stress, Kic = 33.2 + 20.734 e°'0° 2 (T-RI'Ntyr)],

T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature.

The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 3. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 407F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for hcatup and cooldown calculated as described in Section 5 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperatures for the in service hydrostatic leak tests for the Comanche Peak Units I and 2 reactor vessel at 36 EFPY is 152*F. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40'F higher than the pressure-temperature limit curve constitutes the limit for core operation for the reactor vessel.

WCAP- 16346-NP 43 Figures 6-1 through 6-4 define all of the above limits for ensuring prevention of non-ductile failure for the Comanche Peak Units I and 2 reactor vessel for 36 EFPY with and without the "Flange-Notch" requirementt31 . The data points used for the heatup and cooldown pressure-temperature limit curves shown in Figures 6-1 through 6-4 are presented in Tables 6-1 through 6-4.

WCAP- 16346-NP 44

~VCAP-16346-NP 44 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shcll Plate R-1 107-I (from Comanche Peak Unit 1)

LIMITING ART VALUES AT 36 EFPY: I/4T, 92-F 3/4T, 80-F 2500 Leak Test Limit lopertim Version:5.2 Run:16045 Operlim 2250 -__ Critical Limit --

20 Deg. Fm/Hr 2000Critical Limit 2000 ___160 Deg. F/Hr -

Unacceptable Operation l Critical Limit 1750 -- _ 100 Deg. F/Hr ___-

1500 1250 1000

.5 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 6-1 Comanche Peak Units 1 and 2 Reactor Coolant System Ileatup Limitations (Ileatup Rates of 20, 60 and 100°F/hr) Applicable for the First 36 EFPY (wlo the "Flange-Notch" & Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/Kc)

WCAP- 16346-NP 45 WCAP-I 6346-NP 45 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Plate R-1107-1 (froln Comanche Peak Unit 1)

LIMITING ART VALUES AT 36 EFPY: 1/4T, 92WF 3/4T, 80OF 2500 SOperlim Version:5.2 Run:16045 Operlm.xls Version: 5.2 2250 [Unacceptable]

Operation 2000

~Acceptable 1750 Operation 1500 1 250 Cooldown Rates, F/Hr 1000 _s-- steady-state

-20

-40

°60

-100 750 500 250

,,,mpT__tu F,', ..

U 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 6-2 Comanche Peak Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100'F/hr) Applicable for the First 36 EFPY (w/o the "Flange-Notch" & Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/Ke,)

WCAP-16346-NP 46 XVCAP- 16346-NP 46 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Plate R-I 107-1 (from Comanche Peak Unit 1)

LIMITING ART VALUES AT 36 EFPY: I/4T, 92-F 3/4T, 80-F 2500 Leak Test Limit [operlim_Version:5.2 Run: 16045 Operlim 5 -- 20 Deg. F/Hr 2000Critical 2000 / " '-60 F/Hrt Deg. Limi -

Heatup Rate 20 Deg. F/Hr Critical Limit 1750 HauRte10 Deg. F/Hr -

60 Deg. F/Hr 1500-i - - - - _ _ _ _ _ _

Heatup Rate 100 Deg. F/Hr Acceptable 151Operation S1000oo 1

aea Onacciponble 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 6-3 Comanche Peak Units I and 2 Reactor Coolant System lleatup Limitations (llcatup Rates of 20, 60 and 100°F/hr) Applicable for the First 36 EFPY (wi/ the "Flange-Notch" but w/o Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/K 1,)

WCAP- 16346-NP 47 WCAP-I 6346-NP 47 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Plate R-l 107-1 (from Comanche Peak Unit 1)

LIMITING ART VALUES AT 36 EFPY: I/4T, 92-F 3/4T, 80-F 2500 1 '. 1*

Ooperlim Version:5.2 Run:16045 Operlim.xls Version: 5.2 1 I "

I

.11__

2250 $i 1--i Unacceptable Operation I 2000 _

1750 4 4-3-4 4 4-I Acceptable L Operation 1500 LI 1250 1000 steady-state ___ ___ ______

(3 750 ii_____0_

500

-10,---- ---- ,

Bo60u r r ,--0--1, r r I, r r ~ I r ~ ~ *-- , r 250 0 I 0 50 100 1 50 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 6-4 Comanche Peak Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 36 EFPY (w/ the "Flange-Notch" but w/o Margins for Instrumentation Errors) Using 1998 App. G Methodology (w/K1 *)

WCAP- 16346-NP 48 WCAP- 16346-NP 48 TABLE 6-1 36 EFPY Heatup Curve Data Points Using 1998 App. G Methodology (w/Kit, w/o Flange Notch & Uncertainties for Instrumentation Errors) 20 Ileatup Critical. Limit 60 lleatup Critical. Limit 100 Hleatup Critical. Limit T P T P T P T P T P T P (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig)

)Sg I.

60 0 152 0 60 0 152 0 60 0 152 0 60 808 152 808 60 808 152 828 60 803 152 803 65 829 152 852 65 828 152 831 65 803 152 803 70 852 152 878 70 828 152 832 70 803 152 806 75 878 152 906 75 828 152 838 75 803 152 807 8o 906 152 938 80 831 152 841 8o 803 152 814 85 938 152 972 85 838 152 851 85 803 152 815 90 972 152 1011 90 851 152 868 90 803 152 824 95 1011 152 1053 95 868 152 890 95 806 152 827 100 1053 152 1100 100 890 152 916 100 814 152 839 105 1100 152 1152 105 916 152 947 105 824 155 857 110 1152 155 1208 110 947 155 982 110 839 160 879 115 1208 160 1266 115 982 160 1023 115 857 165 905 120 1266 165 1329 120 1023 165 1069 120 879 170 935 125 1329 170 1399 125 1069 170 1120 125 905 175 970 130 1399 175 1477 130 1120 175 1178 130 935 180 1009 135 1477 180 1562 135 1178 180 1242 135 970 185 1054 140 1562 185 1657 140 1242 185 1314 140 1009 190 1104 145 1657 190 1761 145 1314 190 1393 145 1054 195 1160 150 1761 195 1877 150 1393 195 1481 150 1104 200 1223 155 1877 200 2004 155 1481 200 1579 155 1160 205 1293 160 2004 205 2145 160 1579 205 1687 160 1223 210 1370 165 2145 210 2300 165 1687 210 1806 165 1293 215 1457 170 2300 215 2472 170 1806 215 1938 170 1370 220 1552 175 2472 175 1938 220 2084 175 1457 225 1658 180 2084 225 2245 180 1552 230 1775 185 2245 230 2423 185 1658 235 1904 190 2423 190 1775 240 2047 195 1904 245 2205 200 2047 250 2379 205 2205 250 2459 210 2379 Leak Test Limit Temp. (OF) 135 2000 1 Pressure (psig) 152 2485 _ I I I

WCAP- 16346-NP 49 WCAP-1 6346-NP 49 TABLE 6-2 36 EFPY Cooldown Curve Data Points Using 1998 App. G Methodology (wv/Kic, w/o Flange Notch & Uncertainties for Instrumentation Errors)

Steady State 20'F/hr. 401F/hr. 60 °F/hr. 100 0F/hr.

T (OF) P (psig) T (OF) P (psig) T (OF) P (psig) T (IF) P (psig) T (OF) P (psig) 60 0 60 0 60 0 60 0 60 0 60 808 60 775 60 744 60 713 60 658 65 829 65 798 65 769 65 741 65 691 70 852 70. 824 70 797 70 771 70 729 75 878 75 852 75 827 75 805 75 770 80 906 80 883 80 862 80 843 80 817 85 938 85 917 85 899 85 885 85 868 90 972 90 955 90 941 90 931 90 925 95 1011 95 997 95 988 95 982 95 982 100 1053 100 1044 100 1039 100 1039 100 1039 105 1100 105 1095 105 1095 105 1095 105 1095 110 1152 110 1152 110 1152 110 1152 110 1152 115 1209 115 1209 115 1209 115 1209 115 1209 120 1272 120 1272 120 1272 120 1272 120 1272 125 1342 125 1342 125 1342 125 1342 125 1342 130 1419 130 1419 130 1419 130 1419 130 1419 135 1504 135 1504 135 1504 135 1504 135 1504 140 1599 140 1599 140 1599 140 1599 140 1599 145 1703 145 1703 145 1703 145 1703 145 1703 150 1818 150 1818 150 1818 150 1818 150 1818 155 1946 155 1946 155 1946 155 1946 155 1946 160 2086 160 2086 160 2086 160 2086 160 2086 165 2242 165 2242 165 2242 165 2242 165 2242 170 2414 170 2414 170 2414 170 2414 170 2414

WCAP- 16346-NP 50 TABLE 6-3 36 EFPY Hleatup Curve Data Points Using 1998 App. G Methodology (w/Kic & Flange Notch, w/o Uncertainties for Instrumentation Errors) 20 fleatup Critical. Limit 60 Ileatup Critical. Limit 100 Ileatup Critical. Limit T P T P T P T P T P T P (OF) (psig) ('F) (psig) (of) (psi (OF) (psig) (OF) (psig) (OF) (psig) 60 0 151 0 60 0 151 0 60 0 151 0 60 621 151 621 60 621 151 621 60 621 151 621 65 621 151 621 65 621 151 621 65 621 151 621 70 621 151 621 70 621 151 621 70 621 151 621 75 621 151 621 75 621 151 621 75 621 151 621 80 621 151 621 80 621 151 621 80 621 151 621 85 621 151 621 85 621 151 621 85 621 151 621 90 621 151 621 90 621 151 621 90 621 151 621 95 621 151 621 95 621 151 621 95 621 151 621 100 621 151 621 100 621 151 621 100 621 151 621 105 621 151 621 105 621 151 621 105 621 151 621 110 621 155 621 110 621 155 621 110 621 155 621 115 621 160 621 115 621 160 621 115 621 160 621 120 621 165 621 120 621 165 621 120 621 165 621 125 621 170 621 125 621 170 621 125 621 170 621 130 621 175 621 130 621 175 621 130 621 175 621 135 621 180 621 135 621 180 621 135 621 180 621 140 621 185 621 140 621 185 621 140 621 185 621 145 621 190 621 145 621 190 621 145 621 190 621 150 621 195 621 150 621 195 621 150 621 195 621 155 621 200 621 155 621 200 621 155 621 200 1223 160 621 200 2004 160 621 200 1579 160 621 205 1293 160 2004 205 2145 160 1579 205 1687 160 1223 210 1370 165 2145 210 2300 165 1687 210 1806 165 1293 215 1457 170 2300 215 2472 170 1806 215 1938 170 1370 220 1552 175 2472 175 1938 220 2084 175 1457 225 1658 180 2084 225 2245 180 1552 230 1775 185 2245 230 2423 185 1658 235 1904 190 2423 190 1775 240 2047 195 1904 245 2205 200 2047 250 2379 205 2205 210 2379 Leak Test Limit Temp. (OF) 135 2000 Pressure (psig) 152 2485 1 1

WCAP- 16346-NP 51 WCAP-1 6346-NP 51 TABLE 6-4 36 EFPY Cooldown Curve Data Points Using 1998 App. G Methodology (w/Kic & Flange Notch, w/o Uncertainties for Instrumentation Errors)

Steady State 20'F/hr. 40'F/hr. 60*F/hr. 100'F/hr.

T( 0 F) P (psig) T (OF) P (psig) T (OF) P (psig) 0 T"( F) P (psig) T (OF) P (psig) 60 0 60 0 60 0 60 0 60 0 60 621 60 621 60 621 60 621 60 621 65 621 65 621 65 621 65 621 65 621 70 621 70 621 70 621 70 621 70 621 75 621 75 621 75 621 75 621 75 621 80 621 80 621 80 621 80 621 80 621 85 621 85 621 85 621 85 621 85 621 90 621 90 621 90 621 90 621 90 621 95 621 95 621 95 621 95 621 95 621 100 621 100 621 100 621 100 621 100 621 105 621 105 621 105 621 105 621 105 621 110 621 110 621 110 621 110 621 110 621 115 621 115 621 115 621 115 621 115 621 120 621 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 130 621 135 621 135 621 135 621 135 621 135 621 140 621 140 621 140 621 140 621 140 621 145 621 145 621 145 621 145 621 145 621 150 621 150 621 150 621 150 621 150 621 155 621 155 621 155 621 155 621 155 621 160 621 160 621 160 621 160 621 160 621 160 2086 160 2086 160 2086 160 2086 160 2086 165 2242 165 2242 165 2242 165 2242 165 2242 170 2414 170 2414 170 2414 170 2414 170 2414

WCAP-16346-NP 52 7 REFERENCES

1. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S.

Nuclear Regulatory Commission, May 1988.

2. WCAP-14040-NP-A, Revision 4, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J.D. Andrachek, et. al., May 2004.

3. Code of Federal Regulations, 10 CFR Part 50, Appendix Q "Fracture Toughness Requirements,"

U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

4. "Fracture Toughness Requirements", Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.

5. Combustion Engineering Certified Material Test Reports for Comanche Peak Unit I Reactor Vessel Intermediate and Lower Shell Plates:

" Intermediate Shell Plate R-1 107-1: Combustion Engineering, Inc., Metallurgical Research and Development Dept., Contract No. 10773, Job No. 732124-001, 11-27-74.

Intermediate Shell Plate R-1 107-2: Combustion Engineering, Inc., Metallurgical Research and Development Dept., Contract No. 10773, Job No. 732124-003, 11/27/74.

Intermediate Shell Plate R-1 107-3: Combustion Engineering, Inc., Metallurgical Research and Development Dept., Contract No. 10773, Job No. 732124-005, 11/20/74.

" Lower Shell Plate R-1 108-1: Combustion Engineering, Inc., Metallurgical Research and Development Dept., Contract No. 10773, Job No. 732142-001, 9-8-1975.

Lower Shell Plate R-1 108-2: Combustion Engineering, Inc., Metallurgical Research and Development Dept., Contract No. 10773, Job No. 732142-003, 9-5-1975.

Lower Shell Plate R-1 108-3: Combustion Engineering, Inc., Metallurgical Research and Development Dept., Contract No. 10773, Job No. 732142-005, 9-8-1975.

6. TU Electric # TXX-94172, "Comanche Peak Steam Electric Station (CPSES) Docket Nos. 50-445 and 50-446 Generic Letter (GL) 92-01, Revision 1, 'Reactor Vessel Structural Integrity,' Texas Utilities Electric Company, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2 (TAC Nos.

M83451 and M83452)," William J. Cahill Jr., Dated June 20, 1994 (Plts Enclosures).

7. TU Electric # CPSES-9706134, "Comanche Peak Steam Electric Station (CPSES) Docket Nos.

50-445 and 50-446 Response to NRC Request for Additional Information Regarding NRC Generic Letter 92-01, Revision 1, Supplement I," C. Lance Terry, Dated December 31, 1997 (Plus Enclosures).

8. CE Report NPSD-1039, Rev. 2, "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds", CEOG Task 902, By the CE Owners Group, June 1997.

9. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.

WCAP- 16346-NP 53

10. WCAP- 16277-NP, "Analysis of Capsule X from the TXU Energy Comanche Peak Unit 2 Reactor Vessel Radiation Surveillance Program," T.J. Laubbam, et. al., September 2004.

11. WCAP- 15144, "Analysis of Capsule Y from the TU Electric Company Comanche Peak Unit I Reactor Vessel Radiation Surveillance Program," T.J. Laubham, et. al., January 1999.

12.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix Q,"Fracture Toughness Criteria for Protection Against Failure." Dated December 1998, through 2000 Addendum.

13. 1989 Section III, Division I of the ASME Boiler and Pressure Vessel Code, Paragraph NB-233 1, "Material for Vessels"

14. WCAP-9806, Revision 3, "The Nuclear Design and Core Physics Characteristics of the Comanche Peak Unit 1 Nuclear Power Plant Cycle 1," April 1990.

15. WCAP-13094, "The Nuclear Design and Core Physics Characteristics of the Comanche Peak Unit I Nuclear Power Plant Cycle 2," December 1991.

16. WCAP-13613, "The Nuclear Design and Core Physics Characteristics of the Comanche Peak Unit 1 Nuclear Power Plant Cycle 3," January 1993.

17. CPSES-9804980, "Reactor Core Data to Support the Westinghouse Evaluation of the CPSES Unit I Reactor Vessel Material Surveillance Specimen (Capsule "Y")," September 1998.

18. CPSES-200401980, "Transmittal of Design Information for Unit 1 Fluence Analysis," August 2004.

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

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

WCAP-1 6346-NP A-I APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS

WCAP- 16346-NP A-2 A.1 NEUTRON DOSINIETRY Comparisons of measured dosimetry results to both the calculated and least squares adjusted values for all surveillance capsules withdrawn from service to date at Comanche Peak Unit 1 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.¶JAl^' 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 3.2 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.I Sensor Reaction Rate Determinations In this section, the results of the evaluations of the two neutron sensor sets withdrawn to date as part of the Comanche Peak Unit I Reactor Vessel Materials Surveillance Program are presented. The capsule designation, location within the reactor, and time of withdrawal of each of these dosimetry sets were as follows:

Azimuthal Withdrawal Irradiation Capsule ID Location Time Time [EFPY]

U 31.50 Dual End of Cycle 1 0.91 Y 29' Dual End of Cycle 6 6.24 The azimuthal locations included in the above tabulation represent the first octant equivalent azimuthal angle of the geometric center of the respective surveillance capsules.

The passive neutron sensors included in the evaluations of Surveillance Capsules U and Y are summarized as follows:

Reaction Sensor Material Of Interest Capsule U Capsule Y Copper 3' Cu(n,a)°UCo X X Iron 4 5 Fc(n,p)5' Mn X X Nickel 5"Ni(n,p)s'*Co X x 3

Uranium-238 2 3U(n,'137Cs X X Neptunium-237 23 7Np(n,0 13 7 Cs X X Cobalt-Aluminum* 3Co(ny') 6UCo X X

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

WCAP-16346-NP A-3 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 tle 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.

Results from the radiometric counting of the neutron sensors from Capsules U and Y are documented in References A-2 and A-3. 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 U (April 1990 -

October 199 1) and Y (April 1990 - March 1998) was based on the monthly power generation of Comanche Peak Unit I from initial reactor criticality 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 Comanche Peak Unit 1 through the end of Cycle 10 (April 1990 - March 2004) 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 F Y 1 1 Cj [ 1 - e-'] [e 4 ',

Pr.t-

WCAP-1 6346-NP A-A where:

R = Reaction rate averaged over the irradiation period and referenced to operation at a core power level of P,,f (rps/nucleus).

A = Measured specific activity (dps/gm).

No = Number of target element atoms per gram of sensor.

F = Atom fraction of the target isotope in the target element.

Y = Number of product atoms produced per reaction.

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

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

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

= Decay constant of the product isotope (1/see).

tj = Length of irradiation periodj (sec).

td = Decay time following irradiation period j (see).

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 [Pj]/[P,,f] 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 3.2, 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, Cj is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those employing low leakage fuel management, tihe additional Cj 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 2 35 U 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 23SU 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 Comanche Peak Unit 1 fission sensor reaction rates are summarized as follows:

WCAP-16346-NP A-5 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 U and Y 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 23SU impurities, plutonium build-in, and gamma ray induced fission effects.

A.1.2 Least Squares Evaluation of Sensor Sets Least squares adjustment methods provide the capability of combining the measurement data wvith the corresponding neutron transport calculations resulting in a Best Estimate neutron energy spectrum with associated uncertainties. Best Estimates for key exposure parameters such as (p(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, Ri - a , = 1"(Ojig +/- 8 .)(.pg _+/- )q,,

relates a set of measured reaction rates, Ri, to a single neutron spectrum, Pg, through the multigroup dosimeter reaction cross-section, ag, each with an uncertainty 5. 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 Comanche Peak Unit 1 surveillance capsule dosimetry, the FERRET codeAI41 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 (T(E > 1.0 MeV) and dpa) along with associated uncertainties for the two in-vessel capsules withdrawn to date.

WCAP- 16346-NP A-6 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 Comanche Peak Unit 1 application, the calculated neutron spectrum was obtained from the results of plant specific neutron transport calculations described in Section 3.2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section A. 1.1.

The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section librarytA-]. The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations byASTM Standard E 1018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E 706 (liB)".

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

The following provides a summary of the uncertainties associated with the least squares evaluation of the Comanche Peak Unit I 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:

Reaction Uncertainty 6

1Cu(n,a)6oCo 5%

'4Fe(n,p)-Mn 5%

58Ni(n,p)58Co 5%

238 U(nf)137 Cs 10%

2-7 Np(n,f)-3 7Cs 10%

5 9Co(ny)°Co 5%

These uncertainties are given at the I Tlevel.

WCAP-16346-NP A-7 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 Comanche Peak Unit 1 surveillance program, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.

Reaction Uncertainty 63Cu(n, a) 6°Co 4.08-4.16%

"Fe(n,p)-4Mn 3.05-3.11%

5SNi(np)SSCo 23 37 4.49-4.56%

8 U(n,f)1 Cs 0.54-0.64%

237 Np(n,0137Cs 10.32-10.97%

59Co(n,-y)6°Co 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 neutron 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:

=gR 2+R *R *PW

WCAP- 16346-NP A-8 where R, specifies an overall fractional normalization uncertainty and the fractional uncertainties R. and Rg. specify additional random groupvise uncertainties that are correlated with a correlation matrix given by.

P9g, = [1D - 0]6g + 0e" where 2

H = (g-g')

2y2 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 Comanche Peak Unit I calculated spectra was as follows:

Flux Normalization Uncertainty (Rj 15%

Flux Group Uncertainties (Rg, Rg.)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 29%

(E < 0.68 eV) 52%

Short Range Correlation (0)

(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Flux Group Correlation Range (y)

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2

WCAP- 16346-NP A-9 A.1.3 Comparisons of Measurements and Calculations Results of the least squares evaluations of the dosimetry from the Comanche Peak Unit I 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 3.4 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 Ia 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 Ylevel.

Further comparisons of the measurement results (from Tables A-5 and A-6) 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 qp(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 0.98 to 1.21 for the 10 samples included in the data set. The overall average M/C ratio for the entire set of Comanche Peak Unit 1 data is 1.10 with an associated standard deviation of 7.1%.

In the comparisons of best estimate and calculated fast neutron exposure parameters, the corresponding BE/C comparisons for the capsule data sets range from 1.05 to 1.08 for neutron flux (E > 1.0 MeV) and from 1.06 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 1.07 with a standard deviation of 1.9% and 1.06 with a standard deviation of 0.8%, respectively.

WCAP-16346-NP A-10 Based on these comparisons, it is concluded that the calculated fast neutron exposures provided in Section 3.2 of this report arc validated for use in the assessment of the condition of the materials comprising the beltline region of the Comanche Peak Unit I reactor pressure vessel.

WCAP- 16346-NP A-11 WCAP-1 6346-NP A-Il TABLE A-I Nuclear Parameters Used In The Evaluation Of Neutron Sensors Target 90% Response Fission Monitor Reaction of Atom Range Product Yield Material Interest Fraction (NIeV) Half-life (%)

Copper 6 3Cu (n,a) 0.6917 4.9-11.9 5.271 y Iron 54Fe (n,p) 0.0585 2.1-8.5 312.1 d Nickel 58Ni (n,p) 0.6808 1.5 - 8.3 70.82 d 238U (n,f) 1.0000 1.3 -6.9 30.07 y 6.02 Uranium-238 2 37 Neptunium-237 Np (n,f) 1.0000 0.3 - 3.8 30.07 y 6.17 59Co (n,y,) 0.0015 non-threshold 5.271 y Cobalt-Aluminum Note: The 90% response range is defined such that, in the neutron spectrum characteristic of the Comanche Peak Unit I 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 upper limit.

WCAP- 16346-NP A-12 WCAP-1 6346-NP A- 12 TABLE A-2 Monthly Thermal Generation During The First Ten Fuel Cycles Of The Comanche Peak Unit I Reactor (Reactor power of 3411 MWt from startup through the end of Cycle 9, and 3458 MWt for Cycle 10)

Thermal Thermal Thermal Generation Generation Generation Year Month (MWIt-hr) Year Month (iMlWt-hr) Year Month (MWt-hr) 1990 4 127135 1993 4 2326657 1996 4 2149764 1990 5 745699 1993 5 2493823 1996 5 2529852 1990 6 521146 1993 6 2240371 1996 6 2457287 1990 7 1688200 1993 7 2534100 1996 7 2530409 1990 8 1515303 1993 8 2522312 1996 8 2324480 1990 9 1480019 1993 9 2443150 1996 9 2449157 1990 10 1911852 1993 10 410302 1996 10 315929 1990 11 1072991 1993 11 0 1996 11 959823 1990 12 2371846 1993 12 794163 1996 12 2536114 1991 1 2265340 1994 1 2536147 1997 1 2530064 1991 2 2053558 1994 2 2062073 1997 2 2278471 1991 3 1376052 1994 3 2490712 1997 3 2529483 1991 4 0 1994 4 2445196 1997 4 2454782 1991 5 262292 1994 5 2459767 1997 5 2531276 1991 6 2436355 1994 6 2449780 1997 6 2450992 1991 7 2399188 1994 7 2531808 1997 7 2536073 1991 8 2522559 1994 8 2506758 1997 8 2531964 1991 9 2431359 1994 9 2450190 1997 9 2449639 1991 10 156278 1994 10 2532215 1997 10 2196887 1991 11 0 1994 11 2362352 1997 11 2452982 1991 12 1234345 1994 12 2358419 1997 12 2387719 1992 1 2252734 1995 1 2531561 1998 1 2531683 1992 2 2169969 1995 2 2292522 1998 2 2287085 1992 3 2082375 1995 3 214563 1998 3 1618720 1992 4 2250441 1995 4 697465 1998 4 95044 1992 5 2371846 1995 5 2399712 1998 5 2455855 1992 6 2001084 1995 6 1934160 1998 6 2450787 1992 7 2125108 1995 7 2529901 1998 7 2533273 1992 8 2514943 1995 8 2415471 1998 8 2536024 1992 9 2439220 1995 9 2451500 1998 9 2451500 1992 10 1591109 1995 10 2494609 1998 10 2532029 1992 11 0 1995 11 2115554 1998 11 2454119 1992 12 138514 1995 12 2412212 1998 12 2532479 1993 1 1995026 1996 1 1426955 1999 1 2531767 1993 2 2269843 1996 2 1989034 1999 2 2287026 1993 3 2434962 1996 3 2524995 1999 3 2536263

WCAP- 16346-NP A-13 WCAP-l 6346-NP A-13 TABLE A-2 cont'd Monthly Thermal Generation During The First Ten Fuel Cycles Of The Comanche Peak Unit I Reactor (Reactor power of 3411 MWt from startup through the end of Cycle 9, and 3458 MWt for Cycle 10)

Thermal Thermal Generation Generation Year Month (MWt-hr) Year Month (MWt-hir) 1999 4 2450541 2002 4 2449924 1999 5 2531243 2002 5 2531606 1999 6 2454339 2002 6 2452827 1999 7 2502755 2002 7 2535591 1999 8 2520184 2002 8 2531416 1999 9 1927102 2002 9 2204280 1999 10 52213 2002 10 0 1999 11 2402258 2002 11 491755 1999 12 2536712 2002 12 1569092 2000 1 2531882 2003 1 2569381 2000 2 2372279 2003 2 2321037 2000 3 2535722 2003 3 1993314 2000 4 2450958 2003 4 2482903 2000 5 2536074 2003 5 2237613 2000 6 2454250 2003 6 2486937 2000 7 2472397 2003 7 2569748 2000 8 2535713 2003 8 2564875 2000 9 2431919 2003 9 2486184 2000 10 2539103 2003 10 2573428 2000 11 2451983 2003 11 2471970 2000 12 2518969 2003 12 2569728 2001 1 2497694 2004 1 2565329 2001 2 2290038 2004 2 2399714 2001 3 1806552 2004 3 2155520 2001 4 516742 2001 5 2535533 2001 6 2453317 2001 7 2532596 2001 8 1547417 2001 9 2453381 2001 10 2538204 2001 11 2451413 2001 12 2533492 2002 1 2536264 2002 2 2286146 2002 3 2535146

WCAP- 1634 6-NP A-14 V/CAP-I 6346-NP A-14 TABLE A-3 Calculated Cj Factors at the Surveillance Capsule Center Core Midplane Elevation Fuel Cycle Length p(E > 1.0 McV) [n/cm-s]

Cycle [EFPSJ Capsule U Capsule Y I 2.89E+07 1.10E+l1 1.02E+11 2 2.43E+07 6.54E+10 3 2.30E+07 7.22E+10 4 3.70E+07 8.39E+ 10 5 4.24E+07 7.03E+10 6 4.14E+07 6.451E+10 Average 1.10E+11 7.58E+10 Fuel Cycle Length C_

Cycle IEFPSI Capsule U Capsules' 1 2.89E+07 1.000 1.340 2 2.4313+07 0.862 3 2.3013+07 0.952 4 3.70E+07 1.106 5 4.2413+07 0.927 6 4.14E+07 0.851 Average 1.000 1.000

WCAP- 16346-NP A-15 TABLE A-4 Measured Sensor Activities And Reaction Rates Surveillance Capsule U Radially Adjusted Measured Saturated Reaction Activit- Activity Rate Reaction Location (dpslg) (dps/g) (rps/atom) 63 Cu (n,'a) 6 °Co Top 4.75E+04 4.60E+05 7.02E-17 Middle 4.43E+04 4.29E+05 6.54E-17 Bottom 4.31E+04 4.17E+05 6.37E-17 Average 6.64E-17 54 54 Fe (n,p) Mn Top 1.24E+06 4.08E+06 6.46E- 15 Middle 1.30E+06 4.27E+06 6.77E- 15 Bottom 1.21 E+06 3.98E+06 6.30E-15 Average 6.51E-15 5 5 "Ni (n,p) 1Co Top 7.80E+06 6.44E+07 9.21E-15 Middle 7.31 E+06 6.03E+07 8.63E-15 Bottom 7.30E+06 6.02E+07 8.62E- 15 Average 8.82 E- 15 8

23U (n,l) 137Cs (Cd) Middle 1.50E+05 7.31E+06 4.80E-14 235 2 Including U, 39Pu, and y fission corrections: 4.04E-14 2 37 Np (n,O) 137Cs (Cd) Middle 1.28E+06 6.24E+07 3.98E-13 Including y fission correction: 3.94E-13 59 Co (n,-y) 6OCo Top 9.47E+06 9.17E+07 5.98E-12 Top 8.18E+06 7.92E+07 5.17E-12 Middle 9.40E+06 9.10E+07 5.94E-12 Middle 7.97E+06 7.72E+07 5.03E- 12 Bottom 9.20E+06 8.91E+07 5.811E-12 Average 5.59E-12 59 Co (ny) 6°Co (Cd) Top 4.85E+06 4.70E+07 3.06E-12 Middle 4.93E+06 4.77E+07 3.11 E- 12 Average 3.09E-12 Notes: 1) Measured specific activities are indexed to a counting date of April 13, 1992.

2) The average 23-U (n,O reaction rate of 4.04E-14 includes a correction factor of 0.872 to account for plutonium build-in and an additional factor of 0.966 to account for photo-fission effects in the sensor.

3) The average 23 7Np (n,f) reaction rate of 3.94E-13 includes a correction factor of 0.990 to account for photo-fission effects in the sensor.

4) Reaction rates referenced to the Cycle I Rated Reactor Power of 3411 MWt.

WCAP- 16346-NP A-16 TABLE A-4 cont'd Measured Sensor Activities And Reaction Rates Surveillance Capsule Y Radially Adjusted Measured Saturated Reaction Activity Activity Rate Reaction Location (dps/g) (dps/g) (rps/atorn) 63 60 Cu (n,'a) Co Top 1.6413+05 3.3813+05 5.16E-17 Middle 1.4613+05 3.011E+05 4.5913-17 Bottom 1.45E+05 2.9113+05 4.56E-17 Average 4.77E-17 54 Fe (n,p) 54Mn Top 1.77E+06 3.19E+06 5.06E-15 Middle 1.61E+06 2.91EE+06 4.61E-15 Bottom 1.63E+06 2.94E+06 4.66E-15 Average 4.78E-15 5

"Ni (n,p) 5"Co Top 7.4913+06 4.8413+07 6.9313-15 Middle 6.7613+06 4.37E-+07 6.2613-15 Bottom 6.84E+06 4.42E+07 6.33E-15 Average 6.51E-I 5 23SU (nf) 37Cs (Cd) Middle 7.28E+05 5.58E+06 3.67E-14 2 239 Including "U, Pu, andy fission corrections: 2.93E-14 37 2 Np (n,f) 137Cs (Cd) Middle 5.3 1E+06 4.0713+07 2.6013-13 Including y fission correction: 2.57E-13 59 Co (n,,y) 61Co Top 2.3213+07 4.7913+07 3.1213-12 Top 2.74E+07 5.6513+07 3.69E-12 Middle 2.1213+07 4.3713+07 2.8513-12 Middle 2.6012+07 5.36E+07 3.50E-12 Bottom 2.5012+07 5.166E+07 3.36E-12 Bottom 2.65E+07 5.47E+07 3.57E- 12 Average 3.35E-12 60 "Co (n,y) Co (Cd) Top 1.46E+07 3.011E+07 1.97E-12 Middle 1.4 1E+07 2.91 E+07 1.9013- 12 Bottom 1.4813+07 3.0513+07 1.9913- 12 Average 1.95E-12 Notes: 1) Measured specific activities are indexed to a counting date of September 8, 1998.

2) The average 2-"U (n,f) reaction rate of 2.9313-14 includes a correction factor of 0.826 to account for plutonium build-in and an additional factor of 0.967 to account for photo-fission effects in the sensor.

3) The average 237Np (n,f) reaction rate of 2.57E-13 includes a correction factor of 0.990 to account for photo-fission effects in the sensor.

4) Reaction rates referenced to the Cycles 1-6 Rated Reactor Power of 3411 MWt.

WCAP- 16346-NP A-17 WCAP-l 6346-NP A- 17 TABLE A-5 Comparison of Measured, Calculated, and Best Estimate Reaction Rates At The Surveillance Capsule Center Capsule U

1~

Reaction Rate lrns/atoml Best Reaction Measured Calculated Estimate M/C MIBE 63Cu(noa) 60Co 6.64E-17 5.63E-17 6.33E-17 1.18 1.05

-'Fe(n,p)54Mn 6.51E-15 6.42E-15 6.70E-15 1.01 0.97 58Ni(n,p)58Co 8.82E-15 9.03E-15 9.29E-15 0.98 0.95 21'U(n,0t 3 7Cs (Cd) 4.04E-14 3.50E-14 3.65E-14 1.15 1.11 237Np(n,f) 3 7 Cs (Cd) 3.94E-13 3.46E-13 3.78E-13 1.14 1.04 59Co(n,y)60Co 5.59E-12 4.99E-12 5.51E-12 1.12 1.01 59Co(n,,y)6°Co (Cd) 3.09E- 12 3.47E- 12 3.14E- 12 0.89 0.98 Note: See Section A. 1.2 for details describing the Best Estimate (BE) reaction rates.

Capsule Y Reaction Rate [rps/atom]

Best Reaction Measured Calculated Estimate M/C M/BE 63Cu(n,a) 60 Co 4.77E-17 4.16E-17 4.63E-17 1.15 1.03 54Fe(n,p) 54Mn 4.78E-15 4.55E-15 4.87E-15 1.05 0.98 5"Ni(n,p) 58Co 6.51E-15 6.37E-15 6.75E-15 1.02 0.96 23 37

.U(n,f)t 237Np(n,f)137 Cs (Cd) 2.93E-14 2.43E-14 2.61E-14 1.21 1.12 Cs (Cd) 2.57E-13 2.36E-13 2.55E-13 1.09 1.01 59Co(n,,y)6°Co 3.35E-12 3.29E-12 3.3 1E-12 1.02 1.01 59Co(n,'y)60Co (Cd) 1.95E-12 2.30E-12 1.98E-12 0.85 0.98 Note: See Section A. 1.2 for details describing the Best Estimate (BE) reaction rates.

WCAP-1 6346-NP A-18 TABLE A-6 Comparison of Calculated and Best Estimate Exposure Rates At The Surveillance Capsule Center V(E> 1.0 MMcI ln/cm2 -sl Best Uncertainty Capsule ID Calculated Estimate Oa) BE/C U 1.10E+I1 1.16E+I 1 6% 1.05 Y 7.58E+10 8.21E+10 6% 1.08 Note: Calculated results arc based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period and are the average neutron exposure over the irradiation period for each capsule. See Section A. 1.2 for details describing the Best Estimate exposure rates.

Note: Calculated results are based on the synthesized transport calculations taken at the core midplane following the completion of each respective capsules irradiation period and are the average neutron exposure over the irradiation period for each capsule. See Section A. 1.2 for details describing the Best Estimate exposure rates.

WCAP-16346-NP A-19 TABLE A-7 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast Neutron Tluheshold Reactions M/C Ratio Reaction Capsule U Capsule Y 63 Cu(n,a)WCo 1.18 1.15 54 Fe(n,p)S4Mn 1.01 1.05 "Ni(n,p) 5 8Co 5 0.98 1.02

-3 8U(n,p)137 237 37 Cs (Cd) 1.15 1.21 Np(n,f' Cs (Cd) 1.14 1.09 Average 1.09 1.10

% Standard Deviation 8.3 6.7 Note: The overall average M/C ratio for the set of 10 sensor measurements is 1. 10 with an associated standard deviation of 7.1%.

TABLE A-8 Comparison of Best Estimate/Calculated (BE/C) Exposure Rate Ratios BE/C Ratio Capsule ID p(E > 1.0 MeV) dpa/s U 1.05 1.06 Y 1.08 1.07 Average 1.07 1.06

% Standard Deviation 1.9 0.8

WCAP-16346-NP A-20 Appendix A References A-1. 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-13422, "Analysis of Capsule U from the Texas Utilities Electric Company Comanche Peak Unit No. 1 Reactor Vessel Radiation Surveillance Program," July 1992.

A-3. WCAP-15144, "Analysis of Capsule Y from the TU Electric Company Comanche Peak Unit 1 Reactor Vessel Radiation Surveillance Program," January 1999.

A-4. A. Schmittroth, FERRET Data 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.

WCAP-16346-NP B-1 APPENDIX B Thermal Stress Intensity Factors (Kit)

The following page contain the thermal stress intensity factors (K,,) for the maximum heatup and cooldown rates at 36 EFPY. The vessel radius to the 1/4T and Y4T locations are as follows:

0 I/4T Radius = 88.818" 0 3/4T Radius = 93.133"

WCAP-1 6346-NP B-2 TABLE B1I K1, Values for 100 0F/hr Heatup Curve (36 EFPY)

Vessel Temperature I/4T Thermal Vessel Temperature 3/4T Thermal Stress Water @ 1/4T Location for Stress @ 3/4T Location for Intensity Factor Temp. 100*F/hr Ileatup Intensity Factor 100*F/hr Ileatup (KSI SQ. RT. IN.)

(OF) (OF) (KSI SQ. RT. IN.) (OF) 60 56 -0.9955 55 0.4729 65 59 -2.4527 55 1.4376 70 62 -3.7136 56 2.4259 75 65 -4.9121 57 3.3570 80 68 -5.9484 59 4.1917 85 72 -6.8957 61 4.9397 90 76 -7.7187 63 5.6022 95 8o -8.4708 65 6.1958 100 84 -9.1295 68 6.7232 105 88 -9.7286 71 7.1945 110 92 -10.2561 75 7.6153 115 97 -10.7377 78 7.9942 120 101 -11.1644 82 8.3345 125 106 -11.5552 85 8.6415 130 110 -11.9029 89 8.9183 135 115 -12.2227 93 9.1691 140 119 -12.5087 97 9.3963 145 124 -12.7732 102 9.6032 150 129 -13.0109 106 9.7917 155 133 -13.2322 110 9.9643 160 138 -13.4323 115 10.1225 165 143 -13.6197 119 10.2683 170 147 -13.7903 123 10.4028 175 152 -13.9514 128 10.5277 180 157 -14.0989 132 10.6437 185 162 -14.2392 137 10.7522 190 167 -14.3688 142 10.8537 195 172 -14.4928 146 10.9492 200 176 -14.6082 151 11.0394 205 181 -14.7196 156 11.1249 210 186 -14.8239 160 11.2061 0

Note: At the lowest temperatures (T = to 60OF to 90 F), the heatup curve is limited by the 3/4T pressure at T = 90'F. In that temperature range, considering the raw pressures at 1/4T, 3/4T and SS, the curve would be limited by Steady State at T = 60'F and 65°F, then by the 3/4T location there after.

WCAP- 16346-NP B-3 WCAP-1 6346-NP B-3 TABLE B2 Kit Values for 100OF/hr Cooldown Curve (36 EFPY)

Vessel Temperature a 100'F/hr Cooldown Water 1/4T Location for 1/4T Thermal Stress Temp. 100*F/hr Cooldown Intensity Factor (OF) (OF) (KSI SQ. RT. IN.)

170 196 16.6130 165 191 16.5432 160 186 16.4738 155 181 16.4039 150 176 16.3346 145 171 16.2648 140 166 16.1955 135 161 16.1258 130 156 16.0566 125 151 15.9870 120 146 15.9179 115 140 15.8486 110 135 15.7797 105 130 15.7105 100 125 15.6419 95 120 15.5730 90 115 15.5045 85 110 15.4359 80 105 15.3677 75 100 15.2993 70 95 15.2314 65 90 15.1633 60 84 15.0949 Note: At temperatures larger than T = 90', the 100OF/hr cooldowN curve is limited by a smaller cooldown rate or Stead), State.

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